Biocompatibility of root-end filling materials: recent update

Directory of Open Access Journals (Sweden)

Payal Saxena

2013-08-01

Full Text Available The purpose of a root-end filling is to establish a seal between the root canal space and the periradicular tissues. As root-end filling materials come into contact with periradicular tissues, knowledge of the tissue response is crucial. Almost every available dental restorative material has been suggested as the root-end material of choice at a certain point in the past. This literature review on root-end filling materials will evaluate and comparatively analyse the biocompatibility and tissue response to these products, with primary focus on newly introduced materials.

Measurement of enriched uranium and uranium-aluminum fuel materials with the AWCC

International Nuclear Information System (INIS)

Krick, M.S.; Menlove, H.O.; Zick, J.; Ikonomou, P.

1985-05-01

The active well coincidence counter (AWCC) was calibrated at the Chalk River Nuclear Laboratories (CRNL) for the assay of 93%-enriched fuel materials in three categories: (1) uranium-aluminum billets, (2) uranium-aluminum fuel elements, and (3) uranium metal pieces. The AWCC was a standard instrument supplied to the International Atomic Energy Agency under the International Safeguards Project Office Task A.51. Excellent agreement was obtained between the CRNL measurements and previous Los Alamos National Laboratory measurements on similar mockup fuel material. Calibration curves were obtained for each sample category. 2 refs., 8 figs., 15 tabs

Discrimination of high-Z materials in concrete-filled containers using muon scattering tomography

Science.gov (United States)

Frazão, L.; Velthuis, J.; Thomay, C.; Steer, C.

2016-07-01

An analysis method of identifying materials using muon scattering tomography is presented, which uses previous knowledge of the position of high-Z objects inside a container and distinguishes them from similar materials. In particular, simulations were performed in order to distinguish a block of Uranium from blocks of Lead and Tungsten of the same size, inside a concrete-filled drum. The results show that, knowing the shape and position from previous analysis, it is possible to distinguish 5 × 5 × 5 cm3 blocks of these materials with about 4h of muon exposure, down to 2 × 2 × 2 cm3 blocks with 70h of data using multivariate analysis (MVA). MVA uses several variables, but it does not benefit the discrimination over a simpler method using only the scatter angles. This indicates that the majority of discrimination is provided by the angular information. Momentum information is shown to provide no benefits in material discrimination.

Raw material uranium

International Nuclear Information System (INIS)

Arnold, O.

1975-01-01

In this paper some aspects are being considered, in as far as they can contribute to a better understanding of uranium as a raw material and an energy carrier, and as they can indicate the possible ways and means open to the German Federal Republic for securing this highly desirable raw material, without becoming even more dependent on the economic and political views of the producing countries, than it is the case in respect of oil. (orig.) [de

21 CFR 872.3310 - Coating material for resin fillings.

Science.gov (United States)

2010-04-01

... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Coating material for resin fillings. 872.3310... (CONTINUED) MEDICAL DEVICES DENTAL DEVICES Prosthetic Devices § 872.3310 Coating material for resin fillings. (a) Identification. A coating material for resin fillings is a device intended to be applied to the...

Topological materials discovery using electron filling constraints

Science.gov (United States)

Chen, Ru; Po, Hoi Chun; Neaton, Jeffrey B.; Vishwanath, Ashvin

2018-01-01

Nodal semimetals are classes of topological materials that have nodal-point or nodal-line Fermi surfaces, which give them novel transport and topological properties. Despite being highly sought after, there are currently very few experimental realizations, and identifying new materials candidates has mainly relied on exhaustive database searches. Here we show how recent studies on the interplay between electron filling and nonsymmorphic space-group symmetries can guide the search for filling-enforced nodal semimetals. We recast the previously derived constraints on the allowed band-insulator fillings in any space group into a new form, which enables effective screening of materials candidates based solely on their space group, electron count in the formula unit, and multiplicity of the formula unit. This criterion greatly reduces the computation load for discovering topological materials in a database of previously synthesized compounds. As a demonstration, we focus on a few selected nonsymmorphic space groups which are predicted to host filling-enforced Dirac semimetals. Of the more than 30,000 entires listed, our filling criterion alone eliminates 96% of the entries before they are passed on for further analysis. We discover a handful of candidates from this guided search; among them, the monoclinic crystal Ca2Pt2Ga is particularly promising.

Exploration on relationship between uranium and organic materials in carbonate-siliceous pelite type uranium ore deposits

International Nuclear Information System (INIS)

Dong Yongjie

1996-01-01

The author determines the content of uranium and organic carbon of part specimen of surrounding rocks and ores, which sampled from carbonate and black shale type uranium deposits in Xiushui, Jiangxi Province, and Tongcheng, Hubei Province. According to the analytical operation regulations of organic materials, extraction and separation of chloroform pitch is carried out. Internal relationships between uranium and organic derivative is discussed. The conclusion shows that: (1) certain co-relationship between U and organic carbon and chloroform extract is detected; (2) evolutionary processes of organic materials in the exogenetic uranium deposits are not all the same; (3) non-hydrocarbon is closely related to uranium, so it can be regarded as indicator of uranium gathering in exogenetic uranium deposits

Sorption of Np (Ⅴ) on Beishan granite fracture filling materials

International Nuclear Information System (INIS)

Jiang Tao; Wang Bo; Bao Liangjin; Zhou Duo; Long Haoqi; Song Zhixin; Chen Xi

2012-01-01

The sorption behaviors of Np (Ⅴ) on the granite fracture filling materials were studied by batch experiments under anaerobic in Beishan groundwater. The impact of pH of groundwater, CO 3 2- , humic acid and different components of granite fracture filling materials on sorption of Np (Ⅴ) was investigated. The results show that the granite fracture filling materials have strong capacity of Np (Ⅴ) adsorption. The value of K d , for Np (Ⅴ) sorption on the granite fracture filling materials is 843 mL/g. With the increase of pH, the value of K d increases at first and then decreases. K d of Np sorption on granite fracture filling materials in the presence of CO 3 2- and humic acid decreases. The chlorite and feldspar are major contributors to the sorption of Np (Ⅴ) on Beishan granite fracture filling materials. (authors)

Uranium industry annual 1998

International Nuclear Information System (INIS)

1999-01-01

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

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

International Nuclear Information System (INIS)

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

1997-07-01

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

Comparative evaluation of low cost materials as constructed wetland filling media

Science.gov (United States)

Pinho, Henrique J. O.; Vaz, Mafalda M.; Mateus, Dina M. R.

2017-11-01

Three waste materials from civil construction activities were assessed as low cost alternative filling materials used in Constructed Wetlands (CW). CW are green processes for wastewater treatment, whose design includes an appropriate selection of vegetation and filling material. The sustainability of such processes may be incremented using recovered wastes as filling materials. The abilities of the materials to support plant growth and to contribute to pollutants removal from wastewater were assessed and compared to expanded clay, a filling usually used in CW design. Statistical analysis, using one-way ANOVA and Welch's ANOVA, demonstrate that limestone fragments are a better choice of filling material than brick fragments and basalt gravel.

Certification of a uranium dioxide reference material for chemical analyses

International Nuclear Information System (INIS)

Le Duigou, Y.

1984-01-01

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

Uranium as Raw Material for Nuclear Energy

International Nuclear Information System (INIS)

Lelek, V.

2006-01-01

There is lot of information bringing our attention to the problem of limited raw material resources. Fortunately uranium for nuclear energy is very concentrated source and that is why its transport brings no problems and could be realized from anywhere. Second question is if overall resources are available for current nuclear energy development. Data documenting reasons for nowadays price growth are presenting and it is clearly shown that the most probable explanation is that there is gap in new uranium mines preparation and the lot of smaller mines were closed in the period of low uranium prices. Conclusion is that there is at least for the first half of this century even for thermal reactors enough uranium. Situation could be changed if there will massive production of liquid fuel using hydrogen, produced through nuclear heating. Public information about former military uranium resources are also included. Contemporary about one half of US nuclear power-stations is using high enriched uranium diluted with natural uranium - it is expected to continue this way up to 2012. Uranium is complicated market (Authors)

Effect of Different Filling Materials in Anammox Bacteria Enrichment

Directory of Open Access Journals (Sweden)

Dilek ÖZGÜN

2012-12-01

Full Text Available Purpose: Anaerobic ammonium oxidation (Anammox is a process that ammonium as electron donor is oxidized to nitrogen gas using nitrite as electron acceptor. Compared to conventional nitrification-denitrification processes, this process is used less oxygen and no organic material (methanol, glucose. However, the slow growth rate of Anammox bacteria (11-30 days is disadvantages. Therefore, batch reactors have been carried out in these bacteria enrichment. In this study continuously operated upflow anaerobic sludge reactor (UASB using different filling materials disposing of sensitive and slow-growing Anammox bacteria out of the system is purposed. Design and Methods: System is operated up-flow column reactor at 2 days hydraulic retention time (HRT in 45 days. In this study, ceramic stones and Linpor filling material are used. Using synthetic wastewater containing ammonium and nitrite, Ar/CO2 anaerobic conditions (95/5% supplied with gas. System is operated at a temperature 253 C in UASB. Temperature, pH, ammonia-nitrogen and nitrite nitrogen are measured. Results: Both filling material reactors are operated in 45 days. Ceramic stones filling reactor is observed quickly reaches 90% were used reactor ammonium removal. The ammonium nitrogen removal was slower in Linpor filling materials reactor. Nitrite removal is reached up to 90% in both the reactor. When compared to the stoichiometric equation in Linpor was composed of large amounts of nitrate. At the end of 25 days the results were similar to ceramic stone filling reactor with Linpor filling material reactors. Conclusions and Original Value: Anammox process as from nitrogen removal processes was discovered in 1995. Anammox bacteria that make up this process due to very low growth rates of microbial bacteria in the system must be kept in the system. Most of the studies in the literature, these bacteria enrichment stage is started instead of a continuous batch reactor system. In this study

Uranium refining in South Africa. The production of uranium trioxide, considering raw material properties and nuclear purity requirements

International Nuclear Information System (INIS)

Colborn, R.P.; Bayne, D.L.G.; Slabber, M.N.

1980-01-01

Conventional practice results in raw materials being delivered to the uranium refineries in a form more suitable for transportation than for processing, and therefore the refineries are required to treat these raw materials to produce an acceptable intermediate feed stock. During this treatment, it is advantageous to include a purification step to ensure that the feed stock is of the required purity for nuclear grade uranium hexafluoride production, and this usually results in ammonium diuranate slurries of the required quality being produced as the intermediate feed stock. All subsequent processing steps can therefore be standardized and are effectively independent of the origin of the raw materials. It is established practice in South Africa to transport uranium as an ammonium diuranate slurry from the various mines to the Nufcor central processing plant for UOC production, and therefore the process for the production of uranium hexafluoride in South Africa was designed to take cognizance of existing transport techniques and to accept ammonium diuranate slurries as the raw material. The South African refinery will be able to process these slurries directly to uranium trioxide. This paper discusses the conditions under which the various ammonium diuranate raw materials, exhibiting a wide range of properties, can be effectively processed to produce a uranium trioxide of acceptably consistent properties. Mention is also made of the uranium hexafluoride distillation process adopted

Differences in in vitro dissolution properties of settled and airborne uranium material

International Nuclear Information System (INIS)

Scripsick, R.C.; Crist, K.C.; Tillery, M.I.; Soderholm, S.C.

1984-01-01

The dissolution behavior of settled and airborne uranium material produced by firing of depleted uranium munitions was studied using an in vitro dissolution technique. Differences in the composition of bulk and respirable fraction samples of these materials were observed. Dissolution analysis results suggest that under some conditions a rapidly dissolving uranium fraction may be formed. This fraction may play an important role in determining hazard potential associated with inhalation exposure to uranium materials. The fact that a larger rapidly dissolving fraction was observed in the airborne material than in the settled material indicates that dissolution analysis should be performed on appropriate size fraction samples. 20 references, 3 figures, 4 tables

Material control and accounting requirements for uranium enrichment facilities

International Nuclear Information System (INIS)

Ting, P.

1991-01-01

This paper reports that the U.S. Nuclear Regulatory Commission has defined material control and accounting (MC and A) requirement for low-enriched uranium enrichment plants licensed under 10 CFR parts 40 and 70. Following detailed assessment of potential safeguards issues relevant to these facilities, a new MC and A rule was developed. The primary safeguards considerations are detection of the loss of special nuclear material, detection of clandestine production of special nuclear material of low strategic significance for unauthorized use or distribution, and detection of unauthorized production of uranium enriched to ≥10 wt % U-235. The primary safeguards concerns identified were the large absolute limit of error associated with the material balance closing, the inability to shutdown some uranium enrichment technologies to perform a cleanout inventory of the process system, and the flexibility of some of these technologies to produce higher enrichments. Unauthorized production scenarios were identified for some technologies that could circumvent the detection of the production and removal of 5 kilograms of U-235 as high-enriched uranium through conventional material control and accounting programs. Safeguards techniques, including the use of production and process control information, measurements, and technical surveillance, were identified to compensate for these concerns

Disposal and reclamation of southwestern coal and uranium wastes

International Nuclear Information System (INIS)

Wewerka, E.M.

1979-01-01

The types of solid wastes and effluents produced by the southwestern coal and uranium mining and milling industries are considered, and the current methods for the disposal and reclamation of these materials discussed. The major means of disposing of the solid wastes from both industries is by land fill or in some instances ponding. Sludges or aqueous wastes are normally discharged into settling and evaporative ponds. Basic reclamation measures for nearly all coal and uranium waste disposal sites include solids stabilization, compacting, grading, soil preparation, and revegetation. Impermeable liners and caps are beginning to be applied to disposal sites for some of the more harmful coal and uranium waste materials

Thorium and Uranium in the Rock Raw Materials Used For the Production of Building Materials

Science.gov (United States)

Pękala, Agnieszka

2017-10-01

Thorium and uranium are constant components of all soils and most minerals thereby rock raw materials. They belong to the particularly dangerous elements because of their natural radioactivity. Evaluation of the content of the radioactive elements in the rock raw materials seems to be necessary in the early stage of the raw material evaluation. The rock formations operated from deposits often are accumulated in landfills and slag heaps where the concentration of the radioactive elements can be many times higher than under natural conditions. In addition, this phenomenon may refer to buildings where rock raw materials are often the main components of the construction materials. The global control system of construction products draws particular attention to the elimination of used construction products containing excessive quantities of the natural radioactive elements. In the presented study were determined the content of thorium and uranium in rock raw materials coming from the Bełachatów lignite deposit. The Bełchatów lignite deposit extracts mainly lignite and secondary numerous accompanying minerals with the raw material importance. In the course of the field works within the framework of the carried out work has been tested 92 samples of rocks of varied petrographic composition. There were carried out analyses of the content of the radioactive elements for 50 samples of limestone of the Jurassic age, 18 samples of kaolinite clays, and 24 samples of siliceous raw materials, represented by opoka-rocks, diatomites, gaizes and clastic rocks. The measurement of content of the natural radioactive elements thorium and uranium based on measuring the frequency counts of gamma quantum, recorded separately in measuring channels. At the same time performed measurements on volume patterns radioactive: thorium and uranium. The studies were carried out in Mazar spectrometer on the powdered material. Standardly performed ten measuring cycles, after which were calculated

Uranium reference materials

International Nuclear Information System (INIS)

Donivan, S.; Chessmore, R.

1987-07-01

The Technical Measurements Center has prepared uranium mill tailings reference materials for use by remedial action contractors and cognizant federal and state agencies. Four materials were prepared with varying concentrations of radionuclides, using three tailings materials and a river-bottom soil diluent. All materials were ground, dried, and blended thoroughly to ensure homogeneity. The analyses on which the recommended values for nuclides in the reference materials are based were performed, using independent methods, by the UNC Geotech (UNC) Chemistry Laboratory, Grand Junction, Colorado, and by C.W. Sill (Sill), Idaho National Engineering Laboratory, Idaho Falls, Idaho. Several statistical tests were performed on the analytical data to characterize the reference materials. Results of these tests reveal that the four reference materials are homogeneous and that no large systematic bias exists between the analytical methods used by Sill and those used by TMC. The average values for radionuclides of the two data sets, representing an unbiased estimate, were used as the recommended values for concentrations of nuclides in the reference materials. The recommended concentrations of radionuclides in the four reference materials are provided. Use of these reference materials will aid in providing uniform standardization among measurements made by remedial action contractors. 11 refs., 9 tabs

CHARACTERISTICS OF FLORIDA FILL MATERIALS AND SOILS 1990

Science.gov (United States)

The report gives results of laboratory work by the University of Florida in support of the Foundation Fill Data Base project of the Foundation Fill Materials Specifications Task Area of the Florida Radon Research Program (FRRP). Work included determination of radon concentrations...

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

International Nuclear Information System (INIS)

Donivan, S.; Chessmore, R.

1988-12-01

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

Uranium fluoride and metallic uranium as target materials for heavy-element experiments at SHIP

Energy Technology Data Exchange (ETDEWEB)

Kindler, Birgit [Gesellschaft fuer Schwerionenforschung (GSI), Planckstrasse 1, D-64291 Darmstadt (Germany)], E-mail: b.kindler@gsi.de; Ackermann, Dieter; Hartmann, Willi; Hessberger, Fritz Peter; Hofmann, Sigurd; Huebner, Annett; Lommel, Bettina; Mann, Rido; Steiner, Jutta [Gesellschaft fuer Schwerionenforschung (GSI), Planckstrasse 1, D-64291 Darmstadt (Germany)

2008-06-01

In this contribution we describe the production and application of uranium targets for synthesis of heavy elements. The targets are prepared from uranium fluoride (UF{sub 4}) and from metallic uranium with thin carbon foils as backing. Targets of UF{sub 4} were produced by thermal evaporation in a similar way as the frequently applied targets out of Bi, Bi{sub 2}O{sub 3}, Pb, PbS, SmF{sub 3}, and NdF{sub 3,} prepared mostly from isotopically enriched material [Birgit Kindler, et al., Nucl. Instr. and Meth. A 561 (2006) 107; Bettina Lommel, et al., Nucl. Instr. and Meth. A 561 (2006) 100]. In order to use more intensive beams and to avoid scattering of the reaction products in the target, metallic uranium is favorable. However, evaporation of metallic uranium is not feasible at a sustainable yield. Therefore, we established magnetron sputtering of metallic uranium. We describe production and properties of these targets. First irradiation tests show promising results.

The existence state of uranium(VI) in portland cement matrix material immobilization body

International Nuclear Information System (INIS)

Tan Hongbin; Li Yuxiang

2005-01-01

The basis of Portland cement material reaction with uranium, the corrosion of uranium minerals in nature and the state of study on immobilization of uranium by Portland cement matrix material are introduced, and some considerations are presented. (authors)

Toxicity of Uranium Adsorbent Materials using the Microtox Toxicity Test

Energy Technology Data Exchange (ETDEWEB)

Park, Jiyeon [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jeters, Robert T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Gill, Gary A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Kuo, Li-Jung [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Bonheyo, George T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2015-10-01

The Marine Sciences Laboratory at the Pacific Northwest National Laboratory evaluated the toxicity of a diverse range of natural and synthetic materials used to extract uranium from seawater. The uranium adsorbent materials are being developed as part of the U. S. Department of Energy, Office of Nuclear Energy, Fuel Resources Program. The goal of this effort was to identify whether deployment of a farm of these materials into the marine environment would have any toxic effects on marine organisms.

Derivation of uranium residual radioactive material guidelines for the Shpack site

International Nuclear Information System (INIS)

Cheng, J.J.; Yu, C.; Monette, F.; Jones, L.

1991-08-01

Residual radioactive material guidelines for uranium were derived for the Shpack site in Norton, Massachusetts. This site has been identified for remedial action under the Formerly Utilized Sites Remedial Action Program (FUSRAP) of the US Department of Energy (DOE). The uranium guidelines were derived on the basis of the requirement that the 50-year committed effective dose equivalent to a hypothetical individual who lives or works in the immediate vicinity of the Shpack site should not exceed a dose of 100 mrem/yr following decontamination. The DOE residual radioactive material guideline computer code, RESRAD, which implements the methodology described in the DOE manual for implementing residual radioactive material guidelines, was used in this evaluation. Three potential scenarios were considered for the site; the scenarios vary with regard to time spent at the site, sources of water used, and sources of food consumed. The results of the evaluation indicate that the basic dose limit of 100 mrem/yr will not be exceeded for uranium (including uranium-234, uranium-235, and uranium-238) within 1000 years, provided that the soil concentration of combined uranium (uranium-234 and uranium-238) at the Shpack site does not exceed the following levels: 2500 pCi/g for Scenario A (recreationist: the expected scenario); 1100 pCi/g for Scenario B (industrial worker: a plausible scenario); and 53 pCi/g for Scenario C (resident farmer using a well water as the only water source: a possible but unlikely scenario). The uranium guidelines derived in this report apply to the combined activity concentration of uranium-234 and uranium-238 and were calculated on the basis of a dose of 100 mrem/yr. In setting the actual uranium guidelines for the Shpack site, DOE will apply the as low as reasonably achievable (ALARA) policy to the decision-making process, along with other factors, such as whether a particular scenario is reasonable and appropriate. 8 refs., 2 figs., 8 tabs

Mechanical behaviour of dental composite filling materials using digital holography

OpenAIRE

Monteiro, J.M.; Lopes, H.; Vaz, M.A.P.; Campos, J.C. Reis

2010-01-01

One of the most common clinical problems in dentistry is tooth decay. Among the dental filling materials used to repair tooth structure that has been destroyed by decay are dental amalgam and composite materials based on acrylics. Dental amalgam has been used by dentists for the past 150 years as a dental restorative material due to its low cost, ease of application, strength, durability, and bacteriostatic effects. However its safety as a filling material has been questioned due to th...

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

Nickel container of highly-enriched uranium bodies and sodium

Science.gov (United States)

Zinn, Walter H.

1976-01-01

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

Nickel container of highly-enriched uranium bodies and sodium

International Nuclear Information System (INIS)

Zinn, W.H.

1976-01-01

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

Density of loose-fill insulation material exposed to cyclic humidity conditions

DEFF Research Database (Denmark)

Rasmussen, Torben Valdbjørn

the granulated loose-fill material is exposed to a climate that is characterised as cyclic humidity conditions (a constant temperature and a relative humidity alternating between two predetermined constant relative humidity levels). A better understanding of the behaviour of granulated loose-fill material...

Quantifying multiple trace elements in uranium ore concentrates. An interlaboratory comparison

International Nuclear Information System (INIS)

Buerger, S.; Boulyga, S.F.; Penkin, M.V.; Jovanovic, S.; Lindvall, R.; Rasmussen, G.; Riciputi, L.

2014-01-01

An intercomparison was organized, with six laboratories tasked to quantify sixty-nine impurities in two uranium materials. The main technique employed for analysis was inductively coupled plasma mass spectrometry in combination with matrix-matched external calibration. The results presented highlight the current state-of-the-practice; lessons learned include previously unaccounted polyatomic interferences, issues related to sample dissolution, blank correction and calibration, and the challenge of estimating measurement uncertainties. The exercise yielded consensus values for the two analysed materials, suitable for use as laboratory standards to partially fill a gap in the availability of uranium reference materials characterized for impurities. (author)

Electrochemical investigation of tetravalent uranium β-diketones for active materials of all-uranium redox flow battery

International Nuclear Information System (INIS)

Yamamura, Tomoo; Shiokawa, Yoshinobu; Ikeda, Yasuhisa

2002-01-01

For active materials of the all-uranium redox flow battery for the power storage, tetravalent uranium β-diketones were investigated. The electrode reactions of U(ba) 4 and U(btfa) 4 were examined in comparison with that of U(acac) 4 , where ba denotes benzoylacetone, btfa benzoyltrifluoroacetone and acac acetylacetone. The cyclic voltammograms of U(ba) 4 and U(btfa) 4 solutions indicate that there are two series of redox reactions corresponding to the complexes with different coordination numbers of four and three. The electrode kinetics of the U(IV)/U(III) redox reactions for btfa complexes is examined. The obtained result supports that the uranium β-diketone complexes examined in the present study will serve as excellent active materials for negative electrolyte in the redox flow battery. (author)

Dissolution of uranium oxide materials in simulated lung fluid

International Nuclear Information System (INIS)

Scripsick, R.C.; Soderholm, S.C.

1985-01-01

Depleted uranium (DU) oxide aerosols prepared in the laboratory and collected in the field were tested to characterize their dissolution in simulated lung fluid and to determine how dissolution is affected by aerosol preparation. DU, a by-product of the uranium fuel cycle, has been selected by the US military for use in several types of munitions. During development, manufacture, testing, and use of these munitions, opportunities exist for inhalation exposure to various (usually oxide) aerosol forms of DU. The hazard potential associated with such exposures is closely related to the chemical form, the size of the DU aerosol material, and its dissolution properties. Five DU sample materials produced by exposing uranium alloy penetrators to certain controlled oxidation atmospheres were studied (oxidation temperatures ranged from 500 to 900 0 C). In addition, two DU sample materials collected in the field were provided by the US Air Force. All sample materials were generated as aerosols and the respirable fraction was separated and collected. Data suggest that under some conditions a rapidly dissolving U 3 O 8 fraction may be formed concurrent with the production of UO 2

Gamma-spectrometric determination of {sup 232}U in uranium-bearing materials

Energy Technology Data Exchange (ETDEWEB)

Zsigrai, Jozsef [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), 76125 Karlsruhe, P.O. Box 2340 (Germany); Nguyen, Tam Cong [Centre for Energy Research of the Hungarian Academy of Sciences (EK), 1525 Budapest 114, P.O. Box 49 (Hungary); Berlizov, Andrey [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), 76125 Karlsruhe, P.O. Box 2340 (Germany)

2015-09-15

The {sup 232}U content of various uranium-bearing items was measured using low-background gamma spectrometry. The method is independent of the measurement geometry, sample form and chemical composition. Since {sup 232}U is an artificially produced isotope, it carries information about previous irradiation of the material, which is relevant for nuclear forensics, nuclear safeguards and for nuclear reactor operations. A correlation between the {sup 232}U content and {sup 235}U enrichment of the investigated samples has been established, which is consistent with theoretical predictions. It is also shown how the correlation of the mass ratio {sup 232}U/{sup 235}U vs. {sup 235}U content can be used to distinguish materials contaminated with reprocessed uranium from materials made of reprocessed uranium.

Innovations in endodontic filling materials: guttapercha vs Resilon.

Science.gov (United States)

de Souza Filho, Francisco José; Gallina, Giuseppe; Gallottini, Livio; Russo, Riccardo; Cumbo, Enzo Maria

2012-01-01

One of the goals of endodontic treatment is to achieve a complete, tridimensional, hermetic sealing of the root canal system to prevent the entry of microorganisms or their products through both the coronal and apical pathways. Gutta-percha is the most widely used material for root canal filling and despite its numerous properties, such as biocompatibility and thermoplasticity, it has however an important limit: the lack of adhesion to the canal walls. Attempts to address this problem have been made over the years by using endodontic cements capable of bonding to canal dentine but their tendency to resorption in time can compromise the quality of treatment. The first step towards a real adhesive endodontic filling(4) is rather recent; in fact, it goes back to 2003 when, on the occasion of the American Dental Association (ADA) Annual Session, Resilon Research LLC introduced a new canal filling adhesive system based on a thermoplastic synthetic resin material called Resilon™. The real innovation of this system is its capacity of creating a core made of Resilon™ bonded to the cement which adheres to dentine walls previously conditioned with a self-etching primer(4) so no changes in the techniques of canal preparation are required. The aim of this study was to evaluate the capacity of two filling materials (gutta-percha and Resilon) to adapt to the canal anatomy, especially on the apical third, using the continuous wave of condensation technique. Our data suggest that in the third apical the gutta-percha best shows rheological properties that are as important as the bond capability.

Effectiveness of rotatory and reciprocating movements in root canal filling material removal

Directory of Open Access Journals (Sweden)

Emmanuel João Nogueira Leal SILVA

2015-01-01

Full Text Available The purpose of this study was to compare the efficacy of reciprocating and rotary techniques for removing gutta-percha and sealer from root canals. Forty straight and oval single-rooted premolars were prepared up to size 30, filled with gutta-percha and sealer, and then randomly allocated to two experimental retreatment groups: ProTaper Retreatment System (PTRS and WaveOne System (WS. Procedural errors, time of retreatment and apically extruded material were recorded for all the roots. The roots were radiographed after retreatment. The percentage of residual material was calculated using image analysis software. The data were analyzed by Kolmogorov-Smirnov and t tests, with a significance level set at 5%. No system completely removed the root filling material from the root canal. No significant differences were observed between the systems, in terms of residual filling material in any tested third (p > 0.05. WS was faster in removing filling material than PTRS (p< 0.05. Extrusion was observed in 4 cases in PTRS and in 5 cases in WS. No procedural errors were observed in either group. It can be concluded that although no differences were observed in the efficacy of PTRS and WS for removing root filling material, WS was faster than PTRS.

Study of Uranium Concentrations in Water and Organic Material from Streams in Sweden

International Nuclear Information System (INIS)

Ek, J.

1981-12-01

The purpose of the investigation has been to study how uranium concentrations in stream water and organic material are related to various geological parameters such as rock types, average uranium content and radioactivity, fracturing, leachability of uranium from the bedrock, occurrence of uranium mineralisations and thickness and type of Quarternary deposits. The investigation has also taken account of the effects of environmental factors such as climate , precipitation, height above sea level and topography. The background concentration of uranium in organic stream sediment varies from 1 ppm to 45 ppm, with a background value of 10 ppm for all 14 areas considered together. The threshold value for organic stream material varies from 3 ppm U to 303 ppm U with a threshold value of 133 ppm U for all 14 areas considered together. For water, the background concentration varies between the 5 areas from 0.2 ppb U to 0.7 ppb U with a background value of 0.4 ppb U for all 5 areas together. The threshold value varies from 0.3 ppb U to 5.2 ppb U with a threshold value of 2.9 ppb U for all 5 areas together. An investigation of the correlation between uranium concentrations in water and organic stream material from one and the same sampling point shows a positive correlation for high concentrations, but the correlation becomes successively less significant with lower concentrations. Uranium concentrations in organic stream material and water are positively correlated with the following geological parameters:1) Background concentrations of uranium in the bedrock. 2) Abundance of fractures in the bedrock. 3) Leachability of uranium from the bedrock. 4) Presence of uranium mineralisations. For organic stream material, this positive correlation is obtained for both high and low uranium concentrations whereas for water it occurs only with high concentrations. In areas of broken topography and high relief, there is a more clearly defined correlation to the bedrock than in areas of

Experimental study of strontium sorption on fissure filling material

Energy Technology Data Exchange (ETDEWEB)

Eriksen, T E; Cui, Daqing [Royal Institute of Technology, Stockholm (Sweden). Dept. of Chemistry

1994-12-01

We have carried out a comparative study of sorption and desorption of strontium in groundwater on separated magnetic and size fractions of fissure filling material taken from natural fissures in granitic rock. Complete reversibility of the sorption process was demonstrated by identical Freundlich isotherms, isotopic exchangeability and pH dependence of the distribution coefficients Rd. The sorption was found to be strongly pH dependent in the range 3-11. The pH effect can be accommodated in the sorption model by considering the surface areas and surface charges of the minerals in the fissure filling material. 20 refs, 9 figs, 3 tabs.

Experimental study of strontium sorption on fissure filling material

International Nuclear Information System (INIS)

Eriksen, T.E.; Cui, Daqing

1994-12-01

We have carried out a comparative study of sorption and desorption of strontium in groundwater on separated magnetic and size fractions of fissure filling material taken from natural fissures in granitic rock. Complete reversibility of the sorption process was demonstrated by identical Freundlich isotherms, isotopic exchangeability and pH dependence of the distribution coefficients Rd. The sorption was found to be strongly pH dependent in the range 3-11. The pH effect can be accommodated in the sorption model by considering the surface areas and surface charges of the minerals in the fissure filling material. 20 refs, 9 figs, 3 tabs

Experimental Study of Goaf Filling Materials Based on Red Mud

Science.gov (United States)

Mu, Mangen; Gao, Xiaozhen; Guo, Taoming; Hu, Xinping

2018-01-01

Red mud as soild waste is difficult to treatment. Goaf filling materials can make a large use of red mud. By the experimental study,we find that the red mud, fly ash, ground slag and desulfida-tion gypsum can be used to make goaf filling materials based on the principle of alkali excitation and metalion stability.Through the control variable method, we find that the optimal proportion of goaf filling materials based on red mud is red mud 55%, fly ash 30%, cement 7.5%, fly ash 2.5%, desulfurization gypsum 5%, admixture 1%, and water solid ratio=1:1.2.The 28days final material strength was 2.0 MPa,which achives the technical specification requirements.Through the test of SEM, XRD and IR, it is indicated that the strength formation of goaf filling material based on red mud is from the unformed linking hydration products of amorphous alkali excitation system. With curing time from 3 to 7 days, the unformed linking hydration products grown a lot of vitreous hydration products. When hydration reaction basicly finished after 28 days, the hydration products have developed into a large volume of massive vitreous with an extremely dense structure. The Ca2SiO3 mineral phase is significantly reduced, which is participate in hydration reactions. The decrease of Ca2SiO3 indicates that the Si-O bond in the system have been ruptured and reorganized.

77 FR 60482 - Regulatory Guide 5.67, Material Control and Accounting for Uranium Enrichment Facilities...

Science.gov (United States)

2012-10-03

... Accounting for Uranium Enrichment Facilities Authorized To Produce Special Nuclear Material of Low Strategic... Accounting for Uranium Enrichment Facilities Authorized to Produce Special Nuclear Material of Low Strategic... and is applicable to the Paducah GDP and other uranium enrichment facilities that have been licensed...

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

International Nuclear Information System (INIS)

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

1997-01-01

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

Carbon Nanotubes Filled with Ferromagnetic Materials

Directory of Open Access Journals (Sweden)

Albrecht Leonhardt

2010-08-01

Full Text Available Carbon nanotubes (CNT filled with ferromagnetic metals like iron, cobalt or nickel are new and very interesting nanostructured materials with a number of unique properties. In this paper we give an overview about different chemical vapor deposition (CVD methods for their synthesis and discuss the influence of selected growth parameters. In addition we evaluate possible growth mechanisms involved in their formation. Moreover we show their identified structural and magnetic properties. On the basis of these properties we present different application possibilities. Some selected examples reveal the high potential of these materials in the field of medicine and nanotechnology.

Carbon Nanotubes Filled with Ferromagnetic Materials

Science.gov (United States)

Weissker, Uhland; Hampel, Silke; Leonhardt, Albrecht; Büchner, Bernd

2010-01-01

Carbon nanotubes (CNT) filled with ferromagnetic metals like iron, cobalt or nickel are new and very interesting nanostructured materials with a number of unique properties. In this paper we give an overview about different chemical vapor deposition (CVD) methods for their synthesis and discuss the influence of selected growth parameters. In addition we evaluate possible growth mechanisms involved in their formation. Moreover we show their identified structural and magnetic properties. On the basis of these properties we present different application possibilities. Some selected examples reveal the high potential of these materials in the field of medicine and nanotechnology. PMID:28883334

Department of Energy depleted uranium recycle

International Nuclear Information System (INIS)

Kosinski, F.E.; Butturini, W.G.; Kurtz, J.J.

1994-01-01

With its strategic supply of depleted uranium, the Department of Energy is studying reuse of the material in nuclear radiation shields, military hardware, and commercial applications. the study is expected to warrant a more detailed uranium recycle plan which would include consideration of a demonstration program and a program implementation decision. Such a program, if implemented, would become the largest nuclear material recycle program in the history of the Department of Energy. The bulk of the current inventory of depleted uranium is stored in 14-ton cylinders in the form of solid uranium hexafluoride (UF 6 ). The radioactive 235 U content has been reduced to a concentration of 0.2% to 0.4%. Present estimates indicate there are about 55,000 UF 6 -filled cylinders in inventory and planned operations will provide another 2,500 cylinders of depleted uranium each year. The United States government, under the auspices of the Department of Energy, considers the depleted uranium a highly-refined strategic resource of significant value. A possible utilization of a large portion of the depleted uranium inventory is as radiation shielding for spent reactor fuels and high-level radioactive waste. To this end, the Department of Energy study to-date has included a preliminary technical review to ascertain DOE chemical forms useful for commercial products. The presentation summarized the information including preliminary cost estimates. The status of commercial uranium processing is discussed. With a shrinking market, the number of chemical conversion and fabrication plants is reduced; however, the commercial capability does exist for chemical conversion of the UF 6 to the metal form and for the fabrication of uranium radiation shields and other uranium products. Department of Energy facilities no longer possess a capability for depleted uranium chemical conversion

Measuring the fill height of sealed cans with a compound pendulum

International Nuclear Information System (INIS)

Rinard, P.M.

1995-06-01

A compound pendulum has been designed, fabricated, tested, and used to determine the fill height of material in sealed cans. The specific cans that stimulated this work are partially filled with uranium and plutonium oxide. Fill height affects nondestructive assays using fission neutrons, but corrections for various fill heights can be made once the height is known. Heights vary with use as the powder compacts or loosens, so it is necessary to determine the height at the time of the neutron measurement. The pendulum is small and readily portable so it can be taken to the location of the neutron measurement. Tests with open cans filled with sand to various known heights had accuracies generally within 3%. Factors that can affect the accuracy are examined and discussed. Experience in using the pendulum on sealed cans is related

Contribution to the study of nuclear fuel materials with a metallic uranium base

International Nuclear Information System (INIS)

Englander, M.

1957-11-01

In a power reactor destined to supply industrially recoverable thermal energy, the most economical source of heat still consists of natural metallic uranium. However, the nuclear fuel material, most often employed in the form of rods of 20 to 40 mm diameter, is subjected to a series of stresses which lead to irreversible distortions usually incompatible with the substructure of the reactor. As a result the fuel material must possess at the outset a certain number of qualities which must be determined. Investigations have therefore been carried out, first on the technological characters peculiar to each of the three allotropic phases of pure uranium metal, and on their interactions on the stabilisation of the material which consists of either cast uranium or uranium pile-treated in the γ phase. (author) [fr

Inorganic-Organic hybrid materials for uranium extraction from phosphoric acid

International Nuclear Information System (INIS)

El-Mourabit, Sabah

2013-01-01

Phosphate rocks are industrially processed in large quantities to produce phosphoric acid and fertilisers. These rocks contain significant concentration of uranium (50 to 300 ppm) which could be interesting for nuclear industry. This work deals with the valorisation of uranium as a by-product from fertiliser industry. The aim of this study is to develop a hybrid material, constituted of an inorganic solid support grafted with an extractant (complexing molecule), which can extract selectively uranium from phosphoric acid medium. The first step of our approach was to identify an inorganic support which is stable under these particular conditions (strong acidity and complexing medium). The chemical and mechanical stability of different meso-porous materials, such as silica, glass and carbon was studied. In a second phase, we focused on the identification and the optimisation of complexing molecules, specific of uranium in phosphoric acid. These ligands were then grafted on the most stable solids. Finally, the efficiency of these hybrid systems was evaluated through different tests of extraction, selectivity and de-extraction. (author) [fr

Designing strip footing foundations using expanded polystyrene (EPS) as fill material

DEFF Research Database (Denmark)

Psarropoulos, Prodromos; Zania, Varvara; Spyrakos, Konstantinos

2010-01-01

One of the modern uses of expanded polystyrene (EPS) is in strip footings as fill material. The current study investigates the effect of the geofoam filling in the static and seismic design of the base slab founded on strip footings. For this purpose the finite element method is employed, and three......-dimensional as well as two-dimensional parametric analyses are conducted taking into account static and seismic loading conditions. The interaction of the soil–geofoam–foundation system is taken into consideration. The use of EPS as fill material in foundation systems is proven to be not only technically but also...

Uranium

International Nuclear Information System (INIS)

Hamdoun, N.A.

2007-01-01

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

Computed tomography assessment of the efficiency of different techniques for removal of root canal filling material

International Nuclear Information System (INIS)

Dall'agnol, Cristina; Barletta, Fernando Branco; Hartmann, Mateus Silveira Martins

2008-01-01

This study evaluated the efficiency of different techniques for removal of filling material from root canals, using computed tomography (CT). Sixty mesial roots from extracted human mandibular molars were used. Root canals were filled and, after 6 months, the teeth were randomly assigned to 3 groups, according to the root-filling removal technique: Group A - hand instrumentation with K-type files; Group B - reciprocating instrumentation with engine-driven K-type files; and Group C rotary instrumentation with engine-driven ProTaper system. CT scans were used to assess the volume of filling material inside the root canals before and after the removal procedure. In both moments, the area of filling material was outlined by an experienced radiologist and the volume of filling material was automatically calculated by the CT software program. Based on the volume of initial and residual filling material of each specimen, the percentage of filling material removed from the root canals by the different techniques was calculated. Data were analyzed statistically by ANOVA and chi-square test for linear trend (α=0.05). No statistically significant difference (p=0.36) was found among the groups regarding the percent means of removed filling material. The analysis of the association between the percentage of filling material removal (high or low) and the proposed techniques by chi-square test showed statistically significant difference (p=0.015), as most cases in group B (reciprocating technique) presented less than 50% of filling material removed (low percent removal). In conclusion, none of the techniques evaluated in this study was effective in providing complete removal of filling material from the root canals. (author)

Computed tomography assessment of the efficiency of different techniques for removal of root canal filling material

Energy Technology Data Exchange (ETDEWEB)

Dall' agnol, Cristina; Barletta, Fernando Branco [Lutheran University of Brazil, Canoas, RS (Brazil). Dental School. Dept. of Dentistry and Endodontics]. E-mail: fbarletta@terra.com.br; Hartmann, Mateus Silveira Martins [Uninga Dental School, Passo Fundo, RS (Brazil). Postgraduate Program in Dentistry

2008-07-01

This study evaluated the efficiency of different techniques for removal of filling material from root canals, using computed tomography (CT). Sixty mesial roots from extracted human mandibular molars were used. Root canals were filled and, after 6 months, the teeth were randomly assigned to 3 groups, according to the root-filling removal technique: Group A - hand instrumentation with K-type files; Group B - reciprocating instrumentation with engine-driven K-type files; and Group C rotary instrumentation with engine-driven ProTaper system. CT scans were used to assess the volume of filling material inside the root canals before and after the removal procedure. In both moments, the area of filling material was outlined by an experienced radiologist and the volume of filling material was automatically calculated by the CT software program. Based on the volume of initial and residual filling material of each specimen, the percentage of filling material removed from the root canals by the different techniques was calculated. Data were analyzed statistically by ANOVA and chi-square test for linear trend ({alpha}=0.05). No statistically significant difference (p=0.36) was found among the groups regarding the percent means of removed filling material. The analysis of the association between the percentage of filling material removal (high or low) and the proposed techniques by chi-square test showed statistically significant difference (p=0.015), as most cases in group B (reciprocating technique) presented less than 50% of filling material removed (low percent removal). In conclusion, none of the techniques evaluated in this study was effective in providing complete removal of filling material from the root canals. (author)

Radiological assessment of an area with uranium residual material

International Nuclear Information System (INIS)

Perez-Sanchez, Danyl; Cancio, David; Alvarez, Alicia

2008-01-01

As a result of a pilot project developed at the old Spanish 'Junta de Energia Nuclear' to extract uranium from ores, tailings materials were generated. Most of these residual materials were sent back to different uranium mines, but a small amount of it was mixed with conventional building materials and deposited near the old plant until the surrounding ground was flattened. The affected land is included in an area under institutional control and used as recreational area. At the time of processing, uranium isotopes were separated but other radionuclides of the uranium decays series as 230 Th, 226 Ra and daughters remain in the residue. Recently, the analyses of samples taken at different ground's depths confirm their presence. This paper presents the methodology used to calculate the derived concentration level to ensure the reference dose level of 0.1 mSv y-1 used as radiological criteria. In this study, a radiological impact assessment was performed modelling the area as recreational scenario. The modelization study was carried out with the code RESRAD considering as exposure pathways, external irradiation, inadvertent ingestion of soil, inhalation of resuspended particles, and inhalation of outdoor radon ( 222 Rn). As result was concluded that, if the concentration of 226 Ra in the first 15 cm of soil is lower than, 0.34 Bq g-1 , the dose would not exceed the reference dose. Applying this value as a derived concentration level and comparing with the results of measurements on the ground, some areas with a concentration of activity slightly higher than latter were found. In these zones the remediation proposal has been to cover with a layer of 15 cm of clean material. This action represents a reduction of 85% of the dose and ensures compliance with the reference dose. (author)

Characterization of Uranium-Bearing Material by Passive Non-Destructive Gamma Spectrometry

International Nuclear Information System (INIS)

Lakosi, L.; Zsigrai, J.; Nguyen, C.T.

2009-01-01

Characterization of nuclear materials is equally important in nuclear safeguards (inventory verification) and in nuclear security (revealing illicit trafficking). Analysis of materials is a key issue in both fields. Natural (NU), depleted (DU), low-enriched (LEU), and high-enriched uranium (HEU) samples were analysed by high resolution gamma spectrometry (HRGS). Isotopic composition and total U-content of reactor fuel pellets and powder were determined. A unique HRGS method was developed for the first time for determining the production date of the material of unknown origin. Identifying reprocessed uranium proved to be possible by HRGS as well.

Uranium in phosphorus-bearing raw materials and technological problems of its recovery

Energy Technology Data Exchange (ETDEWEB)

Gorecki, H; Gorecka, H [Politechnika Wroclawska (Poland)

1981-01-01

A problem of uranium recovery from phosphorus-bearinq raw materials is discussed. The different methods of uranium recovery from extractive phosphoric acid are briefly described. The information on their applications in the industry is also given.

Nuclear material control and accounting system evaluation in uranium conversion operations

International Nuclear Information System (INIS)

Moreira, Jose Pontes

1994-01-01

The Nuclear Material Control and Accounting Systems in uranium conversion operations are described. The conversion plant, uses ammonium diuranate (ADU), as starting material for the production of uranium hexafluoride. A combination of accountability and verification measurement is used to verify physical inventory quantities. Two types of inspection are used to minimize the measurements uncertainty of the Material Unaccounted For (MUF) : Attribute inspection and Variation inspection. The mass balance equation is the base of an evaluation of a Material Balance Area (MBA). Statistical inference is employed to facilitate rapid inventory taking and enhance material control of Safeguards. The calculation of one sampling plan for a MBA and the methodology of inspection evaluation are also described. We have two kinds of errors : no detection and false delation. (author)

Improving gross count gamma-ray logging in uranium mining with the NGRS probe

Directory of Open Access Journals (Sweden)

Carasco C.

2018-01-01

Full Text Available AREVA Mines and the Nuclear Measurement Laboratory of CEA Cadarache are collaborating to improve the sensitivity and precision of uranium concentration measurement by means of gamma ray logging. The determination of uranium concentration in boreholes is performed with the Natural Gamma Ray Sonde (NGRS based on a NaI(Tl scintillation detector. The total gamma count rate is converted into uranium concentration using a calibration coefficient measured in concrete blocks with known uranium concentration in the AREVA Mines calibration facility located in Bessines, France. Until now, to take into account gamma attenuation in a variety of boreholes diameters, tubing materials, diameters and thicknesses, filling fluid densities and compositions, a semi-empirical formula was used to correct the calibration coefficient measured in Bessines facility. In this work, we propose to use Monte Carlo simulations to improve gamma attenuation corrections. To this purpose, the NGRS probe and the calibration measurements in the standard concrete blocks have been modeled with MCNP computer code. The calibration coefficient determined by simulation, 5.3 s-1.ppmU-1 ± 10%, is in good agreement with the one measured in Bessines, 5.2 s-1.ppmU-1. Based on the validated MCNP model, several parametric studies have been performed. For instance, the rock density and chemical composition proved to have a limited impact on the calibration coefficient. However, gamma self-absorption in uranium leads to a nonlinear relationship between count rate and uranium concentration beyond approximately 1% of uranium weight fraction, the underestimation of the uranium content reaching more than a factor 2.5 for a 50 % uranium weight fraction. Next steps will concern parametric studies with different tubing materials, diameters and thicknesses, as well as different borehole filling fluids representative of real measurement conditions.

Improving gross count gamma-ray logging in uranium mining with the NGRS probe

Science.gov (United States)

Carasco, C.; Pérot, B.; Ma, J.-L.; Toubon, H.; Dubille-Auchère, A.

2018-01-01

AREVA Mines and the Nuclear Measurement Laboratory of CEA Cadarache are collaborating to improve the sensitivity and precision of uranium concentration measurement by means of gamma ray logging. The determination of uranium concentration in boreholes is performed with the Natural Gamma Ray Sonde (NGRS) based on a NaI(Tl) scintillation detector. The total gamma count rate is converted into uranium concentration using a calibration coefficient measured in concrete blocks with known uranium concentration in the AREVA Mines calibration facility located in Bessines, France. Until now, to take into account gamma attenuation in a variety of boreholes diameters, tubing materials, diameters and thicknesses, filling fluid densities and compositions, a semi-empirical formula was used to correct the calibration coefficient measured in Bessines facility. In this work, we propose to use Monte Carlo simulations to improve gamma attenuation corrections. To this purpose, the NGRS probe and the calibration measurements in the standard concrete blocks have been modeled with MCNP computer code. The calibration coefficient determined by simulation, 5.3 s-1.ppmU-1 ± 10%, is in good agreement with the one measured in Bessines, 5.2 s-1.ppmU-1. Based on the validated MCNP model, several parametric studies have been performed. For instance, the rock density and chemical composition proved to have a limited impact on the calibration coefficient. However, gamma self-absorption in uranium leads to a nonlinear relationship between count rate and uranium concentration beyond approximately 1% of uranium weight fraction, the underestimation of the uranium content reaching more than a factor 2.5 for a 50 % uranium weight fraction. Next steps will concern parametric studies with different tubing materials, diameters and thicknesses, as well as different borehole filling fluids representative of real measurement conditions.

Effect of bulk-fill base material on fracture strength of root-filled teeth restored with laminate resin composite restorations.

Science.gov (United States)

Taha, N A; Maghaireh, G A; Ghannam, A S; Palamara, J E

2017-08-01

To evaluate the effect of using a bulk-fill flowable base material on fracture strength and fracture patterns of root-filled maxillary premolars with MOD preparations restored with laminate restorations. Fifty extracted maxillary premolars were selected for the study. Standardized MOD cavities with endodontic treatment were prepared for all teeth, except for intact control. The teeth were divided randomly into five groups (n=10); (Group 1) sound teeth, (Group 2) unrestored teeth; (Group 3) MOD cavities with Vitrebond base and resin-based composite (Ceram. X One Universal); (Group 4) MOD cavities with 2mm GIC base (Fuji IX GP) and resin-based composite (Ceram. X One Universal) open laminate, (Group 5) MOD cavities were restored with 4mm of bulk-fill flowable base material (SDR) and resin-based composite (Ceram. X One Universal). All teeth were thermocycled and subjected to a 45° ramped oblique load in a universal testing machine. Fracture load and fracture patterns were recorded. Data were analyzed using one-way ANOVA and Dunnett's T3 test. Restoration in general increased the fracture strength compared to unrestored teeth. The fracture strength of group 5 (bulk-fill) was significantly higher than the fracture strength of the GIC laminate groups and not significantly different from the intact teeth (355±112N, P=0.118). The type of failure was unfavorable for most of the groups, with the majority being mixed failures. The use of a bulk-fill flowable base material significantly increased the fracture strength of extracted root-filled teeth with MOD cavities; however it did not improve fracture patterns to more favorable ones. Investigating restorative techniques that may improve the longevity of root-filled premolar teeth restored with direct resin restorations. Copyright © 2017 Elsevier Ltd. All rights reserved.

Introduction. Physicochemical aspects of uranium concentrates obtaining from the wastes and raw materials

International Nuclear Information System (INIS)

Mirsaidov, I.U.

2014-01-01

The uranium deposits of Tajikistan played an immensely significant role in the practical solution of a radioactive raw materials problem which appeared during the post-World War II years in the USSR. The pioneer in this field became complex №6 (currently known as 'Vostokredmet'). The first soviet uranium was produced from the ores extracted from the republic's deposits. For 50 years (1945-1995 y.), uranium bearing raw materials from all over the former USSR were delivered to Tajikistan, and uranium oxide was produced, which was later delivered back to Russia for further production of enriched uranium. The total volume of uranium produced in Tajikistan plants was approximately 100 thousands tons. In Sughd region, during that period, more than 55 million tons of uranium waste was accumulated. The total activity of the waste, according to different calculations, is approximately 240-285 TBq. The total amount of waste in dumps and tailings piles is estimated to be more than 170 million tons, most of which are located in the neighborhoods of hydrometallurgical plants and heap leaching locations. Uranium industry wastes in Northern Tajikistan have become attractive for different investors and commercial companies, from secondary reprocessing of mines and tailings' point of view, since the uranium price is increasing. In this regard, research on developing uranium extraction methods from wastes is broadening. The study of the possibility and economic reasonability of reprocessing former year's dumps requires comprehensive examination, and relates not only to uranium extraction but to safe extraction of dumps from tailings as well.

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

Recovering uranium and/or aluminium from refractory silico-aluminous material

International Nuclear Information System (INIS)

Livesey-Goldblatt, E.; Nagy, I.F.; Tunley, T.H.

1983-01-01

A process for recovering uranium and/or aluminium from a refractory silico-aluminous material comprises leaching the material in one or more stages, obtaining a pregnant solution which contains little or no acid and recovering the desired metal from the solution

Recycling and disposal of FUSRAP materials from the Ashland 2 site at a licensed uranium mill

International Nuclear Information System (INIS)

Howard, B.; Conboy, D.; Rehmann, M.; Roberts, H.

1999-01-01

During World War II the Manhattan Engineering District (MED) used facilities near Buffalo, N.Y. to extract natural uranium from ores. Some of the byproduct material left from the ores (MED byproduct), containing low levels of uranium, thorium, and radium, was deposited on a disposal site known as Ashland 2, located in Tonawanda, NY. On behalf of the United States Army Corps of Engineers (USACE, or the Corps), ICF Kaiser Engineers (ICFKE) was tasked to provide the best value clean-up results that meet all of the criteria established in the Record of Decision for the site. International Uranium (USA) Corporation (IUC), the operator of the White Mesa Uranium Mill, a Nuclear Regulatory Commission (NRC)-licensed mill near Blanding, Utah, was selected to perform uranium extraction on the excavated materials, therefore giving the best value as it provided beneficial use of the material consistent with the Resource Conservation and Recovery Act (RCRA) intent to encourage recycling and recovery, while also providing the most cost-effective means of disposal. Challenges overcome to complete this project included (1) identifying the best-value location to accept the material; (2) meeting regulatory requirements with IUC obtaining an NRC license amendment to accept and process the material as an alternate feed; (3) excavating and preparing the material for shipment, then shipping the material to the Mill for uranium recovery; and (4) processing the material, followed by disposal of tailings from the process in the Mill's licensed uranium tailings facility. Excavation from Ashland 2 and processing of the Ashland 2 material at the White Mesa Mill resulted in a cleaner environment at Tonawanda, a cost avoidance of up to $16 million, beneficial recovery of source material, and environmentally protective disposal of byproduct material. (author)

Uranium conversion

International Nuclear Information System (INIS)

Oliver, Lena; Peterson, Jenny; Wilhelmsen, Katarina

2006-03-01

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

Multifunctional composite material based on carbon-filled polyurethane

International Nuclear Information System (INIS)

Malinovskaya, T; Melentyev, S; Pavlov, S

2015-01-01

The research paper deals with the performance of composite resistive material heating coatings based on the polyurethane binder, filled with colloidal-graphite preparation C- 1, which can be used in structures of electric heaters. Frequency dependences of transmission and reflection coefficients, dielectric permeability of composite materials with the various content of carbon fillers (technical carbon, graphite) in polyurethane varnish in ranges of frequencies 26-40 GHz and 110-260 GHz are experimentally investigated. (paper)

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

Uranium and thorium mining and milling: material security and risk assessment

International Nuclear Information System (INIS)

Steinhaeusler, F.; Zaitseva, L.

2005-01-01

Full text: At present physical protection for the front end of the nuclear fuel cycle is typically at a significantly lower level than at any other part of the nuclear fuel cycle. In view of past experiences (Israel, South Africa, Pakistan, India) it is feasible to take into consideration some generic threat scenarios, potentially resulting in loss of control over uranium or thorium, respectively their concentrates, such as: illegal mining of an officially closed uranium- or thorium mine; covert diversion of uranium- or thorium ore whilst officially mining another ore; covert transport of radioactive ore or product, using means of public rail, road, ship, or air transport; covert en route diversion of an authorized uranium- or thorium transport; covert removal of uranium-or thorium ore or concentrate from an abandoned facility. The Stanford-Salzburg database on nuclear smuggling, theft, and orphan radiation sources (DSTO) contains information on trafficking incidents involving mostly uranium, but also some thorium, from 30 countries in five continents with altogether 113 incidents in the period 1991 to 2004. These activities range from uranium transported in backpacks by couriers in Afghanistan, to a terrorist organization purchasing land in order to mine covertly for uranium in Australia, and the clandestine shipment of almost two tons of uranium hexafluoride from Asia to Africa, using the services of a national airline. Potential participants in such illegal operations range from entrepreneurs to members of organized crime, depending on the level of sophistication of the operation. End-users and 'customers' of such illegal operations are suspected to be non-state actors, organizations or governments involved in a covert operation with the ultimate aim to acquire a sufficient amount of nuclear material for a nuclear device. The actual risk for these activities to succeed in the acquisition of an adequate amount of suitable radioactive material depends on one or

The application of waste fly ash and construction-waste in cement filling material in goaf

Science.gov (United States)

Chen, W. X.; Xiao, F. K.; Guan, X. H.; Cheng, Y.; Shi, X. P.; Liu, S. M.; Wang, W. W.

2018-01-01

As the process of urbanization accelerated, resulting in a large number of abandoned fly ash and construction waste, which have occupied the farmland and polluted the environment. In this paper, a large number of construction waste and abandoned fly ash are mixed into the filling material in goaf, the best formula of the filling material which containing a large amount of abandoned fly ash and construction waste is obtained, and the performance of the filling material is analyzed. The experimental results show that the cost of filling material is very low while the performance is very good, which have a good prospect in goaf.

Derivation of uranium residual radioactive material guidelines for the Ventron site

International Nuclear Information System (INIS)

Loureiro, C.; Yu, C.; Jones, L.

1992-03-01

Residual radioactive material guidelines for uranium were derived for the Ventron site in Beverly, Massachusetts. This site has been identified for remedial action under the Formerly Utilized Sites Remedial Action Program of the US Department of Energy (DOE). The derivations for the single radionuclides and the total uranium guidelines were based on the requirement that the 50-year committed effective dose equivalent to a hypothetical individual who lives or works in the immediate vicinity of the Ventron site should not exceed a dose of 100 mrem/yr following remedial action. The DOE residual radioactive material guideline computer code, RESRAD, which implements the methodology described in the DOE manual for implementing residual radioactive material guidelines, was used in this evaluation

Colloid transport in model fracture filling materials

Science.gov (United States)

Wold, S.; Garcia-Garcia, S.; Jonsson, M.

2010-12-01

Colloid transport in model fracture filling materials Susanna Wold*, Sandra García-García and Mats Jonsson KTH Chemical Science and Engineering Royal Institute of Technology, SE-100 44 Stockholm, Sweden *Corresponding author: E-mail: wold@kth.se Phone: +46 8 790 6295 In colloid transport in water-bearing fractures, the retardation depends on interactions with the fracture surface by sorption or filtration. These mechanisms are difficult to separate. A rougher surface will give a larger area available for sorption, and also when a particle is physically hindered, it approaches the surface and enables further sorption. Sorption can be explained by electrostatics were the strongest sorption on minerals always is observed at pH below pHpzc (Filby et al., 2008). The adhesion of colloids to mineral surfaces is related to the surface roughness according to a recent study (Darbha et al., 2010). There is a large variation in the characteristics of water-bearing fractures in bedrock in terms of aperture distribution, flow velocity, surface roughness, mineral distributions, presence of fracture filling material, and biological and organic material, which is hard to implement in modeling. The aim of this work was to study the transport of negatively charged colloids in model fracture filling material in relation to flow, porosity, mineral type, colloid size, and surface charge distribution. In addition, the impact on transport of colloids of mixing model fracture filling materials with different retention and immobilization capacities, determined by batch sorption experiments, was investigated. The transport of Na-montmorillonite colloids and well-defined negatively charged latex microspheres of 50, 100, and 200 nm diameter were studied in either columns containing quartz or quartz mixed with biotite. The ionic strength in the solution was exclusively 0.001 and pH 6 or 8.5. The flow rates used were 0.002, 0.03, and 0.6 mL min-1. Sorption of the colloids on the model fracture

Portland cement versus MTA as a root-end filling material. A pilot study.

Science.gov (United States)

da Silva, Sérgio Ribeiro; da Silva Neto, José Dias; Veiga, Daniela Francescato; Schnaider, Taylor Brandão; Ferreira, Lydia Masako

2015-02-01

To assess periradicular lesions clinically and by computed tomography (CT) after endodontic surgery using either Portland cement or mineral trioxide aggregate (MTA) as a root-end filling material. Three patients diagnosed with periradicular lesions by cone-beam CT underwent endodontic surgery with root-end filling. Patient A was treated with MTA as the root-end filling material, patient B was treated with Portland cement and patient C had two teeth treated, one with MTA and the other with Portland cement. Six months after surgery, the patients were assessed clinically and by CT scan and the obtained results were compared. Periradicular tissue regeneration was observed in all cases, with no significant differences in bone formation when comparing the use of MTA and Portland cement as root-end filling materials. Both mineral trioxide aggregate and Portland cement were successful in the treatment of periradicular lesions.

A study of the material accountancy procedure at the uranium enrichment facility

International Nuclear Information System (INIS)

Shirahashi, J.; Akiba, M.; Omae, M.

1984-01-01

This paper describes an evaluation of material accountancy based on total uranium (U element MUF) to detect diversions of significant quantity in the uranium enrichment facility operating at a stated maximum enrichment level of 5%. Verification that material production is within the declared enrichment can be achieved by the inspection activities associated with limited - frequency unannounced access (LFUA) to cascade areas as treated extensively in HSP. According to the experience of the material accountancy at our facility, the reduction of the material accountancy capability by changing from U-235 isotope MUF to U element MUF is only about half. However, still the U element MUF approach can meet the current IAEA detection goals for the up to about 1000 tswu/a plant

Experience of on-site disposal of production uranium-graphite nuclear reactor.

Science.gov (United States)

Pavliuk, Alexander O; Kotlyarevskiy, Sergey G; Bespala, Evgeny V; Zakharova, Elena V; Ermolaev, Vyacheslav M; Volkova, Anna G

2018-04-01

The paper reported the experience gained in the course of decommissioning EI-2 Production Uranium-Graphite Nuclear Reactor. EI-2 was a production Uranium-Graphite Nuclear Reactor located on the Production and Demonstration Center for Uranium-Graphite Reactors JSC (PDC UGR JSC) site of Seversk City, Tomsk Region, Russia. EI-2 commenced its operation in 1958, and was shut down on December 28, 1990, having operated for the period of 33 years all together. The extra pure grade graphite for the moderator, water for the coolant, and uranium metal for the fuel were used in the reactor. During the operation nitrogen gas was passed through the graphite stack of the reactor. In the process of decommissioning the PDC UGR JSC site the cavities in the reactor space were filled with clay-based materials. A specific composite barrier material based on clays and minerals of Siberian Region was developed for the purpose. Numerical modeling demonstrated the developed clay composite would make efficient geological barriers preventing release of radionuclides into the environment. Copyright © 2018 Elsevier Ltd. All rights reserved.

Uranium conversion; Urankonvertering

Energy Technology Data Exchange (ETDEWEB)

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

2006-03-15

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

ALPHA SPECTROMETRIC EVALUATION OF SRM-995 AS A POTENTIAL URANIUM/THORIUM DOUBLE TRACER SYSTEM FOR AGE-DATING URANIUM MATERIALS

Energy Technology Data Exchange (ETDEWEB)

Beals, D.

2011-12-06

Uranium-233 (t{sub 1/2} {approx} 1.59E5 years) is an artificial, fissile isotope of uranium that has significant importance in nuclear forensics. The isotope provides a unique signature in determining the origin and provenance of uranium-bearing materials and is valuable as a mass spectrometric tracer. Alpha spectrometry was employed in the critical evaluation of a {sup 233}U standard reference material (SRM-995) as a dual tracer system based on the in-growth of {sup 229}Th (t{sub 1/2} {approx} 7.34E3 years) for {approx}35 years following radiochemical purification. Preliminary investigations focused on the isotopic analysis of standards and unmodified fractions of SRM-995; all samples were separated and purified using a multi-column anion-exchange scheme. The {sup 229}Th/{sup 233}U atom ratio for SRM-995 was found to be 1.598E-4 ({+-} 4.50%) using recovery-corrected radiochemical methods. Using the Bateman equations and relevant half-lives, this ratio reflects a material that was purified {approx} 36.8 years prior to this analysis. The calculated age is discussed in contrast with both the date of certification and the recorded date of last purification.

76 FR 41308 - Strata Energy, Inc., Ross In Situ Recovery Uranium Project, Crook County, WY; Notice of Materials...

Science.gov (United States)

2011-07-13

..., Inc., Ross In Situ Recovery Uranium Project, Crook County, WY; Notice of Materials License Application...-4737, or by e-mail to [email protected] . The Ross In Situ Recovery Uranium Project License... source and byproduct materials license at its Ross In Situ Recovery Uranium Project site located in Crook...

Machining of uranium and uranium alloys

International Nuclear Information System (INIS)

Morris, T.O.

1981-01-01

Uranium and uranium alloys can be readily machined by conventional methods in the standard machine shop when proper safety and operating techniques are used. Material properties that affect machining processes and recommended machining parameters are discussed. Safety procedures and precautions necessary in machining uranium and uranium alloys are also covered. 30 figures

Efficacy of ProTaper universal retreatment files in removing filling materials during root canal retreatment.

Science.gov (United States)

Giuliani, Valentina; Cocchetti, Roberto; Pagavino, Gabriella

2008-11-01

The aim of this study was to evaluate the efficacy of the ProTaper Universal System rotary retreatment system and of Profile 0.06 and hand instruments (K-file) in the removal of root filling materials. Forty-two extracted single-rooted anterior teeth were selected. The root canals were enlarged with nickel-titanium (NiTi) rotary files, filled with gutta-percha and sealer, and randomly divided into 3 experimental groups. The filling materials were removed with solvent in conjunction with one of the following devices and techniques: the ProTaper Universal System for retreatment, ProFile 0.06, and hand instruments (K-file). The roots were longitudinally sectioned, and the image of the root surface was photographed. The images were captured in JPEG format; the areas of the remaining filling materials and the time required for removing the gutta-percha and sealer were calculated by using the nonparametric one-way Kruskal-Wallis test and Tukey-Kramer tests, respectively. The group that showed better results for removing filling materials was the ProTaper Universal System for retreatment files, whereas the group of ProFile rotary instruments yielded better root canal cleanliness than the hand instruments, even though there was no statistically significant difference. The ProTaper Universal System for retreatment and ProFile rotary instruments worked significantly faster than the K-file. The ProTaper Universal System for retreatment files left cleaner root canal walls than the K-file hand instruments and the ProFile Rotary instruments, although none of the devices used guaranteed complete removal of the filling materials. The rotary NiTi system proved to be faster than hand instruments in removing root filling materials.

The choice of iron-containing filling for composite radioprotective material

Science.gov (United States)

Matyukhin, P. V.

2018-03-01

The paper presents the data the composition of modern composite building materials including materials which in addition to high physical-mechanical have radio-protective properties. The article presents infrared researches and differential thermal data of fine-grained magnetite and hematite beneficiated iron-ore concentrates. The choice of the most suitable filling for new composite radio-protective building material engineering and development was made basing on the magnetite and hematite data presented in the paper.

75 FR 62153 - Notice of the Nuclear Regulatory Commission Issuance of Materials License SUA-1596 for Uranium...

Science.gov (United States)

2010-10-07

... Commission Issuance of Materials License SUA-1596 for Uranium One Americas, Inc. Moore Ranch In Situ Recovery.... SUPPLEMENTARY INFORMATION: The Nuclear Regulatory Commission (NRC) has issued a license to Uranium One Americas, Inc. (Uranium One) for its Moore Ranch uranium in situ recovery (ISR) facility in Campbell County...

Lacustrine-humate model for primary uranium ore deposits, Grants Uranium Region, New Mexico

International Nuclear Information System (INIS)

Turner-Peterson, C.E.

1985-01-01

Two generations of uranium ore, primary and redistributed, occur in fluvial sandstones of the Upper Jurassic Morrison Formation in the San Juan basin; the two stages of ore formation can be related to the hydrologic history of the basin. Primary ore formed soon after Morrison deposition, in the Late Jurassic to Early Cretaceous, and a model, the lacustrine-humate model, is offered that views primary mineralization as a diagenetic event related to early pore fluid evolution. The basic premise is that the humate, a pore-filling organic material closely associated with primary ore, originated as humic acids dissolved in pore waters of greenish-gray lacustrine mudstones deposited in the mud-flat facies of the Brushy Basin Member and similar K shale beds in the Westwater Canyon Member. During compaction associated with early burial, formation water expelled from lacustrine mudstone units carried these humic acids into adjacent sandstone beds where the organics precipitated, forming the humate deposits that concentrated uranium. During the Tertiary, much later in the hydrologic history of the basin, when Jurassic sediments were largely compacted, oxygenated ground water flowed basinward from uplifted basin margins. This invasion of Morrison sandstone beds by oxidizing ground waters redistributed uranium from primary ores along redox boundaries, forming ore deposits that resemble roll-front-type uranium ores. 11 figures

Preliminary concepts: coordinated safeguards for materials management in a thorium--uranium fuel reprocessing plant

International Nuclear Information System (INIS)

Hakkila, E.A.; Barnes, J.W.; Dayem, H.A.; Dietz, R.J.; Shipley, J.P.

1978-10-01

This report addresses preliminary concepts for coordinated safeguards materials management in a typical generic thorium--uranium-fueled light-water reactor (LWR) fuels reprocessing plant. The reference facility is designed to recover thorium and uranium from first-generation (denatured 235 U) startup fuels, first-recycle and equilibrium (denatured 233 U) thorium--uranium LWR fuels, and to recover the plutonium generated in the 238 U denaturant as well. 12 figures, 3 tables

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

International Nuclear Information System (INIS)

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

1989-01-01

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

Chemical properties of gutta-percha endodontic filling material: investigation of five commercial brands

International Nuclear Information System (INIS)

Silva Junior, Joao Batista A.; Paula, Regina C.M.; Feitosa, Judith P.A.; Gurgel Filho, Eduardo; Teixeira, Fabricio B

2001-01-01

Chemical composition e thermal stability of five brands of gutta-percha endodontic filling material were investigated. Samples with higher amount of organic materials possess higher thermal stability. Investigation of gutta-percha polymer extracted from the endodontic filling by IR and NMR shows that the polymer is predominantly trans-polyisoprene. The thermal stability and molar mass were similar for four brands, however the 'Tanari' brand has got lower molar mass value than the other ones. (author)

Geochemical study of the insoluble organic material (kerogen) in the Oklo uranium ore and the associated Francevillian schists

International Nuclear Information System (INIS)

Vandenbroucke, M.; Rouzaud, J.N.; Oberlin, A.

1978-01-01

The purpose of this study was to describe the organic material associated with uranium ore and ore transformations undergone by it, in terms of the following problems: (1) In the natural reactor zones, evolution of the organic material in the core and as a function of the distance away from it; (2) Comparison of organic materials from a rich and a poor ore; (3) Intercomparison of organic materials in the dispersed and concentrated state; (4) Comparison of organic materials in the uranium ore zones and in the adjacent non-mineralized Francevillian. The organic material from the reactor core could not be isolated by the normal techniques of treatment with acid. It is found in other cases that the organic material is oxidized in the uranium-bearing sediments and that the nearer to the reaction zone, the greater the oxidation, irrespective of the state of dispersion of the organic material in the rock. The uranium content does not affect this phenomenon, which is attributed to the action of the water raised to a high temperature in the vicinity of the reaction zones. On the basis of the present distribution of organic material and uranium the authors suggest a pattern for the formation of the deposit that would take into account localization of the ore in the sandstones and the part played by organic material in the accumulation process. (author)

IRIS Assessment Plan for Uranium (Scoping and Problem Formulation Materials)

Science.gov (United States)

In January 2018, EPA released the IRIS Assessment Plan for Uranium (Oral Reference Dose) (Scoping and Problem Formulation Materials). An IRIS Assessment Plan (IAP) communicates to the public the plan for assessing each individual chemical and includes summary informatio...

COGEMA's UMF [Uranium Management Facility

International Nuclear Information System (INIS)

Lamorlette, G.; Bertrand, J.P.

1988-01-01

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

Potential benefits and impacts on the CRWMS transportation system of filling spent fuel shipping casks with depleted uranium silicate glass

International Nuclear Information System (INIS)

Pope, R.B.; Forsberg, C.W.; DeHart, M.D.; Childs, K.W.; Tang, J.S.

1996-01-01

A new technology, the Depleted Uranium Silicate COntainer Fill System (DUSCOFS), is proposed to improve the performance and reduce the uncertainties of geological disposal of spent nuclear fuel (SNF), thus reducing both radionuclide release rates from the waste package and the potential for repository nuclear criticality events. DUSCOFS may also provide benefits for SNF storage and transport if it is loaded into the container early in the waste management cycle. Assessments have been made of the benefits to be derived by placing depleted uranium silicate (DUS) glass into SNF containers for enhancing repository performance assessment and controlling criticality over geologic times in the repository. Also, the performance, benefits, and impacts which can be derived if the SNF is loaded into a multi-purpose canister with DUS glass at a reactor site have been assessed. The DUSCOFS concept and the benefits to the waste management cycle of implementing DUSCOFS early in the cycle are discussed in this paper

Study of uranium oxidation states in geological material.

Science.gov (United States)

Pidchenko, I; Salminen-Paatero, S; Rothe, J; Suksi, J

2013-10-01

A wet chemical method to determine uranium (U) oxidation states in geological material has been developed and tested. The problem faced in oxidation state determinations with wet chemical methods is that U redox state may change when extracted from the sample material, thereby leading to erroneous results. In order to quantify and monitor U redox behavior during the acidic extraction in the procedure, an analysis of added isotopic redox tracers, (236)U(VI) and (232)U(IV), and of variations in natural uranium isotope ratio ((234)U/(238)U) of indigenous U(IV) and U(VI) fractions was performed. Two sample materials with varying redox activity, U bearing rock and U-rich clayey lignite sediment, were used for the tests. The Fe(II)/Fe(III) redox-pair of the mineral phases was postulated as a potentially disturbing redox agent. The impact of Fe(III) on U was studied by reducing Fe(III) with ascorbic acid, which was added to the extraction solution. We observed that ascorbic acid protected most of the U from oxidation. The measured (234)U/(238)U ratio in U(IV) and U(VI) fractions in the sediment samples provided a unique tool to quantify U oxidation caused by Fe(III). Annealing (sample heating) to temperatures above 500 °C was supposed to heal ionizing radiation induced defects in the material that can disturb U redox state during extraction. Good agreement between two independent methods was obtained for DL-1a material: an average 38% of U(IV) determined by redox tracer corrected wet chemistry and 45% for XANES. Copyright © 2013 Elsevier Ltd. All rights reserved.

Construction of a test embankment using a sand-tire shred mixture as fill material.

Science.gov (United States)

Yoon, Sungmin; Prezzi, Monica; Siddiki, Nayyar Zia; Kim, Bumjoo

2006-01-01

Use of tire shreds in construction projects, such as highway embankments, is becoming an accepted way of beneficially recycling scrap tires. However, in the last decade there was a decline in the use of pure tire shreds as fill materials in embankment construction, as they are susceptible to fire hazards due to the development of exothermic reactions. Tire shred-sand mixtures, on the other hand, were found to be effective in inhibiting exothermic reactions. When compared with pure tire shreds, tire shred-sand mixtures are less compressible and have higher shear strength. However, the literature contains limited information on the use of tire shred-soil mixtures as a fill material. The objectives of this paper are to discuss and evaluate the feasibility of using tire shred-sand mixtures as a fill material in embankment construction. A test embankment constructed using a 50/50 mixture, by volume, of tire shreds and sand was instrumented and monitored to: (a) determine total and differential settlements; (b) evaluate the environmental impact of the embankment construction on the groundwater quality due to leaching of fill material; and (c) study the temperature variation inside the embankment. The findings in this research indicate that mixtures of tire shreds and sand are viable materials for embankment construction.

Setting of loose-fill insulation materials in walls; Saetningsfri indblaesning af loesfyldsisolering i vaegge

Energy Technology Data Exchange (ETDEWEB)

Rasmussen, T.V.

2001-07-01

The report describes material behaviour, which significantly influences the settling of loose-fill insulation materials. The specific application presented here is loose-fill insulation material injected in walls as thermal insulation. The physical formulation of the issue to be discussed is that the mass is kept in position in the cavity by frictional forces, which counteracts the settling but complicates injection. The purpose of this study is to investigate whether there is a possibility that decreased friction will be able to release settling. Cellulose loose-fill material injected in a 0.1 m thick and 1 m wide gypsum wall with a minimum density of 48 kg/m3 was found not to settle if kept at a constant relative humidity, RH 50 %. A minimum density of 53 kg/m3 is necessary if the thickness of the wall is increased from 0.1 m to 0.3 m. If changing the constant environment from RH 50 % to RH 80 % a minimum density of 63 kg/m3 is necessary. Furthermore, results so far show that cellulose loose-fill material spread on the attic floor will have a density after settling of 48 kg /M3 for a constant RH 50 %, corresponding to 43 kg/m3 dry material. The results were found by using a model and tests. (au)

Characteristics of novel root-end filling material using epoxy resin and Portland cement.

Science.gov (United States)

Lee, Sang-Jin; Chung, Jin; Na, Hee-Sam; Park, Eun-Joo; Jeon, Hyo-Jin; Kim, Hyeon-Cheol

2013-04-01

The aim of this study was to evaluate the physical properties and cytotoxicity of a novel root-end filling material (EPC) which is made from epoxy resin and Portland cement as a mineral trioxide aggregate (MTA) substitute. EPC, developed as a root-end filling material, was compared with MTA and a mixture of AH Plus sealer and MTA (AMTA) with regard to the setting time, radio-opacity, and microleakage. Setting times were evaluated using Vicat apparatus. Digital radiographs were taken to evaluate the aluminium equivalent radio-opacity using an aluminium step wedge. Extracted single-rooted teeth were used for leakage test using methylene blue dye. After canal shaping and obturation, the apical 3-mm root was resected, and a root-end cavity with a depth of 3 mm was prepared. The root-end cavities were filled with MTA, AMTA, and EPC for 15 specimens in each of three groups. After setting in humid conditions for 24 h, the specimens were tested for apical leakage. For evaluation of the biocompatibility of EPC, cell (human gingival fibroblast) viability was compared for MTA and Portland cement by MTT assay, and cell morphological changes were compared for MTA and AH Plus by fluorescence microscopy using DAPI and F-actin staining. The setting time, radio-opacity, and microleakage were compared using one-way ANOVA and Scheffe's post hoc comparison, and the cytotoxicity was compared using the nonparametric Kruskal-Wallis rank sum test. Statistical significance was set at 95%. EPC had a shorter setting time and less microleakage compared with MTA (p Portland cement, was found to be a useful material for root-end filling, with favourable radio-opacity, short setting time, low microleakage, and clinically acceptable low cytotoxicity. The novel root-end filling material would be a potentially useful material for a surgical endodontic procedure with favourable properties.

Experimental Study and Application of Inorganic Solidified Foam Filling Material for Coal Mines

Directory of Open Access Journals (Sweden)

Hu Wen

2017-01-01

Full Text Available Spontaneous combustion of residual coal in a gob due to air leakage poses a major risk to mining safety. Building an airtight wall is an effective measure for controlling air leakage. A new type of inorganic solidified foam-filled material was developed and its physical and chemical properties were analyzed experimentally. The compressive strength of this material increased with the amount of sulphoaluminate cement. With an increasing water–cement ratio, the initial setting time was gradually extended while the final setting time firstly shortened and then extended. The change in compressive strength had the opposite tendency. Additionally, as the foam expansion ratio increased, the solidification time tended to decrease but the compressive strength remained approximately constant. With an increase in foam production, the solidification time increased and the compressive strength decreased exponentially. The results can be used to determine the optimal material ratios of inorganic solidified foam-filled material for coal mines, and filling technology for an airtight wall was designed. A field application of the new material demonstrated that it seals crossheadings tightly, leaves no fissures, suppresses air leakage to the gob, and narrows the width of the spontaneous combustion and heat accumulation zone.

Evaluation of physicochemical properties of root-end filling materials using conventional and Micro-CT tests

Directory of Open Access Journals (Sweden)

Fernanda Ferrari Esteves TORRES

Full Text Available Abstract Objective To evaluate solubility, dimensional stability, filling ability and volumetric change of root-end filling materials using conventional tests and new Micro-CT-based methods. Material and Methods Solubility (loss of mass after 7 and 30 days, and dimensional stability (in mm were evaluated in accordance with Carvalho-Junior, et al. 7 (2007. The filling ability and volumetric change (in mm3 were evaluated by Micro-CT (Bruker-MicroCT, Kontich, Belgium using resin models with cavities 3 mm deep and 1 mm in diameter. The cavities were filled with materials to evaluate filling ability, and then scanned by Micro-CT. After 7 and 30 days immersed in distilled water, the filled cavities were scanned again to evaluate the volumetric change. MTA Angelus (MTA, Biodentine (BIO and zinc oxide-eugenol cement (ZOE were evaluated. Data were submitted to analysis of variance (ANOVA and Tukey's test with 5% significance level. Results The results suggested correlated or complementary data between the proposed tests. At 7 days, BIO showed higher solubility and at 30 days, showed higher volumetric change in comparison with MTA (p0.05 at 7 days. At 30 days, they presented similar solubility. BIO and MTA showed higher dimensional stability than ZOE (p<0.05. ZOE and BIO showed higher filling ability (p<0.05. Conclusions ZOE presented a higher dimensional change, and BIO had greater solubility after 7 days. BIO presented filling ability and dimensional stability, but greater volumetric change than MTA after 30 days. Micro-CT can provide important data on the physicochemical properties of materials complementing conventional tests.

Contrast material filling of the peroneal tendon sheath

International Nuclear Information System (INIS)

Zadravecz, Gy.; Grexa, E.

1981-01-01

In case of complaints after fracture of the calcaneus the common sheath of the peroneus tendons was filled up with contrast material. The tendon sheath was punctured Oehind the external ankle. The three-directional radiograms clearly showed the dislocation and compression of the tendons, caused by the exostosis of the calcaneus. The concomitant tendovaginitis caused the complaints. This alteration was observed in 11% of all the calcaneus fractures. (L.E.)

Lake Austin uranium deposit, Western Australia

International Nuclear Information System (INIS)

Heath, A.G.; Deutscher, R.L.; Butt, C.R.M.

1984-01-01

The Lake Austin uranium deposit is a calcrete type deposit in the Yilgarn Block, near Cue, in a catchment area of granitoids and greenstones. The uranium is in valley fill and the sediments of the Lake Austin playa. The mineralization occurs over 1 to 6 meter thickness close to the water table in calcrete overlying clays and/or weathered bedrock. The principal uranium mineral is carnotite. Waters in nearby channels have an uranium content of over 30 ppb. The chloride content of the water increases downstream in the nearby drainages, as does the uranium and vanadium content. (author)

Matpro--version 10: a handbook of materials properties for use in the analysis of light water reactor fuel rod behavior

International Nuclear Information System (INIS)

Reymann, G.A.

1978-02-01

The materials properties correlations and computer subcodes (MATPRO--Version 10) developed for use with various LWR fuel rod behavior analytical programs at the Idaho National Engineering Laboratory are described. Formulations of fuel rod material properties, which are generally semiempirical in nature, are presented for uranium dioxide and mixed uranium--plutonium dioxide fuel, zircaloy cladding, and fill gas mixtures

MATPRO-Version 11: a handbook of materials properties for use in the analysis of light water reactor fuel rod behavior

International Nuclear Information System (INIS)

Hagrman, D.L.; Reymann, G.A.

1979-02-01

This handbook describes the materials properties correlations and computer subcodes (MATPRO-Version 11) developed for use with various LWR fuel rod behavior analytical programs at the Idaho National Engineering Laboratory. Formulations of fuel rod material properties, which are generally semiempirical in nature, are presented for uranium dioxide and mixed uranium--plutonium dioxide fuel, zircaloy cladding, and fill gas mixtures

MATPRO-Version 11: a handbook of materials properties for use in the analysis of light water reactor fuel rod behavior

Energy Technology Data Exchange (ETDEWEB)

Hagrman, D.L.; Reymann, G.A. (comps.)

1979-02-01

This handbook describes the materials properties correlations and computer subcodes (MATPRO-Version 11) developed for use with various LWR fuel rod behavior analytical programs at the Idaho National Engineering Laboratory. Formulations of fuel rod material properties, which are generally semiempirical in nature, are presented for uranium dioxide and mixed uranium--plutonium dioxide fuel, zircaloy cladding, and fill gas mixtures.

Use of highly enriched uranium in the material testing reactor BR2

International Nuclear Information System (INIS)

Beeckmans de West-Meerbeeck, A.

1979-05-01

In the material testing reactor BR2, the use of highly enriched uranium is determined by the consideration of the fast, epithermal and thermal neutron flux effectively available for the experimental devices. The choice of the core configuration is defined by combining the localisation of the experimental devices and of fuel elements of various burnup, such as to satisfy the irradiation conditions of the experimental load, compatible with an economic use of the fuel elements and safe operation of the reactor. Taking into account the present manufacturing technology for MTR fuels (37 Wt % uranium density in the fuel meat) the highly enriched uranium cannot be avoided; if higher concentration of uranium could be realised by some new manufacturing technology, the 235 U density of fuel elements at elimination should be kept at the required level and the enrichment could be reduced accordingly

Use of highly enriched uranium in the material testing reactor BR2

International Nuclear Information System (INIS)

Beeckmans de West-Meerbeeck, A.

1979-05-01

In the material testing reactor BR2, the use of highly enriched uranium is determined by the consideration of the fast, epithermal and thermal neutron flux effectively available for the experimental devices. The choice of the core configuration is defined by combining the localisation of the experimental devices and of fuel elements of various burnup, such as to satisfy the irradiation conditions of the experimental load, compatible with an economic use of the fuel elements and safe operation of the reactor. Taking into account the present manufacturing technology for MTR fuels (37 Wt % uranium density in the fuel meat) the highly enriched uranium cannot be avoided: if higher concentration of uranium could be realised by some new manufacturing technology, the 235 U density of fuel elements at elimination should be kept at the required level and the enrichment could be reduced accordingly. (author)

Contribution to the study of nuclear fuel materials with a metallic uranium base; Contribution a l'etude des materiaux combustibles nucleaires a base d'uranium metallique

Energy Technology Data Exchange (ETDEWEB)

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

1957-11-15

In a power reactor destined to supply industrially recoverable thermal energy, the most economical source of heat still consists of natural metallic uranium. However, the nuclear fuel material, most often employed in the form of rods of 20 to 40 mm diameter, is subjected to a series of stresses which lead to irreversible distortions usually incompatible with the substructure of the reactor. As a result the fuel material must possess at the outset a certain number of qualities which must be determined. Investigations have therefore been carried out, first on the technological characters peculiar to each of the three allotropic phases of pure uranium metal, and on their interactions on the stabilisation of the material which consists of either cast uranium or uranium pile-treated in the {gamma} phase. (author) [French] Dans un reacteur de puissance destine a fournir de l'energie thermique industriellement recuperable, la source de chaleur la plus economique reste constituee par de l'uranium metallique naturel. Or, le materiau combustible nucleaire, employe le plus souvent sous forme de barreaux de 20 a 40 mm de diametre, se trouve soumis a un ensemble de contraintes qui provoque des deformations irreversibles, le plus souvent incompatibles avec l'infrastructure du reacteur. Par consequent, le materiau combustible doit presenter a l'origine un certain nombre de qualites qu'il est necessaire de determiner. Aussi a-t-on d'abord etudie les caracteres technologiques propres a chacune des trois phases allotropiques de l'uranium-metal pur et leurs interactions sur la stabilisation du materiau constitue soit par de l'uranium coule, soit par de l'uranium traite en pile en phase {gamma}. (auteur)

RECOMMENDED FOUNDATION FILL MATERIALS CONSTRUCTION STANDARD OF THE FLORIDA RADON RESEARCH PROGRAM

Science.gov (United States)

The report summarizes the technical basis for a recommended foundation fill materials standard for new construction houses in Florida. he radon-control construction standard was developed by the Florida Radon Research Program (FRRP). ill material standards are formulated for: (1)...

230Th-234U Model-Ages of Some Uranium Standard Reference Materials

International Nuclear Information System (INIS)

Williams, R.W.; Gaffney, A.M.; Kristo, M.J.; Hutcheon, I.D.

2009-01-01

The 'age' of a sample of uranium is an important aspect of a nuclear forensic investigation and of the attribution of the material to its source. To the extent that the sample obeys the standard rules of radiochronometry, then the production ages of even very recent material can be determined using the 230 Th- 234 U chronometer. These standard rules may be summarized as (a) the daughter/parent ratio at time=zero must be known, and (b) there has been no daughter/parent fractionation since production. For most samples of uranium, the 'ages' determined using this chronometer are semantically 'model-ages' because (a) some assumption of the initial 230 Th content in the sample is required and (b) closed-system behavior is assumed. The uranium standard reference materials originally prepared and distributed by the former US National Bureau of Standards and now distributed by New Brunswick Laboratory as certified reference materials (NBS SRM = NBL CRM) are good candidates for samples where both rules are met. The U isotopic standards have known purification and production dates, and closed-system behavior in the solid form (U 3 O 8 ) may be assumed with confidence. We present here 230 Th- 234 U model-ages for several of these standards, determined by isotope dilution mass spectrometry using a multicollector ICP-MS, and compare these ages with their known production history

3D Analysis of D-RaCe and Self-Adjusting File in Removing Filling Materials from Curved Root Canals Instrumented and Filled with Different Techniques

Directory of Open Access Journals (Sweden)

Neslihan Simsek

2014-01-01

Full Text Available The aim of this study was to compare the efficacy of D-RaCe files and a self-adjusting file (SAF system in removing filling material from curved root canals instrumented and filled with different techniques by using microcomputed tomography (micro-CT. The mesial roots of 20 extracted mandibular first molars were used. Root canals (mesiobuccal and mesiolingual were instrumented with SAF or Revo-S. The canals were then filled with gutta-percha and AH Plus sealer using cold lateral compaction or thermoplasticized injectable techniques. The root fillings were first removed with D-RaCe (Step 1, followed by Step 2, in which a SAF system was used to remove the residual fillings in all groups. Micro-CT scans were used to measure the volume of residual filling after root canal filling, reinstrumentation with D-RaCe (Step 1, and reinstrumentation with SAF (Step 2. Data were analyzed using Wilcoxon and Kruskal-Wallis tests. There were no statistically significant differences between filling techniques in the canals instrumented with SAF (P=0.292 and Revo-S (P=0.306. The amount of remaining filling material was similar in all groups (P=0.363; all of the instrumentation techniques left filling residue inside the canals. However, the additional use of SAF was more effective than using D-RaCe alone.

Evaluation of adsorption of uranium from aqueous solution using biochar materials

Energy Technology Data Exchange (ETDEWEB)

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

2015-07-01

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

Evaluation of adsorption of uranium from aqueous solution using biochar materials

International Nuclear Information System (INIS)

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

2015-01-01

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

Comparative evaluation of liner materials for inactive uranium-mill-tailings piles

International Nuclear Information System (INIS)

Buelt, J.L.; Barnes, S.M.

1981-01-01

Under the funding of the Department of Energy's Uranium Mill Tailings Remedial Action (UMTRA) Program, Pacific Northwest Laboratory (PNL) has completed the initial accelerated testing phase of eight candidate liner materials. The tests were designed to comparatively evaluate the long term effectiveness of liner materials as a radionuclide and hazardous chemical leachate barrier. The eight materials tested were selected from a technical review of published literature and industrial specialists. Conditions were then identified that would accelerate the aging processes expected in a uranium tailings environment for 1000 years. High calcium leachates were forced through thin layers of clay liners to accelerate the ion exchange rate of sodium and calcium. Asphalt and synthetic materials were accelerated by exposure to elevate temperatures, high concentrations of oxygen, and increased strengths of aqueous oxidizing agents. By comparing the changes of permeability with time of exposure, the most acceptable materials were then identified. These materials are a catalytically airblown asphalt membrane and natural soil amended with sodium bentonite. Both materials showed an increased resistance to leachate penetration throughout the exposure period with final permeabilities less than 10 -7 cm/s. In addition, the asphalt membrane and sodium bentonite are among the least expensive materials to install at a disposal site. Therefore based on their economic and technical merits, these two materials are being evaluated further in field tests at Grand Junction, Colorado

Preliminary evaluation of the uranium favorability in the Kaiparowits Plateau Region, Garfield and Kane Counties, Utah

International Nuclear Information System (INIS)

Dubyk, W.S.; Young, P.

1978-05-01

The basal sandstone of the Chinle Formation (Upper Triassic) and the Salt Wash Member of the Morrison Formation (Upper Jurassic) in the Kaiparowits Plateau, southcentral Utah, were evaluated in terms of uranium potential. Both surface and subsurface data were utilized. Favorability of the basal Chinle sandstone was based on (1) presence of intermediate-size sandstone-filled channels cut into the Moenkopi; (2) presence of carbonaceous material; (3) an adequate source of uranium; and (4) gamma-ray anomalies from test-hole logs. Favorability of the Salt Wash Member of the Morrison Formation was based on (1) sandstone-mudstone ratios that approach equality, and (2) presence of thick sandstone lenses, carbonaceous material, and halos of light-tan to brown limonite staining. Although the basal Chinle sandstone and the Salt Wash Member of the Morrison contain sizable uranium deposits throughout much of the Colorado Plateau, both units lack characteristics that are favorable for significant uranium deposits in the Kaiparowits Plateau

Nondestructive assay of special nuclear material for uranium fuel-fabrication facilities

International Nuclear Information System (INIS)

Smith, H.A. Jr.; Schillebeeckx, P.

1997-01-01

A high-quality materials accounting system and effective international inspections in uranium fuel-fabrication facilities depend heavily upon accurate nondestructive assay measurements of the facility's nuclear materials. While item accounting can monitor a large portion of the facility inventory (fuel rods, assemblies, storage items), the contents of all such items and mass values for all bulk materials must be based on quantitative measurements. Weight measurements, combined with destructive analysis of process samples, can provide highly accurate quantitative information on well-characterized and uniform product materials. However, to cover the full range of process materials and to provide timely accountancy data on hard-to-measure items and rapid verification of previous measurements, radiation-based nondestructive assay (NDA) techniques play an important role. NDA for uranium fuel fabrication facilities relies on passive gamma spectroscopy for enrichment and U isotope mass values of medium-to-low-density samples and holdup deposits; it relies on active neutron techniques for U-235 mass values of high-density and heterogeneous samples. This paper will describe the basic radiation-based nondestructive assay techniques used to perform these measurements. The authors will also discuss the NDA measurement applications for international inspections of European fuel-fabrication facilities

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

International Nuclear Information System (INIS)

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

1991-01-01

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

Simulation of water flow and retention in earthen-cover materials overlying uranium mill tailings

International Nuclear Information System (INIS)

Simmons, C.S.; Gee, G.W.

1981-09-01

The water retention characteristics of a multilayer earthen cover for uranium mill tailings were simulated under arid weather conditions common to Grand Junction, Colorado. The multilayer system described in this report consists of a layer of wet clay/gravel (radon barrier), which is separated from a surface covering of fill soil by a washed rock material used as a capillary barrier. The capillary barrier is designed to prevent the upward migration of water and salt from the tailings to the soil surface and subsequent loss of water from the wet clay. The flow model, UNSATV, described in this report uses hydraulic properties of the layered materials and historical climatic data for two years (1976 and 1979) to simulate long-term hydrologic response of the multilayer system. Application of this model to simulate the processes of infiltration, evaporation and drainage is described in detail. Simulations over a trial period of one relatively wet and two dry years indicated that the clay-gravel layer remained near saturation, and hence, that the layer was an effective radon barrier. Estimates show that the clay-gravel layer would not dry out (i.e., revert to drying dominated by isothermal vapor-flow conditions) for at least 20 years, provided that the modeled dry-climate period continues

Shear bond strength of bulk-fill and nano-restorative materials to dentin.

Science.gov (United States)

Colak, Hakan; Ercan, Ertugrul; Hamidi, Mehmet Mustafa

2016-01-01

Bulk-fill composite materials are being developed for preparation depths of up to 4 mm in an effort to simplify and improve the placement of direct composite posterior restorations. The aim of our study was to compare shear-bond strength of bulk-fill and conventional posterior composite resins. In this study, 60 caries free extracted human molars were used and sectioned parallel to occlusal surface to expose midcoronal dentin. The specimens were randomly divided into four groups. Total-etch dentine bonding system (Adper Scotchbond 1XT, 3M ESPE) was applied to dentin surface in all the groups to reduce variability in results. Then, dentine surfaces covered by following materials. Group I: SonicFill Bulk-Fill, Group II: Tetric EvoCeram (TBF), Group III: Herculite XRV Ultra, and Group IV: TBF Bulk-Fill, 2 mm × 3 mm cylindrical restorations were prepared by using application apparatus. Shear bond testing was measured by using a universal testing machine. Kruskal-Wallis and Mann-Whitney U-tests were performed to evaluate the data. The highest value was observed in Group III (14.42 ± 4.34) and the lowest value was observed in Group IV (11.16 ± 2.76) and there is a statistically significant difference between these groups (P = 0.046). However, there is no statistically significant difference between the values of other groups. In this study, Group III was showed higher strength values. There is a need for future studies about long-term bond strength and clinical success of these adhesive and bulk-fill systems.

Dating of Uranium Materials by Using Combination of ICP-MS and SIMS

International Nuclear Information System (INIS)

Stebelkov, Y.; Aleshin, M.; Elantyev, I.; Grachev, A.; Zhizhin, K.; Stebelkov, V.

2015-01-01

The age of nuclear and some other radioactive materials can be determined undoubtedly by using ICP-MS techniques. But it can be correct if only one nuclear or another radioactive material is presented in analyzed sample and no isotope-chronograph presented in background particles in significant quantities. For particle analysis, which can be implemented by using SIMS, these restrictions are not valid. Practically one particle always characterizes only one material and does not contain background isotopes-chronographs. But age determination is based on the result of measuring of the content ratio of Th230 and U234. The difference of ionization coefficients of uranium and thorium and dependence of these coefficients on composition of particle does not allow using this method directly for age determination. Nevertheless SIMS is useful for dating of uranium materials, especially if the sample can contain small amounts of different materials. In this case the analysis of different fragments of materials by SIMS can confirm or not confirm the result, had been obtained by ICP-MS. If all detected and analyzed particles and fragments will have the same ratio of ion currents of Th230 and U234, the result of ICP-MS is true. If particles and fragments will have different ratios the result of ICP-MS cannot be related to any of presented materials. But in the last case the ages of different materials can be still estimated if different fragments have different ages, but the same elemental composition. The ''age'' had been determined by ICP-MS can be correlated with the average ratio of ion currents of Th230 and U234 had been determined by SIMS for all analyzed particles or fragments. This correlation determines the ratio of ionization coefficients of uranium and thorium, which should be the same for particles with the same elemental composition. (author)

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

International Nuclear Information System (INIS)

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

1989-01-01

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

Recovery of uranium from crude uranium tetrafluoride

Energy Technology Data Exchange (ETDEWEB)

Ghosh, S K; Bellary, M P; Keni, V S [Chemical Engineering Division, Bhabha Atomic Research Centre, Mumbai (India)

1994-06-01

An innovative process has been developed for recovery of uranium from crude uranium tetrafluoride cake. The process is based on direct dissolution of uranium tetrafluoride in nitric acid in presence of aluminium hydroxide and use of solvent extraction for removal of fluorides and other bulk impurities to make uranium amenable for refining. It is a simple process requiring minimum process step and has advantage of lesser plant corrosion. This process can be applied for processing of uranium tetrafluoride generated from various sources like uranium by-product during thorium recovery from thorium concentrate, first stage product of uranium recovery from phosphoric acid by OPPA process and off grade uranium tetrafluoride material. The paper describes the details of the process developed and demonstrated on bench and pilot scale and its subsequent modification arising out of bulky solid waste generation. The modified process uses a lower quantity of aluminium hydroxide by allowing a lower dissolution of uranium per cycle and recycles the undissolved material to the next cycle, maintaining the overall recovery at high level. This innovation has reduced the solid waste generated by a factor of four at the cost of a slightly larger dissolution vessel and its increased corrosion rate. (author). 4 refs., 1 fig., 3 tabs.

Recovery of uranium from crude uranium tetrafluoride

International Nuclear Information System (INIS)

Ghosh, S.K.; Bellary, M.P.; Keni, V.S.

1994-01-01

An innovative process has been developed for recovery of uranium from crude uranium tetrafluoride cake. The process is based on direct dissolution of uranium tetrafluoride in nitric acid in presence of aluminium hydroxide and use of solvent extraction for removal of fluorides and other bulk impurities to make uranium amenable for refining. It is a simple process requiring minimum process step and has advantage of lesser plant corrosion. This process can be applied for processing of uranium tetrafluoride generated from various sources like uranium by-product during thorium recovery from thorium concentrate, first stage product of uranium recovery from phosphoric acid by OPPA process and off grade uranium tetrafluoride material. The paper describes the details of the process developed and demonstrated on bench and pilot scale and its subsequent modification arising out of bulky solid waste generation. The modified process uses a lower quantity of aluminium hydroxide by allowing a lower dissolution of uranium per cycle and recycles the undissolved material to the next cycle, maintaining the overall recovery at high level. This innovation has reduced the solid waste generated by a factor of four at the cost of a slightly larger dissolution vessel and its increased corrosion rate. (author)

Uranium - raw material reserves for coming generations

International Nuclear Information System (INIS)

Keutner, H.

1981-01-01

Large uranium occurences have been discovered in the South of Mexico. The deposits are situated in the Sierra Mixteca. Reserves of 9.400 tons had been at Mexico's disposal even before these new discoveries. The quantitiy discovered recentyl amounts to 20.000 tons. The uranium reserves available apart from those in centrally controlled economic systems are presently estimated at five million tons. Meanwhile American scientists have found out that all the rivers of the world transport about 16.000 tons of uranium from the continents into the oceans per annum. The energy value of this washed out amount of uranium corresponds to the 25-fold world power demand of today. US scientists have discovered that the oceans can provide uranium for about seven million years of the present world energy demand. While the petroleum reserves decrease worldwide it seems that the exploration of uranium has just been started. (orig.) [de

232Th Mass Determination in a Uranium/Thorium Mixture for Safeguards Purposes

International Nuclear Information System (INIS)

Nangu, M.; Marumo, B.; Mbedzi, E.; Rasweswe, M.; Croft, S.; McElroy, R.; Chapman, J.; Bosko, A.

2015-01-01

In nuclear safeguards it is required that thorium content in safeguarded material should be quantified and reported as appropriate. As such the South African State System of Control and Accounting (SSAC) on discovering a number of safeguarded waste drums which contained considerable quantities of thorium decided to initiate a project to properly quantify their thorium content using a high purity germanium detector and In Situ Object Counting System (ISOCS) efficiency calibration software. These metal waste drums are contained inside overpacks which for health reasons cannot be opened and thus giving rise to the challenge of determining the exact fill heights and the density of the material. Fill heights determined using transmission sources and the material density calculated from them together with the geometry used for the overpacks could be used to further refine the ISOCS calibration geometry and thus improving the quantitative result. In order to have confidence on the ISOCS measurements, it was decided to also validate the ISOCS results through the preparation of similar density standards that would be used for the efficiency calibration in the determination of the 232Th activity in the material. In addition, MGAU v4.2, which was used to determine uranium enrichment in a measured material, also provides an approximate 232Th abundance relative to uranium content. ISOCS measurements of 232Th masses in waste drums were compared to MGAU results. Results of these studies are presented in this paper. (author)

Neutralization of Aerosolized Bio-Agents by Filled Nanocomposite Materials through Thermal and Chemical Inactivation Mechanisms

Science.gov (United States)

2016-06-01

Bio -agents by Filled Nanocomposite Materials through Thermal and Chemical Inactivation Mechanisms Distribution Statement A. Approved for public...of Cincinnati Project Title: Neutralization of Aerosolized Bio -agents by Filled Nanocomposite Materials through Thermal and Chemical Inactivation...fire ball, where they will not effectively interact with any viable bio -aerosol. 1.1.4. Conclusions Cryo-milling is necessary to achieve a

Uranium management activities

International Nuclear Information System (INIS)

Jackson, D.; Marshall, E.; Sideris, T.; Vasa-Sideris, S.

2001-01-01

One of the missions of the Department of Energy's (DOE) Oak Ridge Office (ORO) has been the management of the Department's uranium materials. This mission has been accomplished through successful integration of ORO's uranium activities with the rest of the DOE complex. Beginning in the 1980's, several of the facilities in that complex have been shut down and are in the decommissioning process. With the end of the Cold War, the shutdown of many other facilities is planned. As a result, inventories of uranium need to be removed from the Department facilities. These inventories include highly enriched uranium (HEU), low enriched uranium (LEU), normal uranium (NU), and depleted uranium (DU). The uranium materials exist in different chemical forms, including metals, oxides, solutions, and gases. Much of the uranium in these inventories is not needed to support national priorities and programs. (author)

230Th-234U Model-Ages of Some Uranium Standard Reference Materials

Energy Technology Data Exchange (ETDEWEB)

Williams, R W; Gaffney, A M; Kristo, M J; Hutcheon, I D

2009-05-28

The 'age' of a sample of uranium is an important aspect of a nuclear forensic investigation and of the attribution of the material to its source. To the extent that the sample obeys the standard rules of radiochronometry, then the production ages of even very recent material can be determined using the {sup 230}Th-{sup 234}U chronometer. These standard rules may be summarized as (a) the daughter/parent ratio at time=zero must be known, and (b) there has been no daughter/parent fractionation since production. For most samples of uranium, the 'ages' determined using this chronometer are semantically 'model-ages' because (a) some assumption of the initial {sup 230}Th content in the sample is required and (b) closed-system behavior is assumed. The uranium standard reference materials originally prepared and distributed by the former US National Bureau of Standards and now distributed by New Brunswick Laboratory as certified reference materials (NBS SRM = NBL CRM) are good candidates for samples where both rules are met. The U isotopic standards have known purification and production dates, and closed-system behavior in the solid form (U{sub 3}O{sub 8}) may be assumed with confidence. We present here {sup 230}Th-{sup 234}U model-ages for several of these standards, determined by isotope dilution mass spectrometry using a multicollector ICP-MS, and compare these ages with their known production history.

Uranium

International Nuclear Information System (INIS)

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

2010-01-01

concentration in peat bogs, deposits combined with marine phosphates, with coal and lignite, with black shales, with carbonate rocks, deposits in Precambrian quartz pebble conglomerates, basal-type deposits, deposits in sandstones (tabular, roll-type and tectono-lithologic deposits), breccia chimney filling deposits, deposits in metamorphic rocks, metasomatic deposits, deposits in intrusive rocks, deposits associated with hematite breccia complexes, deposits in granitic rocks, deposits in volcanic rocks, deposits in proterozoic discordances (Athabasca basin, Pine Creek geo-syncline); 4 - French uranium bearing areas and deposits: history of the French uranium mining industry, geological characteristics of French deposits (black shales, sandstones, granites), abroad success of French mining companies (Africa, North America, South America, Australia, Asia); 5 - exploration and exploitation; 6 - uranium economy: perspectives of uranium demand, present day production status, secondary resources, possible resources, market balances, prices and trends, future availability and nuclear perspectives. (J.S.)

Process evaluations for uranium recovery from scrap material

International Nuclear Information System (INIS)

Westphal, B.R.; Benedict, R.W.

1992-01-01

The integral Fast Reactor (IFR) concept being developed by Argonne National Laboratory is based on pyrometallurgical processing of spent nuclear metallic fuel with subsequent fabrication into new reactor fuel by an injection casting sequence. During fabrication, a dilute scrap stream containing uranium alloy fines and broken quartz (Vycor) molds in produced. Waste characterization of this stream, developed by using present operating data and chemical analysis was used to evaluate different uranium recovery methods and possible process variations for the return of the recovered metal. Two methods, comminution with size separation and electrostatic separation, have been tested and can recover over 95% of the metal. Recycling the metal to either the electrochemical process or the injection casting was evaluated for the different economic and process impacts. The physical waste parameters and the important separation process variables are discussed with their effects on the viability of recycling the material. In this paper criteria used to establish the acceptable operating limits is discussed

Cleaning efficacy of reciprocal and rotary systems in the removal of root canal filling material

Science.gov (United States)

Koçak, Mustafa Murat; Koçak, Sibel; Türker, Sevinç Aktemur; Sağlam, Baran Can

2016-01-01

Aim: The aim of this study was to compare the efficacy of hand file, nickel titanium rotary instrument, and two reciprocating instruments for removing gutta-percha and sealer from the root canals. Materials and Methods: Eighty-eight mandibular premolar teeth were used. The root canals were shaped and filled with gutta-percha and a resin-based sealer. The specimens were divided into four groups according to the technique by which the root filling material was removed: Group 1 — Wave One; Group 2 — Reciproc; Group 3 — ProTaper; and Group 4 — Gates-Glidden burs and stainless steel hand file. Then teeth were split longitudinally and photographed. The images were transferred to a computer. The ratio of remaining filling material to the root canal periphery was calculated with the aid of ImageJ software. Statistical analysis was performed using Kruskal–Wallis and Mann–Whitney tests. Results: A significant difference was found among all groups (P hand file. Conclusion: The reciprocating files were found to be significantly more effective in removing the filling material from the canal walls compared to the rotational file and hand file. PMID:27099429

SULPHUR DIOXIDE LEACHING OF URANIUM CONTAINING MATERIAL

Science.gov (United States)

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

1958-12-01

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

Long term radiological impact of a uranium mine restoration

International Nuclear Information System (INIS)

Mora, Veronica; Bordonaba Marisa; Sanchez Guillermo

2008-01-01

During the 1990s, many uranium mines were closed as consequence of low prices of this mineral. It was due to a decrease in the demand for uranium and an increase in the overall supply. The resulting was a further complicated implementation of sites restorations. This report deals with one of the relevant aspects of the radiological protection scope: 'the evaluation of the long term radiological impact in the population due to the uranium mine restoration activities' for the uranium mine sited in Saelices el Chico (Salamanca, Spain). These restoration activities have basically consisted of recovering the original site by filling the old open pits with the material stockpiled in the waste dumps. The main problems associated with this material include radon release and particles emission. The strategy used to solve this problem has been covered these structures with a layer with beds of clay material rock, waste material and a cover tree. The pathways considered for the radiological impact have been: 1) Inhalation; 2) Ingestion of contaminated water, milk, vegetables and meat; 3) External exposure from clouds immersion, grounds concentrations and direct gamma radiation. Three computer codes have been used with the object of evaluating the above-mentioned impact. Two of them are well-known NRC (Nuclear Regulatory Commission) codes: RESRAD 6.30 and MILDOS-AREA. We have also applied DOEFLURA, developed in ENUSA [1, 2, 3]. Four scenarios have been studied: Resident Farmer Scenario, Resident scenario, Livestock pasture scenario and Forest scenario, Estimation of radioactive doses for the member of the public in the different scenarios has been calculated with this programme. A period of 3500 years from now has been studied. (author)

Depleted uranium hexafluoride: The source material for advanced shielding systems

Energy Technology Data Exchange (ETDEWEB)

Quapp, W.J.; Lessing, P.A. [Idaho National Engineering Lab., Idaho Falls, ID (United States); Cooley, C.R. [Department of Technology, Germantown, MD (United States)

1997-02-01

The U.S. Department of Energy (DOE) has a management challenge and financial liability problem in the form of 50,000 cylinders containing 555,000 metric tons of depleted uranium hexafluoride (UF{sub 6}) that are stored at the gaseous diffusion plants. DOE is evaluating several options for the disposition of this UF{sub 6}, including continued storage, disposal, and recycle into a product. Based on studies conducted to date, the most feasible recycle option for the depleted uranium is shielding in low-level waste, spent nuclear fuel, or vitrified high-level waste containers. Estimates for the cost of disposal, using existing technologies, range between $3.8 and $11.3 billion depending on factors such as the disposal site and the applicability of the Resource Conservation and Recovery Act (RCRA). Advanced technologies can reduce these costs, but UF{sub 6} disposal still represents large future costs. This paper describes an application for depleted uranium in which depleted uranium hexafluoride is converted into an oxide and then into a heavy aggregate. The heavy uranium aggregate is combined with conventional concrete materials to form an ultra high density concrete, DUCRETE, weighing more than 400 lb/ft{sup 3}. DUCRETE can be used as shielding in spent nuclear fuel/high-level waste casks at a cost comparable to the lower of the disposal cost estimates. Consequently, the case can be made that DUCRETE shielded casks are an alternative to disposal. In this case, a beneficial long term solution is attained for much less than the combined cost of independently providing shielded casks and disposing of the depleted uranium. Furthermore, if disposal is avoided, the political problems associated with selection of a disposal location are also avoided. Other studies have also shown cost benefits for low level waste shielded disposal containers.

Exploration for uranium and other nuclear materials

International Nuclear Information System (INIS)

Hernandez, E.C.

1975-05-01

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

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

Investigation of residual anode material after electrorefining uranium in molten chloride salt

Energy Technology Data Exchange (ETDEWEB)

Rose, M.A., E-mail: marose@anl.gov [Department of Nuclear Engineering, Purdue University, West Lafayette, IN, 47907 (United States); Nuclear Engineering Division, Argonne National Laboratory, Argonne, IL 60439 (United States); Williamson, M.A.; Willit, J. [Nuclear Engineering Division, Argonne National Laboratory, Argonne, IL 60439 (United States)

2015-12-15

A buildup of material at uranium anodes during uranium electrorefining in molten chloride salts has been observed. Potentiodynamic testing has been conducted using a three electrode cell, with a uranium working electrode in both LiCl/KCl eutectic and LiCl each containing ∼5 mol% UCl{sub 3}. The anodic current response was observed at 50° intervals between 450 °C and 650 °C in the eutectic salt. These tests revealed a buildup of material at the anode in LiCl/KCl salt, which was sampled at room temperature, and analyzed using ICP-MS, XRD and SEM techniques. Examination of the analytical data, current response curves and published phase diagrams has established that as the uranium anode dissolves, the U{sup 3+} ion concentration in the diffusion layer surrounding the electrode rises precipitously to levels, which may at low temperatures exceed the solubility limit for UCl{sub 3} or in the case of the eutectic salt for K{sub 2}UCl{sub 5}. The reduction in current response observed at low temperature in eutectic salt is eliminated at 650 °C, where K{sub 2}UCl{sub 5} is absent due to its congruent melting and only simple concentration polarization effects are seen. In LiCl similar concentration effects are seen though significantly longer time at applied potential is required to effect a reduction in the current response as compared to the eutectic salt.

Uranium analysis in water flowing by the nuclear track detection method on solid dielectric materials

International Nuclear Information System (INIS)

Arambula, H.

1981-01-01

The objective of this experiment was threefold: to study the content of uranium in tap and spring water, to establish a technique for the quantitative analysis for the presence of uranium in liquids, and to test the qualities as detector fission fragments of three solid insulator materials using the nuclear tracks register method. The latter allows for the measurement of concentrations of fissile elements up to 10 -12 gr/gm employing (n, f.f.) reactions. The test samples were of tap water and of water from six fresh water springs located in San Luis Potosi and Guanajuato. Glass, lexan polycarbonate and muscovite mica were the detector materials used. The technique consisted in evaporating the water from the test samples, which had been previously placed upon the detector materials, and in doing the same for the standard control sample solutions having known concentrations of uranium. All the samples were then irradiated with thermal neutrons, and the 235 U, present in the samples, fissioned. The fission fragments produced permanent damage on the detectors, known as latent tracks. A specific corroding chemical was then applied to each detector which caused the latent tracks to dissolve into grooves. Known as etching tracks, these grooves were microscopically visible and could be measured for track density (tracks/mm 2 ). The concentrations of uranium present in the test samples were measured by comparing the track densities of the test samples with those of the standard control samples. The concentration of uranium found in the spring water samples ranged from 0.09 to 0.89 μqr.U/1, and those of tap water, from 0.18 to 0.19 μqr U/1. Lexan polycarbonate and muscovite mica proved to be better, as detectors, than glass. Glass for quantitative analysis, we found not recommendable as a detector material because of its alterable composition in the presence of uranium. (author)

Uranium metallogenic model related to CO2 and hydrocarbon in granite type uranium deposits

International Nuclear Information System (INIS)

Ou Guangxi; Chen Anfu; Cui Jianyong; Xu Yinhuan; Wang Chunhua; Xu Yan

2001-01-01

The report is concerned with the inseparable connections between the uranium migration, enrichment rule and the geochemical characteristics of CO 2 and hydrocarbon gas, as well as the relations between the deposit locations and the gas abnormal distribution in rocky body, which are based on the analysis of some data and phenomena in 11 typical deposits in 2 granite type uranium ore fields, including the observations of 250 rocky fluid inclusion sections and the analyzed data of which 2470 are in gas composition, 200 in uranium content, 50 in thermometry. All the conclusions are drawn from different angles for the first time and this new exploration and advancement fills up the blank of gas geochemistry study in uranium deposits or other metal deposits

High-temperature thermal conductivity of uranium chromite and uranium niobate

International Nuclear Information System (INIS)

Fedoseev, D.V.; Varshavskaya, I.G.; Lavrent'ev, A.V.; Oziraner, S.N.; Kuznetsova, D.G.

1979-01-01

The technique of determining thermal conductivity coefficient of uranium niobate and uranium chromite on heating with laser radiation is described. Determined is the coefficient of free-convective heat transfer (with provision for a conduction component) by means of a standard specimen. The thermal conductivity coefficients of uranium chromite and niobate were measured in the 1300-1700 K temperature range. The results are presented in a diagram form. It has been calculated, that the thermal conductivity coefficient for uranium niobate specimens is greater in comparison with uranium chromite specimens. The thermal conductivity coefficients of the materials mentioned depend on temperature very slightly. Thermal conductivity of the materials considerably depends on their porosity. The specimens under investigation were fabricated by the pressing method and had the following porosity: uranium chromite - 30 %, uranium niobate - 10 %. Calculation results show, that thermal conductivity of dense uranium chromite is higher than thermal conductivity of dense uranium niobate. The experimental error equals approximately 20 %, that is mainly due to the error of measuring the temperature equal to +-25 deg, with a micropyrometer

Materials safeguards and accountability in the low enriched uranium conversion-fabrication sector of the fuel cycle

International Nuclear Information System (INIS)

Schneider, R.A.; Nilson, R.; Jaech, J.L.

1978-01-01

Today materials accounting in the low enriched conversion-fabrication sector of the LWR fuel cycle is of increased importance. Low enriched uranium is rapidly becoming a precious metal with current dollar values in the range of one dollar per gram comparing with gold and platinum at 7-8 dollars per gram. In fact, people argue that its dollar value exceeds its safeguards value. Along with this increased financial incentive for better material control, the nuclear industry is faced with the impending implementation of international safeguards and increased public attention over its ability to control nuclear materials. Although no quantity of low enriched uranium (LEU) constitutes a practical nuclear explosive, its control is important to international safeguards because of plutonium production or further enrichment to an explosive grade material. The purpose of the paper is to examine and discuss some factors in the area of materials safeguards and accountability as they apply to the low enriched uranium conversion-fabrication sector. The paper treats four main topics: basis for materials accounting; our assessment of the proposed new IAEA requirements; adequacy of current practices; and timing and direction of future modifications

A comparison between thorium-uranium and low enrichment uranium cycles in the high temperature reactors

Energy Technology Data Exchange (ETDEWEB)

Cerles, J M

1973-03-15

In a previous report, it was shown that the Uranium cycle could be used as well with multi-hole block (GGA type) as with tubular elements. Now, in a F.S.V. geometry, a comparison is made between Thorium cycle and Uranium cycle. This comparison will be concerned with the physical properties of the materials, the needs of natural Uranium, the fissile material inventory and, at last, an attempt of economical considerations. In this report the cycle will be characterizd by the fertile material. So, we write ''Thorium cycle'' for Highly Enriched Uranium - Thorium cycle and ''Uranium cycle'' for low Enrichment Uranium cycle.

Rheology of Emulsion-Filled Gels Applied to the Development of Food Materials

Directory of Open Access Journals (Sweden)

Ivana M. Geremias-Andrade

2016-08-01

Full Text Available Emulsion-filled gels are classified as soft solid materials and are complex colloids formed by matrices of polymeric gels into which emulsion droplets are incorporated. Several structural aspects of these gels have been studied in the past few years, including their applications in food, which is the focus of this review. Knowledge of the rheological behavior of emulsion-filled gels is extremely important because it can measure interferences promoted by droplets or particle inclusion on the textural properties of the gelled systems. Dynamic oscillatory tests, more specifically, small amplitude oscillatory shear, creep-recovery tests, and large deformation experiments, are discussed in this review as techniques present in the literature to characterize rheological behavior of emulsion-filled gels. Moreover, the correlation of mechanical properties with sensory aspects of emulsion-filled gels appearing in recent studies is discussed, demonstrating the applicability of these parameters in understanding mastication processes.

Push-out bond strength of different tricalcium silicate-based filling materials to root dentin

Directory of Open Access Journals (Sweden)

Jorge Henrique Stefaneli Marques

2018-03-01

Full Text Available Abstract: The aim of this study was to evaluate the bond strength of different triccalcium silicate cements to retrograde cavity using a push out test. Thirty maxillary central incisors were shaped using #80 hand files and sectioned transversally. Root slices were obtained from the apical 4 mm after eliminating the apical extremity. The specimens were embedded in acrylic resin and positioned at 45° to the horizontal plane for preparation of root-end cavities with a diamond ultrasonic retrotip. The samples were divided into three groups according to the root-end filling material (n = 10: MTA Angelus, ProRoot MTA and Biodentine. A gutta-percha cone (#80 was tugged-back at the limit between the canal and the root-end cavity. The root-end cavity was filled and the gutta-percha cone was removed after complete setting of the materials. The specimens were placed in an Instron machine with the root-end filling turned downwards. The push-out shaft was inserted in the space previously occupied by the gutta-percha cone and push out testing was performed at a crosshead speed of 1.0 mm/min. There was no statistically significant difference in resistance to push out by the materials tested (p > 0.01. MTA Angelus and ProRoot MTA showed predominantly mixed failure while Biodentine exhibited mixed and cohesive failures. The tricalcium silicate-based root-end filling materials showed similar bond strength retrograde cavity.

Push-out bond strength of different tricalcium silicate-based filling materials to root dentin.

Science.gov (United States)

Stefaneli Marques, Jorge Henrique; Silva-Sousa, Yara Teresinha Corrêa; Rached-Júnior, Fuad Jacob Abi; Macedo, Luciana Martins Domingues de; Mazzi-Chaves, Jardel Francisco; Camilleri, Josette; Sousa-Neto, Manoel Damião

2018-03-08

The aim of this study was to evaluate the bond strength of different triccalcium silicate cements to retrograde cavity using a push out test. Thirty maxillary central incisors were shaped using #80 hand files and sectioned transversally. Root slices were obtained from the apical 4 mm after eliminating the apical extremity. The specimens were embedded in acrylic resin and positioned at 45° to the horizontal plane for preparation of root-end cavities with a diamond ultrasonic retrotip. The samples were divided into three groups according to the root-end filling material (n = 10): MTA Angelus, ProRoot MTA and Biodentine. A gutta-percha cone (#80) was tugged-back at the limit between the canal and the root-end cavity. The root-end cavity was filled and the gutta-percha cone was removed after complete setting of the materials. The specimens were placed in an Instron machine with the root-end filling turned downwards. The push-out shaft was inserted in the space previously occupied by the gutta-percha cone and push out testing was performed at a crosshead speed of 1.0 mm/min. There was no statistically significant difference in resistance to push out by the materials tested (p > 0.01). MTA Angelus and ProRoot MTA showed predominantly mixed failure while Biodentine exhibited mixed and cohesive failures. The tricalcium silicate-based root-end filling materials showed similar bond strength retrograde cavity.

Standard test method for analysis of isotopic composition of uranium in nuclear-grade fuel material by quadrupole inductively coupled plasma-mass spectrometry

CERN Document Server

American Society for Testing and Materials. Philadelphia

2000-01-01

1.1 This test method is applicable to the determination of the isotopic composition of uranium (U) in nuclear-grade fuel material. The following isotopic weight percentages are determined using a quadrupole inductively coupled plasma-mass spectrometer (Q-ICP-MS): 233U, 234U, 235U, 236U, and 238U. The analysis can be performed on various material matrices after acid dissolution and sample dilution into water or dilute nitric (HNO3) acid. These materials include: fuel product, uranium oxide, uranium oxide alloys, uranyl nitrate (UNH) crystals, and solutions. The sample preparation discussed in this test method focuses on fuel product material but may be used for uranium oxide or a uranium oxide alloy. Other preparation techniques may be used and some references are given. Purification of the uranium by anion-exchange extraction is not required for this test method, as it is required by other test methods such as radiochemistry and thermal ionization mass spectroscopy (TIMS). This test method is also described i...

Preparation, characterization, uranium (VI) biosorption models, and conditions optimization by response surface methodology (RSM) for amidoxime-functionalized marine fungus materials

Energy Technology Data Exchange (ETDEWEB)

Yang, Xuechun; Gao, Yang; Jiang, Min; He, Dianxiong; Liao, Sen; Hou, Dan; Yan, Xueming; Long, Wei; Wu, Yaxin; Tan, Ni [Univ. of South China, Hengyang (China). School of Chemistry and Chemical Engineering

2017-08-01

Amidoxime-functionalized marine fungus Fusarium sp. ZZF51 (ZGDA) was synthesized and studied to adsorb uranium (VI) from the aqueous solution. Different instrumental techniques such as FTIR, SEM, and TGA were employed for the characterization of the manufactured materials, and theirs ability of removal uranium (VI) was optimized using RSM. The experimental results showed the maximum adsorption capacity for the synthesized materials was 230.78 mg g{sup -1} at the following optimization conditions: S-L ratio 150 mg L{sup -1}, pH 5.13, uranium (VI) initial concentration 40 mg L{sup -1}, and equilibrium time 122.40 min. More than 85% of the absorbed uranium (VI) could be desorbed by 0.5 or 1.0 mol L{sup -1} HCl, and the modified mycelium could be reused at least five times. The thermodynamic experimental data of adsorption uranium (VI) could fit better with Langmuir and Freundlich isotherms models, and the pseudo-second-order model was better to interpret the kinetics process. The modified fungus materials exhibited the better sorption capacity for uranium (VI) in comparison with raw biomass should be attributed to the strong chelation of amidoxime to uranium (VI) ions.

Nd:YAG laser in endodontics: filling-material edge bordering on a root channel laser cavity

Science.gov (United States)

Belikov, Andrei V.; Sinelnik, Yuri A.; Moroz, Boris T.; Pavlovskaya, Irina V.

1997-12-01

For the very first time it is represented a study of filling material edge bordering upon root channel cavity modified with a laser. As a filling material it is used a glass ionomer cement. It is demonstrated that Nd:YAG laser radiation effects on increase of grade of edge bordering on the average of 20 - 30% at temperature rise of no more than 2 - 3 degrees in periodontium area in a period of operation.

Contribution to the study of nuclear fuel materials with a metallic uranium base; Contribution a l'etude des materiaux combustibles nucleaires a base d'uranium metallique

Energy Technology Data Exchange (ETDEWEB)

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

1957-11-15

In a power reactor destined to supply industrially recoverable thermal energy, the most economical source of heat still consists of natural metallic uranium. However, the nuclear fuel material, most often employed in the form of rods of 20 to 40 mm diameter, is subjected to a series of stresses which lead to irreversible distortions usually incompatible with the substructure of the reactor. As a result the fuel material must possess at the outset a certain number of qualities which must be determined. Investigations have therefore been carried out, first on the technological characters peculiar to each of the three allotropic phases of pure uranium metal, and on their interactions on the stabilisation of the material which consists of either cast uranium or uranium pile-treated in the {gamma} phase. (author) [French] Dans un reacteur de puissance destine a fournir de l'energie thermique industriellement recuperable, la source de chaleur la plus economique reste constituee par de l'uranium metallique naturel. Or, le materiau combustible nucleaire, employe le plus souvent sous forme de barreaux de 20 a 40 mm de diametre, se trouve soumis a un ensemble de contraintes qui provoque des deformations irreversibles, le plus souvent incompatibles avec l'infrastructure du reacteur. Par consequent, le materiau combustible doit presenter a l'origine un certain nombre de qualites qu'il est necessaire de determiner. Aussi a-t-on d'abord etudie les caracteres technologiques propres a chacune des trois phases allotropiques de l'uranium-metal pur et leurs interactions sur la stabilisation du materiau constitue soit par de l'uranium coule, soit par de l'uranium traite en pile en phase {gamma}. (auteur)

Method of preparation of uranium nitride

Science.gov (United States)

Kiplinger, Jaqueline Loetsch; Thomson, Robert Kenneth James

2013-07-09

Method for producing terminal uranium nitride complexes comprising providing a suitable starting material comprising uranium; oxidizing the starting material with a suitable oxidant to produce one or more uranium(IV)-azide complexes; and, sufficiently irradiating the uranium(IV)-azide complexes to produce the terminal uranium nitride complexes.

Air-Filled Nanopore Based High-Performance Thermal Insulation Materials

OpenAIRE

Gangåssæter, Haakon Fossen; Jelle, Bjørn Petter; Alex Mofid, Sohrab; Gao, Tao

2017-01-01

State-of-the-art thermal insulation solutions like vacuum insulation panels (VIP) and aerogels have low thermal conductivity, but their drawbacks may make them unable to be the thermal insulation solutions that will revolutionize the building industry regarding energy-efficient building envelopes. Nevertheless, learning from these materials may be crucial to make new and novel high-performance thermal insulation products. This study presents a review on the state-of-the-art air-filled thermal...

Application of uranium impurity data for material characterization in nuclear safeguards

International Nuclear Information System (INIS)

Penkin, M.V.; Boulyga, S.F.; Fischer, D.M.

2016-01-01

Samples of materials involved in the conversion of uranium into nuclear-grade products are collected to support the verification of States' declarations and to look for indications of possible undeclared materials and activities. Samples are analysed by several laboratories to determine concentrations of about sixty impurities; the data consistency is addressed through the unified reporting requirements, the use of common reference materials, and via inter-laboratory comparisons. The impurity analysis results, along with other essential parameters, are interpreted to judge sample conformity to the relevant specifications, to evaluate the facility design information, to assess material provenance and intended use. (author)

Dental Glass Ionomer Cements as Permanent Filling Materials? – Properties, Limitations and Future Trends

Directory of Open Access Journals (Sweden)

Ulrich Lohbauer

2009-12-01

Full Text Available Glass ionomer cements (GICs are clinically attractive dental materials that have certain unique properties that make them useful as restorative and luting materials. This includes adhesion to moist tooth structures and base metals, anticariogenic properties due to release of fluoride, thermal compatibility with tooth enamel, biocompatibility and low toxicity. The use of GICs in a mechanically loaded situation, however, has been hampered by their low mechanical performance. Poor mechanical properties, such as low fracture strength, toughness and wear, limit their extensive use in dentistry as a filling material in stress-bearing applications. In the posterior dental region, glass ionomer cements are mostly used as a temporary filling material. The requirement to strengthen those cements has lead to an ever increasing research effort into reinforcement or strengthening concepts.

Surficial uranium deposits in Somalia

International Nuclear Information System (INIS)

Briot, P.

1984-01-01

Surficial uranium deposits in Somalia are of the valley-fill calcrete type and occur in the arid Mudugh Province of the Dusa Mareb-El Bur region. They are located in a belt about 240 km in length which is orientated parallel to the north-south regional tectonic framework. The uranium resources of the region amount to about 5,000 t U 3 O 8 at an average grade of 0.1% U 3 O 8 . Basement rocks constitute a 7,000 m thick succession of Jurassic to Quaternary sediments of the Somalian Basin. Uranium mineralization in the form of carnotite occurs in the uppermost Mercia Series. The origin of the uranium and vanadium is unclear due to a shortage of the favourable source rocks. (author)

US uranium market developments

International Nuclear Information System (INIS)

Krusiewski, S.V.; Patterson, J.A.

1980-01-01

Domestic uranium delivery commitments have risen significantly since January 1979, with the bulk of deliveries scheduled after 1990. Much of the long-term procurement will be obtained from captive production. However, buyers have adjusted their delivery schedules in the near term, deferring some procurement to later years, including a portion of planned captive production. Under current commitments, US imports of foreign uranium in the 1981 to 1985 period will be greater than our exports of domestic uranium. The anticipated supply of domestic uranium through 1985 is clearly more than adequate to fill the probable US demand in the meantime, uranium producers are continuing their efforts to increase future domestic supply by their considerable investments in new or expanded mine and mill facilities. Since January 1980, average contract prices including market-price settlements, for 1980 uranium deliveries have increased slightly, but average market-price settlements made this year have decreased by several dollars. While the general trend of US uranium prices has been upward since we began reporting price data in 1973, some reductions in average prices for future deliveries appeared in 1980. The softening of prices for new procurement can be expected to be increasingly apparent in future surveys

Soil Characterization by Large Scale Sampling of Soil Mixed with Buried Construction Debris at a Former Uranium Fuel Fabrication Facility

International Nuclear Information System (INIS)

Nardi, A.J.; Lamantia, L.

2009-01-01

Recent soil excavation activities on a site identified the presence of buried uranium contaminated building construction debris. The site previously was the location of a low enriched uranium fuel fabrication facility. This resulted in the collection of excavated materials from the two locations where contaminated subsurface debris was identified. The excavated material was temporarily stored in two piles on the site until a determination could be made as to the appropriate disposition of the material. Characterization of the excavated material was undertaken in a manner that involved the collection of large scale samples of the excavated material in 1 cubic meter Super Sacks. Twenty bags were filled with excavated material that consisted of the mixture of both the construction debris and the associated soil. In order to obtain information on the level of activity associated with the construction debris, ten additional bags were filled with construction debris that had been separated, to the extent possible, from the associated soil. Radiological surveys were conducted of the resulting bags of collected materials and the soil associated with the waste mixture. The 30 large samples, collected as bags, were counted using an In-Situ Object Counting System (ISOCS) unit to determine the average concentration of U-235 present in each bag. The soil fraction was sampled by the collection of 40 samples of soil for analysis in an on-site laboratory. A fraction of these samples were also sent to an off-site laboratory for additional analysis. This project provided the necessary soil characterization information to allow consideration of alternate options for disposition of the material. The identified contaminant was verified to be low enriched uranium. Concentrations of uranium in the waste were found to be lower than the calculated site specific derived concentration guideline levels (DCGLs) but higher than the NRC's screening values. The methods and results are presented

Feasibility study on consolidation of Fernald Environmental Management Project depleted uranium materials

International Nuclear Information System (INIS)

1995-01-01

In 1991, the DOE made a decision to close the FMPC located in Fernald, Ohio, and end its production mission. The site was renamed FEMP to reflect Fernald's mission change from uranium production to environmental restoration. As a result of this change, the inventory of strategic uranium materials maintained at Fernald by DOE DP will need to be relocated to other DOE sites. Although considered a liability to the Fernald Plant due to its current D and D mission, the FEMP DU represents a potentially valuable DOE resource. Recognizing its value, it may be important for the DOE to consolidate the material at one site and place it in a safe long-term storage condition until a future DOE programmatic requirement materializes. In August 1995, the DOE Office of Nuclear Weapons Management requested, Lockheed Martin Energy Systems (LMES) to assess the feasibility of consolidating the FEMP DU materials at the Oak Ridge Reservation (ORR). This feasibility study examines various phases associated with the consolidation of the FEMP DU at the ORR. If useful short-term applications for the DU fail to materialize, then long-term storage (up to 50 years) would need to be provided. Phases examined in this report include DU material value; potential uses; sampling; packaging and transportation; material control and accountability; environmental, health and safety issues; storage; project management; noneconomic factors; schedule; and cost

First-Principles Integrated Adsorption Modeling for Selective Capture of Uranium from Seawater by Polyamidoxime Sorbent Materials.

Science.gov (United States)

Ladshaw, Austin P; Ivanov, Alexander S; Das, Sadananda; Bryantsev, Vyacheslav S; Tsouris, Costas; Yiacoumi, Sotira

2018-04-18

Nuclear power is a relatively carbon-free energy source that has the capacity to be utilized today in an effort to stem the tides of global warming. The growing demand for nuclear energy, however, could put significant strain on our uranium ore resources, and the mining activities utilized to extract that ore can leave behind long-term environmental damage. A potential solution to enhance the supply of uranium fuel is to recover uranium from seawater using amidoximated adsorbent fibers. This technology has been studied for decades but is currently plagued by the material's relatively poor selectivity of uranium over its main competitor vanadium. In this work, we investigate the binding schemes between uranium, vanadium, and the amidoxime functional groups on the adsorbent surface. Using quantum chemical methods, binding strengths are approximated for a set of complexation reactions between uranium and vanadium with amidoxime functionalities. Those approximations are then coupled with a comprehensive aqueous adsorption model developed in this work to simulate the adsorption of uranium and vanadium under laboratory conditions. Experimental adsorption studies with uranium and vanadium over a wide pH range are performed, and the data collected are compared against simulation results to validate the model. It was found that coupling ab initio calculations with process level adsorption modeling provides accurate predictions of the adsorption capacity and selectivity of the sorbent materials. Furthermore, this work demonstrates that this multiscale modeling paradigm could be utilized to aid in the selection of superior ligands or ligand compositions for the selective capture of metal ions. Therefore, this first-principles integrated modeling approach opens the door to the in silico design of next-generation adsorbents with potentially superior efficiency and selectivity for uranium over vanadium in seawater.

Influence of uranium hydride oxidation on uranium metal behaviour

International Nuclear Information System (INIS)

Patel, N.; Hambley, D.; Clarke, S.A.; Simpson, K.

2013-01-01

This work addresses concerns that the rapid, exothermic oxidation of active uranium hydride in air could stimulate an exothermic reaction (burning) involving any adjacent uranium metal, so as to increase the potential hazard arising from a hydride reaction. The effect of the thermal reaction of active uranium hydride, especially in contact with uranium metal, does not increase in proportion with hydride mass, particularly when considering large quantities of hydride. Whether uranium metal continues to burn in the long term is a function of the uranium metal and its surroundings. The source of the initial heat input to the uranium, if sufficient to cause ignition, is not important. Sustained burning of uranium requires the rate of heat generation to be sufficient to offset the total rate of heat loss so as to maintain an elevated temperature. For dense uranium, this is very difficult to achieve in naturally occurring circumstances. Areas of the uranium surface can lose heat but not generate heat. Heat can be lost by conduction, through contact with other materials, and by convection and radiation, e.g. from areas where the uranium surface is covered with a layer of oxidised material, such as burned-out hydride or from fuel cladding. These rates of heat loss are highly significant in relation to the rate of heat generation by sustained oxidation of uranium in air. Finite volume modelling has been used to examine the behaviour of a magnesium-clad uranium metal fuel element within a bottle surrounded by other un-bottled fuel elements. In the event that the bottle is breached, suddenly, in air, it can be concluded that the bulk uranium metal oxidation reaction will not reach a self-sustaining level and the mass of uranium oxidised will likely to be small in relation to mass of uranium hydride oxidised. (authors)

Influence of uranium hydride oxidation on uranium metal behaviour

Energy Technology Data Exchange (ETDEWEB)

Patel, N.; Hambley, D. [National Nuclear Laboratory (United Kingdom); Clarke, S.A. [Sellafield Ltd (United Kingdom); Simpson, K.

2013-07-01

This work addresses concerns that the rapid, exothermic oxidation of active uranium hydride in air could stimulate an exothermic reaction (burning) involving any adjacent uranium metal, so as to increase the potential hazard arising from a hydride reaction. The effect of the thermal reaction of active uranium hydride, especially in contact with uranium metal, does not increase in proportion with hydride mass, particularly when considering large quantities of hydride. Whether uranium metal continues to burn in the long term is a function of the uranium metal and its surroundings. The source of the initial heat input to the uranium, if sufficient to cause ignition, is not important. Sustained burning of uranium requires the rate of heat generation to be sufficient to offset the total rate of heat loss so as to maintain an elevated temperature. For dense uranium, this is very difficult to achieve in naturally occurring circumstances. Areas of the uranium surface can lose heat but not generate heat. Heat can be lost by conduction, through contact with other materials, and by convection and radiation, e.g. from areas where the uranium surface is covered with a layer of oxidised material, such as burned-out hydride or from fuel cladding. These rates of heat loss are highly significant in relation to the rate of heat generation by sustained oxidation of uranium in air. Finite volume modelling has been used to examine the behaviour of a magnesium-clad uranium metal fuel element within a bottle surrounded by other un-bottled fuel elements. In the event that the bottle is breached, suddenly, in air, it can be concluded that the bulk uranium metal oxidation reaction will not reach a self-sustaining level and the mass of uranium oxidised will likely to be small in relation to mass of uranium hydride oxidised. (authors)

Development of a novel filling technique. Loading bulk particulate materials into tankers or processes

Energy Technology Data Exchange (ETDEWEB)

Farnish, R.J.; Berry, R.; Bradley, M. [Greenwich Univ., Chatham Maritime, Kent (United Kingdom). Wolfson Centre for Bulk Solids Handling Technology

2008-07-01

The majority of industrial dosing or filling operations demand high filling rates and often good repeatability of discharges. For coarse, free-flowing materials the issues of obtaining a high degree of filling efficiency are substantially less challenging than for less free-flowing or cohesive bulk particulates. Typical equipment arrangements for achieving a controlled (often dual rate) discharge of particles into a relatively small capacity container (flask, sack or big bag) often rely on either a mechanical extraction of material from a buffer (screw feeders) or the manipulation of a constricting arrangement to achieve a turn down in discharge rate. Where less freeflowing or very fine particles are being handled, the introduction of air into the powder is invariably used to modify the bulk condition of the material to a condition where discharge can be initiated and supported (typical examples being powder feed to an impeller packer, or discharge of powder into a rail or road wagon). This article will therefore report on some recent research that has been undertaken by The Wolfson Centre for Bulk Solids Handling Technology, University of Greenwich. (orig.)

Thermal properties of lauric acid filled in carbon nanotubes as shape-stabilized phase change materials.

Science.gov (United States)

Feng, Yanhui; Wei, Runzhi; Huang, Zhi; Zhang, Xinxin; Wang, Ge

2018-03-14

Carbon nanotubes (CNTs) filled with lauric acid (LA) as a kind of shape-stabilized phase change material were prepared and their structures and phase change properties were characterized. The results showed that the melting point and latent heat of LA confined in carbon nanotubes were lower than those of the bulk material, and both decrease as the diameters of CNTs and the filling ratios of LA decrease. Molecular dynamics (MD) simulations indicated that LA molecules form a liquid layer near pore walls and crystallize at the pore center. When the LA filling ratio was reduced to a certain value, all LA molecules were attached to the inner walls of CNTs, hindering their crystallization. A linear relationship between the melting temperature shift and structural properties was obtained based on the modified Gibbs-Thomson equation, which gives a reliable interpretation of the size effect of nanochannels in phase change materials. We also found that the thermal conductivity of the composite CNTs/LA was four times larger than that of pure LA. This study will provide insights into the design of novel composite phase change materials with better thermal properties by the selection of suitable porous materials and tailoring their pore structures.

Three-dimensional evaluation of effectiveness of hand and rotary instrumentation for retreatment of canals filled with different materials.

Science.gov (United States)

Hammad, Mohammad; Qualtrough, Alison; Silikas, Nick

2008-11-01

The aim of this study was to measure the remaining filling volume of different obturation materials from root-filled extracted teeth by using 2 removal techniques. Eighty single-rooted teeth were collected and decoronated, and the root canal was prepared by using the ProTaper nickel-titanium rotary files. The teeth were randomly allocated into 4 groups, and each group was obturated by using a different material. Group 1 was filled with gutta-percha and TubliSeal sealer, group 2 was filled with EndoRez points and EndoRez sealer, group 3 was filled with RealSeal points and RealSeal sealer, and Group 4 was filled with a gutta-percha point and GuttaFlow sealer. Teeth were scanned with a micro-computed tomography scan, and then root fillings were removed by using ProTaper retreatment files or hand K-files. Teeth were scanned again, and volume measurements were carried out with micro-computed tomography software. Statistical analysis showed significant differences between the 2 removal techniques for gutta-percha and for both techniques between gutta-percha and the other groups. The present study showed that all tested filling materials were not completely removed during retreatment by using hand or rotary files. Gutta-percha was more efficiently removed by using hand K-files.

Hybrid waste filler filled bio-polymer foam composites for sound absorbent materials

Science.gov (United States)

Rus, Anika Zafiah M.; Azahari, M. Shafiq M.; Kormin, Shaharuddin; Soon, Leong Bong; Zaliran, M. Taufiq; Ahraz Sadrina M. F., L.

2017-09-01

Sound absorption materials are one of the major requirements in many industries with regards to the sound insulation developed should be efficient to reduce sound. This is also important to contribute in economically ways of producing sound absorbing materials which is cheaper and user friendly. Thus, in this research, the sound absorbent properties of bio-polymer foam filled with hybrid fillers of wood dust and waste tire rubber has been investigated. Waste cooking oil from crisp industries was converted into bio-monomer, filled with different proportion ratio of fillers and fabricated into bio-polymer foam composite. Two fabrication methods is applied which is the Close Mold Method (CMM) and Open Mold Method (OMM). A total of four bio-polymer foam composite samples were produce for each method used. The percentage of hybrid fillers; mixture of wood dust and waste tire rubber of 2.5 %, 5.0%, 7.5% and 10% weight to weight ration with bio-monomer. The sound absorption of the bio-polymer foam composites samples were tested by using the impedance tube test according to the ASTM E-1050 and Scanning Electron Microscope to determine the morphology and porosity of the samples. The sound absorption coefficient (α) at different frequency range revealed that the polymer foam of 10.0 % hybrid fillers shows highest α of 0.963. The highest hybrid filler loading contributing to smallest pore sizes but highest interconnected pores. This also revealed that when highly porous material is exposed to incident sound waves, the air molecules at the surface of the material and within the pores of the material are forced to vibrate and loses some of their original energy. This is concluded that the suitability of bio-polymer foam filled with hybrid fillers to be used in acoustic application of automotive components such as dashboards, door panels, cushion and etc.

Miscellaneous radioactive materials detected during uranium mill tailings surveys

International Nuclear Information System (INIS)

Wilson, M.J.

1993-10-01

The Department of Energy's (DOE) Office of Environmental Restoration and Waste Management directed the Oak Ridge National Laboratory Pollutant Assessments Group in the conduct of radiological surveys on properties in Monticello, Utah, associated with the Mendaciously millsite National Priority List site. During these surveys, various radioactive materials were detected that were unrelated to the Monticello millsite. The existence and descriptions of these materials were recorded in survey reports and are condensed in this report. The radioactive materials detected are either naturally occurring radioactive material, such as rock and mineral collections, uranium ore, and radioactive coal or manmade radioactive material consisting of tailings from other millsites, mining equipment, radium dials, mill building scraps, building materials, such as brick and cinderblock, and other miscellaneous sources. Awareness of the miscellaneous and naturally occurring material is essential to allow DOE to forecast the additional costs and schedule changes associated with remediation activities. Also, material that may pose a health hazard to the public should be revealed to other regulatory agencies for consideration

Ex vivo evaluation of coronal and apical microbial leakage of root canal - Filled with gutta-percha or Resilon/Epiphany root canal filling material

Directory of Open Access Journals (Sweden)

de Almeida-Gomes Fabio

2010-01-01

Full Text Available Objective: This ex vivo study compared coronal and apical microleakage of root canals filled with Resilon/Epiphany (RE or gutta-percha/Grossman sealer (GP, using either lateral condensation (LC or System B (SB technique. Materials and Methods: Specimens in eight experimental groups were obturated using the following materials and techniques: Groups 1 and 3 - GP and LC; groups 2 and 4 - GP and SB; groups 5 and 7 - RE and LC; groups 6 and 8 - RE and SB. Apical and coronal leakages were tested using bacterial methods. For coronal analysis, the number of days required for complete contamination of the root canals was recorded according to observation of the brain heart infusion broth turbidity for 15 weeks. For apical analysis, the teeth were cleaved and the leakage was measured at 30 days. Data were collected for each sample and analyzed statistically with the Chi-square and ANOVA tests. Results: Leakage was found in all groups. The difference between filling materials, obturation techniques, and median time of leakage was not statistically significant for coronal ( P=0.847 and apical ( P=0.5789 leakages. Conclusion: There were no differences between the different filling materials (gutta-percha/Grossman sealer and Resilon/Epiphany and obturation techniques (lateral condensation and system B technique in coronal or apical leakages.

Friction Properties of Laminated Composite Materials of Alpha-Tricalcium Phosphate–Filled Poly (Vinyl Alcohol) Hydrogels

OpenAIRE

Yamamoto, Kanae; Iwai, Tomoaki; Shoukaku, Yutaka

2015-01-01

The aim of this study was to examine the mechanical characteristics of a polyvinyl alcohol hydrogel (PVA-H) as a candidate material for artificial joint cartilage. In the study, PVA-H was filled with α-tricalcium phosphate (α-TCP) in order to improve its mechanical properties. In addition, laminated composite materials with 3 layers were prepared by laminating α-TCP–filled PVA-H and unfilled PVA-H. The samples were prepared with different numbers of repeated freeze–thaw cycles and several con...

Equations of state for enriched uranium and uranium alloy to 3500 MPa

International Nuclear Information System (INIS)

Bai Chaomao; Hai Yuying; Liu Jenlong; Li Zhenrong

1990-04-01

The volume compressions of 6 kinds of cast materials including enriched uranium, poor uranium, U-0.57 wt% Ti, U-0.33 wt% Nb, U-2.85 wt% Nb and U-7.5 wt% Nb-3.3 wt% Zr have been determined by monitoring piston displacements in a piston cylinder apparatus with double strengthening rings to 3500 MPa at room temperature. The dilation of the cylinder vessel and the press deformation were corrected by some experiments. The calculational data free from using the standard sample closed with used standard sample. The volume compressions of enriched uranium and poor uranium are nearly coincident. Pure uranium is more compressible than uranium alloys. These values of enriched uranium are in close agreement with values of Bridgman's pure uranium. The fitting coefficients of Bridgman's polynomial and Anderson's equation of state and isothermal bulk modules for the above materials are given

Quality assurance in the enriched uranium operations NDA facility

Energy Technology Data Exchange (ETDEWEB)

May, P.K.; Ceo, R.N. [Oak Ridge Y-12 Plant, TN (United States)

1997-11-01

The Nondestructive Analysis (NDA) Facility at the Oak Ridge Y-12 Plant has characterized process wastes for Enriched Uranium Operations since 1978. Since that time, over 50,000 items have been analyzed. Analysis results are used to determine whether or not recovery of uranium from process wastes is economically feasible. Our instrument complement includes one large segmented gamma scanner (SGS), two smaller SGS, two solution assay systems (SAS), and Active Well Coincidence Counter (AWCC). The large SGS is used for analyzing High Efficiency Particulate Air (HEPA) filters ant 208-L drums filled with combustible contaminated waste. The smaller SGS are used to analyze 4-L containers of ash and leached residues. The SAS are used to analyze 125 ml bottles of aqueous or organic waste solutions that may contain uranium. The gamma-based NDA techniques are used to identify which process wastes can be discarded, and which must be recycled. The AWCC is used to analyze high-density materials which are not amenable to gamma-ray analysis. 1 ref., 4 figs.

Evolution of Xihulitu basin and its control to uranium ore-formation

International Nuclear Information System (INIS)

Guo Qingyin; Li Ziying; Dong Wenming

2003-01-01

There is a close relationship between basin filling succession and evolution of the basin. Characteristics of basin evolution can be studied by analyzing the basin filling succession. Two major periods are recognized according to the filling succession and subsequent alteration of the Xihulitu Basin. Evolutionary characteristics of each stage of the basin formation and alteration have been discussed in details. The types and special distribution of uranium metallization are controlled by the scale, connection degree and distribution of sandstone units and impermeable mudstone beds. The environment of uranium ore-formation became favorable as the faults modified the hydrodynamic condition. The basin had been uplifted for a long time after it was filled. Intergranular pores are not destroyed due to the weak mechanical compaction, which is beneficial to groundwater penetrating. Montmorillonitization and zeolitization in some sandstone units are strong because of the high content of volcanic fragments. The major uranium metallization is the phreatic oxidation type. The northern zone of the second sub-basin in the central section of the basin is regarded as the first perspective target for subsequent exploration. (authors)

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

Science.gov (United States)

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

2018-03-01

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

Elaboration of vitreous and vitrocrystalline basalt materials containing simulated radioactive ash wastes. Study of some physical properties and leaching behaviour

International Nuclear Information System (INIS)

Lebeau, M.J.

1988-01-01

This work is a preliminary study of a matrix for containment of incinerator ashes. Basalt presents a good capacity for ash incorporation (up to 50%), glasses obtained are homogeneous and the low viscosity at 1300 0 C allows an industrial preparation. Vitrocrystalline products have a dendritic texture which can be controlled by cooling rate and are composed of magnesioferrite, pyroxene, plagioclase and vitreous fraction depending on the filling material ratio. Uranium and thorium, for actinide simulation in ashes, are localized in the glass. Glass leaching decrease with ash content and the alteration film presents a better retention of some elements, such as uranium and thorium. Vitrocrystalline materials are less leachable than glasses. Interesting possibilities are shown for use of basalt as filling material in underground storage, since basalt decrease glass deterioration [fr

Study on filling materials suitable for seawater piping trench closure work at Fukushima Daiichi Nuclear Power Plant

International Nuclear Information System (INIS)

Yanai, Shuji; Hibi, Yasuki; Nishikori, Kazumasa; Sato, Keita

2016-01-01

Highly contaminated water leaking from the reactor buildings and turbine buildings damaged by the 2011 Great East Japan Earthquake has accumulated in the seawater piping trenches of Fukushima Daiichi Nuclear Power Station Units 2, 3, and 4. In November 2014, work commenced to replace and remove this contaminated water by filling the trenches with filling materials, and this work was completed in December 2015. This paper summarizes the contents of this study on various filling materials, including special fillers with long-distance underwater flowability applied to the horizontal tunnel parts of the trenches. (author)

Quantitative assessment of image artifacts from root filling materials on CBCT scans made using several exposure parameters

Energy Technology Data Exchange (ETDEWEB)

Rabelo, Katharina Alves; Cavalcanti, Yuri Wanderley; De Oliveira Pinto, Martina Gerlane; De Melo, Daniela Pita [Dept. of Oral Diagnosis, State University of Paraiba, Campina Grande (Brazil); Melo, Saulo Leonardo Sousa [Dept. of Oral Pathology, Radiology and Medicine, University of Iowa, Iowa City (United States); Campos, Paulo Sergio Flores; De Andrade Freitas Oliveira, Luciana Soares [Federal University of Bahia, Salvador (Brazil)

2017-09-15

To quantify artifacts from different root filling materials in cone-beam computed tomography (CBCT) images acquired using different exposure parameters. Fifteen single-rooted teeth were scanned using 8 different exposure protocols with 3 different filling materials and once without filling material as a control group. Artifact quantification was performed by a trained observer who made measurements in the central axial slice of all acquired images in a fixed region of interest using ImageJ. Hyperdense artifacts, hypodense artifacts, and the remaining tooth area were identified, and the percentages of hyperdense and hypodense artifacts, remaining tooth area, and tooth area affected by the artifacts were calculated. Artifacts were analyzed qualitatively by 2 observers using the following scores: absence (0), moderate presence (1), and high presence (2) for hypodense halos, hypodense lines, and hyperdense lines. Two-way ANOVA and the post-hoc Tukey test were used for quantitative and qualitative artifact analysis. The Dunnet test was also used for qualitative analysis. The significance level was set at P<.05. There were no significant interactions among the exposure parameters in the quantitative or qualitative analysis. Significant differences were observed among the studied filling materials in all quantitative analyses. In the qualitative analyses, all materials differed from the control group in terms of hypodense and hyperdense lines (P<.05). Fiberglass posts did not differ statistically from the control group in terms of hypodense halos (P>.05). Different exposure parameters did not affect the objective or subjective observations of artifacts in CBCT images; however, the filling materials used in endodontic restorations did affect both types of assessments.

DUSCOBS - a depleted-uranium silicate backfill for transport, storage, and disposal of spent nuclear fuel

International Nuclear Information System (INIS)

Forsberg, C.W.; Pope, R.B.; Ashline, R.C.; DeHart, M.D.; Childs, K.W.; Tang, J.S.

1995-01-01

A Depleted Uranium Silicate COntainer Backfill System (DUSCOBS) is proposed that would use small, isotopically-depleted uranium silicate glass beads as a backfill material inside storage, transport, and repository waste packages containing spent nuclear fuel (SNF). The uranium silicate glass beads would fill all void space inside the package including the coolant channels inside SNF assemblies. Based on preliminary analysis, the following benefits have been identified. DUSCOBS improves repository waste package performance by three mechanisms. First, it reduces the radionuclide releases from SNF when water enters the waste package by creating a local uranium silicate saturated groundwater environment that suppresses (1) the dissolution and/or transformation of uranium dioxide fuel pellets and, hence, (2) the release of radionuclides incorporated into the SNF pellets. Second, the potential for long-term nuclear criticality is reduced by isotopic exchange of enriched uranium in SNF with the depleted uranium (DU) in the glass. Third, the backfill reduces radiation interactions between SNF and the local environment (package and local geology) and thus reduces generation of hydrogen, acids, and other chemicals that degrade the waste package system. In addition, the DUSCOBS improves the integrity of the package by acting as a packing material and ensures criticality control for the package during SNF storage and transport. Finally, DUSCOBS provides a potential method to dispose of significant quantities of excess DU from uranium enrichment plants at potential economic savings. DUSCOBS is a new concept. Consequently, the concept has not been optimized or demonstrated in laboratory experiments

Derivation of uranium residual radioactive material guidelines for the Elza Gate Site

International Nuclear Information System (INIS)

Cheng, J.J.; Yu, C.; Devgun, J.S.

1991-02-01

Residual radioactive material guidelines for uranium were derived for a large, homogeneously contaminated area at the Elza Gate Site in Oak Ridge, Tennessee. The derivation of the single-nuclide and total uranium guidelines was based on the requirement that the 50-year committed effective dose equivalent to hypothetical individual who lives or works in the immediate vicinity of the Elza Gate Site should not exceed a dose of 100 mrem/yr following decontamination. The DOE residual radioactive guideline computer code RESRAD was used in this evaluation. Four potential scenarios were considered for the site; the scenarios vary with regard to time spent at the site, sources of water used, and sources of food consumed. The results of the evaluation indicate that the basic dose limit of 100 mrem/yr will not be exceeded for uranium within 1000 years, provided that the soil concentration of uranium at the Elza Gate Site does not exceed the following levels: 1800 pCi/g for Scenario A (industrial worker: the expected scenario); 4000 pCi/g for Scenario B (recreationist: a plausible scenario); 470 pCi/g for Scenario C (resident farmer using pond water as the only water source: a possible but unlikely scenario); and 120 pCi/g for Scenario D (resident farmer using well water as the only water source: a possible but unlikely scenario). The uranium guideline applies to the total activity concentration of uranium isotopes in their natural activity concentration ratio of 1:1: 0.046. These guidelines are calculated on the basis of a dose of 100 mrem/yr. In setting the actual uranium guideline for the Elza Gate Site, the DOE will apply the as low as reasonably achievable (ALARA) policy to the decision-making process, along with other factors, such as determining whether a particular scenario is reasonable and appropriate. 10 refs., 3 figs., 7 tabs

Uranium in soil, forest litter and living plant material above three uranium mineralizations in Northern Sweden

International Nuclear Information System (INIS)

Ek, John

1982-01-01

In order to investigate the feasibility of biogeochemical sampling media in uranium exploration, samples from the most common trees and low bushes together with forest litter were collected over the areas of three uranium mineralizations in Northern Sweden and analyzed for uranium. The results were compared with uranium content of the till and its radioactivity. The average uranium content was low for all sample types and considerably lower in the ash of the organic sample types compared to that of the till. No sample type showed any tendency of having higher uranium concentration above mineralizations compared to background areas. These results suggest that, under conditions prevailing in Sweden, the investigated sample types are not suitable for uranium exploration

Cellular localization of uranium in the renal proximal tubules during acute renal uranium toxicity.

Science.gov (United States)

Homma-Takeda, Shino; Kitahara, Keisuke; Suzuki, Kyoko; Blyth, Benjamin J; Suya, Noriyoshi; Konishi, Teruaki; Terada, Yasuko; Shimada, Yoshiya

2015-12-01

Renal toxicity is a hallmark of uranium exposure, with uranium accumulating specifically in the S3 segment of the proximal tubules causing tubular damage. As the distribution, concentration and dynamics of accumulated uranium at the cellular level is not well understood, here, we report on high-resolution quantitative in situ measurements by high-energy synchrotron radiation X-ray fluorescence analysis in renal sections from a rat model of uranium-induced acute renal toxicity. One day after subcutaneous administration of uranium acetate to male Wistar rats at a dose of 0.5 mg uranium kg(-1) body weight, uranium concentration in the S3 segment of the proximal tubules was 64.9 ± 18.2 µg g(-1) , sevenfold higher than the mean renal uranium concentration (9.7 ± 2.4 µg g(-1) ). Uranium distributed into the epithelium of the S3 segment of the proximal tubules and highly concentrated uranium (50-fold above mean renal concentration) in micro-regions was found near the nuclei. These uranium levels were maintained up to 8 days post-administration, despite more rapid reductions in mean renal concentration. Two weeks after uranium administration, damaged areas were filled with regenerating tubules and morphological signs of tissue recovery, but areas of high uranium concentration (100-fold above mean renal concentration) were still found in the epithelium of regenerating tubules. These data indicate that site-specific accumulation of uranium in micro-regions of the S3 segment of the proximal tubules and retention of uranium in concentrated areas during recovery are characteristics of uranium behavior in the kidney. Copyright © 2015 John Wiley & Sons, Ltd.

Recovery of uranium and molybdenum elements from gebel gattar raw material, eastern desert, Egypt. Vol. 3

Energy Technology Data Exchange (ETDEWEB)

El-Hazek, N T; Mahdy, M A; Mahmoud, H M.K. [Nuclear Materials Authority, Cairo, (Egypt)

1996-03-01

G. Gatter uranium mineralizations are located along the faults and fracture zones crossing G.Gattar granitic pluton and long the contact of the pluton with the hammamat sediments. Also, molybdenum id presented in more than one mode of occurrence. The molybdenum mineralization treated in this work is the dessimenated type. The uranium and molybdenum raw material was subjected to series of leaching experiments including acid and alkaline agitation, alkaline percolation, and acid heap leaching techniques. Recovery of uranium and molybdenum was achieved by anion-exchange method followed by their elution by acidified sodium chloride. Uranium precipitation was performed in the form of ammonium diuranate (Yellow Cake). On the other hand molybdenum was precipitated in the form of molybdenum oxide. A tentative flowsheet for the extraction of both uranium and molybdenum is proposed and discussed. 13 figs., 3 tabs.

Recovery of uranium and molybdenum elements from gebel gattar raw material, eastern desert, Egypt. Vol. 3

International Nuclear Information System (INIS)

El-Hazek, N.T.; Mahdy, M.A.; Mahmoud, H.M.K.

1996-01-01

G. Gatter uranium mineralizations are located along the faults and fracture zones crossing G.Gattar granitic pluton and long the contact of the pluton with the hammamat sediments. Also, molybdenum id presented in more than one mode of occurrence. The molybdenum mineralization treated in this work is the dessimenated type. The uranium and molybdenum raw material was subjected to series of leaching experiments including acid and alkaline agitation, alkaline percolation, and acid heap leaching techniques. Recovery of uranium and molybdenum was achieved by anion-exchange method followed by their elution by acidified sodium chloride. Uranium precipitation was performed in the form of ammonium diuranate (Yellow Cake). On the other hand molybdenum was precipitated in the form of molybdenum oxide. A tentative flowsheet for the extraction of both uranium and molybdenum is proposed and discussed. 13 figs., 3 tabs

Domestic round robin exercise on analysis of uranium for nuclear material handling facilities in Japan

International Nuclear Information System (INIS)

Kato, Yoshiyasu; Nagai, Kohta; Handa, Takamitsu; Inoue, Shin-ichi; Sato, Yoshihiro

2016-01-01

Interlaboratory comparison programme as well as internal quality control system is an effective tool for an analytical laboratory responsible to nuclear material accountancy of a nuclear facility to maintain and enhance its capability for analysis. However, it is a burden on nuclear material handling facilities in Japan to attend interlaboratory comparison programme run by overseas institutions because of high costs and complicated procedure for importing nuclear materials, and therefore facilities which can participate in such international programme would be limited. Nuclear Material Control Center has hence started and organised an annual domestic round robin exercise on analysis of uranium standard materials, funded by the Japan Safeguards Office of the Nuclear Regulation Authority, since 2008 to enhance analytical capability of Japanese Facilities. The outline of the round robin exercise will be given and the results of uranium isotopic and concentration analysis reported by participant facilities from 2008 to 2015 will be summarised in the presentation. (author)

Quantitative assessment of image artifacts from root filling materials on CBCT scans made using several exposure parameters

International Nuclear Information System (INIS)

Rabelo, Katharina Alves; Cavalcanti, Yuri Wanderley; De Oliveira Pinto, Martina Gerlane; De Melo, Daniela Pita; Melo, Saulo Leonardo Sousa; Campos, Paulo Sergio Flores; De Andrade Freitas Oliveira, Luciana Soares

2017-01-01

To quantify artifacts from different root filling materials in cone-beam computed tomography (CBCT) images acquired using different exposure parameters. Fifteen single-rooted teeth were scanned using 8 different exposure protocols with 3 different filling materials and once without filling material as a control group. Artifact quantification was performed by a trained observer who made measurements in the central axial slice of all acquired images in a fixed region of interest using ImageJ. Hyperdense artifacts, hypodense artifacts, and the remaining tooth area were identified, and the percentages of hyperdense and hypodense artifacts, remaining tooth area, and tooth area affected by the artifacts were calculated. Artifacts were analyzed qualitatively by 2 observers using the following scores: absence (0), moderate presence (1), and high presence (2) for hypodense halos, hypodense lines, and hyperdense lines. Two-way ANOVA and the post-hoc Tukey test were used for quantitative and qualitative artifact analysis. The Dunnet test was also used for qualitative analysis. The significance level was set at P .05). Different exposure parameters did not affect the objective or subjective observations of artifacts in CBCT images; however, the filling materials used in endodontic restorations did affect both types of assessments

Comparison of the sealing ability of mineral trioxide aggregate and Portland cement used as root-end filling materials.

Science.gov (United States)

Shahi, Shahriar; Yavari, Hamid R; Rahimi, Saeed; Eskandarinezhad, Mahsa; Shakouei, Sahar; Unchi, Mahsa

2011-12-01

Inadequate apical seal is the major cause of surgical endodontic failure. The root-end filling material used should prevent egress of potential contaminants into periapical tissue. The purpose of this study was to compare the sealing ability of four root-end filling materials: white mineral trioxide aggregate (MTA), gray MTA, white Portland cement (PC) and gray PC by dye leakage test. Ninety-six human single-rooted teeth were instrumented, and obturated with gutta-percha. After resecting the apex, an apical cavity was prepared. The teeth were randomly divided into four experimental groups (A: white MTA, B: gray MTA, C: white PC and D: gray PC; n = 20) and two control groups (positive and negative control groups; n = 8). Root-end cavities in the experimental groups were filled with the experimental materials. The teeth were exposed to Indian ink for 72 hours. The extent of dye penetration was measured with a stereomicroscope at 16× magnification. The negative controls showed no dye penetration and dye penetration was seen in the entire root-end cavity of positive controls. However, there was no statistically significant difference among the four experimental groups (P > 0.05). All retrograde filling materials tested in this study showed the same microleakage in vitro. Given the low cost and apparently similar sealing ability of PC, PC could be considered as a substitute for MTA as a root-end filling material.

Studies on fission tracks and distributions of uranium and rare earths in granite materials

International Nuclear Information System (INIS)

Matsuda, Hiroshi; Sakanoue, Masanobu

1987-01-01

Many materials contain fossil records of the slow spontaneous fission of uranium they contain as an impurity. Fission fragments, heavy charged particles released in each fission event, produce microscopic trails of radiation damage that may persist over geological times and may be developed to a size observable under an optical microscope by a suitable etching treatment. Such tracks are also produced by fissions induced by thermal neutron irradiation of the uranium. When the material is heated sufficiently, it anneals and the the microscopic trails become shorter and narrower. The track density decreases, because the chemical etchant will not reach some of the shortened tracks. Measurements of track densities before and after annealing can be used, along with laboratory studies of annealing rates, to determine the annealing temperature. Also, the track density of induced fissions is related to the concentration of uranium and the fluence of neutrons to which it was exposed. If the track density due to induced fissions can be distinguished from that due to fossil tracks, estimates of either the concentration or the fluence can be made if the other is known. Two such materials (one a fragment of a granite paving stone, the other a piece of stained glass from a cathedral window) that had been exposed to the atomic bomb at Nagasaki were used in the present work. The fossil record in zircons in the granite was used to estimate the temperature to which it had been exposed in the bombing. Induced fissions were used to estimate the concentration of uranium in the zircons. Nonuniform heating and cooling and nearly uniform exposure to the neutrons make the granite sample unsuitable for determining the neutron fluence from the bomb. Induced fissions in the stained glass were used to estimate the concentration of uranium and the thermal neutron fluence from the A-bomb. Annealing of tracks in glass was also studied

Synthesis of Uranium nitride powders using metal uranium powders

International Nuclear Information System (INIS)

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

2012-01-01

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

Uranium industry annual 1994

International Nuclear Information System (INIS)

1995-01-01

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

Selection and specification criteria for fills for cut-and-fill mining

Energy Technology Data Exchange (ETDEWEB)

Thomas, E. G.

1980-05-15

Because of significant differences in placement and loading conditions, the ideal fill material for a cut-and-fill operation has different characteristics to those for a fill for a filled open stoping operation. The differing requirements of the two mining operations must be understood and accounted for in establishing fill selection and specification criteria. Within the paper, aspects of the particular requirements of cut-and-fill mining are analyzed and related to the specific fill tests and properties required. Emphasis is placed upon the role of fill in ground support, though this cannot be isolated from overall fill performance. Where appropriate, test data are introduced and areas requiring continuing research highlighted.

Derivation of guidelines for uranium residual radioactive material in soil at the Colonie Site, Colonie, New York

International Nuclear Information System (INIS)

Dunning, D.

1996-05-01

Residual radioactive material guidelines for uranium in soil were derived for the Colonie site located in Colonie, New York. This site has been designated for remedial action under the Formerly Utilized Sites Remedial Action Program (FUSRAP) of the U.S. Department of Energy (DOE). The site became contaminated with radioactive material as a result of operations conducted by National Lead (NL) Industries from 1958 to 1984; these activities included brass foundry operations, electroplating of metal products, machining of various components using depleted uranium, and limited work with small amounts of enriched uranium and thorium. The Colonie site comprises the former NL Industries property, now designated the Colonie Interim Storage Site (CISS), and 56 vicinity properties contaminated by fallout from airborne emissions; 53 of the vicinity properties were previously remediated between 1984 and 1988. In 1984, DOE accepted ownership of the CISS property from NL Industries. Residual radioactive material guidelines for individual radionuclides and total uranium were derived on the basis of the requirement that the 50-year committed effective dose equivalent to a hypothetical individual who lives or works in the immediate vicinity of the site should not exceed a dose of 30 mrem/yr following remedial action for the current use and likely future use scenarios or a dose of 100 mrem/yr for less likely future use scenarios. The DOE residual radioactive material guideline computer code, RESRAD, was used in this evaluation; RESRAD implements the methodology described in the DOE manual for establishing residual radioactive material guidelines

Derivation of guidelines for uranium residual radioactive material in soil at the B ampersand T Metals Company site, Columbus, Ohio

International Nuclear Information System (INIS)

Kamboj, S.; Nimmagadda, Mm.; Yu, C.

1996-01-01

Guidelines for uranium residual radioactive material in soil were derived for the B ampersand T Metals Company site in Columbus, Ohio. This site has been identified for remedial action under the US Department of Energy's (DOE's) Formerly Utilized Sites Remedial Action Program (FUSRAP). Single-nuclide and total-uranium guidelines were derived on the basis of the requirement that following remedial action, the 50-year committed effective dose equivalent to a hypothetical individual living or working in the immediate vicinity of the site should not exceed a dose constraint of 30 mrem/yr for the current use and likely future use scenarios or a dose limit of 100 n-mrem/yr for less likely future use scenarios. The DOE residual radioactive material guideline computer code, RESRAD, was used in this evaluation. RESRAD implements the methodology described in the DOE manual for establishing residual radioactive material guidelines. Three scenarios were considered; each assumed that for a period of 1,000 years following remedial action, the site would be used without radiological restrictions. The three scenarios varied with regard to the type of site use, time spent at the site by the exposed individual, and sources of food and water consumed. The evaluations indicate that the dose constraint of 30 mrem/yr would not be exceeded for uranium (including uranium-234, uranium-235, and uranium-238) within 1,000 years, provided that the soil concentration of total uranium (uranium-234, uranium-235, and uranium-238) at the B ampersand T Metals site did not exceed 1, I 00 pCi/g for Scenario A (industrial worker, current use) or 300 pCi/g for Scenario B (resident with municipal water supply, a likely future use). The dose limit of 100 mrem/yr would not be exceeded at the site if the total uranium concentration of the soil did not exceed 880 pCi/g for Scenario C (resident with an on-site water well, a plausible but unlikely future use)

Uranium industry annual, 1991

International Nuclear Information System (INIS)

1992-10-01

In the Uranium Industry Annual 1991, data on uranium raw materials activities including exploration activities and expenditures, resources and reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities including domestic uranium purchases, commitments by utilities, procurement arrangements, uranium imports under purchase contracts and exports, deliveries to enrichment suppliers, inventories, secondary market activities, utility market requirements, and uranium for sale by domestic suppliers are presented in Chapter 2. A feature article entitled ''The Uranium Industry of the Commonwealth of Independent States'' is included in this report

10 CFR 74.33 - Nuclear material control and accounting for uranium enrichment facilities authorized to produce...

Science.gov (United States)

2010-01-01

... and special nuclear material in the accounting records are based on measured values; (3) A measurement... 10 Energy 2 2010-01-01 2010-01-01 false Nuclear material control and accounting for uranium... Section 74.33 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) MATERIAL CONTROL AND ACCOUNTING OF SPECIAL...

Development of a stable uranium recovery regulatory framework for uranium recovery activities in the United States

International Nuclear Information System (INIS)

Layton, M.C.; Abrams, C.E.

2000-01-01

The U.S. Nuclear Regulatory Commission (NRC) has historically regulated operations at all uranium and thorium recovery facilities under the authority of the Atomic Energy Act of 1954, as amended. Uranium recovery facilities are those plants, or portions of facilities that process uranium- or thorium-bearing material primarily for its source material content. The uranium recovery industry expressed some concerns over several aspects of the NRC's practices, as described in the NRC's guidance documents. In April 1998, the National Mining Association submitted a report to the Commission, that identified specific concerns with NRC's current position and guidance regarding concurrent jurisdiction at uranium mills; dual regulatory authority at in situ leach facilities; the use of mill tailings impoundments for disposal of radioactive material other than 11e.(2) byproduct material; and the ability to process alternate feed material at uranium mills. The NRC staff addressed most of these concerns in two SECY (staff recommendations) papers that were concurrently provided to the Commission, along with a SECY paper on a draft rulemaking plan relating to these and other issues. The issues addressed in these papers included a new rulemaking, disposal of materials other than 11 e.(2) byproduct material, processing of materials other than natural ores, and improved efficiency for regulating in situ leach uranium facilities. The Commission issued final policy decisions on these issues and directions for NRC staff to implement those decisions in July 2000. (author)

Feasibility studies on supercritical fluid extraction of uranium from phosphoric acid

International Nuclear Information System (INIS)

Dubey, B.P.; Agarwal, A.K.

2014-01-01

Supercritical fluid extraction (SFE) is a promising novel technology for extraction of many materials. Work has been carried out worldwide on SFE of uranium from various matrices. However, there are no references indicating the R and D on uranium extraction from phosphoric acid using this technology. Heavy Water Board is involved in technology development for recovery of uranium from secondary source, hence it was considered prudent to investigate the technology of SFE for this purpose. Various experiments were carried out with both WPA (P 2 O 5 content 28%) and MGPA (P 2 O 5 content 54%) using bench scale facility available with one of the private party. Extraction experiments were carried out using several chelating agents including TBP, D2EHPA, D2EHPA+TBP/TOPO, TTA, TTA+TBP etc. Feasibility studies revealed the hydrodynamics of operation indicating liquid expansion by about three times during flow of super critical (SC) CO 2 . No flooding was observed when the extraction column filled 20% of its volume capacity, no carryover of entrained/extracted liquid with SC CO 2 with MGPA, material balance of inputs and outputs established i.e. 100% recovery of MGPA and chelating agent, No operational problems with raw MGPA (untreated). No significant extraction of impurities from phosphoric acid to SC CO 2 , 40℃ temperature and 160 bar pressure found ideal for extraction experiments since no other materials found extracted at these conditions and no apparent change/deterioration in PA and chelating agents. Experiments established feasibility of SCE with CO 2 , proper recovery of PA and chelating agents, no need for pretreatment/gunk removal from PA; however, extraction of uranium was found inadequate even though ORP of feed acid was boosted by H 2 O 2 addition. Investigations revealed that SCE column created reducing environment in phosphoric acid, which was not favourable for uranium extraction, which resulted in difficulty in extraction of Uranium. HWB has now designed

Contribution of the surface contamination of uranium-materials on the quantitative analysis results by electron probe microbeam analysis

International Nuclear Information System (INIS)

Bonino, O.; Fournier, C.; Fucili, C.; Dugne, O.; Merlet, C.

2000-01-01

The analytical testing of uranium materials is necessary for quality research and development in nuclear industry applications (enrichment, safety studies, fuel, etc). Electron Probe Microbeam Analysis Wavelength Dispersive Spectrometry (EPMA-WDS) is a dependable non-destructive analytical technology. The characteristic X-ray signal is measured to identify and quantify the sample components, and the analyzed volume is about one micron cube. The surface contamination of uranium materials modifies and contributes to the quantitative analysis results of EPMA-WDS. This contribution is not representative of the bulk. A thin oxidized layer appears in the first instants after preparation (burnishing, cleaning) as well as a carbon contamination layer, due to metallographic preparation and carbon cracking under the impact of the electron probe. Several analytical difficulties subsequently arise, including an overlapping line between the carbon Ka ray and the Uranium U NIVOVI ray. Sensitivity and accuracy of the quantification of light elements like carbon and oxygen are also reduced by the presence of uranium. The aim of this study was to improve the accuracy of quantitative analysis on uranium materials by EPMA-WDS by taking account of the contribution of surface contamination. The first part of this paper is devoted to the study of the contaminated surface of the uranium materials U, UFe 2 and U 6 Fe a few hours after preparation. These oxidation conditions are selected so as to reproduce the same contamination surfaces occurring in microprobe analytical conditions. Surface characterization techniques were SIMS and Auger spectroscopy. The contaminated surfaces are shown. They consist of successive layers: a carbon layer, an oxidized iron layer, followed by an iron depletion layer (only in UFe 2 and U 6 Fe), and a ternary oxide layer (U-Fe-O for UFe 2 et U 6 Fe and UO 2+x for uranium). The second part of the paper addresses the estimation of the errors in quantitative

Bulk-Fill Resin Composites

DEFF Research Database (Denmark)

Benetti, Ana Raquel; Havndrup-Pedersen, Cæcilie; Honoré, Daniel

2015-01-01

the restorative procedure. The aim of this study, therefore, was to compare the depth of cure, polymerization contraction, and gap formation in bulk-fill resin composites with those of a conventional resin composite. To achieve this, the depth of cure was assessed in accordance with the International Organization...... for Standardization 4049 standard, and the polymerization contraction was determined using the bonded-disc method. The gap formation was measured at the dentin margin of Class II cavities. Five bulk-fill resin composites were investigated: two high-viscosity (Tetric EvoCeram Bulk Fill, SonicFill) and three low......-viscosity (x-tra base, Venus Bulk Fill, SDR) materials. Compared with the conventional resin composite, the high-viscosity bulk-fill materials exhibited only a small increase (but significant for Tetric EvoCeram Bulk Fill) in depth of cure and polymerization contraction, whereas the low-viscosity bulk...

Uranium Industry Annual, 1992

International Nuclear Information System (INIS)

1993-01-01

The Uranium Industry Annual provides current statistical data on the US uranium industry for the Congress, Federal and State agencies, the uranium and electric utility industries, and the public. The feature article, ''Decommissioning of US Conventional Uranium Production Centers,'' is included. Data on uranium raw materials activities including exploration activities and expenditures, resources and reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities including domestic uranium purchases, commitments by utilities, procurement arrangements, uranium imports under purchase contracts and exports, deliveries to enrichment suppliers, inventories, secondary market activities, utility market requirements, and uranium for sale by domestic suppliers are presented in Chapter 2

Uranium Industry Annual, 1992

Energy Technology Data Exchange (ETDEWEB)

1993-10-28

The Uranium Industry Annual provides current statistical data on the US uranium industry for the Congress, Federal and State agencies, the uranium and electric utility industries, and the public. The feature article, ``Decommissioning of US Conventional Uranium Production Centers,`` is included. Data on uranium raw materials activities including exploration activities and expenditures, resources and reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities including domestic uranium purchases, commitments by utilities, procurement arrangements, uranium imports under purchase contracts and exports, deliveries to enrichment suppliers, inventories, secondary market activities, utility market requirements, and uranium for sale by domestic suppliers are presented in Chapter 2.

Catahoula formation as a source of sedimentary uranium deposits in east Texas

International Nuclear Information System (INIS)

Ledger, E.B.; Tieh, T.T.

1983-01-01

Volcanic glass-rich mudstone and siltstone samples from the Oligocene/Miocene Catahoula formation of Jasper County, Texas, and coeval volcaniclastic rock samples from Trans-Pecos, Texas, have been compared as to U, Th, Zr, Ti, K, Rb, and Sr contents. Uranium is slightly greater in the distal ash (5.85 ppM U) compared to the Trans-Pecos samples (average 5.41 ppM U). Diagenetic and pedogenetic alteration of Catahoula volcanic glass releases uranium to solution and, under favorable conditions, this uranium may accumulate to form ore bodies. Uranium has been produced from such ore bodies in south Texas, but economic deposits are not known in east Texas. Significant differences between south and east Texas include: (1) a greater amount of volcanic debris delivered to south Texas, both as air-fall ash and stream-transported material, (2) delivery of only air-fill ash to east Texas, (3) the possibility of more petroleum-related reductants such as H 2 S in south Texas, and (4) pervasive glass alteration with subsequent uranium release in south Texas due to late calichification. These differences argue against economic deposits of the south Texas type being found in east Texas. If economic deposits occur they are likely to be far downdip making exploration difficult and expensive

Fluorimetric determination of uranium in certain refractory minerals, environmental samples and industrial waste materials

International Nuclear Information System (INIS)

Premadas, A.; Saravanakumar, G.

2005-01-01

A simple sample decomposition and laser fluorimetric determination of uranium at trace level is reported in certain refractory minerals, like ilmenite, rutile, zircon and monazite; environmental samples viz. soil and sediments; industrial waste materials, such as, coal fly ash and red mud. Ilmenite sample is decomposed by heating with ammonium fluoride. Rutile, zircon and monazite minerals are decomposed by fusion using a mixture of potassium bifluoride and sodium fluoride. Environmental and industrial waste materials are brought into solution by treating with a mixture of hydrofluoric and nitric acids. The laser induced fluorimetric determination of uranium is carried out directly in rutile, zircon and in monazite minerals and after separation in other samples. The determination limit was 1 μg x g -1 for ilmenite, soil, sediment, coal fly ash and red mud samples, and it is 5 μg x g -1 for rutile, zircon and monazite. The method is also developed for the optical fluorimetric determination of uranium (determination limit 10 μg x g -1 ) in ilmenite, rutile, zircon and monazite minerals. The methods are simple, accurate, and precise and they require small quantity of sample and can be applied for the routine analysis. (author)

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)

Selection and evaluation of inner material candidates for Spanish high level radioactive waste canisters

International Nuclear Information System (INIS)

Puig, Francesc; Dies, Javier; Sevilla, Manuel; Pablo, Joan de; Pueyo, Juan Jose; Miralles, Lourdes; Martinez-Esparza, Aurora

2007-01-01

This paper summarizes the work carried out to analyse different alternatives related to the inner material selection of the Spanish high level waste canister for long term storage. The preliminary repository design considers granitic or clay formations, compacted bentonite sealing, corrosion allowing steel canisters and glass bead filling between the fuel assemblies and canister walls. This filling material will have the primary role of avoiding the possibility of a criticality event, which becomes an issue of major importance once the container is finally breached by corrosion and flooded by groundwater. In the first place, a complete set of requirements have been devised as evaluation criteria for candidate materials examination and selection; resulting in a compilation of demands significantly deeper and more exhaustive than any other similar work found in literature, including over 20 requirements and some other general aspects that could involve improvements in repository performance. Secondly, eight materials or material families (cast iron or steel, borosilicate glass, spinel, depleted uranium, dehydrated zeolites, hematite, phosphates and olivine) have been chosen and examined in detail, extracting some relevant conclusions. Either cast iron, borosilicate glass, spinel or depleted uranium are considered to look quite promising for the mentioned purpose. (authors)

About Dental Amalgam Fillings

Science.gov (United States)

... and Medical Procedures Dental Devices Dental Amalgam About Dental Amalgam Fillings Share Tweet Linkedin Pin it More ... should I have my fillings removed? What is dental amalgam? Dental amalgam is a dental filling material ...

Bioassay for uranium mill tailings

International Nuclear Information System (INIS)

Tschaeche, A.N.

1986-01-01

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

Ore petrography of a sedimentary uranium deposit, Live Oak County, Texas

International Nuclear Information System (INIS)

Bomber, B.J.; Ledger, E.B.; Tieh, T.T.

1986-01-01

Samples from the McLean 5 open-pit uranium mine, a small high-grade deposit located along a normal fault in the Miocene Oakville sandstone of Live Oak County, Texas, have been studied for uranium abundance, distribution, and nature of occurrence on the microscopic level. The host sandstone is composed of quartz, feldspars, and volcanic rock fragments, cemented by sparry calcite. Authigenic minerals include iron disulfide minerals (dominantly pyrite and some marcasite) and small amounts of clays, Ti oxides, and opal. High-grade ore (to 3% U) occurs along the fault, decreasing to less than 1,000 ppm within 10 m from the fault. The ore mineral is amorphous pitchblende and exhibits botryoidal morphology. The microscopic occurrence of uranium, documented by fission-track mapping of petrographic thin sections, is presented in detail. Uranium occurs abundantly as grain coatings and fillings in intergranular spaces in samples with high uranium content, where calcite cement has been partially or totally leached as mineralization proceeded. Lesser amounts are adsorbed onto leucoxene (microcrystalline anatase), mud clasts, and altered igneous rock fragments. Adsorbed uranium is the major code of occurrence in samples, with lower uranium contents farther from the orebody. Textural relations indicate that iron sulfides formed both before and after mineralization. Initial mineralization was by adsorption onto aggregates of fine particles of Ti oxide and clay minerals of various origins. With dissolution of cement and continued uranium influx, uranium precipitated as grain coatings and pore fillings

In situ leaching process for recording uranium values

International Nuclear Information System (INIS)

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

1977-01-01

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

Ablation - breakthrough technology to reduce uranium mining cost and increase resources

International Nuclear Information System (INIS)

Scriven, D.

2014-01-01

Ablation Technologies, LLC has developed and patented a revolutionary mining technology termed “ablation”. Ablation is a process using only mechanical forces to upgrade sandstone uranium ores. Uranium bearing sandstone orebodies are formed from a uranium enriched solution flowing through an aquifer until it reached some type of a “red/ox” zone forcing the uranium and other heavy metals to come out of solution. The precipitate forms a thin coating on the sand grains and fills the interstitial space between the sand grains but does no penetrate the sand grains. The ablation process knocks the precipitate off the sand grains using the forces of abrasion, elastic compression and rebounding, much like a mud coated tennis ball will sheds the mud when bounced off the ground, and to some extent, sonic waves. This produces a product which collectively is exactly the same as the ore going in but with all the individual components separated. This allows for disgressionary separation, the most important of which is screening. The uranium and heavy metals report to the finer fractions of the material, typically less than 250 mesh. The larger fractions contain less than five percent of the uranium but 90 to 95 percent of the mass. The advantages of making an enriched ore are numerous: • Reduce haulage costs from 90 to 95 percent. • Reduce milling costs by reducing material handling costs, acid consumption and tailings disposal costs. • In addition to reducing overall mining and milling costs, the overall recovery of the recourse is increased because the ablation process is so inexpensive, if the material has to be mined it will be ablated and screened. This basically means ore control is significantly reduced, cutoff grade goes to practically zero and overall resource recovery is significantly increased. • Environmentally, the two major advantages are reduced tailings requirements at the mill site and cleaner waste dumps at the mine site. This paper will show

Uranium ores

International Nuclear Information System (INIS)

Poty, B.; Roux, J.

1998-01-01

The processing of uranium ores for uranium extraction and concentration is not much different than the processing of other metallic ores. However, thanks to its radioactive property, the prospecting of uranium ores can be performed using geophysical methods. Surface and sub-surface detection methods are a combination of radioactive measurement methods (radium, radon etc..) and classical mining and petroleum prospecting methods. Worldwide uranium prospecting has been more or less active during the last 50 years, but the rise of raw material and energy prices between 1970 and 1980 has incited several countries to develop their nuclear industry in order to diversify their resources and improve their energy independence. The result is a considerable increase of nuclear fuels demand between 1980 and 1990. This paper describes successively: the uranium prospecting methods (direct, indirect and methodology), the uranium deposits (economical definition, uranium ores, and deposits), the exploitation of uranium ores (use of radioactivity, radioprotection, effluents), the worldwide uranium resources (definition of the different categories and present day state of worldwide resources). (J.S.)

Laser-induced breakdown spectroscopy measurements of uranium and thorium powders and uranium ore

Energy Technology Data Exchange (ETDEWEB)

Judge, Elizabeth J. [Chemistry Division, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Barefield, James E., E-mail: jbarefield@lanl.gov [Chemistry Division, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Berg, John M. [Manufacturing Engineering and Technology Division, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Clegg, Samuel M.; Havrilla, George J.; Montoya, Velma M.; Le, Loan A.; Lopez, Leon N. [Chemistry Division, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

2013-05-01

Laser-induced breakdown spectroscopy (LIBS) was used to analyze depleted uranium and thorium oxide powders and uranium ore as a potential rapid in situ analysis technique in nuclear production facilities, environmental sampling, and in-field forensic applications. Material such as pressed pellets and metals, has been extensively studied using LIBS due to the high density of the material and more stable laser-induced plasma formation. Powders, on the other hand, are difficult to analyze using LIBS since ejection and removal of the powder occur in the laser interaction region. The capability of analyzing powders is important in allowing for rapid analysis of suspicious materials, environmental samples, or trace contamination on surfaces since it most closely represents field samples (soil, small particles, debris etc.). The rapid, in situ analysis of samples, including nuclear materials, also reduces costs in sample collection, transportation, sample preparation, and analysis time. Here we demonstrate the detection of actinides in oxide powders and within a uranium ore sample as both pressed pellets and powders on carbon adhesive discs for spectral comparison. The acquired LIBS spectra for both forms of the samples differ in overall intensity but yield a similar distribution of atomic emission spectral lines. - Highlights: • LIBS analysis of mixed actinide samples: depleted uranium oxide and thorium oxide • LIBS analysis of actinide samples in powder form on carbon adhesive discs • Detection of uranium in a complex matrix (uranium ore) as a precursor to analyzing uranium in environmental samples.

Filling material for a buried cavity in a collapse area using light-weighted foam and active feldspar

Science.gov (United States)

Cho, Jin Woo; Lee, Ju-hyoung; Kim, Sung-Wook; Choi, Eun-Kyeong

2017-04-01

Concrete which is generally used as filling material for a buried cavity has very high strength but significantly high self-load is considered its disadvantage. If it is used as filling material, the second collapse due to additional load, causing irreversible damage. If light-weighted foam and active feldspar are used to solve this problem, the second collapse can be prevented by reducing of self-load of filling material. In this study, the specimen was produced by mixing light-weighted foam, active feldspar and cement, and changes in the density, unconfined compressive strength and hydraulic conductivity were analyzed. Using the light-weighted foam could enable the adjustment of density of specimen between 0.5 g/cm3 and 1.7 g/cm3, and if the mixing ratio of the light-weighted foam increases, the specimen has more pores and smaller range of cross-sectional area. It is confirmed that it has direct correlation with the density, and if the specimen has more pores, the density of the specimen is lowered. The density of the specimen influences the unconfined compressive strength and the hydraulic conductivity, and it was also confirmed that the unconfined compressive strength could be adjusted between 0.6 MPa and 8 MPa and the hydraulic conductivity could be adjusted between 10-9cm/sec and 10-3cm/sec. These results indicated that we can adjust unconfined compressive strength and hydraulic conductivity of filling materials by changing the mixing amount of lightweight-weighted foam according to the requirements of the field condition. Keywords: filling material, buried cavity, light-weighted foam, feldspar Acknowledgement This research was supported by a Grant from a Strategic Research Project (Horizontal Drilling and Stabilization Technologies for Urban Search and Rescue (US&R) Operation) funded by the Korea Institute of Civil Engineering and Building Technology.

Measurement of thermal diffusivity of depleted uranium metal microspheres

Science.gov (United States)

Humrickhouse-Helmreich, Carissa J.; Corbin, Rob; McDeavitt, Sean M.

2014-03-01

The high void space of nuclear fuels composed of homogeneous uranium metal microspheres may allow them to achieve ultra-high burnup by accommodating fuel swelling and reducing fuel/cladding interactions; however, the relatively low thermal conductivity of microsphere nuclear fuels may limit their application. To support the development of microsphere nuclear fuels, an apparatus was designed in a glovebox and used to measure the apparent thermal diffusivity of a packed bed of depleted uranium (DU) microspheres with argon fill in the void spaces. The developed Crucible Heater Test Assembly (CHTA) recorded radial temperature changes due to an initial heat pulse from a central thin-diameter cartridge heater. Using thermocouple positions and time-temperature data, the apparent thermal diffusivity was calculated. The thermal conductivity of the DU microspheres was calculated based on the thermal diffusivity from the CHTA, known material densities and specific heat capacities, and an assumed 70% packing density based on prior measurements. Results indicate that DU metal microspheres have very low thermal conductivity, relative to solid uranium metal, and rapidly form an oxidation layer even in a low oxygen environment. At 500 °C, the thermal conductivity of the DU metal microsphere bed was 0.431 ± 0.0560 W/m-K compared to the literature value of approximately 32 W/m-K for solid uranium metal.

Measurement of thermal diffusivity of depleted uranium metal microspheres

Energy Technology Data Exchange (ETDEWEB)

Humrickhouse-Helmreich, Carissa J., E-mail: carissahelmreich@tamu.edu [Texas A and M University, Department of Nuclear Engineering, 337 Zachry Engineering Center, 3133 TAMU, College Station, TX 77843 (United States); Corbin, Rob, E-mail: rcorbin@terrapower.com [TerraPower, LLC, 330 120th Ave NE, Suite 100, Bellevue, WA 98005 (United States); McDeavitt, Sean M., E-mail: mcdeavitt@tamu.edu [Texas A and M University, Department of Nuclear Engineering, 337 Zachry Engineering Center, 3133 TAMU, College Station, TX 77843 (United States)

2014-03-15

The high void space of nuclear fuels composed of homogeneous uranium metal microspheres may allow them to achieve ultra-high burnup by accommodating fuel swelling and reducing fuel/cladding interactions; however, the relatively low thermal conductivity of microsphere nuclear fuels may limit their application. To support the development of microsphere nuclear fuels, an apparatus was designed in a glovebox and used to measure the apparent thermal diffusivity of a packed bed of depleted uranium (DU) microspheres with argon fill in the void spaces. The developed Crucible Heater Test Assembly (CHTA) recorded radial temperature changes due to an initial heat pulse from a central thin-diameter cartridge heater. Using thermocouple positions and time–temperature data, the apparent thermal diffusivity was calculated. The thermal conductivity of the DU microspheres was calculated based on the thermal diffusivity from the CHTA, known material densities and specific heat capacities, and an assumed 70% packing density based on prior measurements. Results indicate that DU metal microspheres have very low thermal conductivity, relative to solid uranium metal, and rapidly form an oxidation layer even in a low oxygen environment. At 500 °C, the thermal conductivity of the DU metal microsphere bed was 0.431 ± 0.0560 W/m-K compared to the literature value of approximately 32 W/m-K for solid uranium metal.

Measurement of thermal diffusivity of depleted uranium metal microspheres

International Nuclear Information System (INIS)

Humrickhouse-Helmreich, Carissa J.; Corbin, Rob; McDeavitt, Sean M.

2014-01-01

The high void space of nuclear fuels composed of homogeneous uranium metal microspheres may allow them to achieve ultra-high burnup by accommodating fuel swelling and reducing fuel/cladding interactions; however, the relatively low thermal conductivity of microsphere nuclear fuels may limit their application. To support the development of microsphere nuclear fuels, an apparatus was designed in a glovebox and used to measure the apparent thermal diffusivity of a packed bed of depleted uranium (DU) microspheres with argon fill in the void spaces. The developed Crucible Heater Test Assembly (CHTA) recorded radial temperature changes due to an initial heat pulse from a central thin-diameter cartridge heater. Using thermocouple positions and time–temperature data, the apparent thermal diffusivity was calculated. The thermal conductivity of the DU microspheres was calculated based on the thermal diffusivity from the CHTA, known material densities and specific heat capacities, and an assumed 70% packing density based on prior measurements. Results indicate that DU metal microspheres have very low thermal conductivity, relative to solid uranium metal, and rapidly form an oxidation layer even in a low oxygen environment. At 500 °C, the thermal conductivity of the DU metal microsphere bed was 0.431 ± 0.0560 W/m-K compared to the literature value of approximately 32 W/m-K for solid uranium metal

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

International Nuclear Information System (INIS)

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

2014-01-01

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

Comparison of Endoflas and Zinc oxide Eugenol as root canal filling materials in primary dentition

Directory of Open Access Journals (Sweden)

Nivedita Rewal

2014-01-01

Full Text Available Background: Zinc oxide eugenol has long been the material of choice of pediatric dentists worldwide, although it fails to meet the ideal requirements of root canal filling material for primary teeth. Endoflas, a mixture of zinc oxide eugenol, calcium hydroxide, and iodoform, can be considered to be an effective root canal filling material in primary teeth as compared with zinc oxide eugenol. This study was carried out to compare zinc oxide eugenol with endoflas for pulpectomy in primary dentition. Aim: The objective of the study was to compare clinically and radiographically success rates of zinc oxide eugenol with endoflas for the root canal filling of primary teeth at 3, 6, and 9 months. Design: Fifty primary molars were included in the study with 26 teeth in Group I (Endoflas and 24 in Group II (zinc oxide eugenol. A single visit pulpectomy was carried out. Results: The overall success rate of zinc oxide eugenol was 83% whereas 100% success was found in the case of endoflas. The obtained results were compiled and subjected to statistical analysis using the chi-square test. The difference in the success rate between the two was statistically significant (P < 0.05. Conclusion: Endoflas has shown to have better results than zinc oxide eugenol. It should therefore be the material of choice for root canal treatment in deciduous dentition.

Oxidation of uranium and uranium alloys

International Nuclear Information System (INIS)

Orman, S.

1976-01-01

The corrosion behaviour of uranium in oxygen, water and water + oxygen mixtures is compared and contrasted. A considerable amount of work, much of it conflicting, has been published on the U + H 2 O and U + H 2 O + O 2 systems. An attempt has been made to summarise this data and to explain the reasons for the lack of agreement between the experimental results. The evidence for the mechanism involving OH - ion diffusion as the reacting entity in both the U + H 2 O and U + O 2 + H 2 O reactions is advanced. The more limited corrosion data on some lean uranium alloys and on some higher addition alloys referred to as stainless materials is summarised together with some previously unreported results obtained with these materials at AWRE. The data indicates that in the absence of oxygen the lean alloys behave in a similar manner to uranium and evolve hydrogen in approximately theoretical quantities. But the stainless alloys absorb most of the product hydrogen and assessments of reactivity based on hydrogen evolution would be very inaccurate. The direction that future corrosion work on these materials should take is recommended

Uranium industry annual, 1986

International Nuclear Information System (INIS)

1987-01-01

Uranium industry data collected in the EIA-858 survey provide a comprehensive statistical characterization of annual activities of the industry and include some information about industry plans over the next several years. This report consists of two major sections. The first addresses uranium raw materials activities and covers the following topics: exploration activities and expenditures, resources and reserves, mine production of uranium, production of uranium concentrate, and industry employment. The second major section is concerned with the following uranium marketing activities: uranium purchase commitments, uranium prices, procurement arrangements, uranium imports and exports, enrichment services, inventories, secondary market activities utility market requirements and related topics

Uranium Industry. Annual 1984

International Nuclear Information System (INIS)

Lawrence, M.S.S.

1985-01-01

This report provides a statistical description of activities of the US uranium industry during 1984 and includes a statistical profile of the status of the industry at the end of 1984. It is based on the results of an Energy Information Administration (EIA) survey entitled ''Uranium Industry Annual Survey'' (Form EIA-858). The principal findings of the survey are summarized under two headings - Uranium Raw Materials Activities and Uranium Marketing Activities. The first heading covers exploration and development, uranium resources, mine and mill production, and employment. The second heading covers uranium deliveries and delivery commitments, uranium prices, foreign trade in uranium, inventories, and other marketing activities. 32 figs., 48 tabs

Uranium industry annual 1985

International Nuclear Information System (INIS)

1986-11-01

This report consists of two major sections. The first addresses uranium raw materials activities and covers the following topics: exploration activities and expenditures, resources and reserves, mine production of uranium, production of uranium concentrate, and industry employment. The second major section is concerned with the following uranium marketing activities: uranium purchase commitments, uranium prices, procurement arrangements, uranium imports and exports, enrichment services, inventories, secondary market activities, utility market requirements, and related topics. A glossary and appendices are included to assist the reader in interpreting the substantial array of statistical data in this report and to provide background information about the survey

Uranium retrieval support, storage, and marketing

International Nuclear Information System (INIS)

Jackson, J.D.; Marshall, E.M.

2001-01-01

The United States Department of Energy is implementing a stewardship approach to management of uranium assets. This life-cycle approach to managing uranium addresses current needs in the context of a long-term strategy. In June 1998, the United States Department of Energy established the Uranium Management Group. The mission of the UMG is to safely collect and store commercially viable uranium from various DOE facilities at a central location. The Oak Ridge Operations Office, in Oak Ridge, Tennessee, was given the task to establish a facility for the storage of these uranium materials. Materials collected are non-waste uranium and packaged to allow transport and long-term storage. Coordination of uranium management under the Uranium Management Group offers significant opportunities for sayings through improved planning and efficiency and creates an environmentally sound approach for the storage and reuse of excess uranium. (author)

Uranium retrieval support, storage, and marketing

Energy Technology Data Exchange (ETDEWEB)

Jackson, J.D.; Marshall, E.M. [U.S. Department of Energy, Oak Ridge, Tennessee (United States)

2001-07-01

The United States Department of Energy is implementing a stewardship approach to management of uranium assets. This life-cycle approach to managing uranium addresses current needs in the context of a long-term strategy. In June 1998, the United States Department of Energy established the Uranium Management Group. The mission of the UMG is to safely collect and store commercially viable uranium from various DOE facilities at a central location. The Oak Ridge Operations Office, in Oak Ridge, Tennessee, was given the task to establish a facility for the storage of these uranium materials. Materials collected are non-waste uranium and packaged to allow transport and long-term storage. Coordination of uranium management under the Uranium Management Group offers significant opportunities for sayings through improved planning and efficiency and creates an environmentally sound approach for the storage and reuse of excess uranium. (author)

Uranium deposits of Australia to 1975

International Nuclear Information System (INIS)

Spannari, S.

1979-01-01

This bibliography provides a retrospective account of Australian uranium deposits, particularly the unpublished materials in the Australian Capital Territory. Some abstracts are included. Occurrences, mineralogy, ore genesis, structural controls and the eonomic geology of uranium deposits are covered but the mining of uranium, exploration reports, surveys, environmental aspects and controversial materials are not

Radiological Modeling for Determination of Derived Concentration Levels of an Area with Uranium Residual Material - 13533

Energy Technology Data Exchange (ETDEWEB)

Perez-Sanchez, Danyl [CIEMAT, Avenida Complutense 40, 28040, Madrid (Spain)

2013-07-01

As a result of a pilot project developed at the old Spanish 'Junta de Energia Nuclear' to extract uranium from ores, tailings materials were generated. Most of these residual materials were sent back to different uranium mines, but a small amount of it was mixed with conventional building materials and deposited near the old plant until the surrounding ground was flattened. The affected land is included in an area under institutional control and used as recreational area. At the time of processing, uranium isotopes were separated but other radionuclides of the uranium decay series as Th-230, Ra-226 and daughters remain in the residue. Recently, the analyses of samples taken at different ground's depths confirmed their presence. This paper presents the methodology used to calculate the derived concentration level to ensure that the reference dose level of 0.1 mSv y-1 used as radiological criteria. In this study, a radiological impact assessment was performed modeling the area as recreational scenario. The modelization study was carried out with the code RESRAD considering as exposure pathways, external irradiation, inadvertent ingestion of soil, inhalation of resuspended particles, and inhalation of radon (Rn-222). As result was concluded that, if the concentration of Ra-226 in the first 15 cm of soil is lower than, 0.34 Bq g{sup -1}, the dose would not exceed the reference dose. Applying this value as a derived concentration level and comparing with the results of measurements on the ground, some areas with a concentration of activity slightly higher than latter were found. In these zones the remediation proposal has been to cover with a layer of 15 cm of clean material. This action represents a reduction of 85% of the dose and ensures compliance with the reference dose. (authors)

Potentiometric determination of hexavalent uranium in uranium silicide samples

International Nuclear Information System (INIS)

Arlegui, Oscar

1999-01-01

The Chilean Nuclear Energy Commission's Department of Nuclear Materials has among its projects the production of fuels elements for nuclear reactors, and, therefore, the Chemical Analysis Laboratory must have a rapid and reliable method for uranium analysis, to control the uranium concentration during each stage of the production process. For this reason the Chilean Nuclear Energy Commission's Chemical Analysis Laboratory has validated a potentiometric method, which is a modification of the Davies and Gray method proposed by A.R. Eberle. This method uses the Potentiometric Titration Technique and is based on the direct and rapid reduction of uranium (VI) to Uranium (IV), in a concentrated phosphoric acid medium, with excess iron (II) used as a reducing agent. In this medium the excess iron (II) selectively oxidizes to iron (III) with nitric acid, using molybdenum (IV) as a catalyzer, the nitrous acid that is produced is eliminated by adding amidosulfuric acid. The solution is diluted with 1M sulfuric acid and the uranium (IV) obtained is titrated potentiometrically with potassium dichromate in the presence of vanadilic sulfate to obtain a better defined final titration point. The samples were softened with hydrochloric acid and nitric acid and later 50 ml were estimated in a 20% sulfuric acid medium. The analytical method was validated by comparing it with Certified Reference Material (C.R.M.) from the New Brunswick Laboratory (NBL), Metallic Uranium, CRM 112-A. The F Test and the T Test show that the value calculated is less than the tabulated value so the result is traceable to the reference material. The quantification limit, sensitivity, precision and accuracy were quantified for the method

Development of on-line uranium enrichment monitor of gaseous UF6 for uranium enrichment plant

International Nuclear Information System (INIS)

Lu Xuesheng; Liu Guorong; Jin Huimin; Zhao Yonggang; Li Jinghuai; Hao Xueyuan; Ying Bin; Yu Zhaofei

2013-01-01

An on-line enrichment monitor was developed to measure the enrichment of UF 6 , flowing through the processing pipes in uranium enrichment plant. A Nal (Tl) detector was used to measure the count rates of the 185.7 keV γ-ray emitted from 235 U, and the total quantity of uranium was determined from thermodynamic characteristics of gaseous uranium hexafluoride. The results show that the maximum relative standard deviation is less than 1% when the measurement time is 120 s or more and the pressure is more than 2 kPa in the measurement chamber. Uranium enrichment of gaseous uranium hexafluoride in the output end of cascade can be monitored continuously by using the device. It should be effective for nuclear materials accountability verifications and materials balance verification at uranium enrichment plant. (authors)

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

International Nuclear Information System (INIS)

Forsberg, C.W.

1997-01-01

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

Low grade uranium ores as potential sources of raw material

International Nuclear Information System (INIS)

Venzlaff, H.

1976-01-01

Reports on the uranium requirement and the uranium reserves show that, even if the demand were to be stretched out slightly, the rate of new discoveries of uranium would have to be doubled or even tripled within a few years in order to ensure supply. Despite some spectacular discoveries of large scale deposits in Australia it must be said that only very few truly new uranium provinces have been discovered over the past twenty years. In this situation more attention is now being devoted to low grade uranium depositis, to findings whose concentration does not exceed 1,000 ppm. These deposits contain quantities of uranium many times larger than the deposits that can now be mined at prices up to 30/lb of U 3 O 8 . Even now low grade uranium ore is being mined as a byproduct, with the actual valuable mineral producing most of the income from mining activities. However, if one strikes a balance in this situation, one finds that only part of the requirement can be met in this way. Hence, all possibilities must be exhausted to mine uranium as a byproduct, new techniques of uranium production from low grade ores must be developed, and also conventional prospection must be intensified, if the continuity of supply of the nuclear power stations in the eighties and nineties is to be guaranteed. (orig.) [de

International symposium on uranium production and raw materials for the nuclear fuel cycle - Supply and demand, economics, the environment and energy security. Extended synopses

Energy Technology Data Exchange (ETDEWEB)

NONE

2005-07-01

The IAEA periodically organizes nical meetings and international symposia on all areas of the uranium production cycle. This publication contains 160 extended synopses related to the 2005 international symposium on 'Uranium Production and Raw Materials for the Nuclear Fuel Cycle - Supply and Demand, Economics, the Environment and Energy Security'. They cover all areas of natural uranium resources and production cycle including uranium supply and demand; uranium geology and deposit; uranium exploration; uranium mining and milling; waste management; and environment and regulation. Each synopsis was indexed individually.

Comparison of braided-stream depositional environment and uranium deposits at Saint Anthony underground mine

International Nuclear Information System (INIS)

Baird, C.W.; Martin, K.W.; Lowry, R.M.

1980-01-01

United Nuclear's Saint Anthony mine, located in the Laguna district, produces uranium ore from the Jackpile sandstone unit of the Morrison Formation. The Jackpile sediments were deposited in a fluvial environment characterized by aridity, gentle slope, distant source area, and limited flow volume. Resultant stratigraphy consists of an intricate assemblage of trough and tabular cross-stratification grading to near massive bedding at some locations. Interbedded with the Jackpile sands are green mudstones and siltstones that commonly display irregular thicknesses of less than 2 ft and that are laterally discontinuous. Major penecontemporaneous and postdepositional alteration of originally deposited sands, silts, and clays includes: 1) infiltration and filling of interstices by kaolinitic clays; 2) mobilization and relocation of organic carbonaceous material; and 3) geochemical alteration of mineral constituents and fixation of uranium ions in organic carbonaceous material. Mineralized zones of economic volume display a spatial relationship to bedding features indicative of loosely packed sand deposited in dune and trough foresets. This relationship indicates possible permeability control by initial stratigraphy upon the flow of mineralizing solutions. Additionally, the low-energy foreset environment facilitates the accumulation of low-specific-gravity carbonaceous material necessary for interaction with mineralizing solutions. Large volumes of loosely packed foreset sands accumulate in transverse bars in braided-stream environments. These structures have a great potential for conducting large volumes of mineralizing fluids and hosting economic quantities of uranium ore

Effectiveness of the ProTaper Universal Retreatment™ system for removing a 10-year-old filling material.

Science.gov (United States)

Aguiar, C; Bernart, F; Lima, G; Câmara, A

2012-01-01

The aim of the present study was to evaluate the effectiveness of the ProTaper Universal Retreatment™ system and the manual technique for removing a 10-year-old filling material. Twenty mesiobuccal canals of mandibular first molars with curvature degrees ranging from 20° to 35° were instrumented by the Double-Flared Instrumentation Technique with the NitiFlex files. The root canals were dried with sterile paper points and filled with gutta-percha and a zinc oxide-eugenol-based sealer using a cold lateral condensation technique. Accesses to the cavities were temporarily sealed with Cavit. The teeth were embedded in gauze containing saline solution, which was renewed every 2 weeks, and were stored at 37 °C in 100% humidity for 10 years. Buccolingual and mesiodistal radiographs were taken to evaluate the quality of the root canal filling. The radiographs were mounted in slide frames to allow them to be evaluated by the three examiners, who were specialists in endodontics. Only the specimens in whom there was agreement between the three evaluators were included in the study. The specimens were divided into 2 groups with 10 root canals each, according to gutta-percha removal techniques: Group 1: ProTaper Universal Retreatment™ system; Group 2: Hedström and K-files. Buccolingual and mesiodistal radiographs were taken and evaluated by three examiners who determined the effectiveness of the removal of the filling material. The presence of material was measured according to a score scale. Data were analyzed statistically by Fisher's exact test at 5% significance level. There were no statistical differences between the two methods in the coronal (P=0.211), and middle (P=0.266) thirds, but there were statistical differences between the two methods in the apical third (P=0.038). Not one of the experimental techniques promoted complete removal of the filling materials. The manual technique with K-file and Hedström files achieved better results than the Pro

Remote sensing in uranium exploration

International Nuclear Information System (INIS)

Offield, T.W.

1976-01-01

New types of multispectral data and computer enhancement of images provide a basis for quantitative analysis of ground reflectance, colour discrimination and removal of illumination-geometry effects, not possible with standard aerial photographs. These methods can be designed to take advantage of spectral characteristics of minerals such as hematite and limonite in attempting to discriminate areas of alteration around mineral deposits. The spectral bands of Landsat are not optimum for this discrimination, but several studies show that enhancement of Landsat images permits effective mapping of altered ground in some areas. Red and yellow ground may be confused, a problem where only one of these colours marks alteration related to mineralization. Altered ground in uranium areas has been successfully defined at Cameron, Arizona and Crooks Gap and the Powder River Basin in Wyoming. The Wyoming studies, described in some detail, resulted in unambiguous discrimination of red alteration at Crooks Gap but only partial distinction of red altered ground from yellow-weathering areas in the Powder River Basin. In South Texas, heavy vegetation severely limits the detection of reflectance differences in geological materials or of structural features. Thermal-infrared images of the Texas area aid in detection and mapping of channel-fill deposits, potential loci of uranium mineralization in the Miocene Catahoula Tuff. (author)

Use of depleted uranium silicate glass to minimize release of radionuclides from spent nuclear fuel waste packages

International Nuclear Information System (INIS)

Forsberg, C.W.

1996-01-01

A Depleted Uranium Silicate Container Backfill System (DUSCOBS) is proposed that would use small, isotopically-depleted uranium silicate glass beads as a backfill material inside repository waste packages containing spent nuclear fuel (SNF). The uranium silicate glass beads would fill the void space inside the package including the coolant channels inside SNF assemblies. Based on preliminary analysis, the following benefits have been identified. DUSCOBS improves repository waste package performance by three mechanisms. First, it reduces the radionuclide releases from SNF when water enters the waste package by creating a local uranium silicate saturated groundwater environment that suppresses (a) the dissolution and/or transformation of uranium dioxide fuel pellets and, hence, (b) the release of radionuclides incorporated into the SNF pellets. Second, the potential for long-term nuclear criticality is reduced by isotopic exchange of enriched uranium in SNF with the depleted uranium (DU) in the glass. Third, the backfill reduces radiation interactions between SNF and the local environment (package and local geology) and thus reduces generation of hydrogen, acids, and other chemicals that degrade the waste package system. Finally, DUSCOBS provides a potential method to dispose of significant quantities of excess DU from uranium enrichment plants at potential economic savings. DUSCOBS is a new concept. Consequently, the concept has not been optimized or demonstrated in laboratory experiments

Training manual for uranium mill workers on health protection from uranium

International Nuclear Information System (INIS)

McElroy, N.; Brodsky, A.

1986-01-01

This report provides information for uranium mill workers to help them understand the radiation safety aspects of working with uranium as it is processed from ore to yellowcake at the mills. The report is designed to supplement the radiation safety training provided by uranium mills to their workers. It is written in an easily readable style so that new employees with no previous experience working with uranium or radiation can obtain a basic understanding of the nature of radiation and the particular safety requirements of working with uranium. The report should be helpful to mill operators by providing training material to support their radiation safety training programs

Uranium isotopic analysis of depleted uranium in presence of other radioactive materials by using nondestructive gamma-ray measurements in coaxial and planar Ge detectors

International Nuclear Information System (INIS)

Yucel, H.; Yeltepe, E.; Dikmen, H.; Turhan, Sh.; Vural, M.

2006-01-01

Full text: The isotopic abundance of depleted uranium samples in the presence of other radioactive materials, especially actinide isotopes such as Th 232, Np 237-Pa 233 and Am 241 can be determined from two gamma-ray spectrometric methods. One is the absolute method which employs non-destructive gamma-ray spectrometry for energies below 1001 keV using a coaxial Ge detector calibrated with a set of standards. The other is the multi-group analysis (MGA) method using the low energy region (< 300 keV) with a planar Ge detector intrinsically calibrated with gamma and X-rays of uranium without use of standards. At present absolute method, less intense but cleaner gamma peaks at 163.33 keV (5.08 percent) and 205 keV(5.01 percent) of U 235 are preferred over more intense peaks at 143.76 keV(10.76 percent), possible interference with 143.25 keV(0.44 percent) of Np 237 and 185.705 keV(57.2 percent), possible interference with 186.21 keV(3.51 percent) of Ra 226. In the high energy region the 1001.03 keV(0.837 percent) peak of Pa 234 m is used for the isotopic abundance analysis because the more intense 63.3 keV peak of Th 234 daughter of U 238 parent has a fully multiplet(62.86 keV+63.29 keV) and include the interferences of the 62.70 keV(1.5 percent) peak of Pa 234, the 63.81 keV(0.263 percent) peak of Th 232 and the 63.90 keV(0.011 percent) peak of Np 237. Although the MGA method is quicker and more practical, the more laborious absolute gamma spectrometric method can give more accurate results for the isotopic determination of depleted uranium samples. The relative uranium abundances obtained with the second method (i,e., MGA) are in general inconsistent with the declared values for the uranium samples in the presence of the above mentioned actinides. The reason for these erroneous results is proposed to be the interference of the gamma and X-rays of uranium in the 80-130 keV region used in MGA with those emissions from other radioactive materials present

Effective utilization of geomorphology in uranium exploration: a success story from Meghalaya, northeast India

International Nuclear Information System (INIS)

Mamallan, R.; Awati, A.B.; Gupta, K.R.; Kak, S.N.

1995-01-01

The southern fringe of Meghalaya plateau displays a spectacular development of erosional landforms in the thick sedimentary cover of Cretaceous-Tertiary formations. Mahadek formation, the lower member of this sequence, comprises both continental and marginal marine sediments while all the overlying formations are mainly of marine origin. In the study area all the tertiary formations are eroded away, leaving exposed the continental part of the Mahadek formation, which comprises channel-filled and floodplain sediments. Geomorphologically, both these units express themselves as cuestas but significant textural differences were observed, enabling us to discriminate them in aerial photographs. It is known that the channel-filled sedimentary unit incorporates many favourable geological and geochemical characters to host uranium mineralization. The domiasiat uranium deposit occurs in this unit only. By virtue of its distinct geomorphology, three domains of channel-filled sediments were demarcated in aerial photographs. Follow-up radiometric field checks on one of these domains, near the confluence of Wah Blei and Kynshiang rivers, have led to the discovery of significant uranium occurrences, opening up promising new avenues for uranium exploration in Megahalya. (author). 4 refs., 2 figs

Synthesis of partial stabilized cement-gypsum as new dental retrograde filling material

Energy Technology Data Exchange (ETDEWEB)

Sadhasivam, S. [Institute of Biomedical Engineering, National Taiwan University, Taipei, Taiwan (China); Division of Medical Engineering Research, National Health Research Institute, Zhunan, Miaoli County, Taiwan (China); Chen, Jung-Chih [Institute of Biomedical Engineering, National Taiwan University, Taipei, Taiwan (China); Medical Device Innovation Center, National Cheng Kung University, Tainan,Taiwan (China); Savitha, S. [Institute of Biomedical Engineering, National Taiwan University, Taipei, Taiwan (China); Hsu, Ming-Xiang; Hsu, Chung-King [Institute of Materials Science and Engineering, National Taipei University of Technology, Taipei, Taiwan (China); Lin, Chun-Pin [School of Dentistry and Graduate Institute of Clinical Dentistry, College of Medicine, National Taiwan University and National Taiwan University Hospital, Taipei, Taiwan (China); Lin, Feng-Huei, E-mail: double@ntu.edu.tw [Institute of Biomedical Engineering, National Taiwan University, Taipei, Taiwan (China); Division of Medical Engineering Research, National Health Research Institute, Zhunan, Miaoli County, Taiwan (China)

2012-10-01

The study describes the sol-gel synthesis of a new dental retrograde filling material partial stabilized cement (PSC)-gypsum by adding different weight percentage of gypsum (25% PSC + 75% gypsum, 50% PSC + 50% gypsum and 75% PSC + 25% gypsum) to the PSC. The crystalline phase and hydration products of PSC-gypsum were characterized by X-ray diffraction (XRD) and scanning electron microscopy (SEM) analysis. The handling properties such as setting time, viscosity, tensile strength, porosity and pH, were also studied. The XRD and microstructure analysis demonstrated the formation of hydroxyapatite and removal of calcium dihydrate during its immersion in simulated body fluid (SBF) on day 10 for 75% PSC + 25% gypsum. The developed PSC-gypsum not only improved the setting time but also greatly reduced the viscosity, which is very essential for endodontic surgery. The cytotoxic and cell proliferation studies indicated that the synthesized material is highly biocompatible. The increased alkaline pH of the PSC-gypsum also had a remarkable antibacterial activity. - Highlights: Black-Right-Pointing-Pointer A new dental retrograde filling material PSC-gypsum was developed. Black-Right-Pointing-Pointer PSC-gypsum cement has shown excellent initial and final setting time as 15-35 min. Black-Right-Pointing-Pointer It not only improved the setting time but also retain the viscosity, 2 Pa{center_dot}s. Black-Right-Pointing-Pointer High alkaline pH of the cement had a remarkable antibacterial activity. Black-Right-Pointing-Pointer Cytotoxicity studies revealed that the synthesized material is highly biocompatible.

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.

Derivation of uranium residual radioactive material guidelines for the former Alba Craft Laboratory site, Oxford, Ohio

International Nuclear Information System (INIS)

Nimmagadda, M.; Faillace, E.; Yu, C.

1994-01-01

Residual radioactive material guidelines for uranium were derived for the former Alba Craft Laboratory site in Oxford, Ohio. This site has been identified for remedial action under the Formerly Utilized Sites Remedial Action Program (FUSRAP) of the US Department of Energy (DOE). Single nuclide and total uranium guidelines were derived on the basis of the requirement that the 50-year committed effective dose equivalent to a hypothetical individual who lives or works in the immediate vicinity of the former Alba Craft Laboratory site should not exceed a dose of 30 mrem/yr following remedial action for the current use and likely future use scenarios or a dose of 100 mrem/yr for less likely future use scenarios (Yu et al. 1993). The DOE residual radioactive material guideline computer code, RESRAD, which implements the methodology described in the DOE manual for implementing residual radioactive material guidelines, was used in this evaluation

Effect of calcium/silicon ratio on retention of uranium (VI) in portland cement materials

International Nuclear Information System (INIS)

Tan Hongbin; Li Yuxiang

2005-01-01

Calcium silicate hydrate (CSH) materials of varied calcium to silicon (Ca/Si) ratios were prepared by hydrothermal synthesis at 80 degree C, with calcium oxide and micro-silicon employed. These products were determined to be of gel phase by XRD. Leaching tests with 1% hydrochloric acid indicated that more Uranium (VI) was detained by CSH with lower Ca/Si ratios. Alkali-activated slag cement (with a lower Ca/Si ratio) was found to have a stronger retention capacity than Portland cement (with a higher Ca/Si ratio), at 25 degree C in 102-days leaching tests with simulated solidified forms containing Uranium (VI). The accumulative leaching fraction of Uranium (VI) for Alkali-activated slag cement solidified forms is 17.6% lower than that for Portland cement. The corresponding difference of diffusion coefficients is 40.6%. This could be correlated with the difference of Ca/Si ratios between cements of two kinds. (authors)

Dark localized structures in a cavity filled with a left-handed material

International Nuclear Information System (INIS)

Tlidi, Mustapha; Kockaert, Pascal; Gelens, Lendert

2011-01-01

We consider a nonlinear passive optical cavity filled with left-handed and right-handed materials and driven by a coherent injected beam. We assume that both left-handed and right-handed materials possess a Kerr focusing type of nonlinearity. We show that close to the zero-diffraction regime, high-order diffraction allows us to stabilize dark localized structures in this device. These structures consist of dips in the transverse profile of the intracavity field and do not exist without high-order diffraction. We analyze the snaking bifurcation diagram associated with these structures. Finally, a realistic estimation of the model parameters is provided.

Recovering uranium from phosphoric acid

International Nuclear Information System (INIS)

Anon.

1979-01-01

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

Modified Truncated Multiplicity Analysis to Improve Verification of Uranium Fuel Cycle Materials

International Nuclear Information System (INIS)

LaFleur, A.; Miller, K.; Swinhoe, M.; Belian, A.; Croft, S.

2015-01-01

Accurate verification of 235U enrichment and mass in UF6 storage cylinders and the UO2F2 holdup contained in the process equipment is needed to improve international safeguards and nuclear material accountancy at uranium enrichment plants. Small UF6 cylinders (1.5'' and 5'' diameter) are used to store the full range of enrichments from depleted to highly-enriched UF6. For independent verification of these materials, it is essential that the 235U mass and enrichment measurements do not rely on facility operator declarations. Furthermore, in order to be deployed by IAEA inspectors to detect undeclared activities (e.g., during complementary access), it is also imperative that the measurement technique is quick, portable, and sensitive to a broad range of 235U masses. Truncated multiplicity analysis is a technique that reduces the variance in the measured count rates by only considering moments 1, 2, and 3 of the multiplicity distribution. This is especially important for reducing the uncertainty in the measured doubles and triples rates in environments with a high cosmic ray background relative to the uranium signal strength. However, we believe that the existing truncated multiplicity analysis throws away too much useful data by truncating the distribution after the third moment. This paper describes a modified truncated multiplicity analysis method that determines the optimal moment to truncate the multiplicity distribution based on the measured data. Experimental measurements of small UF6 cylinders and UO2F2 working reference materials were performed at Los Alamos National Laboratory (LANL). The data were analyzed using traditional and modified truncated multiplicity analysis to determine the optimal moment to truncate the multiplicity distribution to minimize the uncertainty in the measured count rates. The results from this analysis directly support nuclear safeguards at enrichment plants and provide a more accurate verification method for UF6

Research on geochronology and uranium source of sandstone-hosted uranium ore-formation in major uranium-productive basins, Northern-China

International Nuclear Information System (INIS)

Xia Yuliang; Liu Hanbin; Lin Jinrong; Fan Guang; Hou Yanxian

2004-12-01

A method is developed for correcting uranium content in uranium ore samples by considering the U-Ra equilibrium coefficient, then a U-Pb isochron is drawn up. By performing the above correction ore-formation ages of sandstone-hosted uranium mineralization which may be more realistic have been obtained. The comparative research on U-Pb isotopic ages of detritic zircon in ore-hosting sandstone and zircon in intermediate-acid igneous rocks in corresponding provenance area indicates that the ore-hosting sandstone is originated from the erosion of intermediate-acid igneous rocks and the latters are the material basis for the formation of the uranium-rich sandstone beds. On the basis of the study on U-Pb isotopic system evolution of the provenance rocks and sandstones from ore-hosting series, it is verified that the uranium sources of the sandstone-hosted uranium deposit are: the intermediate-acid igneous rocks with high content of mobile uranium, and the sandstone bodies pre-concentrated uranium. (authors)

Feasibility studies to establish at the Kazakhstan Ulba metallurgical plant the manufacturing capability to produce low-enriched uranium certified reference materials

Energy Technology Data Exchange (ETDEWEB)

Kuzminski, Jozef [Los Alamos National Laboratory; Nesuhoff, J [NBL; Cratto, P [NBL; Pfennigwerth, G [Y12 NATIONAL SEC. COMPLEX; Mikhailenko, A [ULBA METALLURGICAL PLANT; Maliutina, I [ULBA METALLURGICAL PLANT; Nations, J [GREGG PROTECTION SERVICES

2009-01-01

One of the salient features of the transition plan that the United States Department of Energy/National Nuclear Security Administration (DOE/NNSA) is presently implementing in the Former Soviet Union countries is the availability of uranium certified reference materials for calibration of nondestructive assay (NDA) measurement equipment. To address this challenge, DOE/NNSA and U.S. national laboratories have focused their cooperative efforts on establishing a reliable source for manufacturing, certifying, and supplying of such standards. The Ulba Metallurgical Plant (UMP), Kazakhstan, which processes large quantities of low-enriched uranium to produce ceramic fuel pellets for nuclear-powered reactors, is well situated to become a key supplier of low-enriched uranium certified reference materials for the country and Central Asia region. We have recently completed Phase I of a feasibility study to establish at UMP capabilities of manufacturing these standards. In this paper we will discuss details of a proposed methodology for uranium down-blending, material selection and characterization, and a proposed methodology of measurement by destructive (DA) and non-destructive (NDA) analysis to form a database for material certification by the competent State authorities in the Republic of Kazakhstan. In addition, we will discuss the prospect for manufacturing of such standards at UMP.

Electrochemical investigation of uranium β-diketonates for all-uranium redox flow battery

International Nuclear Information System (INIS)

Yamamura, Tomoo; Shiokawa, Yoshinobu; Yamana, Hajimu; Moriyama, Hirotake

2002-01-01

The redox flow battery using uranium as the negative and the positive active materials in polar aprotic solvents was proposed. In order to establish the guiding principle for the uranium compounds as the active materials, the investigation of uranium β-diketonate complexes was conducted on (i) the solubility of active materials, (ii) the electrode reaction of U(VI) and U(IV) β-diketonate complexes and (iii) the estimation of the open circuit voltage of the battery. The solubilities of higher than 0.8 mol dm -3 of U(VI) complexes and higher than 0.4 mol dm -3 of a U(IV) complex were obtained in the solvents. The electrode reactions of U(pta) 4 , UO 2 (dpm) 2 , UO 2 (fod) 2 and UO 2 (pta) 2 were first studied and the redox potentials of uranium β-diketonates were thermodynamically discussed. The open circuit voltage is estimated more than 1 V by using Hacac or Hdpm. The larger open circuit voltage is expected when a ligand with the larger basicity is used

76 FR 71082 - Strata Energy, Inc., Ross Uranium Recovery Project; New Source Material License Application...

Science.gov (United States)

2011-11-16

... NUCLEAR REGULATORY COMMISSION [Docket No. 40-9091; NRC-2011-0148] Strata Energy, Inc., Ross Uranium Recovery Project; New Source Material License Application; Notice of Intent To Prepare a... intent to prepare a supplemental environmental impact statement. SUMMARY: Strata Energy, Inc. (Strata...

Obtention of uranium-molybdenum alloy ingots microstructure and phase characterization

Energy Technology Data Exchange (ETDEWEB)

Pedrosa, Tercio A.; Braga, Daniel M.; Paula, Joao Bosco de; Brina, Jose Giovanni M.; Ferraz, Wilmar B., E-mail: tap@cdtn.b, E-mail: bragadm@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} < 20 wt%) 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. Several uranium alloys that fill this requirement has been investigated since then. Among these alloys, 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 alloys is being performed at the Nuclear Technology Development Centre (CDTN) and also at the Institute of Energetic and Nuclear Research - IPEN. U-{sup 10}Mo ingots were melted in an induction furnace with protective argon atmosphere. The microstructure of the ingots were characterized through optical and scanning electronic microscopy in the as cast and heat treated conditions. Energy Dispersive Spectrometry and X-Ray Diffraction were used as characterization techniques for elemental analysis and phases determination. It was confirmed the presence of metastable gamma-phase in the as cast condition, surrounded by hypereutectoid alpha-phase (uranium-rich phase), as well as a pearlite-like constituent, composed by alternated lamellas of U{sub 2}Mo compound and alpha-phase, in the heat treated condition. (author)

Derivation of guidelines for uranium residual radioactive material in soil at the New Brunswick Site, Middlesex County, New Jersey

International Nuclear Information System (INIS)

Dunning, D.; Kamboj, S.; Nimmagadda, M.; Yu, C.

1996-02-01

Residual radioactive material guidelines for uranium in soil were derived for the New Brunswick Site, located in Middlesex County, New Jersey. This site has been designated for remedial action under the Formerly Utilized Sites Remedial Action Program of the US Department of Energy (DOE). Residual radioactive material guidelines for individual radionuclides of concern and total uranium were derived on the basis of the requirement that the 50-year committed effective dose equivalent to a hypothetical individual who lives or works in the immediate vicinity of the New Brunswick Site should not exceed a dose of 30 mrem/yr following remedial action for the current-use and likely future-use scenarios or a dose of 100 mrem/yr for less likely future-use scenarios. The DOE residual radioactive material guideline computer code, RESRAD, was used in this evaluation; RESRAD implements the methodology described in the DOE manual for establishing residual radioactive material guidelines. The guidelines derived in this report are intended to apply to the remediation of these remaining residual radioactive materials at the site. The primary radionuclides of concern in these remaining materials are expected to be radium-226 and, to a lesser extent, natural uranium and thorium. The DOE has established generic cleanup guidelines for radium and thorium in soil; however, cleanup guidelines for other radionuclides must be derived on a site-specific basis

Influence of dental filling material type on the concentration of interleukin 9 in the samples of gingival crevicular fluid

Directory of Open Access Journals (Sweden)

Stefanović Vladimir

2016-01-01

Full Text Available Background/Aim. Several cytokines and lymphokines (IL1β, ENA78, IL6, TNFα, IL8 and S100A8 are expressed during dental pulp inflammation. Analysis of gingival crevicu-lar fluid (GCF offers a non-invasive means of studying gen-eral host response in oral cavity. Although GCF levels of various mediators could reflect the state of inflammation both in dental pulp and gingiva adjacent to a tooth, GCF samples of those without significant gingivitis could be inter-preted as reflection of pulpal process. The aim of this study was to investigate IL9 GCF values in patients with dental car-ies and to assess possible influence of various dental fillings materials on local IL9 production. Methods. The study group included 90 patients, aged 18–70, with inclusion and exclusion criteria in the prospective clinical study. Of the 6 types of material used for the restoration of prepared cavities, 3 were intended for temporary and 3 for definitive restora-tion. According to dental fillings weight, all the participants were divided into 3 groups: those with fillings lighter than 0.50 g, those with 0.50–1.00 g, and those with fillings heavier than 1.00 g. Samples were taken from gingival sulcus using the filter paper technique. Clinical parameters were deter-mined by bleeding index, plaque index (Silness-Lou, 0–3, gingival index (0–3, and gingival sulcus depth. Cytokine con-centrations were assessed using commercially available cy-tomix. Results. According to the weight of dental fillings, there was a clear decreament trend of IL9 values meaning that dental defects greater than 1.00 g of dental filling were associated with lower GCF IL9 concentration. The IL9 val-ues correlated with the degree of gingival index and depth of gingival sulcus, being higher with more advanced gingivitis and more pronounced anatomical changes in the tooth edge. Different filling materials exerted various local IL9 responses. Zink polycarbonate cement and amalgam fillings induced

Uranium resource technology, Seminar 3, 1980

International Nuclear Information System (INIS)

Morse, J.G.

1980-01-01

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

Uranium deposits in Africa

International Nuclear Information System (INIS)

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

1979-01-01

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

Broadband giant-refractive-index material based on mesoscopic space-filling curves

Science.gov (United States)

Chang, Taeyong; Kim, Jong Uk; Kang, Seung Kyu; Kim, Hyowook; Kim, Do Kyung; Lee, Yong-Hee; Shin, Jonghwa

2016-08-01

The refractive index is the fundamental property of all optical materials and dictates Snell's law, propagation speed, wavelength, diffraction, energy density, absorption and emission of light in materials. Experimentally realized broadband refractive indices remain 1,800 resulting from a mesoscopic crystal with a dielectric constant greater than three million. This gigantic enhancement effect originates from the space-filling curve concept from mathematics. The principle is inherently very broad band, the enhancement being nearly constant from zero up to the frequency of interest. This broadband giant-refractive-index medium promises not only enhanced resolution in imaging and raised fundamental absorption limits in solar energy devices, but also compact, power-efficient components for optical communication and increased performance in many other applications.

Preparation of new nano magnetic material Fe3O4@g-C3N4 and good adsorption performance on uranium ion

Science.gov (United States)

Long, Wei; Liu, Huijun; Yan, Xueming; Fu, Li

2018-03-01

A new nano magnetic material Fe3O4@g-C3N4 was prepared by deposition reduction method, which performed good adsorption performance to uranium ion. Characterization results showed that the g-C3N4 particles were wrapped around the nano magnetic Fe3O4 particles, and the textural properties of this material was improved, so the adsorption performance to uranium ion was good. Adsorption experiments of this material demonstrated that the optimum pH value was 10, the optimum mass of adsorbent was 6.5 mg and the optimum adsorption time was 150 min in the initial concentration of 140 mg/L uranium ion solution system, and the maximum adsorption capacity was up to 352.1 mg/g and the maximum adsorption rate was more than 90%.

77 FR 33782 - License Amendment To Construct and Operate New In Situ Leach Uranium Recovery Facility; Uranium...

Science.gov (United States)

2012-06-07

... and Operate New In Situ Leach Uranium Recovery Facility; Uranium One Americas; Ludeman AGENCY: Nuclear... provided the first time that a document is referenced. The Ludeman facility In Situ Leach Uranium Recovery... request to amend Source Material License SUA-1341 to construct and operate a new in situ leach uranium...

Liquid uranium contaimment in refractories metals

International Nuclear Information System (INIS)

Duarte, J.L.; Padilha, A.F.

1982-01-01

Tests were performed on metalic materials for liquid uranium containment up to 2100 0 C. The materials Nb, Mo, Ta and W in the form of crucibles were tested at 2100 0 C for one hour in the presence of flowing argon. After testing, the crucibles were etched using HCl and analysed by optical metallography and electron proble microanalysis. The results are discussed in terms of Berthoud equation and indicated that the solubility limit of the crucible material in uranium at the temperature controlls the crucible dissolution by liquid uranium. The various phases formed, the mechanism of dissolution and the possible material for future use are presented and discussed. (Author) [pt

Gravimetric determination of carbon in uranium-plutonium carbide materials

International Nuclear Information System (INIS)

Kavanaugh, H.J.; Dahlby, J.W.; Lovell, A.P.

1979-12-01

A gravimetric method for determining carbon in uranium-plutonium carbide materials was developed to analyze six samples simultaneously. The samples are burned slowly in an oxygen atmosphere at approximately 900 0 C, and the gases generated are passed through Schuetze's oxidizing reagent (iodine pentoxide on silica gel) to assure quantitative oxidation of the CO to CO 2 . The CO 2 is collected on Ascarite and weighed. This method was tested using a tungsten carbide reference material (NBS-SRM-276) and a (U,Pu)C sample. For 42 analyses of the tungsten carbide, which has a certified carbon content of 6.09%, an average value of 6.09% was obtained with a standard deviation of 0.01 7 % or a relative standard deviation of 0.28%. For 17 analyses of the (U,Pu)C sample, an average carbon content of 4.97% was found with a standard deviation of 0.01 2 % or a relative standard deviation of 0.24%

Studies on geneses of Lianshanguan granites and Lianshanguan uranium ore deposit

International Nuclear Information System (INIS)

Zhang Jiafu; Xu Guoqing; Wang Wenguang

1994-02-01

Based on the field work, and through the studies of thin-sections, minerals fluid inclusions, isotope geology, rare-earth elements and U-content in rocks and minerals, it is suggested that Lianshanguan granites are of magmatization genesis with multistage. The genetic model of mineralization of Lianshanguan uranium ore deposit is the magmatization-hydrothermal-filled uranium type. The role of mineralization of uranium ore deposit in that area is discussed. Furthermore, the direction of prospecting and following prospecting criteria for similar deposits in this area are also given

Development and inculcation of new methods in the field of exploring, extraction and reprocessing of uranium raw materials

International Nuclear Information System (INIS)

Smirnov, Yu.V.; Efimova, Z.I.; Skorovarov, D.I.; Ivanov, G.F.

1979-01-01

The situation is briefly overviewed in the field of exploration, prospecting, extraction and processing of uranium raw material and industrial introduction of the recently developed methods. The method of underground leaching, including that from deep-seated deposits, is gaining wide acceptance. Of high importance is the industrial introduction of such promising processes as bacterial leaching, continuous column-sorption with a pseudo-luquefied sorbent layer, direct production of UF 6 in an ore-processing factory. Active works are under way now in the field of multi-purpose utilization of uranium ore. New methods are industrially introduced for the extraction of associated uranium from phosphoric acid solutions, copper ore, sea water

Radiation risk assessment of reprocessed uranium

International Nuclear Information System (INIS)

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

1999-01-01

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

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

International Nuclear Information System (INIS)

Baran, V.; Moltasova, J.

1982-01-01

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

Evolution of ore-bearing material sources of endogenous uranium deposits

International Nuclear Information System (INIS)

Kazansk, V.I.; Laverov, N.P.; Tugarinov, A.I.

1976-01-01

Considered are the regularities of changes in types and conditions of uranium deposit formation in connection with the general development of the earth crust tectonic structures. Out of pre-Kembrian uranium deposits considered are Vitwatersrand conglomerates, hydrothermal deposits in pre-Kembrian iron quartzites in the areas of regional fractures in exocontacts of big multiphase granitoid massifs of Proterozoic age and in the fundament folded structures. The hydrothermal-metamorphogen theory is supported of the origin of uranium-bearing sodium metasomatite of Proterozoic, including uranium deposits in the area of the Atabaska lake. Four genetic classes of Palaeozoic deposits are considered. Four periods are singled out in the development of Palaeozoic uranium provinces. Most of the Palaeozoic deposits are shown to be of polygenous origin. Mesozoic deposits are also polygenous, but the combination of ore substance sources in them is more complex

Development of Novel Sorbents for Uranium Extraction from Seawater

Energy Technology Data Exchange (ETDEWEB)

Lin, Wenbin [Univ. of Chicago, IL (United States); Taylor-Pashow, Kathryn [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2014-01-08

As the uranium resource in terrestrial ores is limited, it is difficult to ensure a long-term sustainable nuclear energy technology. The oceans contain approximately 4.5 billion tons of uranium, which is one thousand times the amount of uranium in terrestrial ores. Development of technologies to recover the uranium from seawater would greatly improve the uranium resource availability, sustaining the fuel supply for nuclear energy. Several methods have been previously evaluated including solvent extraction, ion exchange, flotation, biomass collection, and adsorption; however, none have been found to be suitable for reasons such as cost effectiveness, long term stability, and selectivity. Recent research has focused on the amidoxime functional group as a promising candidate for uranium sorption. Polymer beads and fibers have been functionalized with amidoxime functional groups, and uranium adsorption capacities as high as 1.5 g U/kg adsorbent have recently been reported with these types of materials. As uranium concentration in seawater is only ~3 ppb, great improvements to uranium collection systems must be made in order to make uranium extraction from seawater economically feasible. This proposed research intends to develop transformative technologies for economic uranium extraction from seawater. The Lin group will design advanced porous supports by taking advantage of recent breakthroughs in nanoscience and nanotechnology and incorporate high densities of well-designed chelators into such nanoporous supports to allow selective and efficient binding of uranyl ions from seawater. Several classes of nanoporous materials, including mesoporous silica nanoparticles (MSNs), mesoporous carbon nanoparticles (MCNs), meta-organic frameworks (MOFs), and covalent-organic frameworks (COFs), will be synthesized. Selective uranium-binding liagnds such as amidoxime will be incorporated into the nanoporous materials to afford a new generation of sorbent materials that will be

Development of Novel Sorbents for Uranium Extraction from Seawater

International Nuclear Information System (INIS)

Lin, Wenbin; Taylor-Pashow, Kathryn

2014-01-01

As the uranium resource in terrestrial ores is limited, it is difficult to ensure a long-term sustainable nuclear energy technology. The oceans contain approximately 4.5 billion tons of uranium, which is one thousand times the amount of uranium in terrestrial ores. Development of technologies to recover the uranium from seawater would greatly improve the uranium resource availability, sustaining the fuel supply for nuclear energy. Several methods have been previously evaluated including solvent extraction, ion exchange, flotation, biomass collection, and adsorption; however, none have been found to be suitable for reasons such as cost effectiveness, long term stability, and selectivity. Recent research has focused on the amidoxime functional group as a promising candidate for uranium sorption. Polymer beads and fibers have been functionalized with amidoxime functional groups, and uranium adsorption capacities as high as 1.5 g U/kg adsorbent have recently been reported with these types of materials. As uranium concentration in seawater is only ~3 ppb, great improvements to uranium collection systems must be made in order to make uranium extraction from seawater economically feasible. This proposed research intends to develop transformative technologies for economic uranium extraction from seawater. The Lin group will design advanced porous supports by taking advantage of recent breakthroughs in nanoscience and nanotechnology and incorporate high densities of well-designed chelators into such nanoporous supports to allow selective and efficient binding of uranyl ions from seawater. Several classes of nanoporous materials, including mesoporous silica nanoparticles (MSNs), mesoporous carbon nanoparticles (MCNs), meta-organic frameworks (MOFs), and covalent-organic frameworks (COFs), will be synthesized. Selective uranium-binding liagnds such as amidoxime will be incorporated into the nanoporous materials to afford a new generation of sorbent materials that will be

Processing Uranium-Bearing Materials Containing Coal and Loam

Energy Technology Data Exchange (ETDEWEB)

Civin, V; Prochazka, J [Research and Development Laboratory No. 3 of the Uranium Industry, Prague, Czechoslovakia (Czech Republic)

1967-06-15

Among the ores which are classified as low-grade in the CSSR are mixtures of coal and bentonitic loam of tertiary origin, containing approximately 0.1% U and with a moisture content at times well above 20-30%. The uranium is held mainly by the carbonaceous component. Conventional processing of these materials presents various difficulties which are not easily overcome. During leaching the pulp thickens and frequently becomes pasty, due to the presence of montmorillonites. Further complications arise from the high sorption capacity of the materials (again primarily due to montmorillonites) and poor sedimentation of the viscous pulps. In addition, the materials are highly refractory to the leaching agents. The paper presents experience gained in solving the problems of processing these ores. The following basic routes were explored: (1) separation of the carbonaceous and loamy components: The organic component appears to be the main activity carrier. Processing the concentrated material upon separation of the inactive or less active loam may not only remove the thixotropic behaviour but also substantially reduce the cost of the ore treatment; (2) 'liquifying' the pulps or preventing the thickening of the pulp by addition of suitable agents; (3) joint acid or carbonate processing of the materials in question with current ore types; (4) removal or suppression of thixotropic behaviour by thermal pretreatment of the material; and (5) application of the 'acid cure' method. The first method appears to be the most effective, but it presents considerable difficulties due to the extreme dispersion of the carbonaceous phase and further research is being carried out. Methods 2 and 3 proved to be unacceptable. Method 4, which includes roasting at 300-400{sup o}C, is now being operated on an industrial scale. The final method has also shown definite advantages for particular deposits of high montmorillonite content material. (author)

Uranium determination in phosphatized materials by drop electrode polarography

International Nuclear Information System (INIS)

Sequeira, F.M.C. de; Abrao, A.

1987-01-01

An uranium determination procedure in phosphate rocks and crude phosphoric acid is outlined polarography is used. Uranium is previously separedted by extraction with tri-n-byte phosphate 10%-petroleum ether using aluminium nitrate as salting out agent. (M.L.J.) [pt

Vapor corrosion of aluminum cladding alloys and aluminum-uranium fuel materials in storage environments

International Nuclear Information System (INIS)

Lam, P.; Sindelar, R.L.; Peacock, H.B. Jr.

1997-04-01

An experimental investigation of the effects of vapor environments on the corrosion of aluminum spent nuclear fuel (A1 SNF) has been performed. Aluminum cladding alloys and aluminum-uranium fuel alloys have been exposed to environments of air/water vapor/ionizing radiation and characterized for applications to degradation mode analysis for interim dry and repository storage systems. Models have been developed to allow predictions of the corrosion response under conditions of unlimited corrodant species. Threshold levels of water vapor under which corrosion does not occur have been identified through tests under conditions of limited corrodant species. Coupons of aluminum 1100, 5052, and 6061, the US equivalent of cladding alloys used to manufacture foreign research reactor fuels, and several aluminum-uranium alloys (aluminum-10, 18, and 33 wt% uranium) were exposed to various controlled vapor environments in air within the following ranges of conditions: Temperature -- 80 to 200 C; Relative Humidity -- 0 to 100% using atmospheric condensate water and using added nitric acid to simulate radiolysis effects; and Gamma Radiation -- none and 1.8 x 10 6 R/hr. The results of this work are part of the body of information needed for understanding the degradation of the A1 SNF waste form in a direct disposal system in the federal repository. It will provide the basis for data input to the ongoing performance assessment and criticality safety analyses. Additional testing of uranium-aluminum fuel materials at uranium contents typical of high enriched and low enriched fuels is being initiated to provide the data needed for the development of empirical models

Relative probabilities of the uranium isotopes for thorium x-ray emission and fluorescence of uranium x-rays

International Nuclear Information System (INIS)

Parker, J.L.

1991-01-01

Both thorium x-rays from decaying uranium isotopes and self-fluoresced uranium x-rays are prominent in high-resolution gamma-ray spectra of uranium-bearing materials. Useful application of the information carried by those x-rays has been curtailed because the probabilities of the uranium isotopes for thorium x-ray emission and for uranium x-ray fluorescence have not been known. By analyzing enrichment-meter geometry spectra from uranium oxide standards whose enrichments ranged from 0.7% to 91%, relative values, primarily, have been obtained for the probabilities of both processes. Thorium x-ray emission is very heavily dominated by 235 U. In all ordinarily occurring uranium isotopic distributions, thorium x-rays may be used as a valid 235 U signature. The probability for a thorium K α1 x-ray to be emitted in the decay of a 235 U atom is 0.048 ±0.002. In infinitely thick uranium oxide materials, the relative ratios of effectiveness for self-fluorescence, on a per unit mass basis, are approximately 234 U : 235 U : 236 U : 238 U = 1.13 : 1.00 : 0.52 : 0.028. on a per decay basis, the approximate ratios are 0.00039 : 1.00 : 0.017 : 0.18. These results imply that, contrary to what has often been stated, gamma rays are far more important than alpha particles in the self-fluorescence of uranium. Because of the importance of gamma-ray self-fluorescence, the uranium x-ray yield will be somewhat influenced by the size, shape, and composition of the materials. 4 refs., 1 fig

Uranium ores of Kazakhstan as the most technologic source of a fissionable material

International Nuclear Information System (INIS)

Berikbolov, B.R.

1999-01-01

Kazakhstan as is known has unique deposits of uranium. Its resources composed a third part of the world resources. The most important part of resources having a practical value, is related with depression in southern regions of the Republic. By now more than 15 deposits are discovered and partially explored. These deposits from three uranium provinces - Shu-Sarysu, Syr-Darya and Ili. The ores occur in friable water-bearing sandy horizons of Cretaceous and Paleogene age between waterproof agrillaceous sediments at depth from 100 up to 600 m. Ore bodies thickness changes from 5 to 10 m at uranian average-grade 0.03-0.1 %. Width of band shaped ore bodies changes from tens meters to the one kilometers and extent changes from one kilometer up to many tens kilometers. The important feature of deposits is their suitability for development by progressive in situ leaching (ISL) method. It was demonstrated, that uranium ores are comprehensive and, that is important, a lot of commercially important elements, containing in ores, gives in to extraction at development by the ISL method. The preliminary calculation of expenditures for the extraction of useful byproducts from ordinary sulphate solution have demonstrated rather high profitableness for rhenium, scandium, selenium, rare earth even at the very low contents in solution. It was pointed out, that whole technological chain applied now at industrial scale is oriented to mono-metallic uranium ores, therefore present technology of leaching and recovery of industrial solution does not allow ti extract all valuable components containing in ores. The development of new improved technological chain. beginning with a composition of leaching out reagent and up to applying of miscellaneous sorbing materials, can create new mineral-raw base of rare and dissipated elements and to lower considerably the price of uranium mining from sandstone deposits

Sealing ability of MTA, CPM, and MBPc as root-end filling materials: a bacterial leakage study.

Science.gov (United States)

Medeiros, Paulo Leal; Bernardineli, Norberti; Cavenago, Bruno Cavalini; Torres, Sérgio Aparecido; Duarte, Marco Antonio Hungaro; Bramante, Clovis Monteiro; Marciano, Marina Angélica

2016-04-01

Objectives To evaluate the sealing ability of three root-end filling materials (white MTA, CPM, and MBPc) using an Enterococcus faecalis leakage model. Material and Methods Seventy single-root extracted human teeth were instrumented and root-ends were resected to prepare 3 mm depth cavities. Root-end preparations were filled with white MTA, CPM, and MBPc cements. Enterococcus faecalis was coronally introduced and the apical portion was immersed in BHI culture medium with phenol red indicator. The bacterial leakage was monitored every 24 h for 4 weeks. The statistical analysis was performed using the Wilcoxon-Gehan test (pCPM and the other groups. Conclusions The epoxy resin-based cement MBPc had lower bacterial leakage compared with the calcium silicate-based cements MTA and CPM.

Color of bulk-fill composite resin restorative materials.

Science.gov (United States)

Barutcigil, Çağatay; Barutcigil, Kubilay; Özarslan, Mehmet Mustafa; Dündar, Ayşe; Yilmaz, Burak

2018-03-01

To evaluate the color stability of novel bulk-fill composite resins. Color measurements of a nanohybrid composite resin (Z550) and 3 bulk-fill composite resins (BLK, AFX, XTF; n = 45) were performed before polymerization. After polymerization, color measurements were repeated and specimens were immersed in distilled water or red wine, or coffee. Color change [CIEDE2000 (ΔE 00 )] was calculated after 24 h, 1 and 3 weeks. Data were analyzed with Kruskal-Wallis, Mann-Whitney U and Wilcoxon tests (α = 0.05). Color changes observed after polymerization were significant for all groups. Color changes observed in distilled water for Z550 and AFX were significant. Color changes after stored in red wine and coffee were significant for all groups. Bulk-fill composite resin color change increased over time for all groups in red wine and coffee (P composite resin and bulk-fill composite resins. AFX had the highest color change in distilled water. The color of tested bulk-fill composite resins significantly changed after immersion in beverages and over time. Color change observed with the nanohybrid composite resin after 1 week was stable. Clinicians should keep in mind that tested composite resins may change color when exposed to water and significantly change color immediately after they are polymerized. In addition, the color change continues over time should the patient is a coffee and/or red wine consumer. © 2017 Wiley Periodicals, Inc.

33 CFR 336.1 - Discharges of dredged or fill material into waters of the U.S.

Science.gov (United States)

2010-07-01

... 33 Navigation and Navigable Waters 3 2010-07-01 2010-07-01 false Discharges of dredged or fill material into waters of the U.S. 336.1 Section 336.1 Navigation and Navigable Waters CORPS OF ENGINEERS... ENGINEERS DREDGING PROJECTS INVOLVING THE DISCHARGE OF DREDGED MATERIAL INTO WATERS OF THE U.S. AND OCEAN...

Adjustable Polyurethane Foam as Filling Material for a Novel Spondyloplasty: Biomechanics and Biocompatibility.

Science.gov (United States)

Jiang, Hongzhen; Sitoci-Ficici, Kerim Hakan; Reinshagen, Clemens; Molcanyi, Marek; Zivcak, Jozef; Hudak, Radovan; Laube, Thorsten; Schnabelrauch, Matthias; Weisser, Jürgen; Schäfer, Ute; Pinzer, Thomas; Schackert, Gabriele; Zhang, Xifeng; Wähler, Mario; Brautferger, Uta; Rieger, Bernhard

2018-04-01

To investigate the biomechanics and biocompatibility of polyurethane (PU) foam with adjustable stiffness as a filling material for a novel spondyloplasty that is designed to reduce the risk of postoperative adjacent level fractures. Sixty individual porcine lumbar vertebrae were randomly split into 4 groups: A, B, C, and D. Group A served as unmodified vertebral body controls. Groups B, C, and D consisted of hollowed vertebral bodies. Vertebrae of groups C and D were filled with adjustable PU foams of different stiffness. The compressive strength and stiffness of vertebrae from groups A-D were recorded and analyzed. 3T3 mouse fibroblasts were cultured with preformed PU foams for 4 days to test biocompatibility. The strength and stiffness of the hollowed groups were lower than in group A. However, the differences were not statistically significant between group A and group C (P > 0.05), and were obviously different between group A and group B or group D (P < 0.01 and <0.05, respectively). Moreover, the strength and stiffness after filling foams in group C or group D were significantly greater than in group B (P < 0.01 and <0.05, respectively). Live/dead staining of 3T3 cells confirmed the biocompatibility of the PU foam. The new PU foam shows adaptability regarding its stiffness and excellent cytocompatibility in vitro. The results support the clinical translation of the new PU foams as augmentation material in the development of a novel spondyloplasty. Copyright © 2018 Elsevier Inc. All rights reserved.

Microstructure and wettability of root canal dentine and root canal filling materials after different chemical irrigation

International Nuclear Information System (INIS)

Antonijevic, Djordje; Milovanovic, Petar; Brajkovic, Denis; Ilic, Dragan; Hahn, Michael; Amling, Michael; Rakocevic, Zlatko; Djuric, Marija; Busse, Björn

2015-01-01

Graphical abstract: - Highlights: • Different irrigation solutions and disinfectants were used for treatment of root canal dentine and gutta-percha points. • Materials surface characteristics were assessed using quantitative backscattered electron imaging, reference point indentation, and contact angle analyzer. • The most significant differences in mineralization, indentation, and adhesive outcomes were observed after ethylenediaminetetraacetic acid treatment. • Irrigation solutions confer to superior sealing ability of endodontic filling materials. • Micromechanical characteristics of dentine after irrigation are considerable reduced. - Abstract: The objective of this study was to determine the effects of various irrigation solutions on root canal dentine and gutta-percha surface properties. In addition, the effects of disinfectant chemicals on the wettability and surface morphological properties of the filling materials were evaluated. Ethylenediaminetetraacetic acid (EDTA), citric acid, and ozone were employed as irrigation solutions for dentine and gutta-percha treatment. Thereafter, the samples’ microstructure, degree of mineralization, and mechanical properties were assessed by means of quantitative backscattered electron imaging (qBEI) and reference point indentation (RPI). A contact angle analyzer was used to measure adhesion on the tested materials. Here, EDTA had the most significant affect on both the mechanical properties and the adhesive behavior of dentine. Citric acid did not affect dentine wettability, whereas the indentation properties and the mineralization were reduced. Similar effects were observed when ozone was used. The dentinal tubules were significantly widened in citric acid compared to the ozone group. EDTA causes considerable micromechanical surface alteration of dentine and gutta-percha, but represents the best option in clinical cases where a high adhesiveness of the filling materials is desired.

Microstructure and wettability of root canal dentine and root canal filling materials after different chemical irrigation

Energy Technology Data Exchange (ETDEWEB)

Antonijevic, Djordje; Milovanovic, Petar [Laboratory for Anthropology, Institute for Anatomy, Faculty of Medicine, University of Belgrade, Belgrade (Serbia); Brajkovic, Denis [Department of Dentistry, Faculty of Medical Sciences, University of Kragujevac, Kragujevac (Serbia); Ilic, Dragan [Department of Restorative Medicine, School of Dental Medicine, University of Belgrade, Belgrade (Serbia); Hahn, Michael; Amling, Michael [Department of Osteology and Biomechanics (IOBM), University Medical Center Hamburg-Eppendorf, Lottestr. 55A, 22529 Hamburg (Germany); Rakocevic, Zlatko [Laboratory for Atomic Physics, Institute for Nuclear Science “Vinca”, University of Belgrade, Belgrade (Serbia); Djuric, Marija [Laboratory for Anthropology, Institute for Anatomy, Faculty of Medicine, University of Belgrade, Belgrade (Serbia); Busse, Björn, E-mail: b.busse@uke.uni-hamburg.de [Department of Osteology and Biomechanics (IOBM), University Medical Center Hamburg-Eppendorf, Lottestr. 55A, 22529 Hamburg (Germany)

2015-11-15

Graphical abstract: - Highlights: • Different irrigation solutions and disinfectants were used for treatment of root canal dentine and gutta-percha points. • Materials surface characteristics were assessed using quantitative backscattered electron imaging, reference point indentation, and contact angle analyzer. • The most significant differences in mineralization, indentation, and adhesive outcomes were observed after ethylenediaminetetraacetic acid treatment. • Irrigation solutions confer to superior sealing ability of endodontic filling materials. • Micromechanical characteristics of dentine after irrigation are considerable reduced. - Abstract: The objective of this study was to determine the effects of various irrigation solutions on root canal dentine and gutta-percha surface properties. In addition, the effects of disinfectant chemicals on the wettability and surface morphological properties of the filling materials were evaluated. Ethylenediaminetetraacetic acid (EDTA), citric acid, and ozone were employed as irrigation solutions for dentine and gutta-percha treatment. Thereafter, the samples’ microstructure, degree of mineralization, and mechanical properties were assessed by means of quantitative backscattered electron imaging (qBEI) and reference point indentation (RPI). A contact angle analyzer was used to measure adhesion on the tested materials. Here, EDTA had the most significant affect on both the mechanical properties and the adhesive behavior of dentine. Citric acid did not affect dentine wettability, whereas the indentation properties and the mineralization were reduced. Similar effects were observed when ozone was used. The dentinal tubules were significantly widened in citric acid compared to the ozone group. EDTA causes considerable micromechanical surface alteration of dentine and gutta-percha, but represents the best option in clinical cases where a high adhesiveness of the filling materials is desired.

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)

Uranium industry annual, 1988

International Nuclear Information System (INIS)

1989-01-01

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

Preliminary Mass Spectrometric Analysis of Uranium on Environmental Swipe Materials

International Nuclear Information System (INIS)

Cheong, Chang-Sik; Jeong, Youn-Joong; Ryu, Jong-Sik; Shin, Hyung-Seon; Cha, Hyun-Ju; Ahn, Gil-Hoon; Park, Il-Jin; Min, Gyung-Sik

2006-01-01

It is well-known that uranium and plutonium isotopic compositions of safeguards samples are very useful to investigate the history of nuclear activities. To strengthen the capabilities of environmental sampling analysis in the ROK through MOST/DOE collaboration, round robin test for uranium and plutonium was designed in 2003. As the first round robin test, a set of dried uranium-containing solutions (∼35ng and (∼300ng) was distributed to the participating laboratories in November of 2003, with results reported in April of 2004. The KBSI (Korea Basic Science Institute) and ORNL (Oak Ridge National Laboratory) are currently in the process of analyzing uranium on cotton swipes for the second round robin test. As a preliminary test for the second round, KBSI intends to analyze home-made swipe samples into which international uranium standards are added. Here we describe technical steps of sample preparation and mass spectrometry at KBSI, and report some results of the preliminary test

Uranium toxicology

International Nuclear Information System (INIS)

Ferreyra, Mariana D.; Suarez Mendez, Sebastian

1997-01-01

In this paper are presented the methods and procedures optimized by the Nuclear Regulatory Authority (ARN) for the determination of: natural uranium mass, activity of enriched uranium in samples of: urine, mucus, filters, filter heads, rinsing waters and Pu in urine, adopted and in some cases adapted, by the Environmental Monitoring and Internal Dosimetry Laboratory. The analyzed material corresponded to biological and environmental samples belonging to the staff professionally exposed that work in plants of the nuclear fuel cycle. For a better comprehension of the activities of this laboratory, it is included a brief description of the uranium radiochemical toxicity and the limits internationally fixed to preserve the workers health

IRMM-1000a and IRMM-1000b. Uranium reference materials certified for the production date based on the 230Th/234U radiochronometer. Part II. Certification

International Nuclear Information System (INIS)

Venchiarutti, C.; Richter, S.; Jakopic, R.; Aregbe, Y.; Varga, Z.; Nicholl, A.; Krajko, J.; Mayer, K.

2016-01-01

The IRMM-1000a and IRMM-1000b uranium reference materials, of 20 and 50 mg uranium, respectively, were produced by the European Commission Joint Research Centre's Institute for Reference Materials and Measurements (EC-JRC-IRMM) in collaboration with the Institute for Transuranium Elements (EC-JRC-ITU). They are novel uranium reference materials certified for the production date based on the 230 Th/ 234 U radiochronometer, i.e. the date of the last chemical separation of these two radionuclides. The certified reference value and its uncertainty, homogeneity and stability of the material were established in accordance with the ISO Guide 34:2009 and the 'Guide to the Expression of Uncertainty in Measurement'. (author)

Politics of Uranium

International Nuclear Information System (INIS)

Anon.

1982-01-01

Uranium is the most political of all the elements, the material for the production of both the large amounts of electricity and the most destructive weapons in the world. The problems that its dual potential creates are only now beginning to become evident. Author Norman Moss looks at this situation and sheds light on many of the questions that emerge. The nuclear issue always comes back to how much uranium there is, what can be done with it, and which countries have it. Starting with a concise history of uranium and explaining its technology in terms the nonspecialist can understand, The Politics of Uranium considers the political issues that technical arguments obscure. It tells the little-known story of the international uranium cartel, explains the entanglements of governments with the uranium trade, and describes the consequences of wrong decisions and blunders-especially the problems of nuclear waste. It also examines the intellectual and emotional roots of the anti-nuclear movement

Uranium exploration in Australia

International Nuclear Information System (INIS)

Battey, G.C.; Hawkins, B.W.

1977-01-01

As a result of exploration which recommenced in 1966 Australia's uranium reserves increased from 6,200 tonnes in 1967 to 227,000 tonnes uranium by June 1976. Most discoveries in the early 1950's were made by prospectors. The increase in reserves during the past decade is the result of exploration by companies utilising improved technology in areas selected as geologically favourable. These reserves were established at relatively low cost. In the Alligator Rivers Uranium Province the ''vein'' type deposits at Jabiluka, Ranger, Koongarra and Nabarlek contain 17% of the world's reserves. Most of these discoveries resulted from the investigation of airborne radiometric anomalies but cover over the prospective host rocks will necessitate the future use of costlier and more indirect exploration techniques. There was exploration for sandstone type uranium deposits in most of Australia's sedimentary basins. The greatest success was achieved in the Lake Frome Basin in South Australia. Other deposits were found in the Ngalia and Amadeus Basins in Central Australia and in the Westmoreland area, N.W. Queensland. A major uranium deposit was found in an unusual environment at Yeelirrie, Western Australia where carnotite occurs in a caliche and clay host which fills a shallow, ancient drainage channel. Although caliche occurrences are relatively widespread on the Precambrian shield no other economic deposit has been found. Recent discoveries in the Georgetown area of Queensland indicate the presence of another uranium province but it is too early to assess its potential. The ore occurs in clastic sediments at the base of a volcanic sequence overlying a Precambrian basement. Several companies which have established large uranium reserves have a number of additional attractive prospects. Exploration activity in Australia in 1975 was at a lower level than in previous years, but the potential for discovering further deposits is considered to be high

Efficacy of ultrasonic activation of NaOCl and orange oil in removing filling material from mesial canals of mandibular molars with and without isthmus

Directory of Open Access Journals (Sweden)

Mirela Sangoi BARRETO

2015-01-01

Full Text Available Objectives The aim of this study was to evaluate the volume of remaining filling material after passive ultrasonic irrigation (PUI of sodium hypochlorite (NaOCl and orange oil in mesial canals of mandibular molars, with and without isthmus. Material and Methods Thirty mesial roots of mandibular molars were divided according to the presence or absence of isthmus. Canals were prepared and filled (Micro-CT #1. Filling was removed using rotary instruments, and specimens were sub-divided into three groups according to the irrigation procedures: Conventional – conventional irrigation with NaOCl, PUI/NaOCl – PUI of NaOCl (three activations, 20 seconds each, and PUI/orange oil – PUI of orange oil (Micro-CT#2. Specimens were enlarged using the X2 and X3 ProTaper Next instruments and submitted to the same irrigation protocols (Micro-CT #3. Results No differences were found between the experimental groups in each stage of assessment (P>0.05. The volume of residual filling material was similar to that in Micro-CT #2 and Micro-CT #3, but lower than that observed in Micro-CT #1 (P<0.05. When groups were pooled according to the presence or absence of an isthmus, volume of residual filling material was higher in specimens presenting isthmus (P<0.05. Conclusions PUI of NaOCl or orange oil did not improve filling removal. Isthmus consists in an anatomical obstacle that impairs the removal of filling material.

Rock fill in a KBS-3 repository. Rock material for filling of shafts and ramps in a KBS-3V repository in the closure phase

International Nuclear Information System (INIS)

Pusch, Roland

2008-09-01

The content of large blocks in blasted rock makes it impossible to fill and compact the material effectively unless those larger than about 500 mm are removed. Tunnel Boring Machine (TBM) muck gives flat chips, that are usually not longer than a couple of decimeters, and serves better as backfill. The granulometrical composition of both types can be more suitable for effective compaction by crushing, which is hence a preferable process. Use of unsorted, unprocessed blasted rock can only be accepted if the density and physical properties, like self-compaction, are not important. Crushing of blasted rock and TBM muck for backfilling can be made in one or two steps depending on the required gradation. Placement of rock fill is best made by use of tractors with blades that push the material forwards over already placed and compacted material. The dry density of well graded rock fill effectively compacted by very heavy vibratory rollers can be as high as 2,400 kg/m3. For road compaction by ordinary vibratory rollers common dry density values are in the interval 2,050 to 2,200 kg m 3 . Blasted rock dumped and moved on site by tractors can get an average dry density of 1,600-1,800 kg/m3 without compaction. Crushed, blasted rock and TBM muck placed by tractors in horizontal layers and compacted by 5-10 t vibrating rollers in the lower part of the rooms, and moved by tractors to form inclined layers compacted by vibrating plates in the upper part, would get a dry density of 1,900-2,000 kg/m 3 . Flushing water over the rock fill in conjunction with the compaction work gives more effective densification than dry compaction. Based on recorded settlement of Norwegian rock fill dams constructed with water flushing it is estimated that the self-compaction of a 5 m high backfill of crushed rock or TBM muck causes a settlement of the top of the backfill of about 8 mm while a 200 m high shaft fill would undergo compression by more than half a meter. Repeated, strong earthquakes may

Rock fill in a KBS-3 repository. Rock material for filling of shafts and ramps in a KBS-3V repository in the closure phase

Energy Technology Data Exchange (ETDEWEB)

Pusch, Roland (Geodevelopment International AB/SWECO AB, Lund (Sweden))

2008-09-15

The content of large blocks in blasted rock makes it impossible to fill and compact the material effectively unless those larger than about 500 mm are removed. Tunnel Boring Machine (TBM) muck gives flat chips, that are usually not longer than a couple of decimeters, and serves better as backfill. The granulometrical composition of both types can be more suitable for effective compaction by crushing, which is hence a preferable process. Use of unsorted, unprocessed blasted rock can only be accepted if the density and physical properties, like self-compaction, are not important. Crushing of blasted rock and TBM muck for backfilling can be made in one or two steps depending on the required gradation. Placement of rock fill is best made by use of tractors with blades that push the material forwards over already placed and compacted material. The dry density of well graded rock fill effectively compacted by very heavy vibratory rollers can be as high as 2,400 kg/m3. For road compaction by ordinary vibratory rollers common dry density values are in the interval 2,050 to 2,200 kg m3. Blasted rock dumped and moved on site by tractors can get an average dry density of 1,600-1,800 kg/m3 without compaction. Crushed, blasted rock and TBM muck placed by tractors in horizontal layers and compacted by 5-10 t vibrating rollers in the lower part of the rooms, and moved by tractors to form inclined layers compacted by vibrating plates in the upper part, would get a dry density of 1,900-2,000 kg/m3. Flushing water over the rock fill in conjunction with the compaction work gives more effective densification than dry compaction. Based on recorded settlement of Norwegian rock fill dams constructed with water flushing it is estimated that the self-compaction of a 5 m high backfill of crushed rock or TBM muck causes a settlement of the top of the backfill of about 8 mm while a 200 m high shaft fill would undergo compression by more than half a meter. Repeated, strong earthquakes may

Analysis of nuclear reaction products and materials; Preliminary treatment of uranium analysis

International Nuclear Information System (INIS)

Soedyartomo.

1976-01-01

Pre-treatment of samples is necessary to be done in order to achieve the efficient steps and accurate results of uranium analysis. The pre-treatment is particularly affected by the type of sample, the uranium concentration predicated in the sample, and the uranium analytical method which will be applied. A brief discussion about the pre-treatment of uranium analysis in the uranium ore processing and the reprocessing of spent fuel is given. (author)

Uranium thiolate complexes

International Nuclear Information System (INIS)

Leverd, Pascal C.

1994-01-01

This research thesis proposes a new approach to the chemistry of uranium thiolate complexes as these compounds are very promising for various uses (in bio-inorganic chemistry, in some industrial processes like oil desulphurization). It more particularly addresses the U-S bond or more generally bonds between polarizable materials and hard metals. The author thus reports the study of uranium organometallic thiolates (tricyclo-penta-dienic and mono-cyclo-octa-tetraenylic complexes), and of uranium homoleptic thiolates (tetra-thiolate complexes, hexa-thiolate complexes, reactivity of homoleptic thiolate complexes) [fr

Reactive transport of Sr, Cs and Tc through a column packed with fracture-filling material

International Nuclear Information System (INIS)

Cui, D.; Eriksen, T.

1998-01-01

The migration behaviour of the fission products 90 Sr, 137 Cs and 99 Tc in a granite fracture-groundwater system was studied in column experiments using crushed Stripa granite fracture-filling material (125-250 μm) as a sorbent. Based on breakthrough curves of Sr 2+ , Cs + , TcO 4 - and anthraquinonsulphonate, the distribution ratio K d (cm 3 /g) values for Sr 2+ and Cs + were estimated to be 1.5 and 2 respectively; non-sorbing behaviour of TcO 4 - under oxic conditions was observed. The diffusion of Sr 2+ in the material is shown to be a kinetic-controlling step in the desorption process that followed the initial leaching. Under reducing conditions, it was found that TcO 4 - was reduced to TcO 2 .nH 2 O(s) by the Fe(II)-containing fracture-filling material and Tc(IV) aq was rapidly sorbed by the material. This observation proves that the migration of technetium under reducing conditions in deep groundwater systems will not only be controlled by the solubility of TcO 2 . nH 2 O(s) and the speed of groundwater flow but also by the rapid sorption of Tc(IV) aq on all available mineral surfaces. (orig.)

Recovery of uranium by a reverse osmosis process

International Nuclear Information System (INIS)

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

1980-01-01

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

Drawing of uranium in {gamma} phase; Filage de l'uranium en phase gamma

Energy Technology Data Exchange (ETDEWEB)

Stohr, J A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Chevigny, R [Conpagnie, Pechiney, 73 - Chambery (France)

1955-07-01

It describes the study of working and drawing of uranium in {gamma} phase. In a first part, the forgeable characteristics of uranium metal in the three different phases ({alpha}, {beta} and {gamma}) are compared by using BASTIEN-PORTEVIN method. The different experiments are testing the behaviour metal in each phase under different stresses and a temperature gradient as slow and shock traction, slow and shock compression, resilience, flexibility. Results show that optimum conditions for uranium drawing is uranium in phase {gamma}. In a second part, it described the drawing method and process. The uranium rods obtained by this technique are of very good quality. In addition, the material wear is very low which permits a low production cost. Finally, the uranium rod physical properties are studied. (M.P.)

Uranium Raw Material for the Nuclear Fuel Cycle: Exploration, Mining, Production, Supply and Demand, Economics and Environmental Issues (URAM-2009). Proceedings of an International Symposium

Energy Technology Data Exchange (ETDEWEB)

NONE

2014-05-15

This IAEA symposium is a third in a series which began in 2000 to discuss issues related to uranium raw materials. The symposia covered all areas of the uranium production cycle — including uranium geology, exploration, mining; milling and refining of uranium concentrates; and safety, environmental, social, training and regulatory issues — and reported on uranium supply and demand, and market scenarios. The first symposium was held in October 2000 — a time of extremely depressed market prices for uranium and of mines being closed — and primarily addressed environmental and safety issues in the uranium production cycle. By the time the second symposium was held in June 2005, the uranium market had started to improve after nearly two decades of depressed activity because of increased demand due to rising expectations for nuclear power expansion. Since then, there has been a dramatic rise in the uranium spot price, which in turn has promoted a significant increase in uranium exploration activities all over the world. The international symposium on Uranium Raw Material for the Nuclear Fuel Cycle (URAM-2009) was held at the IAEA, Vienna, 22–26 June 2009, at a time when nuclear energy was emerging as a viable alternative to meet the ever increasing demand of electricity in a sustainable manner, without degrading the environment. However, the global recession and credit crunch could impact the growth of the uranium industry. Since 2000, the identified uranium resource base has grown by more than 75%, exploration efforts have continued to increase in greenfield as well as brownfield sites, annual uranium production has risen, and the issue of social licensing and uranium stewardship has become increasingly important for public acceptance of the uranium industry. Some 210 delegates from 33 States and four international organizations participated in the symposium. In total, 120 technical papers were presented in the oral and poster sessions, and an exhibition on

Uranium Raw Material for the Nuclear Fuel Cycle: Exploration, Mining, Production, Supply and Demand, Economics and Environmental Issues (URAM-2009). Proceedings of an International Symposium

International Nuclear Information System (INIS)

2014-05-01

This IAEA symposium is a third in a series which began in 2000 to discuss issues related to uranium raw materials. The symposia covered all areas of the uranium production cycle — including uranium geology, exploration, mining; milling and refining of uranium concentrates; and safety, environmental, social, training and regulatory issues — and reported on uranium supply and demand, and market scenarios. The first symposium was held in October 2000 — a time of extremely depressed market prices for uranium and of mines being closed — and primarily addressed environmental and safety issues in the uranium production cycle. By the time the second symposium was held in June 2005, the uranium market had started to improve after nearly two decades of depressed activity because of increased demand due to rising expectations for nuclear power expansion. Since then, there has been a dramatic rise in the uranium spot price, which in turn has promoted a significant increase in uranium exploration activities all over the world. The international symposium on Uranium Raw Material for the Nuclear Fuel Cycle (URAM-2009) was held at the IAEA, Vienna, 22–26 June 2009, at a time when nuclear energy was emerging as a viable alternative to meet the ever increasing demand of electricity in a sustainable manner, without degrading the environment. However, the global recession and credit crunch could impact the growth of the uranium industry. Since 2000, the identified uranium resource base has grown by more than 75%, exploration efforts have continued to increase in greenfield as well as brownfield sites, annual uranium production has risen, and the issue of social licensing and uranium stewardship has become increasingly important for public acceptance of the uranium industry. Some 210 delegates from 33 States and four international organizations participated in the symposium. In total, 120 technical papers were presented in the oral and poster sessions, and an exhibition on

Uranium Raw Material for the Nuclear Fuel Cycle: Exploration, Mining, Production, Supply and Demand, Economics and Environmental Issues (URAM-2009). Proceedings of an International Symposium

International Nuclear Information System (INIS)

2014-06-01

This IAEA symposium is a third in a series which began in 2000 to discuss issues related to uranium raw materials. The symposia covered all areas of the uranium production cycle — including uranium geology, exploration, mining; milling and refining of uranium concentrates; and safety, environmental, social, training and regulatory issues — and reported on uranium supply and demand, and market scenarios. The first symposium was held in October 2000 — a time of extremely depressed market prices for uranium and of mines being closed — and primarily addressed environmental and safety issues in the uranium production cycle. By the time the second symposium was held in June 2005, the uranium market had started to improve after nearly two decades of depressed activity because of increased demand due to rising expectations for nuclear power expansion. Since then, there has been a dramatic rise in the uranium spot price, which in turn has promoted a significant increase in uranium exploration activities all over the world. The international symposium on Uranium Raw Material for the Nuclear Fuel Cycle (URAM-2009) was held at the IAEA, Vienna, 22–26 June 2009, at a time when nuclear energy was emerging as a viable alternative to meet the ever increasing demand of electricity in a sustainable manner, without degrading the environment. However, the global recession and credit crunch could impact the growth of the uranium industry. Since 2000, the identified uranium resource base has grown by more than 75%, exploration efforts have continued to increase in greenfield as well as brownfield sites, annual uranium production has risen, and the issue of social licensing and uranium stewardship has become increasingly important for public acceptance of the uranium industry. Some 210 delegates from 33 States and four international organizations participated in the symposium. In total, 120 technical papers were presented in the oral and poster sessions, and an exhibition on

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

Derivation of residual radioactive material guidelines for uranium in soil at the Middlesex Sampling Plant Site, Middlesex, New Jersey

International Nuclear Information System (INIS)

Dunning, D.E.

1995-02-01

Residual radioactive material guidelines for uranium in soil were derived for the Middlesex Sampling Plant (MSP) site in Middlesex, New Jersey. This site has been designated for remedial action under the Formerly Utilized Sites Remedial Action Program (FUSRAP) of the US Department of Energy. The site became contaminated from operations conducted in support of the Manhattan Engineer District (MED) and the Atomic Energy Commission (AEC) between 1943 and 1967. Activities conducted at the site included sampling, storage, and shipment of uranium, thorium, and beryllium ores and residues. Uranium guidelines for single radioisotopes and total uranium were derived on the basis of the requirement that the 50-year committed effective dose equivalent to a hypothetical individual living or working in the immediate vicinity of the MSP site should not exceed a dose of 30 mrem/yr following remedial action for the current-use and likely future-use scenarios or a dose of 100 mrem/yr for less likely future-use scenarios. The RESRAD computer code, which implements the methodology described in the DOE manual for establishing residual radioactive material guidelines, was used in this evaluation. Four scenarios were considered for the site. These scenarios vary regarding future land use at the site, sources of water used, and sources of food consumed

Application of the 226Ra-230Th-234U and 227Ac-231Pa-235U radiochronometers to uranium certified reference materials

International Nuclear Information System (INIS)

Rolison, J.M.; Treinen, K.C.; McHugh, K.C.; Gaffney, A.M.; Williams, R.W.

2017-01-01

Uranium certified reference materials (CRM) issued by New Brunswick Laboratory were subjected to dating using four independent uranium-series radiochronometers. In all cases, there was acceptable agreement between the model ages calculated using the 231 Pa- 235 U, 230 Th- 234 U, 227 Ac- 235 U or 226 Ra- 234 U radiochronometers and either the certified 230 Th- 234 U model date (CRM 125-A and CRM U630), or the known purification date (CRM U050 and CRM U100). The agreement between the four independent radiochronometers establishes these uranium certified reference materials as ideal informal standards for validating dating techniques utilized in nuclear forensic investigations in the absence of standards with certified model ages for multiple radiochronometers. (author)

Modeling the mobility of uranium from NORM-rich bedrock using multivariate statistical techniques - The mobility of uranium from U-containing bedrock materials as a function of pH: Implications for tunnel construction

Energy Technology Data Exchange (ETDEWEB)

Helmers, Tari; Fjermestad, Halldis; Salbu, Brit; Skipperud, Lindis [Norwegian University of Life Sciences, P.O. Box 5003, 1432 Aas (Norway); Meland, Sondre; Hagelia, Per [Norwegian Public Roads Administration, P.O. Box 8142, 0033 Oslo (Norway)

2014-07-01

According to amendments made to the Norwegian Pollution Control Act in 2011, naturally occurring radioactive material is now to be considered as an environmental contaminant, in addition to organic pollutants and trace metals. Environmental contamination is strongly correlated with the mobility and bioavailability of metals and radionuclides in natural systems. In order to determine the risk of environmental contamination from e.g. uranium (U) in alum shale areas, it is of particular importance to determine the mobility of U and trace metals found in the rock materials and their binding mechanisms. By determining the speciation and mobility of uranium and trace metals, better predictions can be made on the transport of contaminants in the environment from intervention like road and tunnel construction. The substrate media analyzed in this work was collected from a future tunnel construction site that is being built in the Gran municipality on National road Rv4 in Norway. The bedrock in the Gran municipality is rich in U-bearing minerals. Therefore, there is high potential for environmental contamination from the rock material removed for tunnel construction purposes. The present work focuses upon the effects of pH and the contact time (substrate media: solution) on the mobility of uranium. In order to identify the effects of pH and contact time on mobility, sample cores collected from an area rich in alum shale were subjected to an extended leaching experiment. In this experiment, the substrate materials were treated with five different pH solutions and were analyzed for different contact times. In addition, the results were compared to data from a sequential extraction experiment. In the leaching experiment, the mobilization of uranium in all of the substrate material was affected by the pH of solution. All of the samples were capable of quickly buffering pH solutions with a pH as low as 4 to neutral-alkaline conditions, attributed to the carbonate minerals

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.)

Geological principles of exploration for sandstone-hosted uranium deposits

International Nuclear Information System (INIS)

Le Roux, J.P.

1982-10-01

Although the importance of sandstone-hosted uranium deposits has seemingly faded in recent years due to the discovery of large, high -grade deposits elsewhere, a forecasted energy shortage in the near future will probably necessitate a new look at sedimentary basins as a source of uranium. Back-arc basins adjacent to calcalkaline source areas are especially favourable if they are filled with fluvial, post-Devonian sediments. Syn- and post-depositional tectonics play an important role in the sedimentation-mineralisation process and should be investigated. The oxidation-reduction state of the sandstones is a valid prospecting tool. Sedimentological environments govern the permeability and vegetal matter content of sandstones and directly control uranium mineralisation

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

Energy Technology Data Exchange (ETDEWEB)

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

1960-12-15

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

Status Report from Czechoslovakia [Processing of Low-Grade Uranium Ores

Energy Technology Data Exchange (ETDEWEB)

Civin, V; Belsky, M [Research and Development Laboratory No.3 of the Uranium Industry, Prague, Czechoslovakia (Czech Republic)

1967-06-15

The present paper deals with the fundamental problems and the main routes followed in processing low-grade uranium ores in CSSR. In this connection it may be useful to discuss the definition of low-grade ore. In our country this term is applied to uraniferous material with a very low content of uranium (of the order of 0.01%) whose treatment causes no particular difficulty. However, the same term is also used to designate those materials whose processibility lies on the verge of economic profitability. In our view, this classification, of an ore using two independent criteria (i.e. uranium content and processing economy) is useful from the standpoint of technology. The treatment of both such ore types is as a rule carried out by specific technological processes. Consequently, low-grade uranium ores can be divided into two groups: (1) Ores with a low uranium content. To this category belong in our country uraniferous materials which originate as a by-product of technological processes used in processing other materials. This is primarily gangue and tailings of various physical or physico-chemical pretreatment operations to which the ore is subjected at the mining site. Mention should be made in this connection of mine waters, which represent a useful complementary source of uranium despite their low uranium content (of the order of milligrams per litre). (2) Ores whose economical treatment is problematic. To this category belong deposits of conventional ore types with a uranium content on the limit of profitable treatment. Also, those deposits containing atypical materials possessing such properties which impair the economy of their treatment. This includes ores with a considerable amount of components which are difficult to separate and which at the same time consume the leaching agents. Finally, it covers uranium-bearing materials in refractory forms which are difficult to dissolve and also some special materials, such as lignites, uranium-bearing shales, loams

Physico-mechanical characteristics of commercially available bulk-fill composites.

Science.gov (United States)

Leprince, Julian G; Palin, William M; Vanacker, Julie; Sabbagh, Joseph; Devaux, Jacques; Leloup, Gaetane

2014-08-01

Bulk-fill composites have emerged, arguably, as a new "class" of resin-based composites, which are claimed to enable restoration in thick layers, up to 4mm. The objective of this work was to compare, under optimal curing conditions, the physico-mechanical properties of most currently available bulk-fill composites to those of two conventional composite materials chosen as references, one highly filled and one flowable "nano-hybrid" composite. Tetric EvoCeram Bulk Fill (Ivoclar-Vivadent), Venus Bulk Fill (Heraeus-Kulzer), SDR (Dentsply), X-tra Fil (VOCO), X-tra Base (VOCO), Sonic Fill (Kerr), Filtek Bulk Fill (3M-Espe), Xenius (GC) were compared to the two reference materials. The materials were light-cured for 40s in a 2mm×2mm×25mm Teflon mould. Degree of conversion was measured by Raman spectroscopy, Elastic modulus and flexural strength were evaluated by three point bending, surface hardness using Vickers microindentation before and after 24h ethanol storage, and filler weight content by thermogravimetric analysis. The ratio of surface hardness before and after ethanol storage was considered as an evaluation of polymer softening. Data were analyzed by one-way ANOVA and post hoc Tukey's test (p=0.05). The mechanical properties of the bulk-fill composites were mostly lower compared with the conventional high viscosity material, and, at best, comparable to the conventional flowable composite. Linear correlations of the mechanical properties investigated were poor with degree of conversion (0.090.8). Softening in ethanol revealed differences in polymer network density between material types. The reduction of time and improvement of convenience associated with bulk-fill materials is a clear advantage of this particular material class. However, a compromise with mechanical properties compared with more conventional commercially-available nano-hybrid materials was demonstrated by the present work. Given the lower mechanical properties of most bulk-fill materials

Practical aspects of monitoring and dosimetry of long-lived dust in uranium mines and mills - determination of the annual limit on intake for uranium and uranium/thorium ore dust

International Nuclear Information System (INIS)

Duport, P.; Horvath, F.

1989-01-01

Based on the recommendations of ICRP Publication 26, the dosimetric and metabolic data of ICRP Publication 30, and using available information on the physical and solubility characteristics of uranium and uranium/thorium ore, the ALI values for airborne ore dust were calculated. Four hypothetical types of ore were considered: uranium ore with no radon emanation, uranium ore with 50% radon emanation, uranium/thorium ore with neither 222 Rn nor thoron emanation, and uranium/thorium ore with 50% 22 Rn and 220 Rn emanation. Furthermore, the ALI values were calculated assuming the radionuclides present in the ore were all: (a) solubility class Y: (b) solubility class W; and (c) equal parts of classes Y and W. The ALI values were also calculated for Activity Median Aerodynamic Diameters (AMAD) ranging from 1 to 10 μm. The results of the calculations show that the solubility class of the radionuclides is the single most important factor that governs ALI values. The ALI value for uranium and uranium-thorium ore dust is proportional to (AMAD) 0.5 for class Y materials, (AMAD) 0.2 for a mixture of equal parts of class Y and class W materials, and is independent of the AMAD for class W materials. A series of graphs is given from which it is possible to evaluate the ALI for airborne ore dust when the AMAD of the dust and the solubility characteristics are known approximately. (author)

How to go on with Czech uranium: does current uranium mining in the Czech Republic cover Czech nuclear power plants' needs?

International Nuclear Information System (INIS)

Michalek, Bedrich

2009-01-01

The overview covers the history of uranium mining in the Czech Republic (description of the deposits and overview of their exploitation) and current needs for uranium and the status of uranium resources in the Czech Republic (uranium mining at the Rozna deposit, overview of exploitation of the deposit, uranium ore reserves, possibilities of future use of the Rozna deposit, the Brzkov and Horni Veznice deposits, and the use of mine waters as a secondary uranium source). It is concluded that in view of the current development of uses of raw materials for the power sector worldwide and increasing dependence of many countries (including the Czech Republic) on imports of such raw materials (often from politically unstable countries) it is strategically important to maintain domestic uranium mining to cover the needs of the Czech power sector. Uranium reserves and preconditions for their mining still exist in this country. (P.A.)

Analysis of Hazards Associated with a Process Involving Uranium Metal and Uranium Hydride Powders

Energy Technology Data Exchange (ETDEWEB)

Bullock, J.S.

2000-05-01

An analysis of the reaction chemistry and operational factors associated with processing uranium and uranium hydride powders is presented, focusing on a specific operation in the Development Division which was subjected to the Job Hazard Analysis (JHA) process. Primary emphasis is on the thermodynamic factors leading to pyrophoricity in common atmospheres. The discussion covers feed powders, cold-pressed and hot-pressed materials, and stray material resulting from the operations. The sensitivity of the various forms of material to pyrophoricity in common atmospheres is discussed. Operational recommendations for performing the work described are given.

Uranium production from low grade Swedish shale

International Nuclear Information System (INIS)

Carlsson, O.

1977-01-01

In view of the present nuclear programmes a steep increase in uranium demand is foreseen which will pose serious problems for the uranium industry. The annual additions to uranium ore reserves must almost triple within the next 15 years in order to support the required production rates. Although there are good prospects for the discovery of further conventional deposits of uranium there is a growing interest in low grade uranium deposits. Large quantities of uranium exist in black shales, phosphates, granites, sea water and other unconventional sources. There are however factors which limit the utilization of these low grade materials. These factors include the extraction costs, the environmental constrains on mining and milling of huge amounts of ore, the development of technologies for the beneficiation of uranium and, in the case of very low grade materials, the energy balance. The availability of by-product uranium is limited by the production rate of the main product. The limitations differ very much according to types of ores, mining and milling methods and the surroundings. As an illustration a description is given of the Swedish Ranstad uranium shale project, its potential, constraints and technical solutions

International symposium on uranium raw material for the nuclear fuel cycle: Exploration, mining, production, supply and demand, economics and environmental issues (URAM-2009). Book of abstracts

International Nuclear Information System (INIS)

2009-01-01

The International Symposium on Uranium Raw Material for the Nuclear Fuel Cycle: Exploration, Mining, Production, Supply and Demand, Economics and Environmental Issues (URAM-2009) addressed all aspects of the uranium fuel cycle, from the availability of raw materials to the long-term sustainability of nuclear power. The revival of the uranium industry in recent years has caused a dramatic increase in uranium exploration and mining activities in several countries. URAM-2009 was intended to bring together scientists, exploration and mining geologists, engineers, operators, regulators and fuel cycle specialists to exchange information and discuss updated research and current issues in uranium geology and deposits, exploration, mining and processing, production economics, and environmental and legal issues. Contributed papers covered uranium markets and economics (including supply and demand); social licensing in the uranium production cycle; uranium exploration (including uranium geology and deposits); uranium mining and processing; environmental and regulatory issues; human resources development. There was a poster session throughout the symposium, as well as an exhibition of topical photographs. A workshop on recent developments in Technical Cooperation Projects relevant to the Uranium Production Cycle area was also organized. On the last day of the symposium, there was an experts' Panel Discussion. The presentations and discussions at URAM-2009 (a) led to a better understanding of the adequacy of uranium sources (both primary and secondary) to meet future demand, (b) provided information on new exploration concepts, knowledge and technologies that will potentially lead to the discovery and development of new uranium resources, (c) described new production technology having the potential to more efficiently and economically exploit new uranium resources; (d) documented the environmental compatibility of uranium production and the overall effectiveness of the final

The study of necessity of verification-methods for Depleted Uranium

International Nuclear Information System (INIS)

Park, J. B.; Ahn, S. H.; Ahn, G. H.; Chung, S. T.; Shin, J. S.

2006-01-01

ROK has tried to establish management system for depleted uranium from 2004, and ROK achieved some results in this field including management software, management skill, and the list of company using the nuclear material. But, the studies for the depleted uranium are insufficient exclude the studies of KAERI. In terms of SSAC, we have to study more about whether the depleted uranium is really dangerous material or not and how is the depleted uranium diverted to the nuclear weapon. The depleted uranium was controlled by the item counting in the national system for the small quantity nuclear material. We don't have unique technical methods to clarify the depleted uranium on-the-spot inspection not laboratory scale. Therefore, I would like to suggest of the necessity of the verification methods for depleted uranium. Furthermore, I would like to show you the methods of the verification of the depleted uranium in national system up to now

Drawing of uranium in {gamma} phase; Filage de l'uranium en phase gamma

Energy Technology Data Exchange (ETDEWEB)

Stohr, J.A. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Chevigny, R. [Conpagnie, Pechiney, 73 - Chambery (France)

1955-07-01

It describes the study of working and drawing of uranium in {gamma} phase. In a first part, the forgeable characteristics of uranium metal in the three different phases ({alpha}, {beta} and {gamma}) are compared by using BASTIEN-PORTEVIN method. The different experiments are testing the behaviour metal in each phase under different stresses and a temperature gradient as slow and shock traction, slow and shock compression, resilience, flexibility. Results show that optimum conditions for uranium drawing is uranium in phase {gamma}. In a second part, it described the drawing method and process. The uranium rods obtained by this technique are of very good quality. In addition, the material wear is very low which permits a low production cost. Finally, the uranium rod physical properties are studied. (M.P.)

Experimental investigation of different fill materials in arch bridges with particular focus on Pearl-Chain Bridges

DEFF Research Database (Denmark)

Lund, Mia Schou Møller; Hansen, Kurt Kielsgaard; Hertz, Kristian Dahl

2016-01-01

Pearl-Chain Bridge technology is a recently developed prefabricated arch solution for road and railway bridges allowing faster, more environmentally friendly, and cheaper bridge construction. This study compared the strength and durability properties of three different types of fill material...

Tanks 18 And 19-F Structural Flowable Grout Fill Material Evaluation And Recommendations

International Nuclear Information System (INIS)

Stefanko, D.; Langton, C.

2011-01-01

Cementitious grout will be used to close Tanks 18-F and 19-F. The functions of the grout are to: (1) physically stabilize the final landfill by filling the empty volume in the tanks with a non compressible material; (2) provide a barrier for inadvertent intrusion into the tank; (3) reduce contaminant mobility by (a) limiting the hydraulic conductivity of the closed tank and (b) reducing contact between the residual waste and infiltrating water; and (4) providing an alkaline, chemically reducing environment in the closed tank to control speciation and solubility of selected radionuclides. The objective of this work was to identify a single (all-in-one) grout to stabilize and isolate the residual radionuclides in the tank, provide structural stability of the closed tank and serve as an inadvertent intruder barrier. This work was requested by V. A. Chander, High Level Waste (HLW) Tank Engineering, in HLW-TTR-2011-008. The complete task scope is provided in the Task Technical and QA Plan, SRNL-RP-2011-00587 Revision 0. The specific objectives of this task were to: (1) Identify new admixtures and dosages for formulating a zero bleed flowable tank fill material selected by HLW Tank Closure Project personnel based on earlier tank fill studies performed in 2007. The chemical admixtures used for adjusting the flow properties needed to be updated because the original admixture products are no longer available. Also, the sources of cement and fly ash have changed, and Portland cements currently available contain up to 5 wt. % limestone (calcium carbonate). (2) Prepare and evaluate the placement, compressive strength, and thermal properties of the selected formulation with new admixture dosages. (3) Identify opportunities for improving the mix selected by HLW Closure Project personnel and prepare and evaluate two potentially improved zero bleed flowable fill design concepts; one based on the reactor fill grout and the other based on a shrinkage compensating flowable fill mix

TANKS 18 AND 19-F STRUCTURAL FLOWABLE GROUT FILL MATERIAL EVALUATION AND RECOMMENDATIONS

Energy Technology Data Exchange (ETDEWEB)

Stefanko, D.; Langton, C.

2011-11-01

Cementitious grout will be used to close Tanks 18-F and 19-F. The functions of the grout are to: (1) physically stabilize the final landfill by filling the empty volume in the tanks with a non compressible material; (2) provide a barrier for inadvertent intrusion into the tank; (3) reduce contaminant mobility by (a) limiting the hydraulic conductivity of the closed tank and (b) reducing contact between the residual waste and infiltrating water; and (4) providing an alkaline, chemically reducing environment in the closed tank to control speciation and solubility of selected radionuclides. The objective of this work was to identify a single (all-in-one) grout to stabilize and isolate the residual radionuclides in the tank, provide structural stability of the closed tank and serve as an inadvertent intruder barrier. This work was requested by V. A. Chander, High Level Waste (HLW) Tank Engineering, in HLW-TTR-2011-008. The complete task scope is provided in the Task Technical and QA Plan, SRNL-RP-2011-00587 Revision 0. The specific objectives of this task were to: (1) Identify new admixtures and dosages for formulating a zero bleed flowable tank fill material selected by HLW Tank Closure Project personnel based on earlier tank fill studies performed in 2007. The chemical admixtures used for adjusting the flow properties needed to be updated because the original admixture products are no longer available. Also, the sources of cement and fly ash have changed, and Portland cements currently available contain up to 5 wt. % limestone (calcium carbonate). (2) Prepare and evaluate the placement, compressive strength, and thermal properties of the selected formulation with new admixture dosages. (3) Identify opportunities for improving the mix selected by HLW Closure Project personnel and prepare and evaluate two potentially improved zero bleed flowable fill design concepts; one based on the reactor fill grout and the other based on a shrinkage compensating flowable fill mix

Radiopacity evaluation of new root canal filling materials by digitalization of images.

Science.gov (United States)

Tanomaru-Filho, Mário; Jorge, Erica Gouveia; Guerreiro Tanomaru, Juliane Maria; Gonçalves, Marcelo

2007-03-01

The aim of this study was to evaluate the radiopacity of five root canal filling materials (AH Plus, Intrafill, Roeko Seal, Epiphany, and EndoRez). Following the International Organization of Standardization 6876/2001, five circular specimens (10 x 1 mm) were made from each material. After the material set, radiographs were made using occlusal film and a graduated aluminum step-wedge varying in thickness from 2 to 16 mm. The dental X-ray unit (GE1000) was set at 50 Kvp, 10 mA, 18 pulses/second, and distance of 33.5 cm. The radiographs were digitized, and the radiopacity was compared with the aluminum step-wedge, using WIXWIN-2000 software (Gendex). Data (mm Al) were analyzed using ANOVA and Tukey tests. AH Plus and Epiphany were the most radiopaque materials (9.8 and 8.8 mm Al, respectively), followed by EndoRez (7.2 mm Al). Roeko Seal and Intrafill presented the lowest radiopacity values (5.7 and 6.1 mm Al, respectively). Although the materials evaluated demonstrated different radiopacities, all had values above the minimum recommended by the International Organization of Standardization.

Long-term management and use of depleted uranium

International Nuclear Information System (INIS)

Max, A.

2001-01-01

The products resulting from the process of enrichment of natural uranium, or reprocessed uranium, are enriched uranium products as the light fraction and depleted uranium (uranium tails) as the heavy fraction. If the source material is natural uranium, the mass ratios of uranium products and uranium tails can be derived relatively easily from the required enrichment level of the uranium product (product assay (% of U-235)) and the selected depletion level of the uranium tails (tails assay (% of U-235)). The paper discusses among other aspects the dependence of the tails mass on the required enrichment level of the relevant uranium product, for various tails assays. (orig./CB) [de

Preparation, characterization and certification of uranium isotope reference materials

International Nuclear Information System (INIS)

Oliveira Junior, Olivio Pereira de

2006-01-01

This work describes the preparation, characterization and certification of a set of uranium isotope reference materials ranging from 0.5 to 20.0 % of 235 U in mass. The most important concepts of metrology in chemical measurements were applied so that the certified quantities in these materials could be directly traceable to the International System of Units (SI). As a consequence of this approach, these materials can be used in the instruments calibration, estimation of measurement uncertainty, method validation, assessment of performance of analysts, quality control routines and interlaboratory comparison programmes. The most advanced methods and techniques in mass spectrometry, that is, gas source mass spectrometry (GSMS), thermal ionisation mass spectrometry (TIMS) and inductively coupled plasma mass spectrometry (ICPMS) were investigated to identify which are the dominant components in the uncertainty and to quantify its contribution to the final value of the measurement uncertainty of the isotopic ratio. The results obtained were then compared to verify which are the methods and techniques associated to the lowest measurement uncertainty values. The isotope amount ratio n( 235 U)/n( 238 U) was certified in the materials produced to expanded uncertainties ranging from 0.02 to 0.10 % and the ratios n( 234 U)/n( 238 U) and n( 236 U)/n( 238 U), to uncertainties ranging from 0.03 to 2.20 %. These values fully comply to the requirements of the isotopic characterization of nuclear fuel as well as the analysis of environmental samples for nuclear safeguards. (author)

Geophysical methods in uranium mining

International Nuclear Information System (INIS)

Koehler, K.

1989-01-01

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

The determination of minor isotope abundances in naturally occurring uranium materials. The tracing power of isotopic signatures for uranium

International Nuclear Information System (INIS)

Ovaskainen, R.

1999-01-01

The mass spectrometric determination of minor abundant isotopes, 234 U and 236 U in naturally occurring uranium materials requires instruments of high abundance sensitivity and the use of highly sensitive detection systems. In this study the thermal ionisation mass spectrometer Finnigan MAT 262RPQ was used. It was equipped with 6 Faraday cups and a Secondary Electron Multiplier (SEM), which was operated in pulse counting mode for the detection of extremely low ion currents. The dynamic measurement range was increased considerably combining these two different detectors. The instrument calibration was performed carefully. The linearity of each detector, the deadtime of the ion counting detector, the detector normalisation factor, the baseline of each detector and the mass discrimination in the ion source were checked and optimised. A measurement technique based on the combination of a Gas Source Mass Spectrometry (GSMS) and a Thermal Ionisation Mass Spectrometry (TIMS) was developed for the accurate determination of isotopic composition in naturally occurring uranium materials. Because the expected ratio of n( 234 U)/n( 238 U) exceeded the dynamic measurement range of the Faraday detectors of the TIMS instrument, an experimental design using a combination of two detectors was developed. The n( 234 U)/n( 235 U) and n( 236 U)/n( 235 U) ratios were determined using ion counting in combination with the decelerating device. The n( 235 U)/n( 238 U) ratio was determined by the Faraday detector. This experimental design allowed the detector cross calibration to be circumvented. Precisions of less than 1 percent for the n( 234 U)/n( 235 U) ratios and 5-25 percent for the n( 236 U)/n( 235 U) ratios were achieved. The purpose of the study was to establish a register of isotopic signatures for natural uranium materials. The amount ratio, and isotopic composition of 18 ore concentrates, collected by the International Atomic Energy Agency (IAEA) from uranium milling and mining

International uranium production. A South African perspective

International Nuclear Information System (INIS)

Sinclair-Smith, D.

1984-01-01

Between 1981 and 1983 South Africa experienced a decline in its uranium resources of 23% in the less than $80/kg U category and 12% in the less than $130/kg U category. In 1983 only $5 million was spent on exploration, with activities being concentrated in the Witwatersrand Basin as a byproduct of gold exploration. South Africa has maintained a production level of around 6000 mt U in 1981, 1982 and 1983. One unusual feature of the South African uranium scene is the ability to selectively dump relatively high grade uranium tailings after the extraction of gold and to rework this material as well as material dumped prior to the emergence of the uranium industry. Uranium from this source amounted to some 28% of total production in 1983. (L.L.) (2 tabs., 6 figs.)

METHOD OF PRODUCING URANIUM METAL BY ELECTROLYSIS

Science.gov (United States)

Piper, R.D.

1962-09-01

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

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

International Nuclear Information System (INIS)

Kryvoruchko, Antonina P.; Atamanenkoa, Irina D.

2007-01-01

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

The remediation of abandoned workings of a mining area in Ningxiang uranium mine

International Nuclear Information System (INIS)

Liu Yaochi; Zhou Xinghuo; Liu Bing

2004-01-01

The typical mining under buildings and river was used in a mining area of Ningxiang uranium mine. After the mining ended, 32.1% of the 2.68 m 3 abandoned workings did not fill because of limitation of the cut-and-fill mining method at that time. To remedy this, the mine used new filling methods. After the remedial action, the filling coefficient of pits reached 100%, and that of tunnels reached 86%. It can be proved by the monitoring data that the subsiding of surface has been effectively controlled at the abandoned workings

METHOD OF APPLYING NICKEL COATINGS ON URANIUM

Science.gov (United States)

Gray, A.G.

1959-07-14

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

High-etch-rate bottom-antireflective coating and gap-fill materials using dextrin derivatives in via first dual-Damascene lithography process

Science.gov (United States)

Takei, Satoshi; Sakaida, Yasushi; Shinjo, Tetsuya; Hashimoto, Keisuke; Nakajima, Yasuyuki

2008-03-01

The present paper describes a novel class of bottom antireflective coating (BARC) and gap fill materials using dextrin derivatives. The general trend of interconnect fabrication for such a high performance LSI is to apply cupper (Cu)/ low-dielectric-constant (low-k) interconnect to reduce RC delay. A via-first dual damascene process is one of the most promising processes to fabricate Cu/ low-k interconnect due to its wide miss-alignment margin. The sacrificial materials containing dextrin derivatives under resist for lithography were developed in via-first dual damascene process. The dextrin derivatives in this study was obtained by the esterification of the hydroxyl groups of dextrin resulting in improved solubility in the resist solvents such as propylene glycol monomethylether, propylene glycol monomethylether acetate, and ethyl lactate due to avoid the issue of defects that were caused by incompatability. The etch rate of our developed BARC and gap fill materials using dextrin derivatives was more than two times faster than one of the ArF resists evaluated in a CF4 gas condition using reactive ion etching. The improved etch performance was also verified by comparison with poly(hydroxystyrene), acrylate-type materials and latest low-k materials as a reference. In addition to superior etch performance, these materials showed good resist profiles and via filling performance without voids in via holes.

Effect of the insulation materials filling on the thermal performance of sintered hollow bricks under the air-conditioning intermittent operation

Directory of Open Access Journals (Sweden)

Chaoping Hou

2018-06-01

Full Text Available Wall insulation performance is an important factor affecting building energy consumption and indoor comfortable level. This study proposes that the insulation materials are filled into the cavities of the sintered hollow brick to replace the single insulation layer. The physical models of typical walls were built by the hollow bricks filled with expanding polystyrene board (EPS in cavities and wall thermal performance is numerically analyzed by the Finite Volume Method under air-conditioning intermittent operation, which conforms to the actual operation rules of air-conditioning. Results show that filling EPS in cavities is beneficial to improve the thermal performance of the bricks, and the larger the EPS filling ratio, the higher the thermal performance improvement. The EPS filling ratio increase has the higher sensitivity on inner surface heat flow under the low EPS filling ratio, and filling EPS in the external cavities is optimum with the decrement rate 5.92% higher than filling EPS in internal cavities for the EPS filling ratio of 20%, while filling EPS in internal and external cavities simultaneously is optimum with decrement rate 2.45%–6.87% higher than that with filling EPS in the internal cavities for the EPS filling ratio of 40%–80%. Keywords: Insulation filling ratio, Insulation filling location, Thermal performance, Sintered hollow bricks

The uranium International trade

International Nuclear Information System (INIS)

Gonzalez U, L.A.

1993-01-01

The aim of this thesis is the understanding of how the present dynamic of uranium International trade is developed, the variables which fall into, the factors that are affecting and conditioning it, in order to clarify which are going to be the outlook in the future of this important resource in front of the present ecological situation and the energetic panorama of XXI Century. For this purpose, as starting point, the uranium is considered as a strategic material which importance take root in its energetic potential as alternate energy source, and for this reason in Chapter I, the general problem of raw materials, its classification and present situation in the global market is presented. In Chapter II, by means of a historical review, is explain what uranium is, how it was discovered, and how since the end of the past Century and during the last three decades of present, uranium pass of practically unknown element, to the position of a strategic raw material, which by degrees, generate an International market, owing to its utilization as a basic resource in the generation of energy. Chapter III, introduce us in the roll played by uranium, since its warlike applications until its utilization in nuclear reactors for the generation of electricity. Also is explain the reason for this change in the perception at global level. Finally, in Chapter IV we enter upon specifically in the present conditions of the International market of this mineral throughout the trends of supply and demand, the main producers, users, price dynamics, and the correlation among these economical variables and other factors of political, social and ecological nature. All of these with the purpose to found out, if there exist, a meaning of the puzzle that seems to be the uranium International trade

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

International Nuclear Information System (INIS)

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

1979-02-01

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

Evaluation of Solid Geologic Reference Materials for Uranium-Series Measurements via LA-ICPMS

Science.gov (United States)

Matthews, K. A.; Goldstein, S. J.; Norman, D. E.; Nunn, A. J.; Murrell, M. T.

2008-12-01

Uranium-series geochemistry and geochronology have a wide range of applications in paleoclimatology, volcanology and other disciplines. To further explore these fields, the geoanalytical community has now begun to exploit recent advances in in situ, micron-scale sampling via laser ablation-ICPMS. Unfortunately, improvements in instrumentation have generally outpaced development of the appropriate geologic reference materials required for in situ U-series work. We will report results for uranium and thorium isotopic ratios and elemental concentrations measured in a suite of solid standards from the USGS (e.g., BCR-2G, BHVO-2G, GSD-1G, MACS-1, NKT-2G), as well as those from the MPI-DING series (e.g., ATHO-G, T1-G, StHs6/80-G). Specifically created for microanalysis, two of these standards are synthetic (GSD-1G, MACS-1) and the remainder are naturally-sourced glasses. They cover a range of compositions, ages (± secular equilibrium), elemental concentrations and expected isotopic ratios. The U-series isotopics of some powdered source materials have been characterized (e.g., BCR-2, BHVO-2), although there is no confirmation of the same ratios in the glass. Bulk measurement of these solid standards via TIMS and solution multicollector-ICPMS can then be used to assess the performance of LA-ICPMS techniques which require matrix-matched solid standards for correction of U-series elemental and isotopic ratios. These results from existing, widely-available reference materials will also facilitate quantification and comparison of U-series data among laboratories in the broader geoscience community.

Characterizing uranium oxide reference particles for isotopic abundances and uranium mass by single particle isotope dilution mass spectrometry

International Nuclear Information System (INIS)

Kraiem, M.; Richter, S.; Erdmann, N.; Kühn, H.; Hedberg, M.; Aregbe, Y.

2012-01-01

Highlights: ► A method to quantify the U mass in single micron particles by ID-TIMS was developed. ► Well-characterized monodisperse U-oxide particles produced by an aerosol generator were used. ► A linear correlation between the mass of U and the volume of particle(s) was found. ► The method developed is suitable for determining the amount of U in a particulate reference material. - Abstract: Uranium and plutonium particulate test materials are becoming increasingly important as the reliability of measurement results has to be demonstrated to regulatory bodies responsible for maintaining effective nuclear safeguards. In order to address this issue, the Institute for Reference Materials and Measurements (IRMM) in collaboration with the Institute for Transuranium Elements (ITU) has initiated a study to investigate the feasibility of preparing and characterizing a uranium particle reference material for nuclear safeguards, which is finally certified for isotopic abundances and for the uranium mass per particle. Such control particles are specifically required to evaluate responses of instruments based on mass spectrometric detection (e.g. SIMS, TIMS, LA-ICPMS) and to help ensuring the reliability and comparability of measurement results worldwide. In this paper, a methodology is described which allows quantifying the uranium mass in single micron particles by isotope dilution thermal ionization mass spectrometry (ID-TIMS). This methodology is characterized by substantial improvements recently achieved at IRMM in terms of sensitivity and measurement accuracy in the field of uranium particle analysis by TIMS. The use of monodisperse uranium oxide particles prepared using an aerosol generation technique developed at ITU, which is capable of producing particles of well-characterized size and isotopic composition was exploited. The evidence of a straightforward correlation between the particle volume and the mass of uranium was demonstrated in this study

Opportunities of influence of plasma streams formed in IKA with continuos nor king gas filling en the surface of materials

International Nuclear Information System (INIS)

Useinov, B.M.; Useinova, A.M.; Amrenova, A.U.; Pusankov, S.A.; Sartin, S.A.; Virko, P.G.

2001-01-01

The results of the investigation of influence of plasma stream formed in IKA with continuous working gas filling on the surface of stainless steel 12X18H10T and aluminum are given in this article. It is shown here that the effect of influence of plasma stream on the surface of materials depends on the way of working gas filling. There is the comparison of influence of plasma stream formed in plasma accelerator with impulse and continuous working gas filling

Welding of a powder metallurgy uranium alloy

International Nuclear Information System (INIS)

Holbert, R.K.; Doughty, M.W.; Alexander-Morrison, G.M.

1989-01-01

The interest at the Oak Ridge Y-12 Plant in powder metallurgy (P/M) uranium parts is due to the potential cost savings in the fabrication of the material, to achieving a more homogeneous product, and to the reduction of uranium scrap. The joining of P/M uranium-6 wt-% niobium (U-6Nb) alloys by the electron beam (EB) welding process results in weld porosity. Varying the EB welding parameters did not eliminate the porosity. Reducing the oxygen and nitrogen content in this P/M uranium material did minimize the weld porosity, but this step made the techniques of producing the material more difficult. Therefore, joining wrought and P/M U-6Nb rods with the inertia welding technique is considered. Since no gases will be evolved with the solid-state welding process and the weld area will be compacted, porosity should not be a problem in the inertia welding of uranium alloys. The welds that are evaluated are wrought-to-wrought, wrought-to-P/M, and P/M-to-P/M U-6Nb samples

Retrieval of buried depleted uranium from the T-1 trench

International Nuclear Information System (INIS)

Burmeister, M.; Castaneda, N.; Hull, C.; Barbour, D.; Quapp, W.J.

1998-01-01

The Trench 1 remediation project will be conducted this year to retrieve depleted uranium and other associated materials from a trench at Rocky Flats Environmental Technology Site. The excavated materials will be segregated and stabilized for shipment. The depleted uranium will be treated at an offsite facility which utilizes a novel approach for waste minimization and disposal through utilization of a combination of uranium recycling and volume efficient uranium stabilization

Uranium trace and alpha activity characterization of coal and fly ash using particle track etch technique

International Nuclear Information System (INIS)

Chakravarti, S.K.

1991-01-01

Uranium is extensively found in carbonaceous components of sedimentary rocks and is considered to be accumulated in coals during the coalification process through the geological times. Burning of coal is mainly responsible for a manifold increase in the concentration of radioactive nuclides in atmosphere precipitates. Fly ash being an incombustible residue and formed from 90% of the inorganic material in coal, escapes into the atmosphere and constitutes a potential hazard. Also its use as one of the pozzolanic materials in the products of concrete, bricks etc and filling of ground cavities is even more hazardous because of the wall radioactivity, besides emission and diffusion of radon. This paper reports a simple method called Particle Track Etch (PTE) technique, for trace determination of uranium content in coal and fly ash samples by making use of low cost and versatile plastic detectors known as Solid State Nuclear Track Detectors (SSNTDs). Total alpha activity has also been estimated using these SSNTDs. The values of uranium concentration in coal samples are found to range from 1.1 to 3.6 ppm (uniform component) and 33 to 46 ppm (non-uniform part) whereas in fly ash, it varies from 8 to 11 ppm (uniform) and 55 to 71 ppm in non-uniform range. It is also observed that the alpha activity is a function of uranium concentration for most of the natural samples of coal studied except for mixtures of fly ash samples where relationship is found to be on higher side. (author). 13 refs., 2 tabs., 1 fig

Uranium distribution in Brazilian granitic rocks. Identification of uranium provinces

International Nuclear Information System (INIS)

Tassinari, C.G.G.

1993-01-01

The research characterized and described uranium enriched granitoids in Brazil. They occur in a variety of tectonic environments and are represented by a variety granite types of distinct ages. It may be deduced that in general they have been generated by partial melting process of continental crust. However, some of them, those with tonality composition, indicate a contribution from mantle derived materials, thus suggesting primary uranium enrichment from the upper mantle. Through this study, the identification and characterization of uranium enriched granite or uranium provinces in Brazil can be made. This may also help identify areas with potential for uranium mineralization although it has been note that uranium mineralization in Brazil are not related to the uranium enrichment process. In general the U-anomalous granitoids are composed of granites with alkaline composition and granite ''sensu strictu'' which comprise mainly of syenites, quartz-syenites and biotite-hornblende granites, with ages between 1,800 - 1,300 M.a. The U-anomalous belongings to this period present high Sr initial ratios values, above 0.706, and high Rb contents. Most of the U-enriched granitoids occur within ancient cratonic areas, or within Early to Mid-Proterozoic mobile belts, but after their cratonization. Generally, these granitoids are related to the border zones of the mobile belts or deep crustal discontinuity. Refs, 12 figs, 3 tabs

Characterization of uranium isotopic abundances in depleted uranium metal assay standard 115

International Nuclear Information System (INIS)

Mathew, K.J.; Singleton, G.L.; Essex, R.M.; Hasozbek, A.; Orlowicz, G.; Soriano, M.

2013-01-01

Certified reference material (CRM) 115, Uranium (Depleted) Metal (Uranium Assay Standard), was analyzed using a TRITON Thermal Ionization Mass Spectrometer to characterize the uranium isotope-amount ratios. The certified 235 U/ 238 U 'major' isotope-amount ratio of 0.0020337 (12) in CRM 115 was determined using the total evaporation (TE) and the modified total evaporation (MTE) analytical techniques. In the MTE method, the total evaporation process is interrupted on a regular basis to allow correction of background from peak tailing, internal calibration of the secondary electron multiplier detector versus the Faraday cups, peak-centering, and ion source re-focusing. For the 'minor' 234 U/ 238 U and 236 U/ 238 U isotope-amount ratio measurements using MTE, precision and accuracy comparable to conventional analyses are achieved, without compromising the quality of the 235 U/ 238 U isotope-amount ratios. Characterized values of the 234 U/ 238 U and 236 U/ 238 U isotope-amount ratios in CRM 115 are 0.000007545 (10) and 0.000032213 (84), respectively. The 233 U/ 238 U isotope-amount ratio in CRM 115 is estimated to be -9 . The homogeneity of the CRM 115 materials is established through the absence of any statistically significant unit-to-unit variation in the uranium isotope-amount ratios. The measurements leading to the certification of uranium isotope-amount ratios are discussed. (author)

Uranium industry annual 1990, September 1991

International Nuclear Information System (INIS)

1991-01-01

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

Sandstone-type uranium deposits

International Nuclear Information System (INIS)

Austin, S.R.; D'Andrea, R.F. Jr.

1978-01-01

Three overall factors are necessary for formation of uranium deposits in sandstone: a source of uranium, host rocks capable of transmitting uranium-bearing solutions, and a precipitant. Possible sources of uranium in sandstone-type deposits include groundwaters emanating from granitic highlands, arkosic sediments, tuffaceous material within or overlying the host rocks, connate fluids, and overlying black shales. The first three sources are considered the most likely. Host rocks are generally immature sandstones deposited in alluvial-fan, intermontane-basin or marginal-marine environments, but uranium deposits do occur in well-winnowed barrier-bar or eolian sands. Host rocks for uranium deposits generally show coefficients of permeability on the order of 1 to 100 gal/day/ft 2 . Precipitants are normally agents capable of reducing uranium from the uranyl to the uranous state. The association of uranium with organic matter is unequivocal; H 2 S, a powerful reductant, may have been present at the time of formation of some deposits but may go unnoticed today. Vanadium can serve to preserve the tabular characteristics of some deposits in the near-surface environment, but is considered an unlikely primary precipitant for uranium. Uranium deposits in sandstone are divided into two overall types: peneconcordant deposits, which occur in locally reducing environments in otherwise oxidized sandstones; and roll-type deposits, which occur at the margin of an area where an oxidized groundwater has permeated an otherwise reduced sandstone. Uranium deposits are further broken down into four subclasses; these are described

The Dissolution of Uranium Oxides in HB-Line Phase 1 Dissolvers

International Nuclear Information System (INIS)

Gray, J.H.

2003-01-01

A series of characterization and dissolution studies has been performed to define flowsheet conditions for the dissolution of uranium oxide materials in dissolvers. The samples selected for analysis were uranium oxide materials. The selection of these uranium oxide materials for characterization and dissolution studies was based on high enriched uranium content and trace levels of plutonium. Test results from the characterization study identified ferric oxide (Fe2O3) and iron/chromium/nickel (Fe/Cr/Ni) particles as impurities along with the tri-uranium oxide (U3O8) and uranium trioxide (UO3). The weight percent uranium in this material was found to vary depending on the impurity content. The trace impurity plutonium appears to be associated with the Fe/Cr/Ni particles. A small amount of absorbed moisture and waters of hydration is present. Most of the uranium oxides easily dissolved in low-molar nitric acid solutions without fluoride within one to two hours at solution temperature s between 60-80 degrees C. A small amount of residue remained following this dissolution step. To assure complete dissolution of uranium from these oxide materials, an additional dissolution step at 90 degrees C to boiling for at least one to two hours has been suggested. Only trace amounts of iron associated with Fe2O3 and Fe/Cr/Ni particles will dissolve during the dissolution steps. Neither hydrogen nor heat will be generated during the dissolution of these uranium oxide materials in nitric acid solutions. Some brown nitrogen dioxide (NO2) fumes will be generated during the dissolution of U3O8

Proportioning equipment for vibration filling and compacting of grain materials in pipe containers, especially of fuel elements

International Nuclear Information System (INIS)

Pinkas, V.; Filip, Z.; Beranek, J.

1981-01-01

The equipment consists of a base plate to which are attached the fastening collar fo the pipe container and the guide column with the height-adjustable support. The filling pipe is fixed to the support. The proportioning equipment prevents particles of grain material from segregation, thus allowing to achieve homogeneity of the material in the whole volume to be compacted. It also allows determining the height of the column of material in the pipe container without destructive effects on the stacked material. The equipment is designed for the manufacture of shortened fuel elements. (J.B.)

Preparation and certification of a uranium isotope certified reference material JAERI-U5

International Nuclear Information System (INIS)

Tamura, Shuzo; Hashitani, Hiroshi

1982-06-01

The Committee on Analytical Chemistry of Nuclear Fuels and Reactor Materials, JAERI had planned to prepare a new reference material, JAERI-U5 for uranium isotopic measurement since 1978. The reference material is composed of 6 samples of different enrichment in the range of 0.2 to 4.5 wt. percents of 235 U. The preparation includes dissolution of raw materials, blending of solutions, precipitation of ammonium diuranate, drying and ignition to U 3 O 8 . A mass-spectrometric collaborative analysis was carried out by well-trained two laboratories in this country, JAERI and PNC. The certified values were decided from the result of the collaborative work. As the measurements were based on NBS SRM's, JAERI-U5 should be called tertiary standard. The materials are packed in bottles of low-potassium-content glass for a possible use in non-destructive gamma-rays spectrometry. The reference material has been distributing from JAERI with a price of yen 140,000 per set (6 samples of 2 g of each) since 1979. (author)

The Davies-Gray titration for the assay of uranium in nuclear materials: a performance study

International Nuclear Information System (INIS)

Bickel, M.

1997-01-01

An interlaboratory comparison in two phases was organized to assess the precision and accuracy concerning the assay of uranium in nuclear materials by the potentiometric titration method. This contribution presents the results of this exercise in terms of method performance. Variations to be expected between different laboratories and within a single laboratory are estimated. In general, the method proved again very reliable. (orig.)

Comparative evaluation of microleakage of various retrograde filling materials: An in vitro study.

Science.gov (United States)

Galhotra, Virat; Sofat, Anjali; Pandit, Inder K; Gambhir, Ramandeep Singh; Srivastava, Nikhil; Gugnani, Neeraj

2013-07-01

The present study is envisaged to evaluate and compare the microleakage of mineral trioxide aggregate (MTA) with commonly used retrograde filling materials, like light-cured composite with dentin-bonding agents, light-cured glass ionomer cement (LC GIC) and resin-modified zinc oxide eugenol. Ninety freshly extracted non-carious single-rooted human anterior teeth were used in the study. They were randomly divided into four experimental groups and two control groups of 15 each. Following the biomechanical preparation, all teeth were obturated and then the apices of the obturated teeth were resected by removing 3 mm of each apex at 90° to the long axis of the tooth with a straight fissure bur in a high-speed air-rotor handpiece with water coolant. A 3-mm-deep root end cavity was prepared and the root end fillings were placed as per the manufacturer's instructions and according to the groups divided. The samples were then immersed in 1% methylene blue at room temperature for 72 h, 96 h and 1 week and the dye penetration was measured. All the four materials used in the study showed some microleakage throughout the experimental period. The sealing ability in terms of microleakage can be summarized as: MTA > Composite resin with dentin bonding agent > LC GIC > Resin modified zinc oxide eugenol.

Tanks 18 And 19-F Structural Flowable Grout Fill Material Evaluation And Recommendations

International Nuclear Information System (INIS)

Langton, C. A.; Stefanko, D. B.

2013-01-01

Cementitious grout will be used to close Tanks 18-F and 19-F. The functions of the grout are to: 1) physically stabilize the final landfill by filling the empty volume in the tanks with a non-compressible material; 2) provide a barrier for inadvertent intrusion into the tank; 3) reduce contaminant mobility by a) limiting the hydraulic conductivity of the closed tank and b) reducing contact between the residual waste and infiltrating water; and 4) providing an alkaline, chemically reducing environment in the closed tank to control speciation and solubility of selected radionuclides. The objective of this work was to identify a single (all-in-one) grout to stabilize and isolate the residual radionuclides in the tank, provide structural stability of the closed tank and serve as an inadvertent intruder barrier. This work was requested by V. A. Chander, High Level Waste (HLW) Tank Engineering, in HLW-TTR-2011-008. The complete task scope is provided in the Task Technical and QA Plan, SRNL-RP-2011-00587 Revision 0. The specific objectives of this task were to: 1) Identify new admixtures and dosages for formulating a zero bleed flowable tank fill material selected by HLW Tank Closure Project personnel based on earlier tank fill studies performed in 2007. The chemical admixtures used for adjusting the flow properties needed to be updated because the original admixture products are no longer available. Also, the sources of cement and fly ash have changed, and Portland cements currently available contain up to 5 wt. % limestone (calcium carbonate). 2) Prepare and evaluate the placement, compressive strength, and thermal properties of the selected formulation with new admixture dosages. 3) Identify opportunities for improving the mix selected by HLW Closure Project personnel and prepare and evaluate two potentially improved zero bleed flowable fill design concepts; one based on the reactor fill grout and the other based on a shrinkage compensating flowable fill mix design. 4

Tanks 18 And 19-F Structural Flowable Grout Fill Material Evaluation And Recommendations

Energy Technology Data Exchange (ETDEWEB)

Langton, C. A.; Stefanko, D. B.

2013-04-23

Cementitious grout will be used to close Tanks 18-F and 19-F. The functions of the grout are to: 1) physically stabilize the final landfill by filling the empty volume in the tanks with a non-compressible material; 2) provide a barrier for inadvertent intrusion into the tank; 3) reduce contaminant mobility by a) limiting the hydraulic conductivity of the closed tank and b) reducing contact between the residual waste and infiltrating water; and 4) providing an alkaline, chemically reducing environment in the closed tank to control speciation and solubility of selected radionuclides. The objective of this work was to identify a single (all-in-one) grout to stabilize and isolate the residual radionuclides in the tank, provide structural stability of the closed tank and serve as an inadvertent intruder barrier. This work was requested by V. A. Chander, High Level Waste (HLW) Tank Engineering, in HLW-TTR-2011-008. The complete task scope is provided in the Task Technical and QA Plan, SRNL-RP-2011-00587 Revision 0. The specific objectives of this task were to: 1) Identify new admixtures and dosages for formulating a zero bleed flowable tank fill material selected by HLW Tank Closure Project personnel based on earlier tank fill studies performed in 2007. The chemical admixtures used for adjusting the flow properties needed to be updated because the original admixture products are no longer available. Also, the sources of cement and fly ash have changed, and Portland cements currently available contain up to 5 wt. % limestone (calcium carbonate). 2) Prepare and evaluate the placement, compressive strength, and thermal properties of the selected formulation with new admixture dosages. 3) Identify opportunities for improving the mix selected by HLW Closure Project personnel and prepare and evaluate two potentially improved zero bleed flowable fill design concepts; one based on the reactor fill grout and the other based on a shrinkage compensating flowable fill mix design. 4

Experimental critical parameters of enriched uranium solution in annular tank geometries

International Nuclear Information System (INIS)

Rothe, R.E.

1996-04-01

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

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.

Disposition Options for Uranium-233

International Nuclear Information System (INIS)

Beahm, E.C.; Dole, L.R.; Forsberg, C.W.; Icenhour, A.S.; Storch, S.N.

1999-01-01

The U.S. Department of Energy (DOE) Fissile Materials Disposition Program (MD), in support of the U.S. arms-control and nonproliferation policies, has initiated a program to disposition surplus weapons-usable fissile material by making it inaccessible and unattractive for use in nuclear weapons. Weapons-usable fissile materials include plutonium, high-enriched uranium (HEU), and uranium-233 (sup 233)U. In support of this program, Oak Ridge National Laboratory led DOE's contractor efforts to identify and characterize options for the long-term storage and disposal of excess (sup 233)U. Five storage and 17 disposal options were identified and are described herein

Initial settlements of rock fills on soft clay

OpenAIRE

Pedersen, Truls Martens

2012-01-01

Rock fills that hit the seabed will remold the underlying material. If this material is a clay with sufficiently low shear strength, it will adopt rheological properties, causing flow through the rock fill, and contributing to the initial settlements of the rock fill in addition to conventional consolidation theory. The settlements of the rocks depend upon the height of the rock fill and how the rocks have been laid out. This is due to the viscosity of the clay, and the fact that clay is thix...

Efficacy of CM-Wire, M-Wire, and Nickel-Titanium Instruments for Removing Filling Material from Curved Root Canals: A Micro-Computed Tomography Study.

Science.gov (United States)

Rodrigues, Clarissa Teles; Duarte, Marco Antonio Hungaro; de Almeida, Marcela Milanezi; de Andrade, Flaviana Bombarda; Bernardineli, Norberti

2016-11-01

The aim of this ex vivo study was to evaluate the removal of filling material after using CM-wire, M-wire, and nickel-titanium instruments in both reciprocating and rotary motions in curved canals. Thirty maxillary lateral incisors were divided into 9 groups according to retreatment procedures: Reciproc R25 followed by Mtwo 40/.04 and ProDesign Logic 50/.01 files; ProDesign R 25/.06 followed by ProDesign Logic 40/.05 and ProDesign Logic 50/.01 files; and Gates-Glidden drills, Hedström files, and K-files up to apical size 30 followed by K-file 40 and K-file 50 up to the working length. Micro-computed tomography scans were performed before and after each reinstrumentation procedure to evaluate root canal filling removal. Statistical analysis was performed with Kruskal-Wallis, Friedman, and Wilcoxon tests (P < .05). No significant differences in filling material removal were found in the 3 groups of teeth. The use of Mtwo and ProDesign Logic 40/.05 rotary files did not enhance filling material removal after the use of reciprocating files. The use of ProDesign Logic 50/.01 files significantly reduced the amount of filling material at the apical levels compared with the use of reciprocating files. Association of reciprocating and rotary files was capable of removing a large amount of filling material in the retreatment of curved canals, irrespective of the type of alloy of the instruments. The use of a ProDesign Logic 50/.01 file for apical preparation significantly reduced the amount of remnant material in the apical portion when compared with reciprocating instruments. Copyright © 2016 American Association of Endodontists. Published by Elsevier Inc. All rights reserved.

Airborne uranium, its concentration and toxicity in uranium enrichment facilities

International Nuclear Information System (INIS)

Thomas, J.; Mauro, J.; Ryniker, J.; Fellman, R.

1979-02-01

The release of uranium hexafluoride and its hydrolysis products into the work environment of a plant for enriching uranium by means of gas centrifuges is discussed. The maximum permissible mass and curie concentration of airborne uranium (U) is identified as a function of the enrichment level (i.e., U-235/total U), and chemical and physical form. A discussion of the chemical and radiological toxicity of uranium as a function of enrichment and chemical form is included. The toxicity of products of UF 6 hydrolysis in the atmosphere, namely, UO 2 F 2 and HF, the particle size of toxic particulate material produced from this hydrolysis, and the toxic effects of HF and other potential fluoride compounds are also discussed. Results of an investigation of known effects of humidity and temperature on particle size of UO 2 F 2 produced by the reaction of UF 6 with water vapor in the air are reported. The relationship of the solubility of uranium compounds to their toxic effects was studied. Identification and discussion of the standards potentially applicable to airborne uranium compounds in the working environment are presented. The effectiveness of High Efficiency Particulate (HEPA) filters subjected to the corrosive environment imposed by the presence of hydrogen fluoride is discussed

Uranium deposits in the Eureka Gulch area, Central City district, Gilpin County, Colorado

Science.gov (United States)

Sims, P.K.; Osterwald, F.W.; Tooker, E.W.

1954-01-01

The Eureka Gulch area of the Central City district, Gilpin County, Colo., was mined for ores of gold, silver, copper, lead, and zinc; but there has been little mining activity in the area since World War I. Between 1951 and 1953 nine radioactive mine dumps were discovered in the area by the U.S. Geological Survey and by prospectors. the importance of the discoveries has not been determined as all but one of the mines are inaccessible, but the distribution, quantity, and grade of the radioactive materials found on the mine dumps indicate that the area is worth of additional exploration as a possible source of uranium ore. The uranium ans other metals are in and near steeply dipping mesothermal veins of Laramide age intrusive rocks. Pitchblende is present in at least four veins, and metatorbernite, associated at places with kosolite, is found along two veins for a linear distance of about 700 feet. The pitchblends and metatorbernite appear to be mutually exclusive and seem to occur in different veins. Colloform grains of pitchblende were deposited in the vein essentially contemporaneously with pyrite. The pitchblende is earlier in the sequence of deposition than galena and sphalerite. The metatorbernite replaces altered biotite-quartz-plagioclase gneiss and altered amphibolite, and to a lesser extent forms coatings on fractures in these rocks adjacent to the veins; the kasolite fills vugs in highly altered material and in altered wall rocks. Much of the pitchblende found on the dumps has been partly leached subsequent to mining and is out of equilibrium. Selected samples of metatorbernite-bearing rock from one mine dump contain as much as 6.11 percent uranium. The pitchblende is a primary vein mineral deposited from uranium-bearing hydrothermal solutions. The metatorbernite probably formed by oxidation, solution, and transportation of uranium from primary pitchblende, but it may be a primary mineral deposited directly from fluids of different composition from these

The uranium waste fluid processing examination by liquid and liquid extraction method using the emulsion flow method

International Nuclear Information System (INIS)

Kanda, Nobuhiro; Daiten, Masaki; Endo, Yuji; Yoshida, Hideaki; Mita, Yutaka; Naganawa, Hirochika; Nagano, Tetsushi; Yanase, Nobuyuki

2015-03-01

Spent centrifuges which had used for the development of the uranium enrichment technology are stored in the uranium enrichment facility located in Ningyo-toge Environmental Center, Japan Atomic Energy Agency (JAEA). Our technology of the centrifugal machine processing are supposed to separate the radioactive material adhered on surface of inner parts of centrifuges by the wet way decontamination method using the ultrasonic bath filled dilute sulfuric acid and water, and it is generated the neutralization sediment (sludge) by the processing of the radioactive waste fluid with the decontamination. JAEA had been considering the applicability of a streamlining and reduction of the processing of the sludge by decreases radioactive concentration including the sludge through the removes uranium from the radioactive waste fluid. As part of considerations, JAEA have been promoting technological developments of the uranium extraction separation using The Emulsion Flow Extraction Method (a theory propounded by JAEA-Nuclear Science and Engineering Center) in close coordination and cooperation between with JAEA-Nuclear Science and Engineering Center and Ningyo-toge Environmental Center from 2007 fiscal year. This report describes the outline of the application test using actual waste fluid of dilute sulfuric acid and water by developed the examination system introducing the emulsion flow extraction method. (author)

Healing process following application of set or fresh mineral trioxide aggregate as a root-end filling material.

Science.gov (United States)

Habibi, Mehdi; Ghoddusi, Jamileh; Habibi, Ataollah; Mohtasham, Nooshin

2011-01-01

An unsuccessful attempt to reach the apical area or to place the retrograde material is a major difficulty in periradicular surgery. The aim of this study was to compare the histological evaluation of the healing process following an orthograde versus a retrograde application of mineral trioxide aggregate (MTA) as a root-end filling material during apical surgery on cats' teeth in order to find out whether orthograde placement of MTA before surgery can be used instead of retrograde placement during surgery. In this experimental study, 24 canine teeth in 12 mature and healthy cats were filled with either MTA or gutta-percha in an orthograde manner. Two weeks later, the teeth with MTA were surgically exposed and resected to the set-MTA within the canals. The teeth previously filled by gutta-percha were also surgically exposed, and retrograde cavities were prepared at the root ends and filled with fresh-MTA. After 8 weeks, the animals were euthanized by vital perfusion. Six-micron histological slices were prepared from samples, stained by Hematoxylin & Eosin, and histologically studied by means of a light microscope. The collected data was analyzed by the Chi-square and the T-test. One of the samples in the fresh-MTA group was omitted during processing because of inappropriate sectioning. In the set-MTA group, 5 out of 12 showed chronic abscess, while in the fresh-MTA group, 2 out of 11 were discovered to have chronic abscess; however, no significant difference was observed (P>.05). Hard tissue healing (cementum, bone, cementum + bone formation) in the set-MTA and fresh-MTA groups were 7 out of 12 and 9 out of 11, respectively. While healing seemed more likely to occur in the fresh-MTA group, the difference was statistically insignificant (P>.05). The magnitude of bone, cementum, or bone and cementum formation showed slight differences between the two groups; however, the figures failed to show any marked differences (P>.05). Orthograde placement of MTA could be used

Effect of ultrasonic agitation on push-out bond strength and adaptation of root-end filling materials.

Science.gov (United States)

Alcalde, Murilo Priori; Vivan, Rodrigo Ricci; Marciano, Marina Angélica; Duque, Jussaro Alves; Fernandes, Samuel Lucas; Rosseto, Mariana Bailo; Duarte, Marco Antonio Hungaro

2018-05-01

This study evaluated the effect of ultrasonic agitation of mineral trioxide aggregate (MTA), calcium silicate-based cement (CSC), and Sealer 26 (S26) on adaptation at the cement/dentin interface and push-out bond strength. Sixty maxillary canines were divided into 6 groups ( n = 10): MTA, S26, and CSC, with or without ultrasonic activation (US). After obturation, the apical portions of the teeth were sectioned, and retrograde cavities were prepared and filled with cement by hand condensation. In the US groups, the cement was activated for 60 seconds: 30 seconds in the mesio-distal direction and 30 seconds in the buccal-lingual direction, using a mini Irrisonic insert coupled with the ultrasound transducer. After the materials set, 1.5-mm thick sections were obtained from the apexes. The presence of gaps and the bond between cement and dentin were analyzed using low-vacuum scanning electron microscopy. Push-out bond strength was measured using a universal testing machine. Ultrasonic agitation increased the interfacial adaptation of the cements. The S26 US group showed a higher adaptation value than MTA ( p < 0.05). US improved the push-out bond strength for all the cements ( p < 0.05). The US of retrograde filling cements enhanced the bond to the dentin wall of the root-end filling materials tested.

Geological evolution and uranium mineralisation of Chhinjra area, Kulu district, Himachal Pradesh

Energy Technology Data Exchange (ETDEWEB)

Sen, D B; Kumar, Suresh; Gangadharan, G R [Department of Atomic Energy, New Delhi (India). Atomic Minerals Div.

1995-08-01

Several shear-controlled and fracture-filled/disseminated type uranium occurrences are known in the Rampur window. This paper presents the geology and genetic aspects of fracture-filled type of mineralisation in Chhinjra area on the basis of recent stratigraphical, geochronological and tectonic data. Based on the angular unconformity between Manikaran quartzites and overlying chlorite phyllites, the geological evolution of Chhinjra area has been reconstructed in two stages: pre-unconformity and post-unconformity. Each stage is characterised by different phases of deformation with typical structural style and accompanying mineralisation processes. Four major tectonic events can be recognised here, namely 2500 Ma, 1200 Ma, 700 Ma and 55 Ma. Each event has left its imprint on the rocks as well as uranium mineralisation of Chhinjra area. (author). 14 refs., 4 figs.

Reactivity feedback coefficients of a low enriched uranium fuelled material test research reactor at end-of-life

International Nuclear Information System (INIS)

Muhammad, Farhan

2011-01-01

Highlights: → The isotopic concentration in the fuel changes as soon as it starts its operation. → The neutronic properties of a reactor also change with fuel burnup. → The reactivity feedbacks at end-of-life of a material test reactor fuelled with low enriched uranium fuel are calculated. → Codes used include WIMS-D4 and CITATION. - Abstract: The reactivity feedback coefficients at end-of-life of a material test reactor fuelled with low enriched uranium fuel were calculated. The reactor used for the study was the IAEA's 10 MW benchmark reactor. Simulations were carried out to calculate the different reactivity feedback coefficients including Doppler feedback coefficient, reactivity coefficient for change of water temperature and reactivity coefficient for change of water density. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It was observed that the magnitude of all the reactivity feedback coefficients increased at end of life of the reactor by almost 2-5%.

Uranium in South Africa

International Nuclear Information System (INIS)

Ford, M.A.

1993-01-01

The history, sources, mineralogy, extraction metallurgy, conversion, and enrichment of uranium in South Africa is reviewed. Over the past 40 years extraction plants were built at 27 sites, and over 140 kt of uranium have been produced. Older plants have had to adapt to changing market conditions, no single technology has had the opportunity to become entrenched, and the costs have been reduced to a third of those of the original flowsheet. The research efforts aimed at developing the country's nuclear raw materials have been particularly rewarding, as they have enabled South Africa to become a world leader in the extraction of uranium from low-grade ores and to develop methods for uranium enrichment and the production of nuclear fuels. 43 refs., 7 figs., 4 tabs

The determination of minor isotope abundances in naturally occurring uranium materials. The tracing power of isotopic signatures for uranium

Energy Technology Data Exchange (ETDEWEB)

Ovaskainen, R

1999-11-01

The mass spectrometric determination of minor abundant isotopes, {sup 234}U and {sup 236}U in naturally occurring uranium materials requires instruments of high abundance sensitivity and the use of highly sensitive detection systems. In this study the thermal ionisation mass spectrometer Finnigan MAT 262RPQ was used. It was equipped with 6 Faraday cups and a Secondary Electron Multiplier (SEM), which was operated in pulse counting mode for the detection of extremely low ion currents. The dynamic measurement range was increased considerably combining these two different detectors. The instrument calibration was performed carefully. The linearity of each detector, the deadtime of the ion counting detector, the detector normalisation factor, the baseline of each detector and the mass discrimination in the ion source were checked and optimised. A measurement technique based on the combination of a Gas Source Mass Spectrometry (GSMS) and a Thermal Ionisation Mass Spectrometry (TIMS) was developed for the accurate determination of isotopic composition in naturally occurring uranium materials. Because the expected ratio of n({sup 234}U)/n({sup 238}U) exceeded the dynamic measurement range of the Faraday detectors of the TIMS instrument, an experimental design using a combination of two detectors was developed. The n({sup 234}U)/n({sup 235}U) and n({sup 236}U)/n({sup 235}U) ratios were determined using ion counting in combination with the decelerating device. The n({sup 235}U)/n({sup 238}U) ratio was determined by the Faraday detector. This experimental design allowed the detector cross calibration to be circumvented. Precisions of less than 1 percent for the n({sup 234}U)/n({sup 235}U) ratios and 5-25 percent for the n({sup 236}U)/n({sup 235}U) ratios were achieved. The purpose of the study was to establish a register of isotopic signatures for natural uranium materials. The amount ratio, and isotopic composition of 18 ore concentrates, collected by the International

Detection of tiny amounts of fissile materials in large-sized containers with radioactive waste

Science.gov (United States)

Batyaev, V. F.; Skliarov, S. V.

2018-01-01

The paper is devoted to non-destructive control of tiny amounts of fissile materials in large-sized containers filled with radioactive waste (RAW). The aim of this work is to model an active neutron interrogation facility for detection of fissile ma-terials inside NZK type containers with RAW and determine the minimal detectable mass of U-235 as a function of various param-eters: matrix type, nonuniformity of container filling, neutron gen-erator parameters (flux, pulse frequency, pulse duration), meas-urement time. As a result the dependence of minimal detectable mass on fissile materials location inside container is shown. Nonu-niformity of the thermal neutron flux inside a container is the main reason of the space-heterogeneity of minimal detectable mass in-side a large-sized container. Our experiments with tiny amounts of uranium-235 (<1 g) confirm the detection of fissile materials in NZK containers by using active neutron interrogation technique.

Detection of tiny amounts of fissile materials in large-sized containers with radioactive waste

Directory of Open Access Journals (Sweden)

Batyaev V.F.

2018-01-01

Full Text Available The paper is devoted to non-destructive control of tiny amounts of fissile materials in large-sized containers filled with radioactive waste (RAW. The aim of this work is to model an active neutron interrogation facility for detection of fissile ma-terials inside NZK type containers with RAW and determine the minimal detectable mass of U-235 as a function of various param-eters: matrix type, nonuniformity of container filling, neutron gen-erator parameters (flux, pulse frequency, pulse duration, meas-urement time. As a result the dependence of minimal detectable mass on fissile materials location inside container is shown. Nonu-niformity of the thermal neutron flux inside a container is the main reason of the space-heterogeneity of minimal detectable mass in-side a large-sized container. Our experiments with tiny amounts of uranium-235 (<1 g confirm the detection of fissile materials in NZK containers by using active neutron interrogation technique.

Uranium. Resources, production and demand

International Nuclear Information System (INIS)

1997-01-01

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

235Uranium isotope abundance certified reference material for gamma spectrometry EC nuclear reference material 171 certification report

International Nuclear Information System (INIS)

De Bievre, P.; Eschbach, H.L.; Lesser, R.; Meyer, H.; Audenhove, Van J.

1986-01-01

This certification report contains the information necessary for the final certification of EC nuclear reference material 171. It is also intended to inform the user of the reference material concerned on technical/scientific details which are not given in the certificate. The report describes the reference material which consists of sets of U 3 O 8 samples with five different 235 U/U abundances, filled in cylindrical aluminium cans. The can bottom serves as window for emitted gamma radiation. The report describes how the 235 U/U abundances were characterized, how the other properties relevant for gamma measurements were determined and gives all connected results as well as those from the verification measurements. Appendix A represents the draft certificate. 32 refs

Management of depleted uranium

International Nuclear Information System (INIS)

2001-01-01

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

Development of metal uranium fuel and testing of construction materials (I-VI); Part I

International Nuclear Information System (INIS)

Mihajlovic, A.

1965-11-01

This project includes the following tasks: Study of crystallisation of metal melt and beta-alpha transforms in uranium and uranium alloys; Study of the thermal treatment influence on phase transformations and texture in uranium alloys; Radiation damage of metal uranium; Project related to irradiation of metal uranium in the reactor; Development of fuel element for nuclear reactors

Effectiveness of ProTaper Universal retreatment instruments used with rotary or reciprocating adaptive motion in the removal of root canal filling material.

Science.gov (United States)

Capar, I D; Arslan, H; Ertas, H; Gök, T; Saygılı, G

2015-01-01

To compare the effectiveness of ProTaper Universal retreatment instruments with continuous rotation and adaptive motion (AM; a modified reciprocating motion that combines rotational and reciprocating motion) in the removal of filling material. Mesiobuccal root canals in 36 mandibular first molars were instrumented up to size F2 with the ProTaper Universal instrument (Dentsply Maillefer, Ballaigues, Switzerland) and filled using sealer and ProTaper Universal F2 gutta-percha cones. Gutta-percha was then down-packed and the root canal backfilled using the extruder hand-piece of the Elements Obturation System (SybronEndo, Orange, CA, USA). The teeth were assigned to two groups (n = 18), and removal of the root fillings was performed using one of the following techniques: group 1) ProTaper Universal retreatment files used with rotational motion (RM) and group 2) ProTaper Universal retreatment files used with adaptive motion (AM) (600° clockwise/0° counter-clockwise to 370° clockwise/50° counter-clockwise). The teeth were sectioned, and both halves were analysed at 8 × magnification. The percentage of remaining filling material was recorded. The data were analysed statistically using the Student's t-test at a 95% confidence level (P ProTaper Universal retreatment files with adaptive motion removed more filling materials from root canals than the rotational movement. © 2014 International Endodontic Journal. Published by John Wiley & Sons Ltd.

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.

Two main types of uranium deposit within phanerozoic formations of Ukraine

International Nuclear Information System (INIS)

Shumlyanskiy, V.A.

1997-01-01

The two main types of uranium deposits occurring within Phanaerozoic formations of Ukraine are described. They consist of uraniferous bearing bitumen in the Upper Carboniferous to Lower Triassic red beds, and infiltration (roll front type) uranium ores, occurring in the sediments filling ancient Paleogene river valleys. The first deposit type include black to dark brown beds of disseminated to massive bitumen occurring respectively as ozyantraxolite and oxykerite. These beds include uranium, as well as other metals. This uranium mineralization is dated at 195 to 200 million years old. The second type includes infiltration deposits in Paleogene coal bearing sediments, with the uranium mineralization occurring in the upper part of the sequence. The sediments occur within paleovallyes eroded into the underlying crystalline basement of the Ukraine shield and its weathered crust. The paleovalleys extend to a depth of 70 to 90 metres. The coal bearing sediments are overlain by sediments of younger age. Several uranium deposits of the second type are known, including a few identified as being of industrial grade. (author). 7 figs

Stripping of Uranium (IV) from D2EHPA + TBP system with ammonium oxalate and its recovery as uranium peroxide

International Nuclear Information System (INIS)

Singh, D.K.; Singh, H.

2014-01-01

Uranium is an important fissile material for the generation of electricity by nuclear reactors. To obtain uranium as a final product meeting the stringent nuclear specifications, many process steps are involved starting from ore processing to the precipitation of yellow cake. Solvent extraction is one of the process industrially adopted worldwide to achieve such purity of uranium from leach liquor and usually uses amine or organophosphorus types of extractant depending upon the composition of feed material. In solvent extraction technique, stripping is a prominent hydrometallurgical operation which brings the metal values of interest in aqueous solution for further treatment. In the case of uranium, stripping is dependent on its oxidation state. For hexavalent state generally carbonate solutions are used, where as in the case of tetravalent form salt solution such as ammonium oxalate is effective. Use of ammonium oxalate as stripping agent for tetravalent uranium from pyrophosphoric acid has been reported in patent however the details are not disclosed. In the present investigation an effort has been made to investigate the stripping behaviour of uranium from a synthetically loaded synergistic solvent mixture of uranium in tetravalent state

Analyses of uranium in some phosphate commercial products

International Nuclear Information System (INIS)

Kamel, N.H.M.; Sohsah, M.; Mohammad, H.M.; Sadek, M.

2005-01-01

The raw materials used in manufacturing of phosphate fertilizer products were derived from rocks. Rocks contain a remarkable of natural radioactivity. Uranium and phosphorous were originally initiated at the same time of the initiated rocks. The purpose of this research is to investigate solubility of uranium phosphate species at the phosphate fertilizer samples, samples including; raw phosphate material, single super phosphates (SSP) granules and powdered, triple super phosphates (TSP) and phosphogypsum samples were obtained from Abu-Zabal factory in Egypt. Solubility of uranium phosphate species was estimated. It was found that, less than half of the uranium phosphate species are soluble in water. The soluble uranium may be enter into the food chains by plant. Therefore, restriction should be done in order to limit contamination of land and the public

Use of vacuum in processing of uranium

International Nuclear Information System (INIS)

Saify, M.T.; Rai, C.B.; Singh, S.P.; Singh, R.P.

2003-01-01

Full text: Natural uranium in the form of metal and alloys with suitable heat treatment are being used as fuel in research and some of the power reactors. The fuel is required to satisfy the purity specification from the criteria of neutron economy, corrosion resistance and fabricability. Uranium and its alloys fall under the category of reactive materials. They readily react with atmospheric air to form oxides. If molten uranium is exposed to atmosphere, it reacts violently with atmospheric gases and moisture, leading to explosion in extreme cases. Hence, protective inert atmosphere or high vacuum is required in processing of the materials especially during the melting and casting operation. Vacuum is preferred for melting and remelting of metals and alloys to remove the gaseous and high volatile impurities, to improve the mechanical properties of the material. Also, under vacuum sound castings are produced for further processing by mechanical working or use in casting forms. The addition of reactive alloying elements in uranium is efficiently carried out under vacuum. The paper highlights vacuum systems deployed and applications of vacuum in various operations involved in the processing of uranium and its alloys

Uranium and the War: The effects of depleted uranium weapons in Iraq

International Nuclear Information System (INIS)

Jon williams

2007-01-01

The U.S. Army revealed in March 2003 that it dropped between 320 and 390 tons of depleted uranium during the Gulf War-the first time the material was ever used in combat-and it is estimated that more still has been dropped during the current invasion, though there have been no official counts as yet. Nuclear weapons and nuclear power plants require highly radioactive uranium, so the uranium 238 is removed from the naturally occurring uranium by a process known as enrichment. Depleted uranium is the by-product of the uranium enrichment process. Depleted uranium was a major topic of discussion during a Feb. 24 forum at Santa Cruz with speakers from the Iraq Veterans Against War (IVAW). The panel consisted of five members of the IVAW chapter in Olympia, Washington who visited Santa Cruz as part of a speaking tour of the west coast. These members of the IVAW believe that their experiences in the Gulf War were the beginnings of what will be a long-term health problem in the region. A study conducted by the Pentagon in 2002 predicted that every future battlefield will be contaminated with depleted uranium. Up-to-date health information from Iraq is difficult to come by. But a November report from Al-jazeera concluded that the cancer rate in Iraq has increased tenfold, and the number of birth defects has multiplied fivefold times since the 1991 war. The increase is believed to be caused by depleted uranium.

New Technique for Speciation of Uranium in Sediments Following Acetate-Stimulated Bioremediation

Energy Technology Data Exchange (ETDEWEB)

2011-06-22

Acetate-stimulated bioremediation is a promising new technique for sequestering toxic uranium contamination from groundwater. The speciation of uranium in sediments after such bioremediation attempts remains unknown as a result of low uranium concentration, and is important to analyzing the stability of sequestered uranium. A new technique was developed for investigating the oxidation state and local molecular structure of uranium from field site sediments using X-Ray Absorption Spectroscopy (XAS), and was implemented at the site of a former uranium mill in Rifle, CO. Glass columns filled with bioactive Rifle sediments were deployed in wells in the contaminated Rifle aquifer and amended with a hexavalent uranium (U(VI)) stock solution to increase uranium concentration while maintaining field conditions. This sediment was harvested and XAS was utilized to analyze the oxidation state and local molecular structure of the uranium in sediment samples. Extended X-Ray Absorption Fine Structure (EXAFS) data was collected and compared to known uranium spectra to determine the local molecular structure of the uranium in the sediment. Fitting was used to determine that the field site sediments did not contain uraninite (UO{sub 2}), indicating that models based on bioreduction using pure bacterial cultures are not accurate for bioremediation in the field. Stability tests on the monomeric tetravalent uranium (U(IV)) produced by bioremediation are needed in order to assess the efficacy of acetate-stimulation bioremediation.