National Library of Energy BETA

Sample records for thoron isotopic uranium

  1. Measurement of Radon, Thoron, Isotopic Uranium and Thorium to Determine Occupational and Environmental Exposure and Risk at Fernald Feed Material Production Center

    SciTech Connect (OSTI)

    Naomi H. Harley, Ph.D.

    2004-07-01

    To develop a new and novel area and personal radon/thoron detector for both radon isotopes to better measure the exposure to low airborne concentrations of these gases at Fernald. These measurements are to be used to determine atmospheric dispersion and exposure to radon and thoron prior to and during retrieval and removal of the 4000 Ci of radium in the two silos at Fernald.

  2. Uranium Isotopic Assay Instrument

    SciTech Connect (OSTI)

    Anheier, Norman C.; Wojcik, Michael D.; Bushaw, Bruce A.

    2006-12-01

    The isotopic assay instrument under development at Pacific Northwest National Laboratory (PNNL) is capable of rapid prescreening to detect small and rare particles containing high concentrations of uranium in a heterogeneous sample. The isotopic measurement concept is based on laser vaporization of solid samples followed with sensitive isotope specific detection using either uranium atomic fluorescence emission or uranium atomic absorbance. Both isotopes are measured concurrently, following a single ablation laser pulse, using two external-cavity violet diode lasers. The simultaneous measurement of both isotopes enables the correlation of the fluorescence and absorbance signals on a shot-to-shot basis. This measurement approach demonstrated negligible channel crosstalk between isotopes. Rapid sample scanning provides high spatial resolution isotopic fluorescence and absorbance sample imagery of heterogeneous samples. Laser ablation combined with measurements of laser-induced fluorescence (LALIF) and through-plume laser absorbance (LAPLA) was applied to measure gadolinium isotope ratios in solid samples. Gadolinium has excitation wavelengths very close to the transitions of interest in uranium. Gadolinium has seven stable isotopes, and the natural 152Gd:160Gd ratio of 0.009 is in the range of what will be encountered for 235U:238U isotopic ratios. LAPLA measurements were demonstrated clearly using 152Gd (0.2% isotopic abundance) with a good signal-to-noise ratio. The ability to measure gadolinium abundances at this level indicates that measurements of 235U/238U isotopic ratios for natural (0.72%), depleted (0.25%), and low enriched uranium samples will be feasible.

  3. Uranium isotopes fingerprint biotic reduction

    SciTech Connect (OSTI)

    Stylo, Malgorzata; Neubert, Nadja; Wang, Yuheng; Monga, Nikhil; Romaniello, Stephen J.; Weyer, Stefan; Bernier-Latmani, Rizlan

    2015-04-20

    Knowledge of paleo-redox conditions in the Earth’s history provides a window into events that shaped the evolution of life on our planet. The role of microbial activity in paleo-redox processes remains unexplored due to the inability to discriminate biotic from abiotic redox transformations in the rock record. The ability to deconvolute these two processes would provide a means to identify environmental niches in which microbial activity was prevalent at a specific time in paleo-history and to correlate specific biogeochemical events with the corresponding microbial metabolism. Here, we demonstrate that the isotopic signature associated with microbial reduction of hexavalent uranium (U), i.e., the accumulation of the heavy isotope in the U(IV) phase, is readily distinguishable from that generated by abiotic uranium reduction in laboratory experiments. Thus, isotope signatures preserved in the geologic record through the reductive precipitation of uranium may provide the sought-after tool to probe for biotic processes. Because uranium is a common element in the Earth’s crust and a wide variety of metabolic groups of microorganisms catalyze the biological reduction of U(VI), this tool is applicable to a multiplicity of geological epochs and terrestrial environments. The findings of this study indicate that biological activity contributed to the formation of many authigenic U deposits, including sandstone U deposits of various ages, as well as modern, Cretaceous, and Archean black shales. In addition, engineered bioremediation activities also exhibit a biotic signature, suggesting that, although multiple pathways may be involved in the reduction, direct enzymatic reduction contributes substantially to the immobilization of uranium.

  4. Uranium isotopes fingerprint biotic reduction

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Stylo, Malgorzata; Neubert, Nadja; Wang, Yuheng; Monga, Nikhil; Romaniello, Stephen J.; Weyer, Stefan; Bernier-Latmani, Rizlan

    2015-04-20

    Knowledge of paleo-redox conditions in the Earth’s history provides a window into events that shaped the evolution of life on our planet. The role of microbial activity in paleo-redox processes remains unexplored due to the inability to discriminate biotic from abiotic redox transformations in the rock record. The ability to deconvolute these two processes would provide a means to identify environmental niches in which microbial activity was prevalent at a specific time in paleo-history and to correlate specific biogeochemical events with the corresponding microbial metabolism. Here, we demonstrate that the isotopic signature associated with microbial reduction of hexavalent uranium (U),more » i.e., the accumulation of the heavy isotope in the U(IV) phase, is readily distinguishable from that generated by abiotic uranium reduction in laboratory experiments. Thus, isotope signatures preserved in the geologic record through the reductive precipitation of uranium may provide the sought-after tool to probe for biotic processes. Because uranium is a common element in the Earth’s crust and a wide variety of metabolic groups of microorganisms catalyze the biological reduction of U(VI), this tool is applicable to a multiplicity of geological epochs and terrestrial environments. The findings of this study indicate that biological activity contributed to the formation of many authigenic U deposits, including sandstone U deposits of various ages, as well as modern, Cretaceous, and Archean black shales. In addition, engineered bioremediation activities also exhibit a biotic signature, suggesting that, although multiple pathways may be involved in the reduction, direct enzymatic reduction contributes substantially to the immobilization of uranium.« less

  5. RESOLUTION OF URANIUM ISOTOPES WITH KINETIC PHOSPHORESCENCE ANALYSIS

    SciTech Connect (OSTI)

    Miley, Sarah M.; Hylden, Anne T.; Friese, Judah I.

    2013-04-01

    This study was conducted to test the ability of the Chemchek™ Kinetic Phosphorescence Analyzer Model KPA-11 with an auto-sampler to resolve the difference in phosphorescent decay rates of several different uranium isotopes, and therefore identify the uranium isotope ratios present in a sample. Kinetic phosphorescence analysis (KPA) is a technique that provides rapid, accurate, and precise determination of uranium concentration in aqueous solutions. Utilizing a pulsed-laser source to excite an aqueous solution of uranium, this technique measures the phosphorescent emission intensity over time to determine the phosphorescence decay profile. The phosphorescence intensity at the onset of decay is proportional to the uranium concentration in the sample. Calibration with uranium standards results in the accurate determination of actual concentration of the sample. Different isotopes of uranium, however, have unique properties which should result in different phosphorescence decay rates seen via KPA. Results show that a KPA is capable of resolving uranium isotopes.

  6. Uranium molecular laser isotope separation

    SciTech Connect (OSTI)

    Jensen, R.J.; Sullivan, A.

    1982-01-01

    The Molecular Laser Isotope Separation program is moving into the engineering phase, and it is possible to determine in some detail the plant cost terms involved in the process economics. A brief description of the MLIS process physics is given as a motivation to the engineering and economics discussion. Much of the plant cost arises from lasers and the overall optical system. In the paper, the authors discuss lasers as operating units and systems, along with temporal multiplexing and Raman shifting. Estimates of plant laser costs are given.

  7. Uranium accountancy in Atomic Vapor Laser Isotope Separation

    SciTech Connect (OSTI)

    Carver, R.D.

    1986-01-01

    The AVLIS program pioneers the large scale industrial application of lasers to produce low cost enriched uranium fuel for light water reactors. In the process developed at Lawrence Livermore National Laboratory, normal uranium is vaporized by an electron beam, and a precisely tuned laser beam selectively photo-ionizes the uranium-235 isotopes. These ions are moved in an electromagnetic field to be condensed on the product collector. All other uranium isotopes remain uncharged and pass through the collector section to condense as tails. Tracking the three types of uranium through the process presents special problems in accountancy. After demonstration runs, the uranium on the collector was analyzed for isotopic content by Battelle Pacific Northwest Laboratory. Their results were checked at LLNL by analysis of parallel samples. The differences in isotopic composition as reported by the two laboratories were not significant.

  8. Evolution of isotopic composition of reprocessed uranium during the multiple recycling in light water reactors with natural uranium feed

    SciTech Connect (OSTI)

    Smirnov, A. Yu. Sulaberidze, G. A.; Alekseev, P. N.; Dudnikov, A. A.; Nevinitsa, V. A. Proselkov, V. N.; Chibinyaev, A. V.

    2012-12-15

    A complex approach based on the consistent modeling of neutron-physics processes and processes of cascade separation of isotopes is applied for analyzing physical problems of the multiple usage of reprocessed uranium in the fuel cycle of light water reactors. A number of scenarios of multiple recycling of reprocessed uranium in light water reactors are considered. In the process, an excess absorption of neutrons by the {sup 236}U isotope is compensated by re-enrichment in the {sup 235}U isotope. Specific consumptions of natural uranium for re-enrichment of the reprocessed uranium depending on the content of the {sup 232}U isotope are obtained.

  9. Uranium isotopes in ground water as a prospecting technique

    SciTech Connect (OSTI)

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

    1980-02-01

    The isotopic concentrations of dissolved uranium were determined for 300 ground water samples near eight known uranium accumulations to see if new approaches to prospecting could be developed. It is concluded that a plot of /sup 234/U//sup 238/U activity ratio (A.R.) versus uranium concentration (C) can be used to identify redox fronts, to locate uranium accumulations, and to determine whether such accumulations are being augmented or depleted by contemporary aquifer/ground water conditions. In aquifers exhibiting flow-through hydrologic systems, up-dip ground water samples are characterized by high uranium concentration values (> 1 to 4 ppB) and down-dip samples by low uranium concentration values (less than 1 ppB). The boundary between these two regimes can usually be identified as a redox front on the basis of regional water chemistry and known uranium accumulations. Close proximity to uranium accumulations is usually indicated either by very high uranium concentrations in the ground water or by a combination of high concentration and high activity ratio values. Ground waters down-dip from such accumulations often exhibit low uranium concentration values but retain their high A.R. values. This serves as a regional indicator of possible uranium accumulations where conditions favor the continued augmentation of the deposit by precipitation from ground water. Where the accumulation is being dispersed and depleted by the ground water system, low A.R. values are observed. Results from the Gulf Coast District of Texas and the Wyoming districts are presented.

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

    SciTech Connect (OSTI)

    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. % {sup 234}U, 19.8336 {+-} 0.0059 wt. % {sup 235}U, 0.1337 {+-} 0.0006 wt. % {sup 236}U, and 79.9171 {+-} 0.0057 wt. % {sup 238}U.

  11. Utilizing Isotopic Uranium Ratios in Groundwater Evaluations at FUSRAP Sites

    SciTech Connect (OSTI)

    Frederick, W.T.; Keil, K.G.; Rhodes, M.C.; Peterson, J.M.; MacDonell, M.M.

    2007-07-01

    The U.S. Army Corps of Engineers Buffalo District is evaluating environmental radioactive contamination at several Formerly Utilized Sites Remedial Action Program (FUSRAP) sites throughout New York, Pennsylvania, Ohio, and Indiana. The investigations follow the process defined in the Comprehensive Environmental Response, Compensation and Liability Act (CERCLA). Groundwater data from the Niagara Falls Storage Site (NFSS) in Lewiston, New York were evaluated for isotopic uranium ratios, specifically uranium-234 versus uranium-238 (U- 234 and U-238, respectively), and the results were presented at Waste Management 2006. Since uranium naturally occurs in all groundwater, it can be difficult to distinguish where low-concentration impacts from past releases differ from the high end of a site-specific natural background range. In natural groundwater, the ratio of U-234 to U-238 exceeds 1 (unity) due to the alpha particle recoil effect, in which U-234 is preferentially mobilized to groundwater from adjacent rock or soil. This process is very slow and may take hundreds to thousands of years before a measurable increase is seen in the natural isotopic ratio. If site releases are the source of uranium being measured in groundwater, the U-234 to U-238 ratio is commonly closer to 1, which normally reflects FUSRAP-related, uranium-contaminated wastes and soils. This lower ratio occurs because not enough residence time has elapsed since the 1940's and 1950's for the alpha particle recoil effect to have significantly altered the contamination-derived ratio. An evaluation of NFSS-specific and regional groundwater data indicate that an isotopic ratio of 1.2 has been identified as a signature value to help distinguish natural groundwater, which may have a broad background range, from zones impacted by past releases. (authors)

  12. Separation of uranium isotopes by chemical exchange

    DOE Patents [OSTI]

    Ogle, P.R. Jr.

    1974-02-26

    A chemical exchange method is provided for separating /sup 235/U from / sup 238/U comprising contacting a first phase containing UF/sub 6/ with a second phase containing a compound selected from the group consisting of NOUF/sub 6/, NOUF/sub 7/, and NO/sub 2/UF/sub 7/ until the U Fsub 6/ in the first phase becomes enriched in the /sup 235/U isotope. (Official Gazette)

  13. Third minima in thorium and uranium isotopes in a self-consistent...

    Office of Scientific and Technical Information (OSTI)

    Third minima in thorium and uranium isotopes in a self-consistent theory Title: Third ... Sponsoring Org: USDOE Country of Publication: United States Language: English Word Cloud ...

  14. Utilizing Isotopic Uranium Ratios in Groundwater Evaluations at NFSS

    SciTech Connect (OSTI)

    Rhodes, M.C.; Keil, K.G.; Frederick, W.T.; Papura, T.R.; Leithner, J.S.; Peterson, J.M.; MacDonell, M.M.

    2006-07-01

    The U.S. Army Corps of Engineers (USACE) Buffalo District is currently evaluating environmental contamination at the Niagara Falls Storage Site (NFSS) under the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) as part of its Formerly Utilized Sites Remedial Action Program (FUSRAP). The NFSS is located in the Town of Lewiston in western New York and has been used to store uranium-contaminated materials since 1944. Most of the radioactive materials are currently contained in an on-site structure, but past contamination remains in soil and groundwater. As a naturally occurring radionuclide, uranium is present in all groundwater. Because contamination levels at the site are quite low, it can be difficult to distinguish zones that have been impacted by the past releases from those at the high end of the natural background range. The differences in the isotopic ratio of uranium-234 (U-234) to uranium-238 (U-238) between natural groundwater systems and affected areas are being used in an innovative way to better define the nature and extent of groundwater contamination at NFSS. In natural groundwater, the ratio of U-234 to U-238 exceeds 1 due to the alpha particle recoil effect, in which U-234 is preferentially mobilized to groundwater from adjacent rock or soil. This process is very slow, and it can be hundreds to thousands of years before a measurable impact is seen in the isotopic ratio. Thus, as a result of the recoil effect, the ratio of U-234 to U-238 will be higher in natural groundwater than in contaminated groundwater. This means that if site releases were the source of the uranium being measured in groundwater at NFSS, the ratio of U-234 to U-238 would be expected to be very close to 1 (the same ratio that exists in wastes and soil at the site), because not enough time has elapsed for the alpha particle recoil effect to have significantly altered that ratio. From an evaluation of site and regional groundwater data, an isotopic ratio

  15. Femtosecond Laser Ablation Multicollector ICPMS Analysis of Uranium Isotopes in NIST Glass

    SciTech Connect (OSTI)

    Duffin, Andrew M.; Springer, Kellen WE; Ward, Jesse D.; Jarman, Kenneth D.; Robinson, John W.; Endres, Mackenzie C.; Hart, Garret L.; Gonzalez, Jhanis J.; Oropeza, Dayana; Russo, Richard; Willingham, David G.; Naes, Benjamin E.; Fahey, Albert J.; Eiden, Gregory C.

    2015-02-06

    We have utilized femtosecond laser ablation coupled to multi-collector inductively couple plasma mass spectrometry to measure the uranium isotopic content of NIST 61x (x=0,2,4,6) glasses. The uranium content of these glasses is a linear two-component mixing between isotopically natural uranium and the isotopically depleted spike used in preparing the glasses. Laser ablation results match extremely well, generally within a few ppm, with solution analysis following sample dissolution and chemical separation. In addition to isotopic data, sample utilization efficiency measurements indicate that over 1% of ablated uranium atoms reach a mass spectrometer detector, making this technique extremely efficient. Laser sampling also allows for spatial analysis and our data indicate that rare uranium concentration inhomogeneities exist in NIST 616 glass.

  16. Isotopic Analysis of Uranium in NIST SRM Glass by Femtosecond Laser Ablation

    SciTech Connect (OSTI)

    Duffin, Andrew M.; Hart, Garret L.; Hanlen, Richard C.; Eiden, Gregory C.

    2013-05-19

    We employed femtosecond Laser Ablation Multicollector Inductively Coupled Mass Spectrometry for the 11 determination of uranium isotope ratios in a series of standard reference material glasses (NIST 610, 612, 614, and 12 616). This uranium concentration in this series of SRM glasses is a combination of isotopically natural uranium in 13 the materials used to make the glass matrix and isotopically depleted uranium added to increase the uranium 14 elemental concentration across the series. Results for NIST 610 are in excellent agreement with literature values. 15 However, other than atom percent 235U, little information is available for the remaining glasses. We present atom 16 percent and isotope ratios for 234U, 235U, 236U, and 238U for all four glasses. Our results show deviations from the 17 certificate values for the atom percent 235U, indicating the need for further examination of the uranium isotopes in 18 NIST 610-616. Our results are fully consistent with a two isotopic component mixing between the depleted 19 uranium spike and natural uranium in the bulk glass.

  17. Innovative lasers for uranium isotope separation. [Progress report

    SciTech Connect (OSTI)

    Brake, M.L.; Gilgenbach, R.M.

    1991-06-01

    Copper vapor lasers have important applications to uranium atomic vapor laser isotope separation (AVLIS). The authors have spent the first two years of their project investigating two innovative methods of exciting/pumping copper vapor lasers which have the potential to improve the efficiency and scaling of large laser systems used in uranium isotope separation. Experimental research has focused on the laser discharge kinetics of (1) microwave, and (2) electron beam excitation/pumping of large-volume copper vapor lasers. During the first year, the experiments have been designed and constructed and initial data has been taken. During the second year these experiments have been diagnosed. Highlights of some of the second year results as well as plans for the future include the following: Microwave resonant cavity produced copper vapor plasmas at 2.45 GHz, have been investigated. A CW (0--500 W) signal heats and vaporizes the copper chloride to provide the atomic copper vapor. A pulsed (5 kW, 0.5--5kHz) signal is added to the incoming CW signal via a hybrid mixer to excite the copper states to the laser levels. An enhancement of the visible radiation has been observed during the pulsed pardon of the signal. Electrical probe measurements have been implemented on the system to verify the results of the electromagnetic model formulated last year. Laser gain measurements have been initiated with the use of a commercial copper vapor laser. Measurements of the spatial profile of the emission are also currently being made. The authors plan to increase the amount of pulsed microwave power to the system by implementing a high power magnetron. A laser cavity will be designed and added to this system.

  18. Laser isotope separation: Uranium. (Latest citations from the NTIS bibliographic database). Published Search

    SciTech Connect (OSTI)

    1995-12-01

    The bibliography contains citations concerning the technology and assessment of laser separation of uranium isotopes, compounds, oxides, and alloys. Topics include uranium enrichment plants, isotope enriched materials, gaseous diffusion, centrifuge enrichment, reliability and safety, and atomic vapor separation. Citations also discuss commercial enrichment, market trends, licensing, international competition, and waste management. (Contains 50-250 citations and includes a subject term index and title list.) (Copyright NERAC, Inc. 1995)

  19. Neutron-Induced Fission Cross Section Measurements for Uranium Isotopes and

    Office of Scientific and Technical Information (OSTI)

    Other Actinides at LANSCE (Conference) | SciTech Connect Neutron-Induced Fission Cross Section Measurements for Uranium Isotopes and Other Actinides at LANSCE Citation Details In-Document Search Title: Neutron-Induced Fission Cross Section Measurements for Uranium Isotopes and Other Actinides at LANSCE A well established program of neutron-induced fission cross section measurement at Los Alamos Neutron Science Center (LANSCE) is supporting the Fuel Cycle Research program (FC R&D). The

  20. Laser separation of uranium chosen for scaleup. [Atomic vapor laser isotope separation, molecular laser isotope separation, and plasma separation process

    SciTech Connect (OSTI)

    Rawls, R.L.

    1982-05-17

    Atomic vapor laser isotope separation (AVLIS) has been selected by the Department of Energy to go into large-scale engineering development and demonstration over two other advanced technologies, molecular laser isotope separation and plasma separation. DOE will continue to support development of another uranium enrichment technology, gas centrifugation. By or around 1990, the most promising gas centrifuge technique will be compared to the further developed AVLIS process, and a selection will be made between the two to replace the current technology, gaseous diffusion. The AVLIS process, plasma separation, and molecular laser isotope separation use the elective absorption of radiation of a particular energy level by the /sup 235/U isotope. The plasma separation process selectively energizes /sup 235/U by ion cyclotron resonance. The AVLIS and molecular laser isotope separation processes both use a carefully tuned laser to excite /sup 235/U isotope selectively. In the AVLIS process, uranium metal feed material is melted and vaporized to form an atomic uranium vapor stream. When this vapor stream passes through the beam of copper vapor lasers, the /sup 235/U atoms absorb the light and become ionized. These ionized atoms are collected by electromagnetic fields while the neutral /sup 238/U atoms pass through the magnetic field and are collected as tailings. The AVLIS process has the potential for significantly reducing the cost of enriching uranium. The status of dvelopment, cost, advantages and drawbacks of the five processes, (gaseous diffusion, gas centrifugation, AVLIS, molecular laser separation, plasma separation) are discussed. (ATT)

  1. Laser separation of uranium chosen for scaleup. [Atomic vapor laser isotope separation, molecular laser isotope separation plasma separation process

    SciTech Connect (OSTI)

    Rawls, R.L.

    1982-05-17

    Atomic vapor laser isotope separation (AVLIS) has been selected by the Department of Energy to go into large-scale engineering development and demonstration over two other advanced technologies, molecular laser isotope separation and plasma separation. DOE will continue to support development of another uranium enrichment technology, gas centrifugation. By or around 1990, the most promising gas centrifuge technique will be compared to the further developed AVLIS process, and a selection will be made between the two to replace the current technology, gaseous diffusion. The AVLIS process, plasma separation, and molecular laser isotope separation use the selective absorption of radiation of a particular energy level by the /sup 235/U isotope. The plasma separation process selectively energizes /sup 235/U by ion cyclotron resonance. The AVLIS and molecular laser isotope separation processes both use a carefully tuned laser to excite /sup 235/U isotope selectively. In the AVLIS process, uranium metal feed material is melted and vaporized to from an atomic uranium vapor stream. When this vapor stream passes through the beam of copper vapor lasers, the /sup 235/U atoms absorb the light and become ionized. These ionized atoms are collected by electromagnetic fields while the neutral /sup 238/U atoms pass through the magnetic field and are collected as tailings. The AVLIS process has the potential for significantly reducing the cost of enriching uranium. The status of development, cost, advantages and drawbacks of the five processes (gaseous diffusion, gas centrifugation, AVLIS, molecular laser separation, plasma separation) are discussed. (ATT)

  2. METHOD OF SEPARATING ISOTOPES OF URANIUM IN A CALUTRON

    DOE Patents [OSTI]

    Jenkins, F.A.

    1958-05-01

    Mass separation devices of the calutron type and the use of uranium hexachloride as a charge material in the calutron ion source are described. The method for using this material in a mass separator includes heating the uranium hexachloride to a temperature in the range of 60 to 100 d C in a vacuum and thereby forming a vapor of the material. The vaporized uranium hexachloride is then ionized in a vapor ionizing device for subsequent mass separation processing.

  3. Method and apparatus for storing hydrogen isotopes. [stored as uranium hydride in a block of copper

    DOE Patents [OSTI]

    McMullen, J.W.; Wheeler, M.G.; Cullingford, H.S.; Sherman, R.H.

    1982-08-10

    An improved method and apparatus for storing isotopes of hydrogen (especially tritium) are provided. The hydrogen gas is stored as hydrides of material (for example uranium) within boreholes in a block of copper. The mass of the block is critically important to the operation, as is the selection of copper, because no cooling pipes are used. Because no cooling pipes are used, there can be no failure due to cooling pipes. And because copper is used instead of stainless steel, a significantly higher temperature can be reached before the eutectic formation of uranium with copper occurs, (the eutectic of uranium with the iron in stainless steel forms at a significantly lower temperature).

  4. Investigating Uranium Isotopic Distributions in Environmental Samples Using AMS and MC-ICPMS

    SciTech Connect (OSTI)

    Buchholz, B A; Brown, T A; Hamilton, T F; Hutcheon, I D; Marchetti, A A; Martinelli, R E; Ramon, E C; Tumey, S J; Williams, R W

    2005-12-09

    Major, minor, and trace uranium isotopes were measured at Lawrence Livermore National Laboratory in environmentally acquired samples using different instruments to span large variations in concentrations. Multi-collector inductively-coupled plasma mass spectrometry (MC-ICPMS) can be used to measure major and minor isotopes: {sup 238}U, {sup 235}U, {sup 234}U and {sup 236}U. Accelerator mass spectrometry (AMS) can be used to measure minor and trace isotopes: {sup 234}U, {sup 236}U, and {sup 233}U. The main limit of quantification for minor or trace uranium isotopes is the abundance sensitivity of the measurement technique; i.e., the ability to measure a minor or trace isotope of mass M in the presence of a major isotope at M{+-}1 mass units. The abundance sensitivity for {sup 236}U/{sup 235}U isotope ratio measurements using MC-ICPMS is around {approx}2x10{sup -6}. This compares with a {sup 236}U/{sup 235}U abundance sensitivity of {approx}1x10{sup -7} for the current AMS system, with the expectation of 2-3 orders of magnitude improvement in sensitivity with the addition of another high energy filter. Comparing {sup 236}U/{sup 234}U from MC-ICPMS and AMS produced agreement within {approx}10% for samples at {sup 236}U levels high enough to be measurable by both techniques.

  5. Status of Uranium Atomic Vapor Laser Isotope Separation Program

    SciTech Connect (OSTI)

    Chen, Hao-Lin; Feinberg, R.M.

    1993-06-01

    This report discusses demonstrations of plant-scale hardware embodying AVLIS technology which were completed in 1992. These demonstrations, designed to provide key economic and technical bases for plant deployment, produced significant quantities of low enriched uranium which could be used for civilian power reactor fuel. We are working with industry to address the integration of AVLIS into the fuel cycle. To prepare for deployment, a conceptual design and cost estimate for a uranium enrichment plant were also completed. The U-AVLIS technology is ready for commercialization.

  6. Uranium atomic vapor laser isotope separation (AVL1S)

    SciTech Connect (OSTI)

    Beeler, R.G.; Heestand, G.M.

    1992-12-01

    The high cost associated with gaseous diffusion technology has fostered world-wide competition in the uranium enrichment market. Enrichment costs based on AVLIS technology are projected to be a factor of about three to five times lower. Full scale AVLIS equipment has been built and its performance is being demonstrated now at LLNL. An overview of the AVLIS process will be discussed and key process paramenters will be identified. Application of AVLIS technologies to non-uranium systems will also be highlighted. Finally, the vaporization process along with some key parameters will be discussed.

  7. Uranium Elemental and Isotopic Constraints on Groundwater Flow Beneath the Nopal I Uranium Deposit, Pena Blanca, Mexico

    SciTech Connect (OSTI)

    S.J. Goldstein; M.T. Murrell; A.M. Simmons

    2005-07-11

    The Nopal I uranium deposit in Chihuahua, Mexico, is an excellent analogue for evaluating the fate of spent fuel, associated actinides, and fission products over long time scales for the proposed Yucca Mountain high-level nuclear waste repository. In 2003, three groundwater wells were drilled directly adjacent to (PB-1) and 50 m on either side of the uranium deposit (PB-2 and PB-3) in order to evaluate uranium-series transport in three dimensions. After drilling, uranium concentrations were elevated in all of the three wells (0.1-18 ppm) due to drilling activities and subsequently decreased to {approx}5-20% of initial values over the next several months. The {sup 234}U/{sup 238}U activity ratios were similar for PB-1 and PB-2 (1.005 to 1.079) but distinct for PB-3 (1.36 to 1.83) over this time period, suggesting limited mixing between groundwater from these wells over these short time and length scales. Regional groundwater wells located up to several km from the deposit also have distinct uranium isotopic characteristics and constrain mixing over larger length and time scales. We model the decreasing uranium concentrations in the newly drilled wells with a simple one-dimensional advection-dispersion model, assuming uranium is introduced as a slug to each of the wells and transported as a conservative tracer. Using this model for our data, the relative uranium concentrations are dependent on both the longitudinal dispersion as well as the mean groundwater flow velocity. These parameters have been found to be correlated in both laboratory and field studies of groundwater velocity and dispersion (Klotz et al., 1980). Using typical relationships between velocity and dispersion for field and laboratory studies along with the relationship observed from our uranium data, both velocity (1-10 n/yr) and dispersion coefficient (1E-5 to 1E-2 cm{sup 2}/s) can be derived from the modeling. As discussed above, these relatively small flow velocities and dispersivities agree with

  8. Investigation of the Photochemical Method for Uranium Isotope Separation

    DOE R&D Accomplishments [OSTI]

    Urey, H. C.

    1943-07-10

    To find a process for successful photochemical separation of isotopes several conditions have to be fulfilled. First, the different isotopes have to show some differences in the spectrum. Secondly, and equally important, this difference must be capable of being exploited in a photochemical process. Parts A and B outline the physical and chemical conditions, and the extent to which one might expect to find them fulfilled. Part C deals with the applicability of the process.

  9. Uranium Isotopic Ratio Measurements of U3O8 Reference Materials by Atom Probe Tomography

    SciTech Connect (OSTI)

    Fahey, Albert J.; Perea, Daniel E.; Bartrand, Jonah AG; Arey, Bruce W.; Thevuthasan, Suntharampillai

    2016-01-01

    We report results of measurements of isotopic ratios obtained with atom probe tomography on U3O8 reference materials certified for their isotopic abundances of uranium. The results show good agreement with the certified values. High backgrounds due to tails from adjacent peaks complicate the measurement of the integrated peak areas as well as the fact that only oxides of uranium appear in the spectrum, the most intense of which is doubly charged. In addition, lack of knowledge of other instrumental parameters, such as the dead time, may bias the results. Isotopic ratio measurements can be performed at the nanometer-scale with the expectation of sensible results. The abundance sensitivity and mass resolving power of the mass spectrometer are not sufficient to compete with magnetic-sector instruments but are not far from measurements made by ToF-SIMS of other isotopic systems. The agreement of the major isotope ratios is more than sufficient to distinguish most anthropogenic compositions from natural.

  10. Isotopic Tracking of Hanford 300 Area Derived Uranium in the Columbia River

    SciTech Connect (OSTI)

    Christensen, John N.; Dresel, P. Evan; Conrad, Mark E.; Patton, Gregory W.; DePaolo, Donald J.

    2010-10-31

    Our objectives in this study are to quantify the discharge rate of uranium (U) to the Columbia River from the Hanford Site's 300 Area, and to follow that U down river to constrain its fate. Uranium from the Hanford Site has variable isotopic composition due to nuclear industrial processes carried out at the site. This characteristic makes it possible to use high-precision isotopic measurements of U in environmental samples to identify even trace levels of contaminant U, determine its sources, and estimate discharge rates. Our data on river water samples indicate that as much as 3.2 kg/day can enter the Columbia River from the 300 Area, which is only a small fraction of the total load of dissolved natural background U carried by the Columbia River. This very low-level of Hanford derived U can be discerned, despite dilution to < 1 percent of natural background U, 350 km downstream from the Hanford Site. These results indicate that isotopic methods can allow the amounts of U from the 300 Area of the Hanford Site entering the Columbia River to be measured accurately to ascertain whether they are an environmental concern, or are insignificant relative to natural uranium background in the Columbia River.

  11. Separation Of Uranium And Plutonium Isotopes For Measurement By Multi Collector Inductively Coupled Plasma Mass Spectroscopy

    SciTech Connect (OSTI)

    Martinelli, R E; Hamilton, T F; Williams, R W; Kehl, S R

    2009-03-29

    Uranium (U) and plutonium (Pu) isotopes in coral soils, contaminated by nuclear weapons testing in the northern Marshall Islands, were isolated by ion-exchange chromatography and analyzed by mass spectrometry. The soil samples were spiked with {sup 233}U and {sup 242}Pu tracers, dissolved in minerals acids, and U and Pu isotopes isolated and purified on commercially available ion-exchange columns. The ion-exchange technique employed a TEVA{reg_sign} column coupled to a UTEVA{reg_sign} column. U and Pu isotope fractions were then further isolated using separate elution schemes, and the purified fractions containing U and Pu isotopes analyzed sequentially using multi-collector inductively coupled plasma mass spectrometer (MCICP-MS). High precision measurements of {sup 234}U/{sup 235}U, {sup 238}U/{sup 235}U, {sup 236}U/{sup 235}U, and {sup 240}Pu/{sup 239}Pu in soil samples were attained using the described methodology and instrumentation, and provide a basis for conducting more detailed assessments of the behavior and transfer of uranium and plutonium in the environment.

  12. Process for producing enriched uranium having a {sup 235}U content of at least 4 wt. % via combination of a gaseous diffusion process and an atomic vapor laser isotope separation process to eliminate uranium hexafluoride tails storage

    DOE Patents [OSTI]

    Horton, J.A.; Hayden, H.W. Jr.

    1995-05-30

    An uranium enrichment process capable of producing an enriched uranium, having a {sup 235}U content greater than about 4 wt. %, is disclosed which will consume less energy and produce metallic uranium tails having a lower {sup 235}U content than the tails normally produced in a gaseous diffusion separation process and, therefore, eliminate UF{sub 6} tails storage and sharply reduce fluorine use. The uranium enrichment process comprises feeding metallic uranium into an atomic vapor laser isotope separation process to produce an enriched metallic uranium isotopic mixture having a {sup 235} U content of at least about 2 wt. % and a metallic uranium residue containing from about 0.1 wt. % to about 0.2 wt. % {sup 235} U; fluorinating this enriched metallic uranium isotopic mixture to form UF{sub 6}; processing the resultant isotopic mixture of UF{sub 6} in a gaseous diffusion process to produce a final enriched uranium product having a {sup 235}U content of at least 4 wt. %, and up to 93.5 wt. % or higher, of the total uranium content of the product, and a low {sup 235}U content UF{sub 6} having a {sup 235}U content of about 0.71 wt. % of the total uranium content of the low {sup 235}U content UF{sub 6}; and converting this low {sup 235}U content UF{sub 6} to metallic uranium for recycle to the atomic vapor laser isotope separation process. 4 figs.

  13. Process for producing enriched uranium having a .sup.235 U content of at least 4 wt. % via combination of a gaseous diffusion process and an atomic vapor laser isotope separation process to eliminate uranium hexafluoride tails storage

    DOE Patents [OSTI]

    Horton, James A.; Hayden, Jr., Howard W.

    1995-01-01

    An uranium enrichment process capable of producing an enriched uranium, having a .sup.235 U content greater than about 4 wt. %, is disclosed which will consume less energy and produce metallic uranium tails having a lower .sup.235 U content than the tails normally produced in a gaseous diffusion separation process and, therefore, eliminate UF.sub.6 tails storage and sharply reduce fluorine use. The uranium enrichment process comprises feeding metallic uranium into an atomic vapor laser isotope separation process to produce an enriched metallic uranium isotopic mixture having a .sup.235 U content of at least about 2 wt. % and a metallic uranium residue containing from about 0.1 wt. % to about 0.2 wt. % .sup.235 U; fluorinating this enriched metallic uranium isotopic mixture to form UF.sub.6 ; processing the resultant isotopic mixture of UF.sub.6 in a gaseous diffusion process to produce a final enriched uranium product having a .sup.235 U content of at least 4 wt. %, and up to 93.5 wt. % or higher, of the total uranium content of the product, and a low .sup.235 U content UF.sub.6 having a .sup.235 U content of about 0.71 wt. % of the total uranium content of the low .sup.235 U content UF.sub.6 ; and converting this low .sup.235 U content UF.sub.6 to metallic uranium for recycle to the atomic vapor laser isotope separation process.

  14. Rapid fusion method for the determination of refractory thorium and uranium isotopes in soil samples

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Maxwell, Sherrod L.; Hutchison, Jay B.; McAlister, Daniel R.

    2015-02-14

    Recently, approximately 80% of participating laboratories failed to accurately determine uranium isotopes in soil samples in the U.S Department of Energy Mixed Analyte Performance Evaluation Program (MAPEP) Session 30, due to incomplete dissolution of refractory particles in the samples. Failing laboratories employed acid dissolution methods, including hydrofluoric acid, to recover uranium from the soil matrix. The failures illustrate the importance of rugged soil dissolution methods for the accurate measurement of analytes in the sample matrix. A new rapid fusion method has been developed by the Savannah River National Laboratory (SRNL) to prepare 1-2 g soil sample aliquots very quickly, withmore » total dissolution of refractory particles. Soil samples are fused with sodium hydroxide at 600 ºC in zirconium crucibles to enable complete dissolution of the sample. Uranium and thorium are separated on stacked TEVA and TRU extraction chromatographic resin cartridges, prior to isotopic measurements by alpha spectrometry on cerium fluoride microprecipitation sources. Plutonium can also be separated and measured using this method. Batches of 12 samples can be prepared for measurement in <5 hours.« less

  15. Rapid fusion method for the determination of refractory thorium and uranium isotopes in soil samples

    SciTech Connect (OSTI)

    Maxwell, Sherrod L.; Hutchison, Jay B.; McAlister, Daniel R.

    2015-02-14

    Recently, approximately 80% of participating laboratories failed to accurately determine uranium isotopes in soil samples in the U.S Department of Energy Mixed Analyte Performance Evaluation Program (MAPEP) Session 30, due to incomplete dissolution of refractory particles in the samples. Failing laboratories employed acid dissolution methods, including hydrofluoric acid, to recover uranium from the soil matrix. The failures illustrate the importance of rugged soil dissolution methods for the accurate measurement of analytes in the sample matrix. A new rapid fusion method has been developed by the Savannah River National Laboratory (SRNL) to prepare 1-2 g soil sample aliquots very quickly, with total dissolution of refractory particles. Soil samples are fused with sodium hydroxide at 600 ºC in zirconium crucibles to enable complete dissolution of the sample. Uranium and thorium are separated on stacked TEVA and TRU extraction chromatographic resin cartridges, prior to isotopic measurements by alpha spectrometry on cerium fluoride microprecipitation sources. Plutonium can also be separated and measured using this method. Batches of 12 samples can be prepared for measurement in <5 hours.

  16. Rapid fusion method for the determination of refractory thorium and uranium isotopes in soil samples

    SciTech Connect (OSTI)

    Maxwell, Sherrod L.; Hutchison, Jay B.; McAlister, Daniel R.

    2015-02-14

    Recently, approximately 80% of participating laboratories failed to accurately determine uranium isotopes in soil samples in the U.S Department of Energy Mixed Analyte Performance Evaluation Program (MAPEP) Session 30, due to incomplete dissolution of refractory particles in the samples. Failing laboratories employed acid dissolution methods, including hydrofluoric acid, to recover uranium from the soil matrix. The failures illustrate the importance of rugged soil dissolution methods for the accurate measurement of analytes in the sample matrix. A new rapid fusion method has been developed by the Savannah River National Laboratory (SRNL) to prepare 1-2 g soil sample aliquots very quickly, with total dissolution of refractory particles. Soil samples are fused with sodium hydroxide at 600 C in zirconium crucibles to enable complete dissolution of the sample. Uranium and thorium are separated on stacked TEVA and TRU extraction chromatographic resin cartridges, prior to isotopic measurements by alpha spectrometry on cerium fluoride microprecipitation sources. Plutonium can also be separated and measured using this method. Batches of 12 samples can be prepared for measurement in <5 hours.

  17. Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual report for FY 2009

    SciTech Connect (OSTI)

    Chandler, David; Freels, James D; Ilas, Germina; Miller, James Henry; Primm, Trent; Sease, John D; Guida, Tracey; Jolly, Brian C

    2010-02-01

    This report documents progress made during FY 2009 in studies of converting the High Flux Isotope Reactor (HFIR) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in reactor performance from the current level. Results of selected benchmark studies imply that calculations of LEU performance are accurate. Studies are reported of the application of a silicon coating to surrogates for spheres of uranium-molybdenum alloy. A discussion of difficulties with preparing a fuel specification for the uranium-molybdenum alloy is provided. A description of the progress in developing a finite element thermal hydraulics model of the LEU core is provided.

  18. Establishing Specifications for Low Enriched Uranium Fuel Operations Conducted Outside the High Flux Isotope Reactor Site

    SciTech Connect (OSTI)

    Pinkston, Daniel; Primm, Trent; Renfro, David G; Sease, John D

    2010-10-01

    The National Nuclear Security Administration (NNSA) has funded staff at Oak Ridge National Laboratory (ORNL) to study the conversion of the High Flux Isotope Reactor (HFIR) from the current, high enriched uranium fuel to low enriched uranium fuel. The LEU fuel form is a metal alloy that has never been used in HFIR or any HFIR-like reactor. This report provides documentation of a process for the creation of a fuel specification that will meet all applicable regulations and guidelines to which UT-Battelle, LLC (UTB) the operating contractor for ORNL - must adhere. This process will allow UTB to purchase LEU fuel for HFIR and be assured of the quality of the fuel being procured.

  19. Quantifying Uranium Isotope Ratios Using Resonance Ionization Mass Spectrometry: The Influence of Laser Parameters on Relative Ionization Probability

    SciTech Connect (OSTI)

    Isselhardt, B H

    2011-09-06

    Resonance Ionization Mass Spectrometry (RIMS) has been developed as a method to measure relative uranium isotope abundances. In this approach, RIMS is used as an element-selective ionization process to provide a distinction between uranium atoms and potential isobars without the aid of chemical purification and separation. We explore the laser parameters critical to the ionization process and their effects on the measured isotope ratio. Specifically, the use of broad bandwidth lasers with automated feedback control of wavelength was applied to the measurement of {sup 235}U/{sup 238}U ratios to decrease laser-induced isotopic fractionation. By broadening the bandwidth of the first laser in a 3-color, 3-photon ionization process from a bandwidth of 1.8 GHz to about 10 GHz, the variation in sequential relative isotope abundance measurements decreased from >10% to less than 0.5%. This procedure was demonstrated for the direct interrogation of uranium oxide targets with essentially no sample preparation. A rate equation model for predicting the relative ionization probability has been developed to study the effect of variation in laser parameters on the measured isotope ratio. This work demonstrates that RIMS can be used for the robust measurement of uranium isotope ratios.

  20. DESIGN STUDY FOR A LOW-ENRICHED URANIUM CORE FOR THE HIGH FLUX ISOTOPE REACTOR, ANNUAL REPORT FOR FY 2010

    SciTech Connect (OSTI)

    Cook, David Howard; Freels, James D; Ilas, Germina; Jolly, Brian C; Miller, James Henry; Primm, Trent; Renfro, David G; Sease, John D; Pinkston, Daniel

    2011-02-01

    This report documents progress made during FY 2010 in studies of converting the High Flux Isotope Reactor (HFIR) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current level. Studies are reported of support to a thermal hydraulic test loop design, the implementation of finite element, thermal hydraulic analysis capability, and infrastructure tasks at HFIR to upgrade the facility for operation at 100 MW. A discussion of difficulties with preparing a fuel specification for the uranium-molybdenum alloy is provided. Continuing development in the definition of the fuel fabrication process is described.

  1. Third Minima in Thorium and Uranium Isotopes in a Self-Consistent Theory

    SciTech Connect (OSTI)

    McDonnell, J. D.

    2013-01-01

    Background: Well-developed third minima, corresponding to strongly elongated and reflection-asymmetric shapes associated with dimolecular configurations, have been predicted in some non-self-consistent models to impact fission pathways of thorium and uranium isotopes. These predictions have guided the interpretation of resonances seen experimentally. On the other hand, self-consistent calculations consistently predict very shallow potential-energy surfaces in the third minimum region.

    Purpose: We investigate the interpretation of third-minimum configurations in terms of dimolecular (cluster) states. We study the isentropic potential-energy surfaces of selected even-even thorium and uranium isotopes at several excitation energies. In order to understand the driving effects behind the presence of third minima, we study the interplay between pairing and shell effects.

    Methods: We use the finite-temperature superfluid nuclear density functional theory. We consider two Skyrme energy density functionals: a traditional functional SkM and a recent functional UNEDF1 optimized for fission studies.

    Results: We predict very shallow or no third minima in the potential-energy surfaces of 232Th and 232U. In the lighter Th and U isotopes with N = 136 and 138, the third minima are better developed. We show that the reflection-asymmetric configurations around the third minimum can be associated with dimolecular states involving the spherical doubly magic 132Sn and a lighter deformed Zr or Mo fragment. The potential-energy surfaces for 228,232Th and 232U at several excitation energies are presented. We also study isotopic chains to demonstrate the evolution of the depth of the third minimum with neutron number.

    Conclusions: We show that the neutron shell effect that governs the existence of the dimolecular states around the third minimum is consistent with the spherical-to-deformed shape transition in the Zr andMo isotopes around N = 58.We demonstrate that the depth of

  2. Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2008

    SciTech Connect (OSTI)

    Primm, Trent [ORNL; Chandler, David [ORNL; Ilas, Germina [ORNL; Miller, James Henry [ORNL; Sease, John D [ORNL; Jolly, Brian C [ORNL

    2009-03-01

    This report documents progress made during FY 2008 in studies of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in reactor performance from the current level. Results of selected benchmark studies imply that calculations of LEU performance are accurate. Scoping experiments with various manufacturing methods for forming the LEU alloy profile are presented.

  3. Fuel Grading Study on a Low-Enriched Uranium Fuel Design for the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Ilas, Germina; Primm, Trent

    2009-11-01

    An engineering design study that would enable the conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium to low-enriched uranium fuel is ongoing at Oak Ridge National Laboratory. The computational models used to search for a low-enriched uranium (LEU) fuel design that would meet the requirements for the conversion study, and the recent results obtained with these models during FY 2009, are documented and discussed in this report. Estimates of relevant reactor performance parameters for the LEU fuel core are presented and compared with the corresponding data for the currently operating high-enriched uranium fuel core. These studies indicate that the LEU fuel design would maintain the current performance of the HFIR with respect to the neutron flux to the central target region, reflector, and beam tube locations.

  4. Site evaluations for the uranium-atomic vapor laser isotope separation (U-AVLIS) production plant

    SciTech Connect (OSTI)

    Wolsko, T.; Absil, M.; Cirillo, R.; Folga, S.; Gillette, J.; Habegger, L.; Whitfield, R.

    1991-07-01

    This report describes a uranium-atomic vapor laser isotope separation (U-AVLIS) production plant siting study conducted during 1990 to identify alternative plant sites for examination in later environmental impact studies. A siting study methodology was developed in early 1990 and was implemented between June and December. This methodology had two parts. The first part -- a series of screening analyses that included exclusionary and other criteria -- was conducted to identify a reasonable number of candidates sites. This slate of candidate sites was then subjected to more rigorous and detailed comparative analysis for the purpose of developing a short list of reasonable alternative sites for later environmental examination. To fully appreciate the siting study methodology, it is important to understand the U-AVLIS program and site requirements. 16 refs., 29 figs., 54 tabs.

  5. Uranium, thorium isotopic analyses and uranium-series ages of calcite and opal, and stable isotopic compositions of calcite from drill cores UE25a No. 1, USW G-2 and USW G-3/GU-3, Yucca Mountain, Nevada

    SciTech Connect (OSTI)

    Szabo, B.J.; Kyser, T.K.

    1985-12-31

    Fracture and cavity filling calcite and opal in the unsaturated zone of three drill cores at Yucca Mountain were analyzed for uranium and stable isotope contents, and were dated by the uranium-series method. Stable isotope data indicate that the water from which the calcite precipitated was meteoric in origin. The decrease in {sup 18}O and increase in {sup 13}C with depth are interpreted as being due to the increase in temperature in drill holes corresponding to an estimated maximum geothermal gradient of 43{sup 0} per km. Of the eighteen calcite and opal deposits dated, four of the calcite and all four of the opal deposits yield dates older than 400,000 years and ten of the remaining calcite deposits yield dates between 26,000 and 310,000 years. The stable isotope and uranium data together with the finite uranium-series dates of precipitation suggest a complex history of fluid movements, rock and water interactions, and episodes of fracture filling during the last 310,000 years. 10 refs., 7 figs., 4 tabs.

  6. Innovative lasers for uranium isotope separation. Final report, September 1, 1989--April 1, 1993

    SciTech Connect (OSTI)

    Brake, M.L.; Gilgenbach, R.M.

    1993-07-01

    Copper vapor laser have important applications to uranium atomic vapor laser isotope separation (AVLIS). We have investigated two innovative methods of exciting/pumping copper vapor lasers which have the potential to improve the efficiency and scaling of large laser systems used in uranium isotope separation. Experimental research has focused on the laser discharge kinetics of (1) microwave, and (2) electron beam excitation/pumping of large-volume copper vapor lasers. Microwave resonant cavity produced copper vapor plasmas at 2.45 GHz, have been investigated in three separate experimental configurations. The first examined the application of CW (0-500W) power and was found to be an excellent method for producing an atomic copper vapor from copper chloride. The second used a pulsed (5kW, 0.5--5 kHz) signal superimposed on the CW signal to attempt to produce vaporization, dissociation and excitation to the laser states. Enhanced emission of the optical radiation was observed but power densities were found to be too low to achieve lasing. In a third experiment we attempted to increase the applied power by using a high power magnetron to produce 100 kW of pulsed power. Unfortunately, difficulties with the magnetron power supply were encountered leaving inconclusive results. Detailed modeling of the electromagnetics of the system were found to match the diagnostics results well. An electron beam pumped copper vapor system (350 kV, 1.0 kA, 300 ns) was investigated in three separate copper chloride heating systems, external chamber, externally heated chamber and an internally heated chamber. Since atomic copper spectral lines were not observed, it is assumed that a single pulse accelerator is not capable of both dissociating the copper chloride and exciting atomic copper and a repetitively pulsed electron beam generator is needed.

  7. FRAM isotopic analysis of uranium in thick-walled containers using high energy gamma rays and planar HPGe detectors.

    SciTech Connect (OSTI)

    Sampson, Thomas E.; Hypes, P. A.; Vo, Duc T.

    2002-01-01

    We describe the use of the Los Alamos FRAM isotopic analysis software to make the first reported measurements on thick-walled UF{sub 6} cylinders using small planar HPGe detectors of the type in common use at the IAEA. Heretofore, planar detector isotopic analysis measurements on uranium have used the 100-keV region and can be defeated by 10 mm of steel absorber. The analysis of planar detector measurements through 13-16 mm of steel shows that FRAM can successfully carry out these measurements and analysis in the 120-1024 keV energy range, a range previously thought to be the sole province of more efficient coaxial detectors. This paper describes the measurement conditions and results and also compares the results to other FRAM measurements with coaxial HPGe detectors. The technique of gamma-ray isotopic analysis of arbitrary samples is desirable for measuring the isotopic composition of uranium in UF{sub 6} cylinders because it does not require calibration with standards or knowledge of the cylinder wall thickness. The International Atomic Energy Agency (IAEA) uses the MGAU (Multi Group Analysis Uranium) uranium isotopic analysis software with planar high purity germanium (HPGe) detectors to measure the isotopic composition of uranium. Measurements on UF{sub 6} cylinders with 13-16-mm thick steel walls are usually unsuccessful because of the strong absorption of the 89-100 keV gamma rays and x-rays that MGAU requires for the measurement. This paper describes the use of the Los Alamos FRAM isotopic analysis software to make these measurements on UF{sub 6} cylinders. Uranium measurements with FRAM typically cover the energy range from 120-1001 keV and can easily be made through the walls of UF{sub 6} cylinders. While these measurements are usually performed with efficient coaxial HPGe detectors, this paper reports the first successful measurements using small planar HPGe detectors of the type in common use at the IAEA.

  8. uranium

    National Nuclear Security Administration (NNSA)

    to prepare surplus plutonium for disposition, and readiness to begin the Second Uranium Cycle, to start processing spent nuclear fuel.

    H Canyon is also being...

  9. Determination of the origin of elevated uranium at a Former Air Force Landfill using non-parametric statistics analysis and uranium isotope ratio analysis

    SciTech Connect (OSTI)

    Weismann, J.; Young, C.; Masciulli, S.; Caputo, D.

    2007-07-01

    factors so that gross alpha action levels can be applied to future long-term landfill monitoring to track radiological conditions at lower cost. Ratios of isotopic uranium results were calculated to test whether the elevated uranium displayed signatures indicative of military use. Results of all ratio testing strongly supports the conclusion that the uranium found in groundwater, surface water, and sediment at OU 2 is naturally-occurring and has not undergone anthropogenic enrichment or processing. U-234:U-238 ratios also show that a disequilibrium state, i.e., ratio greater than 1, exists throughout OU 2 which is indicative of long-term aqueous transport in aged aquifers. These results all support the conclusion that the elevated uranium observed at OU 2 is due to the high concentrations in the regional watershed. Based on the results of this monitoring program, we concluded that the elevated uranium concentrations measured in OU 2 groundwater, surface water, and sediment are due to the naturally-occurring uranium content of the regional watershed and are not the result of waste burials in the former landfill. Several lines of evidence indicate that natural uranium has been naturally concentrated beneath OU 2 in the geologic past and the higher of uranium concentrations in down-gradient wells is the result of geochemical processes and not the result of a uranium ore disposal. These results therefore provide the data necessary to support radiological closure of OU 2. (authors)

  10. PREPARING THE HIGH FLUX ISOTOPE REACTOR FOR CONVERSION TO LOW ENRICHED URANIUM FUEL ? RETURN TO 100 MW

    SciTech Connect (OSTI)

    Smith, Kevin Arthur [ORNL; Primm, Trent [ORNL

    2009-01-01

    The feasibility of low-enriched uranium (LEU) fuel as a replacement for the current, high enriched uranium (HEU) fuel for the High Flux Isotope Reactor (HFIR) has been under study since 2006. Reactor performance studies have been completed for conceptual plate designs and show that maintaining reactor performance while converting to LEU fuel requires returning the reactor power to 100 MW from 85 MW. The analyses required to up-rate the reactor power and the methods to perform these analyses are discussed. Comments regarding the regulatory approval process are provided along with a conceptual schedule.

  11. Isotopic evidence for reductive immobilization of uranium across a roll-front mineral deposit

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Brown, Shaun T.; Basu, Anirban; Christensen, John N.; Reimus, Paul; Heikoop, Jeffrey; Simmons, Ardyth; Woldegabriel, Giday; Maher, Kate; Weaver, Karrie; Clay, James; et al

    2016-05-20

    We use uranium (U) isotope ratios to detect and quantify the extent of natural U reduction in groundwater across a roll front redox gradient. Our study was conducted at the Smith Ranch-Highland in situ recovery (ISR) U mine in eastern Wyoming, USA, where economic U deposits occur in the Paleocene Fort Union formation. To evaluate the fate of aqueous U in and adjacent to the ore body, we investigated the chemical composition and isotope ratios of groundwater samples from the roll-front type ore body and surrounding monitoring wells of a previously mined area. The 238U/235U of groundwater varies by approximatelymore » 3‰ and is correlated with U concentrations. Fluid samples down-gradient of the ore zone are the most depleted in 238U and have the lowest U concentrations. Activity ratios of 234U/238U are ~5.5 up-gradient of the ore zone, ~1.0 in the ore zone, and between 2.3 and 3.7 in the down-gradient monitoring wells. High-precision measurements of 234U/238U and 238U/235U allow for development of a conceptual model that evaluates both the migration of U from the ore body and the extent of natural attenuation due to reduction. We find that the premining migration of U down-gradient of the delineated ore body is minimal along eight transects due to reduction in or adjacent to the ore body, whereas two other transects show little or no sign of reduction in the down-gradient region. Lastly, these results suggest that characterization of U isotopic ratios at the mine planning stage, in conjunction with routine geochemical analyses, can be used to identify where more or less postmining remediation will be necessary.« less

  12. Polyatomic interferences on high precision uranium isotope ratio measurements by MC-ICP-MS: Applications to environmental sampling for nuclear safeguards

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Pollington, Anthony D.; Kinman, William S.; Hanson, Susan K.; Steiner, Robert E.

    2015-09-04

    Modern mass spectrometry and separation techniques have made measurement of major uranium isotope ratios a routine task; however accurate and precise measurement of the minor uranium isotopes remains a challenge as sample size decreases. One particular challenge is the presence of isobaric interferences and their impact on the accuracy of minor isotope 234U and 236U measurements. Furthermore, we present techniques used for routine U isotopic analysis of environmental nuclear safeguards samples and evaluate polyatomic interferences that negatively impact accuracy as well as methods to mitigate their impacts.

  13. Polyatomic interferences on high precision uranium isotope ratio measurements by MC-ICP-MS: Applications to environmental sampling for nuclear safeguards

    SciTech Connect (OSTI)

    Pollington, Anthony D.; Kinman, William S.; Hanson, Susan K.; Steiner, Robert E.

    2015-09-04

    Modern mass spectrometry and separation techniques have made measurement of major uranium isotope ratios a routine task; however accurate and precise measurement of the minor uranium isotopes remains a challenge as sample size decreases. One particular challenge is the presence of isobaric interferences and their impact on the accuracy of minor isotope 234U and 236U measurements. Furthermore, we present techniques used for routine U isotopic analysis of environmental nuclear safeguards samples and evaluate polyatomic interferences that negatively impact accuracy as well as methods to mitigate their impacts.

  14. Uranium isotope exchange between gaseous UF{sub 6} and solid UF{sub 5}

    SciTech Connect (OSTI)

    Yato, Yumio; Kishimoto, Yoichiro; Sasao, Nobuyuki; Suto, Osamu; Funasaka, Hideyuki

    1996-08-01

    Based on a collision model, a new rate equation is derived for uranium isotope exchange between gaseous UF{sub 6} and solid UF{sub 5} by considering the number of UF{sub 5} molecules on the solid surface to be dependent on time. The reaction parameters included in the equation are determined from the experimental data and compared with the previous ones. A remarkable agreement is found between the particle sizes of UF{sub 5} estimated from the reaction parameter and from the direct observation with an electron microscope. The rate equation given in this work fully satisfies the related mass conservation and furthermore includes explicitly the terms related to the UF{sub 6} density and the mean size of UF{sub 5} particles, both of which are considered to cause an important effect on the reaction. This remarkable feature facilitates the simulation studies on this reaction under various conditions. The long term behavior of a simulated exchange reaction is studied under the condition considered to be close to that in a recovery zone of the MLIS process. The result indicates that the reaction is virtually limited to the solid surface under this conditions and thus the depletion of {sup 235}UF{sub 5} concentration averaged over the whole UF{sub 5} particles is not significant even after 200 h of the exchange reaction.

  15. Isolation and Puification of Uranium Isotopes for Measurement by Mass-Spectrometry (233, 234, 235, 236, 238U) and Alpha Spectrometry (232U)

    SciTech Connect (OSTI)

    Marinelli, R; Hamilton, T; Brown, T; Marchetti, A; Williams, R; Tumey, S

    2006-05-30

    This report describes a standardized methodology used by researchers from the Center for Accelerator Mass Spectrometry (CAMS) (Energy and Environment Directorate) and the Environmental Radiochemistry Group (Chemistry and Materials Science Directorate) at the Lawrence Livermore National Laboratory (LLNL) for the full isotopic analysis of uranium from solution. The methodology has largely been developed for use in characterizing the uranium composition of selected nuclear materials but may also be applicable to environmental studies and assessments of public, military or occupational exposures to uranium using in-vitro bioassay monitoring techniques. Uranium isotope concentrations and isotopic ratios are measured using a combination of Multi Collector Inductively Coupled Plasma Mass Spectrometry (MC ICP-MS), Accelerator Mass Spectrometry (AMS) and Alpha Spectrometry.

  16. U235: a gamma ray analysis code for uranium isotopic determination...

    Office of Scientific and Technical Information (OSTI)

    Subject: 44 INSTRUMENTATION, INCLUDING NUCLEAR AND PARTICLE DETECTORS; 99 MATHEMATICS, COMPUTERS, INFORMATION SCIENCE, MANAGEMENT, LAW, MISCELLANEOUS; GAMMA RADIATION; URANIUM 235; ...

  17. Activities of isotopes of uranium, thorium, radium, and lead of Antarctic coals

    SciTech Connect (OSTI)

    Gilbert, G.E.; Casella, V.R.; Bishop, C.T.

    1984-06-11

    Five coal samples from the Transantarctic Mountains of Antarctica were analyzed for the primordial radionuclides uranium-238, thorium-232, and uranium-235, the parents of the naturally occurring uranium, thorium, and actinium disintegration series, respectively. In most of the samples, the radionuclides uranium-234, thorium-230 radium-226, and lead-210 of the uranium series and radium-228 and thorium-228 of the thorium series were also measured and found to be in secular equilibrium with their parents. Activities of uranium-238, thorium-232, and uranium-235 ranged from about 0.1 to 1.0 pCi/g, 0.09 to 1.1 pCi/g, and 0.006 to 0.06 pCi/g, respectively. These results are comparable to corresponding activities reported for United States coals. 10 references, 3 tables.

  18. Determination of total and isotopic uranium by inductively coupled plasma-mass spectrometry at the Fernald Environmental Management Project

    SciTech Connect (OSTI)

    Miller, F.L.; Bolin, R.N.; Feller, M.T.; Danahy, R.J.

    1995-04-01

    At the Fernald Environmental Management Project (FEMP) in southwestern Ohio, ICP-mass spectrometry (ICP-MS), with sample introduction by peristaltic pumping, is used to determine total and isotopic uranium (U-234, U-235, U-236 and U-238) in soil samples. These analyses are conducted in support of the environmental cleanup of the FEMP site. Various aspects of the sample preparation and instrumental analysis will be discussed. Initial sample preparation consists of oven drying to determine moisture content, and grinding and rolling to homogenize the sample. This is followed by a nitric/hydrofluoric acid digestion to bring the uranium in the sample into solution. Bismuth is added to the sample prior to digestion to monitor for losses. The total uranium (U-238) content of this solution and the U{sup 235}/U{sup 238} ratio are measured on the first pass through the ICP-MS. To determine the concentration of the less abundant U{sup 234} and U{sup 236} isotopes, the digestate is further concentrated by using Eichrom TRU-Spec extraction columns before the second pass through the ICP-MS. Quality controls for both the sample preparation and instrumental protocols will also be discussed. Finally, an explanation of the calculations used to report the data in either weight percent or activity units will be given.

  19. Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2006

    SciTech Connect (OSTI)

    Primm, R. T.; Ellis, R. J.; Gehin, J. C.; Clarno, K. T.; Williams, K. A.; Moses, D. L.

    2006-11-01

    Neutronics and thermal-hydraulics studies show that, for equivalent operating power [85 MW(t)], a low-enriched uranium (LEU) fuel cycle based on uranium-10 wt % molybdenum (U-10Mo) metal foil with radially, continuously graded fuel meat thickness results in a 15% reduction in peak thermal flux in the beryllium reflector of the High Flux Isotope Reactor (HFIR) as compared to the current highly enriched uranium (HEU) cycle. The uranium-235 content of the LEU core is almost twice the amount of the HEU core when the length of the fuel cycle is kept the same for both fuels. Because the uranium-238 content of an LEU core is a factor of 4 greater than the uranium-235 content, the LEU HFIR core would weigh 30% more than the HEU core. A minimum U-10Mo foil thickness of 84 ?m is required to compensate for power peaking in the LEU core although this value could be increased significantly without much penalty. The maximum U-10Mo foil thickness is 457?m. Annual plutonium production from fueling the HFIR with LEU is predicted to be 2 kg. For dispersion fuels, the operating power for HFIR would be reduced considerably below 85 MW due to thermal considerations and due to the requirement of a 26-d fuel cycle. If an acceptable fuel can be developed, it is estimated that $140 M would be required to implement the conversion of the HFIR site at Oak Ridge National Laboratory from an HEU fuel cycle to an LEU fuel cycle. To complete the conversion by fiscal year 2014 would require that all fuel development and qualification be completed by the end of fiscal year 2009. Technological development areas that could increase the operating power of HFIR are identified as areas for study in the future.

  20. Low-Enriched Uranium Fuel Conversion Activities for the High Flux Isotope Reactor, Annual Report for FY 2011

    SciTech Connect (OSTI)

    Renfro, David G; Cook, David Howard; Freels, James D; Griffin, Frederick P; Ilas, Germina; Sease, John D; Chandler, David

    2012-03-01

    This report describes progress made during FY11 in ORNL activities to support converting the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum (UMo) alloy. With both radial and axial contouring of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current levels achieved with HEU fuel. Studies are continuing to demonstrate that the fuel thermal safety margins can be preserved following conversion. Studies are also continuing to update other aspects of the reactor steady state operation and accident response for the effects of fuel conversion. Technical input has been provided to Oregon State University in support of their hydraulic testing program. The HFIR conversion schedule was revised and provided to the GTRI program. In addition to HFIR conversion activities, technical support was provided directly to the Fuel Fabrication Capability program manager.

  1. Low-Enriched Uranium Fuel Design with Two-Dimensional Grading for the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Ilas, Germina [ORNL; Primm, Trent [ORNL

    2011-05-01

    An engineering design study of the conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel is ongoing at Oak Ridge National Laboratory. The computational models developed during fiscal year 2010 to search for an LEU fuel design that would meet the requirements for the conversion and the results obtained with these models are documented and discussed in this report. Estimates of relevant reactor performance parameters for the LEU fuel core are presented and compared with the corresponding data for the currently operating HEU fuel core. The results obtained indicate that the LEU fuel design would maintain the current performance of the HFIR with respect to the neutron flux to the central target region, reflector, and beam tube locations under the assumption that the operating power for the reactor fueled with LEU can be increased from the current value of 85 MW to 100 MW.

  2. NGSI FY15 Final Report. Innovative Sample Preparation for in-Field Uranium Isotopic Determinations

    SciTech Connect (OSTI)

    Yoshida, Thomas M.; Meyers, Lisa

    2015-11-10

    Our FY14 Final Report included an introduction to the project, background, literature search of uranium dissolution methods, assessment of commercial off the shelf (COTS) automated sample preparation systems, as well as data and results for dissolution of bulk quantities of uranium oxides, and dissolution of uranium oxides from swipe filter materials using ammonium bifluoride (ABF). Also, discussed were reaction studies of solid ABF with uranium oxide that provided a basis for determining the ABF/uranium oxide dissolution mechanism. This report details the final experiments for optimizing dissolution of U3O8 and UO2 using ABF and steps leading to development of a Standard Operating Procedure (SOP) for dissolution of uranium oxides on swipe filters.

  3. Uranium

    SciTech Connect (OSTI)

    Gabelman, J.W.; Chenoweth, W.L.; Ingerson, E.

    1981-10-01

    The uranium production industry is well into its third recession during the nuclear era (since 1945). Exploration is drastically curtailed, and many staffs are being reduced. Historical market price production trends are discussed. A total of 3.07 million acres of land was acquired for exploration; drastic decrease. Surface drilling footage was reduced sharply; an estimated 250 drill rigs were used by the uranium industry during 1980. Land acquisition costs increased 8%. The domestic reserve changes are detailed by cause: exploration, re-evaluation, or production. Two significant discoveries of deposits were made in Mohave County, Arizona. Uranium production during 1980 was 21,850 short tons U/sub 3/O/sub 8/; an increase of 17% from 1979. Domestic and foreign exploration highlights were given. Major producing areas for the US are San Juan basin, Wyoming basins, Texas coastal plain, Paradox basin, northeastern Washington, Henry Mountains, Utah, central Colorado, and the McDermitt caldera in Nevada and Oregon. 3 figures, 8 tables. (DP)

  4. Technical Report on the Behavior of Trace Elements, Stable Isotopes, and Radiogenic Isotopes During the Processing of Uranium Ore to Uranium Ore Concentrate

    SciTech Connect (OSTI)

    Marks, N. E.; Borg, L. E.; Eppich, G. R.; Gaffney, A. M.; Genneti, V. G.; Hutcheon, I. D.; Kristo, M. J.; Lindvall, R. E.; Ramon, C.; Robel, M.; Roberts, S. K.; Schorzman, K. C.; Sharp, M. A.; Singleton, M. J.; Williams, R. W.

    2015-07-09

    The goals of this SP-1 effort were to understand how isotopic and elemental signatures behave during mining, milling, and concentration and to identify analytes that might preserve geologic signatures of the protolith ores. The impurities that are preserved through the concentration process could provide useful forensic signatures and perhaps prove diagnostic of sample origin.

  5. Portable apparatus for the measurement of environmental radon and thoron

    DOE Patents [OSTI]

    Negro, Vincent C.

    2001-01-01

    The radometer is a portable instrument for the measurement of the concentration of atmospheric radon/thoron in a test area. A constant velocity pump pulls the air from the outside at a constant flow rate. If the air is too moist, some or all of the sample is passed through a desiccant filter prior to encountering an electrostatic filter. The electrostatic filter prevents any charged particles from entering the sampling chamber. Once the sample has entered the chamber, the progeny of the decay of radon/thoron are collected on a detector and measured. The measured data is compiled by a computer and displayed.

  6. Atomic vapor laser isotope separation

    SciTech Connect (OSTI)

    Stern, R.C.; Paisner, J.A.

    1986-08-15

    The atomic vapor laser isotope separation (AVLIS) process for the enrichment of uranium is evaluated. (AIP)

  7. Establishing a Cost Basis for Converting the High Flux Isotope Reactor from High Enriched to Low Enriched Uranium Fuel

    SciTech Connect (OSTI)

    Primm, Trent; Guida, Tracey

    2010-02-01

    Under the auspices of the Global Threat Reduction Initiative Reduced Enrichment for Research and Test Reactors Program, the National Nuclear Security Administration /Department of Energy (NNSA/DOE) has, as a goal, to convert research reactors worldwide from weapons grade to non-weapons grade uranium. The High Flux Isotope Reactor (HFIR) at Oak Ridge National Lab (ORNL) is one of the candidates for conversion of fuel from high enriched uranium (HEU) to low enriched uranium (LEU). A well documented business model, including tasks, costs, and schedules was developed to plan the conversion of HFIR. Using Microsoft Project, a detailed outline of the conversion program was established and consists of LEU fuel design activities, a fresh fuel shipping cask, improvements to the HFIR reactor building, and spent fuel operations. Current-value costs total $76 million dollars, include over 100 subtasks, and will take over 10 years to complete. The model and schedule follows the path of the fuel from receipt from fuel fabricator to delivery to spent fuel storage and illustrates the duration, start, and completion dates of each subtask to be completed. Assumptions that form the basis of the cost estimate have significant impact on cost and schedule.

  8. Isotopes

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Office of Science » Nuclear Physics » Isotopes Isotopes Isotopes produced at Los Alamos National Laboratory are saving lives, advancing cutting-edge research and keeping the U.S. safe. Get Expertise Eva Birnbaum (505) 665-7167 Email Wolfgang Runde (505) 667-3350 Email Isotope Production and Applications isotopes Isotopes produced at IPF are critical for medical diagnosis and disease treatment. These positron emission tomography images were made possible using isotopes produced at LANL.

  9. Isotopes

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Eva Birnbaum (505) 665-7167 Email Wolfgang Runde (505) 667-3350 Email Isotope Production and Applications isotopes Isotopes produced at IPF are critical for medical diagnosis and ...

  10. Application of copper vapour lasers for controlling activity of uranium isotopes

    SciTech Connect (OSTI)

    Barmina, E V; Sukhov, I A; Lepekhin, N M; Priseko, Yu S; Filippov, V G; Simakin, Aleksandr V; Shafeev, Georgii A

    2013-06-30

    Beryllium nanoparticles are generated upon ablation of a beryllium target in water by a copper vapour laser. The average size of single crystalline nanoparticles is 12 nm. Ablation of a beryllium target in aqueous solutions of uranyl chloride leads to a significant (up to 50 %) decrease in the gamma activity of radionuclides of the uranium-238 and uranium-235 series. Data on the recovery of the gamma activity of these nuclides to new steady-state values after laser irradiation are obtained. The possibility of application of copper vapour lasers for radioactive waste deactivation is discussed. (laser applications and other topics in quantum electronics)

  11. Demonstration of femtosecond laser ablation inductively coupled plasma mass spectrometry for uranium isotopic measurements in U-10Mo nuclear fuel foils

    SciTech Connect (OSTI)

    Havrilla, George Joseph; Gonzalez, Jhanis

    2015-06-10

    The use of femtosecond laser ablation inductively coupled plasma mass spectrometry was used to demonstrate the feasibility of measuring the isotopic ratio of uranium directly in U-10Mo fuel foils. The measurements were done on both the flat surface and cross sections of bare and Zr clad U-10Mo fuel foil samples. The results for the depleted uranium content measurements were less than 10% of the accepted U235/238 ratio of 0.0020. Sampling was demonstrated for line scans and elemental mapping over large areas. In addition to the U isotopic ratio measurement, the Zr thickness could be measured as well as trace elemental composition if required. A number of interesting features were observed during the feasibility measurements which could provide the basis for further investigation using this methodology. The results demonstrate the feasibility of using fs-LA-ICP-MS for measuring the U isotopic ratio in U-10Mo fuel foils.

  12. Innovative lasers for uranium isotope separation. Progress report for the period September 1, 1989--May 31, 1990

    SciTech Connect (OSTI)

    Brake, M.L.; Gilgenbach, R.M.

    1990-06-01

    Copper vapor lasers have important applications to uranium atomic vapor laser isotope separation (AVLIS). The authors have spent the first year of the project investigating two innovative methods of exciting/pumping copper vapor lasers which have the potential to improve the efficiency and scaling of large laser systems used in uranium isotope separation. Experimental research has focused on the laser discharge kinetics of (1) microwave and (2) electron beam excitation/pumping of large-volume copper vapor lasers. During the first year, the experiments have been designed and constructed and initial data has been taken. Highlights of some of the first year results as well as plans for the future include the following: Microwave resonant cavity produced copper vapor plasmas at 2.45 GHz, both pulsed (5 kW, 5kHz) and CW (0--500 Watts) have been investigated using heated copper chloride as the copper source. The visible emitted light has been observed and intense lines at 510.6 nm and 578.2 nm have been observed. Initial measurements of the electric field strengths have been taken with probes, the plasma volume has been measured with optical techniques, and the power has been measured with power meters. A self-consistent electromagnetic model of the cavity/plasma system which uses the above data as input shows that the copper plasma has skin depths around 100 cm, densities around 10{sup 12} {number_sign}/cc, collisional frequencies around 10{sup 11}/sec., conductivities around 0.15 (Ohm-meter){sup {minus}1}. A simple model of the heat transfer predicts temperatures of {approximately}900 K. All of these parameters indicate that microwave discharges may be well suited as a pump source for copper lasers. These preliminary studies will be continued during the second year with additional diagnostics added to the system to verify the model results. Chemical kinetics of the system will also be added to the model.

  13. Design Study for a Low-enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2007

    SciTech Connect (OSTI)

    Primm, Trent; Ellis, Ronald James; Gehin, Jess C; Ilas, Germina; Miller, James Henry; Sease, John D

    2007-11-01

    This report documents progress made during fiscal year 2007 in studies of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) fuel to low enriched uranium fuel (LEU). Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. A high volume fraction U/Mo-in-Al fuel could attain the same neutron flux performance as with the current, HEU fuel but materials considerations appear to preclude production and irradiation of such a fuel. A diffusion barrier would be required if Al is to be retained as the interstitial medium and the additional volume required for this barrier would degrade performance. Attaining the high volume fraction (55 wt. %) of U/Mo assumed in the computational study while maintaining the current fuel plate acceptance level at the fuel manufacturer is unlikely, i.e. no increase in the percentage of plates rejected for non-compliance with the fuel specification. Substitution of a zirconium alloy for Al would significantly increase the weight of the fuel element, the cost of the fuel element, and introduce an as-yet untried manufacturing process. A monolithic U-10Mo foil is the choice of LEU fuel for HFIR. Preliminary calculations indicate that with a modest increase in reactor power, the flux performance of the reactor can be maintained at the current level. A linearly-graded, radial fuel thickness profile is preferred to the arched profile currently used in HEU fuel because the LEU fuel media is a metal alloy foil rather than a powder. Developments in analysis capability and nuclear data processing techniques are underway with the goal of verifying the preliminary calculations of LEU flux performance. A conceptual study of the operational cost of an LEU fuel fabrication facility yielded the conclusion that the annual fuel cost to the HFIR would increase significantly from the current, HEU fuel cycle. Though manufacturing can be accomplished with existing technology

  14. Utilization of non-weapons-grade plutonium and highly enriched uranium with breeding of the {sup 233}U isotope in the VVER reactors using thorium and heavy water

    SciTech Connect (OSTI)

    Marshalkin, V. E. Povyshev, V. M.

    2015-12-15

    A method for joint utilization of non-weapons-grade plutonium and highly enriched uranium in the thorium–uranium—plutonium oxide fuel of a water-moderated reactor with a varying water composition (D{sub 2}O, H{sub 2}O) is proposed. The method is characterized by efficient breeding of the {sup 233}U isotope and safe reactor operation and is comparatively simple to implement.

  15. Environmental site description for a Uranium Atomic Vapor Laser Isotope Separation (U-AVLIS) production plant at the Paducah Gaseous Diffusion Plant site

    SciTech Connect (OSTI)

    Marmer, G.J.; Dunn, C.P.; Moeller, K.L.; Pfingston, J.M.; Policastro, A.J.; Yuen, C.R.; Cleland, J.H.

    1991-09-01

    Uranium enrichment in the United States has utilized a diffusion process to preferentially enrich the U-235 isotope in the uranium product. The U-AVLIS process is based on electrostatic extraction of photoionized U-235 atoms from an atomic vapor stream created by electron-beam vaporization of uranium metal alloy. The U-235 atoms are ionized when precisely tuned laser light -- of appropriate power, spectral, and temporal characteristics -- illuminates the uranium vapor and selectively photoionizes the U-235 isotope. A programmatic document for use in screening DOE site to locate a U-AVLIS production plant was developed and implemented in two parts. The first part consisted of a series of screening analyses, based on exclusionary and other criteria, that identified a reasonable number of candidate sites. These sites were subjected to a more rigorous and detailed comparative analysis for the purpose of developing a short list of reasonable alternative sites for later environmental examination. This environmental site description (ESD) provides a detailed description of the PGDP site and vicinity suitable for use in an environmental impact statement (EIS). The report is based on existing literature, data collected at the site, and information collected by Argonne National Laboratory (ANL) staff during a site visit. 65 refs., 15 tabs.

  16. Environmental assessment for the demonstration of uranium-atomic vapor laser isotope separation (U-AVLIS) at Lawrence Livermore National Laboratory

    SciTech Connect (OSTI)

    Not Available

    1991-05-01

    The U.S. Department of Energy (DOE), Office of Nuclear Energy, proposes to use full-scale lasers and separators to demonstrate uranium enrichment as part of the national Uranium-Atomic Vapor Laser Isotope Separation (U-AVLIS) Program. Demonstration of uranium enrichment is planned to be conducted in Building 490 of the Lawrence Livermore National Laboratory (LLNL), near Livermore, California in 1991 and 1992. The collective goal of the U-AVLIS Program is to develop and demonstrate an integrated technology for low-cost enrichment of uranium for nuclear reactor fuel. Alternatives to the proposed LLNL demonstration activity are no action, use of alternative LLNL facilities, and use of an alternative DOE site. This EA describes the existing LLNL environment and surroundings that could be impacted by the proposed action. Potential impacts to on- site and off-site environments predicted during conduct of the Uranium Demonstration System (UDS) at LLNL and alternative actions are reported in this EA. The analysis covers routine activities and potential accidents. 81 refs., 8 figs., 6 tabs.

  17. Neutron-induced fission cross section measurements for uranium isotopes {sup 236}U and {sup 234}U at LANSCE

    SciTech Connect (OSTI)

    Laptev, A. B.; Tovesson, F.; Hill, T. S.

    2013-04-19

    A well established program of neutron-induced fission cross section measurement at Los Alamos Neutron Science Center (LANSCE) is supporting the Fuel Cycle Research program (FC R and D). The incident neutron energy range spans from sub-thermal up to 200 MeV by combining two LANSCE facilities, the Lujan Center and the Weapons Neutron Research facility (WNR). The time-of-flight method is implemented to measure the incident neutron energy. A parallel-plate fission ionization chamber was used as a fission fragment detector. The event rate ratio between the investigated foil and a standard {sup 235}U foil is converted into a fission cross section ratio. In addition to previously measured data new measurements include {sup 236}U data which is being analyzed, and {sup 234}U data acquired in the 2011-2012 LANSCE run cycle. The new data complete the full suite of Uranium isotopes which were investigated with this experimental approach. Obtained data are presented in comparison with existing evaluations and previous data.

  18. Measurement of uranium and plutonium in solid waste by passive photon or neutron counting and isotopic neutron source interrogation

    SciTech Connect (OSTI)

    Crane, T.W.

    1980-03-01

    A summary of the status and applicability of nondestructive assay (NDA) techniques for the measurement of uranium and plutonium in 55-gal barrels of solid waste is reported. The NDA techniques reviewed include passive gamma-ray and x-ray counting with scintillator, solid state, and proportional gas photon detectors, passive neutron counting, and active neutron interrogation with neutron and gamma-ray counting. The active neutron interrogation methods are limited to those employing isotopic neutron sources. Three generic neutron sources (alpha-n, photoneutron, and /sup 252/Cf) are considered. The neutron detectors reviewed for both prompt and delayed fission neutron detection with the above sources include thermal (/sup 3/He, /sup 10/BF/sub 3/) and recoil (/sup 4/He, CH/sub 4/) proportional gas detectors and liquid and plastic scintillator detectors. The instrument found to be best suited for low-level measurements (< 10 nCi/g) is the /sup 252/Cf Shuffler. The measurement technique consists of passive neutron counting followed by cyclic activation using a /sup 252/Cf source and delayed neutron counting with the source withdrawn. It is recommended that a waste assay station composed of a /sup 252/Cf Shuffler, a gamma-ray scanner, and a screening station be tested and evaluated at a nuclear waste site. 34 figures, 15 tables.

  19. Isotopes of uranium and thorium, lead-210, and polonium-210 in the lungs of coal miners of Appalachia and the lungs and livers of residents of central Ohio

    SciTech Connect (OSTI)

    Gilbert, G.E.; Casella, V.R.; Bishop, C.T.; Aguirre, A.G.

    1985-10-21

    The lungs of twelve and the livers of three residents of central Ohio and the lungs of four coal miners of Appalachia were analyzed for uranium-238, uranium-234, thorium-230, lead-210, polonium-210, and thorium-232. Mean and median lung concentrations of uranium-238 and of uranium-234 in the lungs of central Ohioans were essentially the same and were equal to 4 fCi/g dry. Mean concentrations of these isotopes in the lungs of Appalachian coal miners were also essentially the same and were equal to 9 fCi/g. Little uranium was found in the liver. The median concentration of thorium-230 in the lungs of central Ohioans was also 4 fCi/g dry; however, the mean concentration was 8 fCi/g due to the relatively high concentration values in a few persons. The mean concentrations of this isotope in the lungs of central Ohioans and Appalachian coal miners were essentially the same; i.e. 8 fCi/g. The mean and median concentrations of thorium-232 in the lungs of central Ohioans were assentially the same and equal to 4 fCi/g. The mean concentration of this isotope in the lungs of Appalachian coal miners was 9 fCi/g. Little thorium was found in the liver. The mean concentrations of lead-210 in the lungs of the two populations were nearly equal and about 23 fCi/g dry. The mean liver/lung ratio of this isotope was essentially two, and the concentrations appeared to be positively correlated with smoking. Polonium-210 concentrations in the lungs were distributed into three sets of values which are described here as low (2-4 fCi/g), medium (20-30 fCi/g), and high (>100 fCi/g), and also appeared to be correlated with smoking. Mean liver concentrations of this irotope were nearly equal to the mean liver concentrations of lead-210 (50 as opposed to 47 fCi/g). 18 refs., 6 tabs.

  20. Uranium enrichment

    SciTech Connect (OSTI)

    Not Available

    1991-04-01

    This book presents the GAO's views on the Department of Energy's (DOE) program to develop a new uranium enrichment technology, the atomic vapor laser isotope separation process (AVLIS). Views are drawn from GAO's ongoing review of AVLIS, in which the technical, program, and market issues that need to be addressed before an AVLIS plant is built are examined.

  1. The Development of Low-Level Measurement Capabilities for Total and Isotopic Uranium in Environmental Samples at Brazilian and Argentine Laboratories by ABACC

    SciTech Connect (OSTI)

    Guidicini, Olga M.; Olsen, Khris B.; Hembree, Doyle M.; Carter, Joel A.; Whitaker, Michael; Hayes, Susan M.

    2005-07-01

    In June 1998, the Brazilian-Argentine Agency for Accounting and Control of Nuclear Materials (ABACC), with assistance from the U.S. Department of Energy (DOE), began a program to assess environmental sampling and analysis capabilities at laboratories in Argentina and Brazil. The program began with staff training conducted in South America and the United States by Oak Ridge National Laboratory (ORNL) and Pacific Northwest National Laboratory (PNNL). Both laboratories are participating members of DOE’s Network of Analytical Laboratories (NWAL) that support IAEA’s environmental sampling program. During the initial planning meeting, representatives from ABACC and all the participating analytical laboratories supporting ABACC were briefed on how the first exercise would be managed and on key aspects necessary to analyze low-level environmental samples for uranium. Subsequent to this training, a laboratory evaluation exercise (Exercise 1) was conducted using standard swipe samples prepared for this exercise by the International Atomic Energy Agency (IAEA). The results of Exercise 1 determined that sample contamination was a major factor in the analysis, and a thorough review of laboratory procedures was required to reduce the level of contamination to acceptable levels. Following modification of sample preparation procedures, the laboratories performed Exercise 2, an analysis of a National Institute of Standards and Technology (NIST) Standard Reference Material (SRM) 1547, Peach Leaves. The results of Exercise 2 demonstrated that several laboratories were capable of accurately determining the total uranium and uranium isotopic distribution in the peach leaves. To build on these successes, Exercise 3 was performed using a series of standard swipe samples prepared by the IAEA and distributed to laboratories supporting ABACC and to PNNL and ORNL. The results of Exercise 3 demonstrate that ABACC now has support laboratories in both Argentina and Brazil, which are capable

  2. Modeling non-steady state radioisotope transport in the vadose zone--A case study using uranium isotopes at Pena Blanca, Mexico

    SciTech Connect (OSTI)

    Ku, T. L.; Luo, S.; Goldstein, S. J.; Murrell, M. T.; Chu, W. L.; Dobson, P. F.

    2009-06-01

    Current models using U- and Th-series disequilibria to study radioisotope transport in groundwater systems mostly consider a steady-state situation. These models have limited applicability to the vadose zone (UZ) where the concentration and migratory behavior of radioisotopes in fluid are often transitory. We present here, as a first attempt of its kind, a model simulating the non-steady state, intermittent fluid transport in vadose layers. It provides quantitative constraints on in-situ migration of dissolved and colloidal radioisotopes in terms of retardation factor and rock-water interaction (or water transit) time. For uranium, the simulation predicts that intermittent flushing in the UZ leads to a linear relationship between reciprocal U concentration and {sup 234}U/{sup 238}U ratio in percolating waters, with the intercept and slope bearing information on the rates of dissolution and {alpha}-recoil of U isotopes, respectively. The general validity of the model appears to be borne out by the measurement of uranium isotopes in UZ waters collected at various times over a period during 1995-2006 from a site in the Pena Blanca mining district, Mexico, where the Nopal I uranium deposit is located. Enhanced {sup 234}U/{sup 238}U ratios in vadose-zone waters resulting from lengthened non-flushing time as prescribed by the model provide an interpretative basis for using {sup 234}U/{sup 238}U in cave calcites to reconstruct the regional changes in hydrology and climate. We also provide a theoretical account of the model's potential applications using radium isotopes.

  3. Manhattan Project: Early Uranium Research, 1939-1941

    Office of Scientific and Technical Information (OSTI)

    ... Retaining programmatic responsibilities for uranium research in the new organizational setup, the Uranium Committee recommended that all four isotope separation methods and the ...

  4. Preliminary Assessment of the Impact on Reactor Vessel dpa Rates Due to Installation of a Proposed Low Enriched Uranium (LEU) Core in the High Flux Isotope Reactor (HFIR)

    SciTech Connect (OSTI)

    Daily, Charles R.

    2015-10-01

    An assessment of the impact on the High Flux Isotope Reactor (HFIR) reactor vessel (RV) displacements-per-atom (dpa) rates due to operations with the proposed low enriched uranium (LEU) core described by Ilas and Primm has been performed and is presented herein. The analyses documented herein support the conclusion that conversion of HFIR to low-enriched uranium (LEU) core operations using the LEU core design of Ilas and Primm will have no negative impact on HFIR RV dpa rates. Since its inception, HFIR has been operated with highly enriched uranium (HEU) cores. As part of an effort sponsored by the National Nuclear Security Administration (NNSA), conversion to LEU cores is being considered for future HFIR operations. The HFIR LEU configurations analyzed are consistent with the LEU core models used by Ilas and Primm and the HEU balance-of-plant models used by Risner and Blakeman in the latest analyses performed to support the HFIR materials surveillance program. The Risner and Blakeman analyses, as well as the studies documented herein, are the first to apply the hybrid transport methods available in the Automated Variance reduction Generator (ADVANTG) code to HFIR RV dpa rate calculations. These calculations have been performed on the Oak Ridge National Laboratory (ORNL) Institutional Cluster (OIC) with version 1.60 of the Monte Carlo N-Particle 5 (MCNP5) computer code.

  5. Environmental site description for a Uranium Atomic Vapor Laser Isotope Separation (U-AVLIS) production plant at the Portsmouth Gaseous Diffusion Plant site

    SciTech Connect (OSTI)

    Marmer, G.J.; Dunn, C.P.; Filley, T.H.; Moeller, K.L.; Pfingston, J.M.; Policastro, A.J.; Cleland, J.H.

    1991-09-01

    Uranium enrichment in the United States has utilized a diffusion process to preferentially enrich the U-235 isotope in the uranium product. In the 1970s, the US Department of Energy (DOE) began investigating more efficient and cost-effective enrichment technologies. In January 1990, the Secretary of Energy approved a plan for the demonstration and deployment of the Uranium Atomic Vapor Laser isotope Separation (U-AVLIS) technology with the near-term goal to provide the necessary information to make a deployment decision by November 1992. Initial facility operation is anticipated for 1999. A programmatic document for use in screening DOE sites to locate a U-AVLIS production plant was developed and implemented in two parts. The first part consisted of a series of screening analyses, based on exclusionary and other criteria, that identified a reasonable number of candidate sites. The final evaluation, which included sensitivity studies, identified the Oak Ridge Gaseous Diffusion Plant (ORGDP) site, the Paducah Gaseous Diffusion Plant (PGDP) site, and the Portsmouth Gaseous Diffusion Plant (PORTS) site as having significant advantages over the other sites considered. This environmental site description (ESD) provides a detailed description of the PORTS site and vicinity suitable for use in an environmental impact statement (EIS). This report is based on existing literature, data collected at the site, and information collected by Argonne National Laboratory (ANL) staff during site visits. The organization of the ESD is as follows. Topics addressed in Sec. 2 include a general site description and the disciplines of geology, water resources, biotic resources, air resources, noise, cultural resources, land use. Socioeconomics, and waste management. Identification of any additional data that would be required for an EIS is presented in Sec. 3.

  6. Behavior of natural uranium, thorium, and radium isotopes in the Wolfcamp brine aquifers, Palo Duro Basin, Texas

    SciTech Connect (OSTI)

    Laul, J.C.; Smith, M.R.; Hubbard, N.

    1984-10-01

    Previously reported results for Palo Duro deep brines show that Ra is highly soluble and not retarded. Relative to Ra, U and Th are highly sorbed. Uranium, like thorium, is in the +4 valence state, indicating a reducing environment. Additional data reported here support these results. However, one Wolfcamp brine sample gives somewhat different results. Radium appears to be somewhat sorbed. Uranium is largely in the +6 valence state, indicating a less reducing condition. In all brines, kinetics for sorption (/sup 228/Th) and desorption (/sup 224/Ra) are rapid. This Wolfcamp brine was tested for the effects of colloids for Ra, U, and Th concentrations. No effects were found.

  7. METHOD FOR PURIFYING URANIUM

    DOE Patents [OSTI]

    Kennedy, J.W.; Segre, E.G.

    1958-08-26

    A method is presented for obtaining a compound of uranium in an extremely pure state and in such a condition that it can be used in determinations of the isotopic composition of uranium. Uranium deposited in calutron receivers is removed therefrom by washing with cold nitric acid and the resulting solution, coataining uranium and trace amounts of various impurities, such as Fe, Ag, Zn, Pb, and Ni, is then subjected to various analytical manipulations to obtain an impurity-free uranium containing solution. This solution is then evaporated on a platinum disk and the residue is ignited converting it to U2/sub 3//sub 8/. The platinum disk having such a thin film of pure U/sub 2/O/sub 8/ is suitable for use with isotopic determination techaiques.

  8. Uranium Processing Facility Site Readiness Subproject Completed...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    ... Successfully Establishes Uranium Lease and Takeback Program to Support Critical Medical Isotope Production Apex Gold discussion fosters international cooperation in run-up to ...

  9. Environmental site description for a Uranium Atomic Vapor Laser Isotope Separation (U-AVLIS) production plant at the Oak Ridge Gaseous Diffusion Plant Site

    SciTech Connect (OSTI)

    Not Available

    1991-09-01

    In January 1990, the Secretary of Energy approved a plan for the demonstration and deployment of the Uranium Atomic Vapor Laser Isotope Separation (U-AVLIS) technology, with the near-term goal to provide the necessary information to make a deployment decision by November 1992. The U-AVLIS process is based on electrostatic extraction of photoionized U-235 atoms from an atomic vapor stream created by electron-beam vaporization of uranium metal alloy. A programmatic document for use in screening DOE sites to locate the U-AVLIS production plant was developed and implemented in two parts (Wolsko et al. 1991). The first part consisted of a series of screening analyses, based on exclusionary and other criteria, that identified a reasonable number of candidate sites. These sites were then subjected to a more rigorous and detailed comparative analysis for the purpose of developing a short list of reasonable alternative sites for later environmental examination. This environmental site description (ESD) provides a detailed description of the ORGDP site and vicinity suitable for use in an environmental impact statement (EIS). The report is based on existing literature, data collected at the site, and information collected by Argonne National Laboratory (ANL) staff during a site visit. The organization of the ESD is as follows. Topics addressed in Sec. 2 include a general site description and the disciplines of geology, water resources, biotic resources, air resources, noise, cultural resources, land use, socioeconomics, and waste management. Identification of any additional data that would be required for an EIS is presented in Sec. 3. Following the site description and additional data requirements, Sec. 4 provides a short, qualitative assessment of potential environmental issues. 37 refs., 20 figs., 18 tabs.

  10. Uranium enrichment

    SciTech Connect (OSTI)

    Not Available

    1991-08-01

    This paper reports that in 1990 the Department of Energy began a two-year project to illustrate the technical and economic feasibility of a new uranium enrichment technology-the atomic vapor laser isotope separation (AVLIS) process. GAO believes that completing the AVLIS demonstration project will provide valuable information about the technical viability and cost of building an AVLIS plant and will keep future plant construction options open. However, Congress should be aware that DOE still needs to adequately demonstrate AVLIS with full-scale equipment and develop convincing cost projects. Program activities, such as the plant-licensing process, that must be completed before a plant is built, could take many years. Further, an updated and expanded uranium enrichment analysis will be needed before any decision is made about building an AVLIS plant. GAO, which has long supported legislation that would restructure DOE's uranium enrichment program as a government corporation, encourages DOE's goal of transferring AVLIS to the corporation. This could reduce the government's financial risk and help ensure that the decision to build an AVLIS plant is based on commercial concerns. DOE, however, has no alternative plans should the government corporation not be formed. Further, by curtailing a planned public access program, which would have given private firms an opportunity to learn about the technology during the demonstration project, DOE may limit its ability to transfer AVLIS to the private sector.

  11. Generation of Radixenon Isotopes

    SciTech Connect (OSTI)

    McIntyre, Justin I.; Bowyer, Ted W.; Hayes, James C.; Heimbigner, Tom R.; Morris, Scott J.; Panisko, Mark E.; Pitts, W. K.; Pratt, Sharon L.; Reeder, Paul L.; Thomas, Charles W.

    2003-06-30

    Pacific Northwest National Laboratory has developed an automated system for separating Xe from air and can detect the following radioxenon isotopes, 131mXe, 133mXe, 133Xe, and 135Xe. This report details the techniques used to generate the various radioxenon isotopes that are used for the calibration of the detector as well as other isotopes that have the potential to interfere with the fission produced radioxenon isotopes. Fission production is covered first using highly enriched uranium followed by a description and results from an experiment to produce radioxenon isotopes from neutron activation of ambient xenon.

  12. Uranium Measurement Improvements at the Savannah River Technology Center

    SciTech Connect (OSTI)

    Shick, C. Jr.

    2002-02-13

    Uranium isotope ratio and isotope dilution methods by mass spectrometry are used to achieve sensitivity, precision and accuracy for various applications. This report presents recent progress made at SRTC in the analysis of minor isotopes of uranium. Comparison of routine measurements of NBL certified uranium (U005a) using the SRTC Three Stage Mass Spectrometer (3SMS) and the SRTC Single Stage Mass Spectrometer (SSMS). As expected, the three stage mass spectrometer yielded superior sensitivity, precision, and accuracy for this application.

  13. Multi-parametric approach towards the assessment of radon and thoron progeny exposures

    SciTech Connect (OSTI)

    Mishra, Rosaline E-mail: rosaline.mishra@gmail.com; Sapra, B. K.; Mayya, Y. S.

    2014-02-15

    Conventionally, the dosimetry is carried out using radon and thoron gas concentration measurements and doses have been assigned using assumed equilibrium factors for the progeny species, which is inadequate pertaining to the variations in equilibrium factors and possibly due to significant thoron. In fact, since the true exposures depend upon the intricate mechanisms of progeny deposition in the lung, therefore an integrated approach for the assessment of progeny is essential. In this context, the recently developed deposition based progeny concentration measurement techniques (DTPS: Direct Thoron progeny sensors and DRPS: Direct Radon progeny sensors) appear to be best suited for radiological risk assessments both among occupational workers and general study populations. DTPS and DRPS consist of aluminized mylar mounted LR115 type passive detectors, which essentially detects the alpha particles emitted from the deposited progeny atoms on the detector surface. It gives direct measure of progeny activity concentrations in air. DTPS has a lower limit of detection limit of 0.1?Bq/m{sup 3} whereas that for DRPS is 1 Bq/m{sup 3}, hence are perfectly suitable for indoor environments. These DTPS and DRPS can be capped with 200-mesh type wire-screen to measure the coarse fraction of the progeny concentration and the corresponding coarse fraction deposition velocities as well as the time integrated fine fraction. DTPS and DRPS can also be lodged in an integrated sampler wherein the wire-mesh and filter-paper are arranged in an array in flow-mode, to measure the fine and coarse fraction concentration separately and simultaneously. The details are further discussed in the paper.

  14. Rescuing a Treasure Uranium-233

    SciTech Connect (OSTI)

    Krichinsky, Alan M; Goldberg, Dr. Steven A.; Hutcheon, Dr. Ian D.

    2011-01-01

    Uranium-233 (233U) is a synthetic isotope of uranium formed under reactor conditions during neutron capture by natural thorium (232Th). At high purities, this synthetic isotope serves as a crucial reference for accurately quantifying and characterizing natural uranium isotopes for domestic and international safeguards. Separated 233U is stored in vaults at Oak Ridge National Laboratory. These materials represent a broad spectrum of 233U from the standpoint isotopic purity the purest being crucial for precise analyses in safeguarding uranium. All 233U at ORNL currently is scheduled to be down blended with depleted uranium beginning in 2015. Such down blending will permanently destroy the potential value of pure 233U samples as certified reference material for use in uranium analyses. Furthermore, no replacement 233U stocks are expected to be produced in the future due to a lack of operating production capability and the high cost of returning to operation this currently shut down capability. This paper will describe the efforts to rescue the purest of the 233U materials arguably national treasures from their destruction by down blending.

  15. Invited Article: In situ comparison of passive radon-thoron discriminative monitors at subsurface workplaces in Hungary

    SciTech Connect (OSTI)

    Kvsi, Norbert; Social Organization for Radioecological Cleanliness, Veszprm ; Vigh, Tams; Manganese Mining Process Ltd., rkt ; Nmeth, Csaba; University of Pannonia, Veszprm ; Ishikawa, Tetsuo; Omori, Yasutaka; Janik, Miroslaw; Yonehara, Hidenori

    2014-02-15

    During a one-year long measurement period, radon and thoron data obtained by two different passive radon-thoron discriminative monitors were compared at subsurface workplaces in Hungary, such as mines (bauxite and manganese ore) and caves (medical and touristic). These workplaces have special environmental conditions, such as, stable and high relative humidity (100%), relatively stable temperature (12C21C), low or high wind speed (max. 2.4 ms{sup ?1}) and low or elevated aerosol concentration (13060000 particles m{sup ?3}). The measured radon and thoron concentrations fluctuated in a wide range among the different workplaces. The respective annual average radon concentrations and their standard deviations (in brackets) measured by the passive radon-thoron discriminative monitor with cellulose filter (CF) and the passive radon-thoron discriminative monitor with sponge filter (SF) were: 350(321) Bqm{sup ?3} and 550(497) Bqm{sup ?3} in the bauxite mine; 887(604) Bqm{sup ?3} and 1258(788) Bqm{sup ?3} in the manganese ore mine; 2510(2341) Bqm{sup ?3} and 3403(3075) Bqm{sup ?3} in the medical cave (Hospital Cave of Tapolca); and 6239(2057) Bq m{sup ?3} and 8512(1955) Bqm{sup ?3} in the touristic cave (Lake Cave of Tapolca). The respective average thoron concentrations and their standard deviation (in brackets) measured by CF and SF monitors were: 154(210) Bqm{sup ?3} and 161(148) Bqm{sup ?3} in the bauxite mine; 187(191) Bqm{sup ?3} and 117(147) Bqm{sup ?3} in the manganese-ore mine; 360(524) Bqm{sup ?3} and 371(789) Bqm{sup ?3} in the medical cave (Hospital Cave of Tapolca); and 1420(1184) Bq m{sup ?3} and 1462(3655) Bqm{sup ?3} in the touristic cave (Lake Cave of Tapolca). Under these circumstances, comparison of the radon data for the SF and CF monitors showed the former were consistently 51% higher in the bauxite mine, 38% higher in the manganese ore mine, and 34% higher in the caves. Consequently, correction is required on previously obtained

  16. URANIUM ALLOYS

    DOE Patents [OSTI]

    Colbeck, E.W.

    1959-12-29

    A uranium alloy is reported containing from 0.1 to 5 per cent by weight of molybdenum and from 0.1 to 5 per cent by weight of silicon, the balance being uranium.

  17. Dry phase reactor for generating medical isotopes

    DOE Patents [OSTI]

    Mackie, Thomas Rockwell; Heltemes, Thad Alexander

    2016-05-03

    An apparatus for generating medical isotopes provides for the irradiation of dry-phase, granular uranium compounds which are then dissolved in a solvent for separation of the medical isotope from the irradiated compound. Once the medical isotope is removed, the dissolved compound may be reconstituted in dry granular form for repeated irradiation.

  18. The Role of COMSOL Toward a Low-Enriched Uranium Fuel Design...

    Office of Scientific and Technical Information (OSTI)

    Uranium Fuel Design for the High Flux Isotope Reactor Citation Details In-Document Search Title: The Role of COMSOL Toward a Low-Enriched Uranium Fuel Design for the High Flux ...

  19. Photochemical isotope separation

    DOE Patents [OSTI]

    Robinson, C.P.; Jensen, R.J.; Cotter, T.P.; Greiner, N.R.; Boyer, K.

    1987-04-28

    A process is described for separating isotopes by selective excitation of isotopic species of a volatile compound by tuned laser light. A highly cooled gas of the volatile compound is produced in which the isotopic shift is sharpened and defined. Before substantial condensation occurs, the cooled gas is irradiated with laser light precisely tuned to a desired wavelength to selectively excite a particular isotopic species in the cooled gas. The laser light may impart sufficient energy to the excited species to cause it to undergo photochemical reaction or even to photoionize. Alternatively, a two-photon irradiation may be applied to the cooled gas to induce photochemical reaction or photoionization. The process is particularly applicable to the separation of isotopes of uranium and plutonium. 8 figs.

  20. Laser isotope separation

    DOE Patents [OSTI]

    Robinson, C. Paul; Jensen, Reed J.; Cotter, Theodore P.; Boyer, Keith; Greiner, Norman R.

    1988-01-01

    A process and apparatus for separating isotopes by selective excitation of isotopic species of a volatile compound by tuned laser light. A highly cooled gas of the volatile compound is produced in which the isotopic shift is sharpened and defined. Before substantial condensation occurs, the cooled gas is irradiated with laser light precisely tuned to a desired wavelength to selectively excite a particular isotopic species in the cooled gas. The laser light may impart sufficient energy to the excited species to cause it to undergo photolysis, photochemical reaction or even to photoionize. Alternatively, a two-photon irradiation may be applied to the cooled gas to induce photolysis, photochemical reaction or photoionization. The process is particularly applicable to the separation of isotopes of uranium.

  1. Photochemical isotope separation

    DOE Patents [OSTI]

    Robinson, C. Paul; Jensen, Reed J.; Cotter, Theodore P.; Greiner, Norman R.; Boyer, Keith

    1987-01-01

    A process for separating isotopes by selective excitation of isotopic species of a volatile compound by tuned laser light. A highly cooled gas of the volatile compound is produced in which the isotopic shift is sharpened and defined. Before substantial condensation occurs, the cooled gas is irradiated with laser light precisely tuned to a desired wavelength to selectively excite a particular isotopic species in the cooled gas. The laser light may impart sufficient energy to the excited species to cause it to undergo photochemical reaction or even to photoionize. Alternatively, a two-photon irradiation may be applied to the cooled gas to induce photochemical reaction or photoionization. The process is particularly applicable to the separation of isotopes of uranium and plutonium.

  2. Laser isotope separation

    DOE Patents [OSTI]

    Robinson, C.P.; Reed, J.J.; Cotter, T.P.; Boyer, K.; Greiner, N.R.

    1975-11-26

    A process and apparatus for separating isotopes by selective excitation of isotopic species of a volatile compound by tuned laser light is described. A highly cooled gas of the volatile compound is produced in which the isotopic shift is sharpened and defined. Before substantial condensation occurs, the cooled gas is irradiated with laser light precisely tuned to a desired wavelength to selectively excite a particular isotopic species in the cooled gas. The laser light may impart sufficient energy to the excited species to cause it to undergo photolysis, photochemical reaction or even to photoionize. Alternatively, a two-photon irradiation may be applied to the cooled gas to induce photolysis, photochemical reaction or photoionization. The process is particularly applicable to the separation of isotopes of uranium.

  3. Uranium 238U/235U isotope ratios as indicators of reduction: Results from an in situ biostimulation experiment at Rifle, Colorado, USA

    SciTech Connect (OSTI)

    Bopp IV, C.J.; Lundstrom, C.C.; Johnson, T.M.; Sanford, R.A.; Long, P.E.; Williams, K.H.

    2010-02-01

    The attenuation of groundwater contamination via chemical reaction is traditionally evaluated by monitoring contaminant concentration through time. However, this method can be confounded by common transport processes (e.g. dilution, sorption). Isotopic techniques bypass the limits of concentration methods, and so may provide improved accuracy in determining the extent of reaction. We apply measurements of {sup 238}U/{sup 235}U to a U bioremediation field experiment at the Rifle Integrated Field Research Challenge Site in Rifle, Colorado (USA). An array of monitoring and injection wells was installed on a 100 m{sup 2} plot where U(VI) contamination was present in the groundwater. Acetate-amended groundwater was injected along an up-gradient gallery to encourage the growth of dissimilatory metal reducing bacteria (e.g. Geobacter species). During amendment, U concentration dropped by an order of magnitude in the experiment plot. We measured {sup 238}U/{sup 235}U in samples from one monitoring well by MC-ICP-MS using a double isotope tracer method. A significant {approx}1.00{per_thousand} decrease in {sup 238}U/{sup 235}U occurred in the groundwater as U(VI) concentration decreased. The relationship between {sup 238}U/{sup 235}U and concentration corresponds approximately to a Rayleigh distillation curve with an effective fractionation factor ({alpha}) of 1.00046. We attribute the observed U isotope fractionation to a nuclear field shift effect during enzymatic reduction of U(VI){sub (aq)} to U(IV){sub (s)}.

  4. URANIUM COMPOSITIONS

    DOE Patents [OSTI]

    Allen, N.P.; Grogan, J.D.

    1959-05-12

    This patent relates to high purity uranium alloys characterized by improved stability to thermal cycling and low thermal neutron absorption. The high purity uranium alloy contains less than 0.1 per cent by weight in total amount of any ore or more of the elements such as aluminum, silicon, phosphorous, tin, lead, bismuth, niobium, and zinc.

  5. Uranium Processing Facility | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Uranium Processing Facility

  6. ISOTOPE FRACTIONATION PROCESS

    DOE Patents [OSTI]

    Clewett, G.H.; Lee, DeW.A.

    1958-05-20

    A new method is described for isotopic enrichment of uranium. It has been found that when an aqueous acidic solution of ionic tetravalent uraniunn is contacted with chelate complexed tetravalent uranium, the U/sup 238/ preferentially concentrates in the complexed phase while U/sup 235/ concentrates in the ionic phase. The effect is enhanced when the chelate compound is water insoluble and is dissolved in a water-immiscible organic solvent. Cupferron is one of a number of sultable complexing agents, and chloroform is a suitable organic solvent.

  7. Analysis of palladium coatings to remove hydrogen isotopes from zirconium fuel rods in Canada deuterium uranium-pressurized heavy water reactors; Thermal and neutron diffusion effects

    SciTech Connect (OSTI)

    Stokes, C.L.; Buxbaum, R.E. )

    1992-05-01

    This paper reports that, in pressurized heavy water nuclear reactors of the type standardly used in Canada (Canada deuterium uranium-pressurized heavy water reactors), the zirconium alloy pressure tubes of the core absorb deuterium produced by corrosion reactions. This deuterium weakens the tubes through hydrogen embrittlement. Thin palladium coatings on the outside of the zirconium are analyzed as a method for deuterium removal. This coating is expected to catalyze the reaction D{sub 2} + 1/2O{sub 2} {r reversible} D{sub 2}O when O{sub 2} is added to the annular (insulating) gas in the tubes. Major reductions in the deuterium concentration and, hence, hydrogen embrittlement are predicted. Potential problems such as plating the tube geometry, neutron absorption, catalyst deactivation, radioactive waste production, and oxygen corrosion are shown to be manageable. Also, a simple set of equations are derived to calculate the effect on diffusion caused by neutron interactions. Based on calculations of ordinary and neutron flux induced diffusion, a palladium coating of 1 {times} 10{sup {minus}6} m is recommended. This would cost approximately $60,000 per reactor unit and should more than double reactor lifetime. Similar coatings and similar interdiffusion calculations might have broad applications.

  8. Laser separation of medical isotopes

    SciTech Connect (OSTI)

    Eerkens, J.W.; Puglishi, D.A.; Miller, W.H.

    1996-12-31

    There is an increasing demand for different separated isotopes as feed material for reactor and cyclotron-produced radioisotopes used by a fast-growing radiopharmaceutical industry. One new technology that may meet future demands for medical isotopes is molecular laser isotope separation (MLIS). This method was investigated for the enrichment of uranium in the 1970`s and 1980s by Los Alamos National Laboratory, Isotope Technologies, and others around the world. While South Africa and Japan have continued the development of MLIS for uranium and are testing pilot units, around 1985 the United States dropped the LANL MLIS program in favor of AVLIS (atomic vapor LIS), which uses electron-beam-heated uranium metal vapor. AVLIS appears difficult and expensive to apply to most isotopes of medical interest, however, whereas MLIS technology, which is based on cooled hexafluorides or other gaseous molecules, can be adapted more readily. The attraction of MLIS for radiopharmaceutical firms is that it allows them to operate their own dedicated separators for small-quantity productions of critical medical isotopes, rather than having to depend on large enrichment complexes run by governments, which are only optimal for large-quantity productions. At the University of Missouri, the authors are investigating LIS of molybdenum isotopes using MoF{sub 6}, which behaves in a way similar to UF{sub 6}, studied in the past.

  9. Uranium industry annual 1997

    SciTech Connect (OSTI)

    1998-04-01

    This report provides statistical data on the U.S. uranium industry`s activities relating to uranium raw materials and uranium marketing.

  10. Beneficial Uses of Depleted Uranium

    SciTech Connect (OSTI)

    Brown, C.; Croff, A.G.; Haire, M. J.

    1997-08-01

    Naturally occurring uranium contains 0.71 wt% {sup 235}U. In order for the uranium to be useful in most fission reactors, it must be enriched the concentration of the fissile isotope {sup 235}U must be increased. Depleted uranium (DU) is a co-product of the processing of natural uranium to produce enriched uranium, and DU has a {sup 235}U concentration of less than 0.71 wt%. In the United States, essentially all of the DU inventory is in the chemical form of uranium hexafluoride (UF{sub 6}) and is stored in large cylinders above ground. If this co-product material were to be declared surplus, converted to a stable oxide form, and disposed, the costs are estimated to be several billion dollars. Only small amounts of DU have at this time been beneficially reused. The U.S. Department of Energy (DOE) has begun the Beneficial Uses of DU Project to identify large-scale uses of DU and encourage its reuse for the primary purpose of potentially reducing the cost and expediting the disposition of the DU inventory. This paper discusses the inventory of DU and its rate of increase; DU disposition options; beneficial use options; a preliminary cost analysis; and major technical, institutional, and regulatory issues to be resolved.

  11. JACKETING URANIUM

    DOE Patents [OSTI]

    Saller, H.A.; Keeler, J.R.

    1959-07-14

    The bonding to uranium of sheathing of iron or cobalt, or nickel, or alloys thereof is described. The bonding is accomplished by electro-depositing both surfaces to be joined with a coating of silver and amalgamating or alloying the silver layer with mercury or indium. Then the silver alloy is homogenized by exerting pressure on an assembly of the uranium core and the metal jacket, reducing the area of assembly and heating the assembly to homogenize by diffusion.

  12. Analysis of Uranium and Plutonium by MC-ICPMS

    SciTech Connect (OSTI)

    Williams, R W

    2005-02-23

    This procedure is written as general guidance for the measurement of elemental isotopic composition by plasma-source inorganic mass spectrometry. Analytical methods for uranium and plutonium are given as examples.

  13. In-line assay monitor for uranium hexafluoride

    DOE Patents [OSTI]

    Wallace, S.A.

    1980-03-21

    An in-line assay monitor for determining the content of uranium-235 in a uranium hexafluoride gas isotopic separation system is provided which removes the necessity of complete access to the operating parameters of the system for determining the uranium-235 content. The method and monitor for carrying out the method involve cooling of a radiation pervious chamber connected in fluid communication with the selected point in the system to withdraw a specimen and solidify the specimen in the chamber. The specimen is irradiated by means of an ionizing radiation source of energy different from that of the 185 keV gamma emissions from uranium-235. The uranium-235 content of the specimen is determined from comparison of the accumulated 185 keV energy counts and reference energy counts. The latter is used to measure the total uranium isotopic content of the specimen.

  14. Continuous reduction of uranium tetrafluoride

    SciTech Connect (OSTI)

    DeMint, A.L.; Maxey, A.W.

    1993-10-21

    Operation of a pilot-scale system for continuous metallothermic reduction of uranium tetrafluoride (UF{sub 4} or green salt) has been initiated. This activity is in support of the development of a cost- effective process to produce uranium-iron (U-Fe) alloy feed for the Uranium-Atomic Vapor Laser Isotope Separation (U-AVLIS) program. To date, five runs have been made to reduce green salt (UF{sub 4}) with magnesium. During this quarter, three runs were made to perfect the feeding system, examine feed rates, and determine the need for a crust breaker/stirrer. No material was drawn off in any of the runs; both product metal and by-product salt were allowed to accumulate in the reactor.

  15. URANIUM-SERIES CONSTRAINTS ON RADIONUCLIDE TRANSPORT AND GROUNDWATER FLOW AT NOPAL I URANIUM DEPOSIT, SIERRA PENA BLANCA, MEXICO

    SciTech Connect (OSTI)

    S. J. Goldstein, S. Luo, T. L. Ku, and M. T. Murrell

    2006-04-01

    Uranium-series data for groundwater samples from the vicinity of the Nopal I uranium ore deposit are used to place constraints on radionuclide transport and hydrologic processes at this site, and also, by analogy, at Yucca Mountain. Decreasing uranium concentrations for wells drilled in 2003 suggest that groundwater flow rates are low (< 10 m/yr). Field tests, well productivity, and uranium isotopic constraints also suggest that groundwater flow and mixing is limited at this site. The uranium isotopic systematics for water collected in the mine adit are consistent with longer rock-water interaction times and higher uranium dissolution rates at the front of the adit where the deposit is located. Short-lived nuclide data for groundwater wells are used to calculate retardation factors that are on the order of 1,000 for radium and 10,000 to 10,000,000 for lead and polonium. Radium has enhanced mobility in adit water and fractures near the deposit.

  16. Domestic production of medical isotope Mo-99 moves a step closer

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Domestic production of medical isotope Mo-99 Domestic production of medical isotope Mo-99 moves a step closer Irradiated uranium fuel has been recycled and reused for molybdenum-99 ...

  17. 2015 Domestic Uranium Production Report

    U.S. Energy Information Administration (EIA) Indexed Site

    and Development Drilling","Mine Production of Uranium ","Uranium Concentrate Production ","Uranium Concentrate Shipments ","Employment " "Year","Drilling (million feet)"," ...

  18. DOE/NNSA Successfully Establishes Uranium Lease and Takeback Program to

    National Nuclear Security Administration (NNSA)

    Support Critical Medical Isotope Production | National Nuclear Security Administration | (NNSA) Successfully Establishes Uranium Lease and Takeback Program to Support Critical Medical Isotope Production Tuesday, February 16, 2016 - 12:00am In January 2016, the U.S. Department of Energy's National Nuclear Security Administration (DOE/NNSA) successfully established the Uranium Lease and Take-Back (ULTB) program, as directed in the American Medical Isotopes Production Act of 2012, to support

  19. Isotope separation by laser means

    DOE Patents [OSTI]

    Robinson, C. Paul; Jensen, Reed J.; Cotter, Theodore P.; Greiner, Norman R.; Boyer, Keith

    1982-06-15

    A process for separating isotopes by selective excitation of isotopic species of a volatile compound by tuned laser light. A highly cooled gas of the volatile compound is produced in which the isotopic shift is sharpened and defined. Before substantial condensation occurs, the cooled gas is irradiated with laser light precisely tuned to a desired wavelength to selectively excite a particular isotopic species in the cooled gas. The laser light may impart sufficient energy to the excited species to cause it to undergo photochemical reaction or even to photoionize. Alternatively, a two-photon irradiation may be applied to the cooled gas to induce photochemical reaction or photoionization. The process is particularly applicable to the separation of isotopes of uranium and plutonium.

  20. Uranium industry annual 1996

    SciTech Connect (OSTI)

    1997-04-01

    The Uranium Industry Annual 1996 (UIA 1996) provides current statistical data on the US uranium industry`s activities relating to uranium raw materials and uranium marketing. The UIA 1996 is prepared for use by the Congress, Federal and State agencies, the uranium and nuclear electric utility industries, and the public. Data on uranium raw materials activities for 1987 through 1996 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 2006, including purchases of uranium and enrichment services, enrichment feed deliveries, uranium fuel assemblies, filled and unfilled market requirements, uranium imports and exports, and uranium inventories are shown in Chapter 2. A feature article, The Role of Thorium in Nuclear Energy, is included. 24 figs., 56 tabs.

  1. Isotope Science

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Science and Production 35 years of experience in isotope production, processing, and applications. Llllll Committed to the safe and reliable production of radioisotopes, products, and services. Contact: Kevin John LANL Isotope Program Manager kjohn@lanl.gov 505-667-3602 Sponsored by the Department of Energy National Isotope Program http://www.nuclear.energy.gov/isotopes/nelsotopes2a.html Isotopes for Environmental Science Isotopes produced at Los Alamos National Laboratory are used as

  2. Isotopes Products

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Other isotopes that have recently shipped from LANL's isotope program include cadmium-109 (X-ray fluorescence sources), arsenic-72 (medical research), and sodium-22 (PET sources).

  3. Uranium isotopes fingerprint biotic reduction (Journal Article...

    Office of Scientific and Technical Information (OSTI)

    a window into events that shaped the evolution of life on our planet. The role of microbial activity in paleo-redox processes remains unexplored due to the inability to...

  4. Research and Medical Isotope Reactor Supply | Y-12 National Security...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Research and Medical ... Research and Medical Isotope Reactor Supply Our goal is to fuel research and test reactors with low-enriched uranium. Y-12 tops the short list of the ...

  5. COPPER COATED URANIUM ARTICLE

    DOE Patents [OSTI]

    Gray, A.G.

    1958-10-01

    Various techniques and methods for obtaining coppercoated uranium are given. Specifically disclosed are a group of complex uranium coatings having successive layers of nickel, copper, lead, and tin.

  6. Uranium Marketing Annual Report -

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    2. Maximum anticipated uranium market requirements of owners and operators of U.S. ... Source: U.S. Energy Information Administration: Form EIA-858 "Uranium Marketing Annual ...

  7. Uranium Industry Annual, 1992

    SciTech Connect (OSTI)

    Not Available

    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.

  8. URANIUM EXTRACTION

    DOE Patents [OSTI]

    Harrington, C.D.; Opie, J.V.

    1958-07-01

    The recovery of uranium values from uranium ore such as pitchblende is described. The ore is first dissolved in nitric acid, and a water soluble nitrate is added as a salting out agent. The resulting feed solution is then contacted with diethyl ether, whereby the bulk of the uranyl nitrate and a portion of the impurities are taken up by the ether. This acid ether extract is then separated from the aqueous raffinate, and contacted with water causing back extractioa of the uranyl nitrate and impurities into the water to form a crude liquor. After separation from the ether extract, this crude liquor is heated to about 118 deg C to obtain molten uranyl nitrate hexahydratc. After being slightly cooled the uranyl nitrate hexahydrate is contacted with acid free diethyl ether whereby the bulk of the uranyl nitrate is dissolved into the ethcr to form a neutral ether solution while most of the impurities remain in the aqueous waste. After separation from the aqueous waste, the resultant ether solution is washed with about l0% of its volume of water to free it of any dissolved impurities and is then contacted with at least one half its volume of water whereby the uranyl nitrate is extracted into the water to form an aqueous product solution.

  9. PRODUCTION OF URANIUM TETRACHLORIDE

    DOE Patents [OSTI]

    Calkins, V.P.

    1958-12-16

    A process is descrlbed for the production of uranium tetrachloride by contacting uranlum values such as uranium hexafluoride, uranlum tetrafluoride, or uranium oxides with either aluminum chloride, boron chloride, or sodium alumlnum chloride under substantially anhydrous condltlons at such a temperature and pressure that the chlorldes are maintained in the molten form and until the uranium values are completely converted to uranlum tetrachloride.

  10. PRODUCTION OF URANIUM MONOCARBIDE

    DOE Patents [OSTI]

    Powers, R.M.

    1962-07-24

    A method of making essentially stoichiometric uranium monocarbide by pelletizing a mixture of uranium tetrafluoride, silicon, and carbon and reacting the mixture at a temperature of approximately 1500 to 1700 deg C until the reaction goes to completion, forming uranium monocarbide powder and volatile silicon tetrafluoride, is described. The powder is then melted to produce uranium monocarbide in massive form. (AEC)

  11. In-line assay monitor for uranium hexafluoride

    DOE Patents [OSTI]

    Wallace, Steven A.

    1981-01-01

    An in-line assay monitor for determining the content of uranium-235 in a uranium hexafluoride gas isotopic separation system is provided which removes the necessity of complete access to the operating parameters of the system for determining the uranium-235 content. The monitor is intended for uses such as safeguard applications to assure that weapons grade uranium is not being produced in an enrichment cascade. The method and monitor for carrying out the method involve cooling of a radiation pervious chamber connected in fluid communication with the selected point in the system to withdraw a specimen and solidify the specimen in the chamber. The specimen is irradiated by means of an ionizing radiation source of energy different from that of the 185 keV gamma emissions from the uranium-235 present in the specimen. Simultaneously, the gamma emissions from the uranium-235 of the specimen and the source emissions transmitted through the sample are counted and stored in a multiple channel analyzer. The uranium-235 content of the specimen is determined from the comparison of the accumulated 185 keV energy counts and the reference energy counts. The latter is used to measure the total uranium isotopic content of the specimen. The process eliminates the necessity of knowing the system operating conditions and yet obtains the necessary data without need for large scintillation crystals and sophisticated mechanical designs.

  12. DECONTAMINATION OF URANIUM

    DOE Patents [OSTI]

    Feder, H.M.; Chellew, N.R.

    1958-02-01

    This patent deals with the separation of rare earth and other fission products from neutron bombarded uranium. This is accomplished by melting the uranium in contact with either thorium oxide, maguesium oxide, alumnum oxide, beryllium oxide, or uranium dioxide. The melting is preferably carried out at from 1150 deg to 1400 deg C in an inert atmosphere, such as argon or helium. During this treatment a scale of uranium dioxide forms on the uranium whtch contains most of the fission products.

  13. Apparatus for storing hydrogen isotopes

    DOE Patents [OSTI]

    McMullen, John W.; Wheeler, Michael G.; Cullingford, Hatice S.; Sherman, Robert H.

    1985-01-01

    An improved method and apparatus for storing isotopes of hydrogen (especially tritium) are provided. The hydrogen gas(es) is (are) stored as hydrides of material (for example uranium) within boreholes in a block of copper. The mass of the block is critically important to the operation, as is the selection of copper, because no cooling pipes are used. Because no cooling pipes are used, there can be no failure due to cooling pipes. And because copper is used instead of stainless steel, a significantly higher temperature can be reached before the eutectic formation of uranium with copper occurs, (the eutectic of uranium with the iron in stainless steel forming at a significantly lower temperature).

  14. URANIUM DECONTAMINATION

    DOE Patents [OSTI]

    Buckingham, J.S.; Carroll, J.L.

    1959-12-22

    A process is described for reducing the extractability of ruthenium, zirconium, and niobium values into hexone contained in an aqueous nitric acid uranium-containing solution. The solution is made acid-deficient, heated to between 55 and 70 deg C, and at that temperature a water-soluble inorganic thiosulfate is added. By this, a precipitate is formed which carries the bulk of the ruthenium, and the remainder of the ruthenium as well as the zirconium and niobium are converted to a hexone-nonextractable form. The rutheniumcontaining precipitate can either be removed from the solu tion or it can be dissolved as a hexone-non-extractable compound by the addition of sodium dichromate prior to hexone extraction.

  15. Uranium industry annual 1994

    SciTech Connect (OSTI)

    1995-07-05

    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.

  16. Uranium industry annual 1998

    SciTech Connect (OSTI)

    1999-04-22

    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.

  17. Uranium Oxide Aerosol Transport in Porous Graphite

    SciTech Connect (OSTI)

    Blanchard, Jeremy; Gerlach, David C.; Scheele, Randall D.; Stewart, Mark L.; Reid, Bruce D.; Gauglitz, Phillip A.; Bagaasen, Larry M.; Brown, Charles C.; Iovin, Cristian; Delegard, Calvin H.; Zelenyuk, Alla; Buck, Edgar C.; Riley, Brian J.; Burns, Carolyn A.

    2012-01-23

    The objective of this paper is to investigate the transport of uranium oxide particles that may be present in carbon dioxide (CO2) gas coolant, into the graphite blocks of gas-cooled, graphite moderated reactors. The transport of uranium oxide in the coolant system, and subsequent deposition of this material in the graphite, of such reactors is of interest because it has the potential to influence the application of the Graphite Isotope Ratio Method (GIRM). The GIRM is a technology that has been developed to validate the declared operation of graphite moderated reactors. GIRM exploits isotopic ratio changes that occur in the impurity elements present in the graphite to infer cumulative exposure and hence the reactors lifetime cumulative plutonium production. Reference Gesh, et. al., for a more complete discussion on the GIRM technology.

  18. Paragenesis and Geochronology of the Nopal I Uranium Deposit, Mexico

    SciTech Connect (OSTI)

    M. Fayek; M. Ren

    2007-02-14

    Uranium deposits can, by analogy, provide important information on the long-term performance of radioactive waste forms and radioactive waste repositories. Their complex mineralogy and variable elemental and isotopic compositions can provide important information, provided that analyses are obtained on the scale of several micrometers. Here, we present a structural model of the Nopal I deposit as well as petrography at the nanoscale coupled with preliminary U-Th-Pb ages and O isotopic compositions of uranium-rich minerals obtained by Secondary Ion Mass Spectrometry (SIMS). This multi-technique approach promises to provide ''natural system'' data on the corrosion rate of uraninite, the natural analogue of spent nuclear fuel.

  19. Process for electroslag refining of uranium and uranium alloys

    DOE Patents [OSTI]

    Lewis, P.S. Jr.; Agee, W.A.; Bullock, J.S. IV; Condon, J.B.

    1975-07-22

    A process is described for electroslag refining of uranium and uranium alloys wherein molten uranium and uranium alloys are melted in a molten layer of a fluoride slag containing up to about 8 weight percent calcium metal. The calcium metal reduces oxides in the uranium and uranium alloys to provide them with an oxygen content of less than 100 parts per million. (auth)

  20. URANIUM RECOVERY PROCESS

    DOE Patents [OSTI]

    Bailes, R.H.; Long, R.S.; Olson, R.S.; Kerlinger, H.O.

    1959-02-10

    A method is described for recovering uranium values from uranium bearing phosphate solutions such as are encountered in the manufacture of phosphate fertilizers. The solution is first treated with a reducing agent to obtain all the uranium in the tetravalent state. Following this reduction, the solution is treated to co-precipitate the rcduced uranium as a fluoride, together with other insoluble fluorides, thereby accomplishing a substantially complete recovery of even trace amounts of uranium from the phosphate solution. This precipitate usually takes the form of a complex fluoride precipitate, and after appropriate pre-treatment, the uranium fluorides are leached from this precipitate and rccovered from the leach solution.

  1. PRODUCTION OF PURIFIED URANIUM

    DOE Patents [OSTI]

    Burris, L. Jr.; Knighton, J.B.; Feder, H.M.

    1960-01-26

    A pyrometallurgical method for processing nuclear reactor fuel elements containing uranium and fission products and for reducing uranium compound; to metallic uranium is reported. If the material proccssed is essentially metallic uranium, it is dissolved in zinc, the sulution is cooled to crystallize UZn/sub 9/ , and the UZn/sub 9/ is distilled to obtain uranium free of fission products. If the material processed is a uranium compound, the sollvent is an alloy of zinc and magnesium and the remaining steps are the same.

  2. Method of recovering uranium hexafluoride

    DOE Patents [OSTI]

    Schuman, S.

    1975-12-01

    A method of recovering uranium hexafluoride from gaseous mixtures which comprises adsorbing said uranium hexafluoride on activated carbon is described.

  3. Enriching stable isotopes: Alternative use for Urenco technology

    SciTech Connect (OSTI)

    Rakhorst, H.; de Jong, P.G.T.; Dawson, P.D.

    1996-12-31

    The International Urenco Group utilizes a technologically advanced centrifuge process to enrich uranium in the fissionable isotope {sup 235}U. The group operates plants in the United Kingdom, the Netherlands, and Germany and currently holds a 10% share of the multibillion dollar world enrichment market. In the early 1990s, Urenco embarked on a strategy of building on the company`s uniquely advanced centrifuge process and laser isotope separation (LIS) experience to enrich nonradioactive isotopes colloquially known as stable isotopes. This paper summarizes the present status of Urenco`s stable isotopes business.

  4. Uranium-series constraints on radionuclide transport and groundwater flow at the Nopal I uranium deposit, Sierra Pena Blanca, Mexico

    SciTech Connect (OSTI)

    Goldstein, S.J.; Abdel-Fattah, A.I.; Murrell, M.T.; Dobson, P.F.; Norman, D.E.; Amato, R.S.; Nunn, A. J.

    2009-10-01

    Uranium-series data for groundwater samples from the Nopal I uranium ore deposit were obtained to place constraints on radionuclide transport and hydrologic processes for a nuclear waste repository located in fractured, unsaturated volcanic tuff. Decreasing uranium concentrations for wells drilled in 2003 are consistent with a simple physical mixing model that indicates that groundwater velocities are low ({approx}10 m/y). Uranium isotopic constraints, well productivities, and radon systematics also suggest limited groundwater mixing and slow flow in the saturated zone. Uranium isotopic systematics for seepage water collected in the mine adit show a spatial dependence which is consistent with longer water-rock interaction times and higher uranium dissolution inputs at the front adit where the deposit is located. Uranium-series disequilibria measurements for mostly unsaturated zone samples indicate that {sup 230}Th/{sup 238}U activity ratios range from 0.005-0.48 and {sup 226}Ra/{sup 238}U activity ratios range from 0.006-113. {sup 239}Pu/{sup 238}U mass ratios for the saturated zone are <2 x 10{sup -14}, and Pu mobility in the saturated zone is >1000 times lower than the U mobility. Saturated zone mobility decreases in the order {sup 238}U{approx}{sup 226}Ra > {sup 230}Th{approx}{sup 239}Pu. Radium and thorium appear to have higher mobility in the unsaturated zone based on U-series data from fractures and seepage water near the deposit.

  5. NICKEL COATED URANIUM ARTICLE

    DOE Patents [OSTI]

    Gray, A.G.

    1958-10-01

    Nickel coatings on uranium and various methods of obtaining such coatings are described. Specifically disclosed are such nickel or nickel alloy layers as barriers between uranium and aluminum- silicon, chromium, or copper coatings.

  6. Uranium Marketing Annual Report -

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    5. Shipments of uranium feed by owners and operators of U.S. civilian nuclear power ... Source: U.S. Energy Information Administration: Form EIA-858 "Uranium Marketing Annual ...

  7. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    Inventories of uranium by owner as of end of year, 2011-15 thousand pounds U3O8 equivalent Inventories at the end of the year Owner of uranium inventory 2011 2012 2013 2014 P2015 ...

  8. Uranium Marketing Annual Report

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Uranium sellers to owners and operators of U.S. civilian nuclear power reactors, 2013-15 2013 2014 2015 American Fuel Resources, LLC Advance Uranium Asset Management Ltd. AREVA ...

  9. Uranium Marketing Annual Report -

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    0. Contracted purchases of uranium from suppliers by owners and operators of U.S. civilian ... Source: U.S. Energy Information Administration, Form EIA-858 "Uranium Marketing Annual ...

  10. Uranium Marketing Annual Report

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    a. Foreign purchases, foreign sales, and uranium inventories owned by U.S. suppliers and ... Foreign sales U.S. supplier owned uranium inventories Owners and operators of U.S. ...

  11. Uranium Marketing Annual Report -

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    Uranium in fuel assemblies loaded into U.S. civilian nuclear power reactors by year, 2011-15 thousand pounds U3O8 equivalent Origin of uranium 2011 2012 2013 2014 P2015 ...

  12. METHOD FOR PURIFYING URANIUM

    DOE Patents [OSTI]

    Knighton, J.B.; Feder, H.M.

    1960-04-26

    A process is given for purifying a uranium-base nuclear material. The nuclear material is dissolved in zinc or a zinc-magnesium alloy and the concentration of magnesium is increased until uranium precipitates.

  13. Uranium-Series Constraints on Subrepository Water Flow at Yucca Mountain, Nevada

    SciTech Connect (OSTI)

    L.A. Neymark; J.B. Paces; S.J. Chipera; D.T. Vaniman

    2006-03-10

    Mineral abundances and whole-rock chemical and uranium-series isotopic compositions were measured in unfractured and rubble core samples from borehole USWSD-9 in the same layers of variably zeolitized tuffs that underlie the proposed nuclear waste repository at Yucca Mountain, Nevada. Uranium concentrations and isotopic compositions also were measured in pore water from core samples from the same rock units and rock leachates representing loosely bound U adsorbed on mineral surfaces or contained in readily soluble secondary minerals. The chemical and isotopic data were used to evaluate differences in water-rock interaction between fractured and unfractured rock and between fracture surfaces and rock matrix. Samples of unfractured and rubble fragments (about 1 centimeter) core and material from fracture surfaces show similar amounts of uranium-series disequilibrium, recording a complex history of sorption and loss of uranium over the past 1 million years. The data indicate that fractures in zeolitized tuffs may not have had greater amounts of water-rock interaction than the rock matrix. The data also show that rock matrix from subrepository units is capable of scavenging uranium with elevated uranium-234/uranium-238 from percolating water and that retardation of radionuclides and dose reduction may be greater than currently credited to this aspect of the natural barrier. Uranium concentrations of pore water and the rock leachates are used to estimate long-term in situ uranium partition coefficient values greater than 7 milliliters per gram.

  14. Uranium industry annual 1995

    SciTech Connect (OSTI)

    1996-05-01

    The Uranium Industry Annual 1995 (UIA 1995) provides current statistical data on the U.S. uranium industry`s activities relating to uranium raw materials and uranium marketing. The UIA 1995 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 period 1986 through 2005 as collected on the Form EIA-858, ``Uranium Industry Annual Survey``. Data collected on the ``Uranium Industry Annual Survey`` provide a comprehensive statistical characterization of the industry`s plans and commitments for the near-term future. Where aggregate data are presented in the UIA 1995, care has been taken to protect the confidentiality of company-specific information while still conveying accurate and complete statistical data. Data on uranium raw materials activities for 1986 through 1995 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 2005, including purchases of uranium and enrichment services, enrichment feed deliveries, uranium fuel assemblies, filled and unfilled market requirements, uranium imports and exports, and uranium inventories are shown in Chapter 2. The methodology used in the 1995 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. For the reader`s convenience, metric versions of selected tables from Chapters 1 and 2 are presented in Appendix D along with the standard conversion factors used. A glossary of technical terms is at the end of the report. 14 figs., 56 tabs.

  15. PROCESS OF PURIFYING URANIUM

    DOE Patents [OSTI]

    Seaborg, G.T.; Orlemann, E.F.; Jensen, L.H.

    1958-12-23

    A method of obtaining substantially pure uranium from a uranium composition contaminated with light element impurities such as sodium, magnesium, beryllium, and the like is described. An acidic aqueous solution containing tetravalent uranium is treated with a soluble molybdate to form insoluble uranous molybdate which is removed. This material after washing is dissolved in concentrated nitric acid to obtaln a uranyl nitrate solution from which highly purified uranium is obtained by extraction with ether.

  16. PREPARATION OF URANIUM HEXAFLUORIDE

    DOE Patents [OSTI]

    Lawroski, S.; Jonke, A.A.; Steunenberg, R.K.

    1959-10-01

    A process is described for preparing uranium hexafluoride from carbonate- leach uranium ore concentrate. The briquetted, crushed, and screened concentrate is reacted with hydrogen fluoride in a fluidized bed, and the uranium tetrafluoride formed is mixed with a solid diluent, such as calcium fluoride. This mixture is fluorinated with fluorine and an inert diluent gas, also in a fluidized bed, and the uranium hexafluoride obtained is finally purified by fractional distillation.

  17. PRODUCTION OF URANIUM TETRAFLUORIDE

    DOE Patents [OSTI]

    Shaw, W.E.; Spenceley, R.M.; Teetzel, F.M.

    1959-08-01

    A method is presented for producing uranium tetrafluoride from the gaseous hexafluoride by feeding the hexafluoride into a high temperature zone obtained by the recombination of molecularly dissociated hydrogen. The molal ratio of hydrogen to uranium hexnfluoride is preferably about 3 to 1. Uranium tetrafluoride is obtained in a finely divided, anhydrous state.

  18. Final Uranium Leasing Program Programmatic Environmental Impact...

    Energy Savers [EERE]

    Uranium Leasing Program Programmatic Environmental Impact Statement (PEIS) Final Uranium Leasing Program Programmatic Environmental Impact Statement (PEIS) Uranium Leasing ...

  19. Isotope separation

    DOE Patents [OSTI]

    Bartlett, Rodney J.; Morrey, John R.

    1978-01-01

    A method and apparatus is described for separating gas molecules containing one isotope of an element from gas molecules containing other isotopes of the same element in which all of the molecules of the gas are at the same electronic state in their ground state. Gas molecules in a gas stream containing one of the isotopes are selectively excited to a different electronic state while leaving the other gas molecules in their original ground state. Gas molecules containing one of the isotopes are then deflected from the other gas molecules in the stream and thus physically separated.

  20. U.S. Uranium Reserves Estimates

    Gasoline and Diesel Fuel Update (EIA)

    Major U.S. Uranium Reserves

  1. URANIUM RECOVERY PROCESS

    DOE Patents [OSTI]

    Yeager, J.H.

    1958-08-12

    In the prior art processing of uranium ores, the ore is flrst digested with nitric acid and filtered, and the uranium values are then extracted tom the filtrate by contacting with an organic solvent. The insoluble residue has been processed separately in order to recover any uranium which it might contain. The improvement consists in contacting a slurry, composed of both solution and residue, with the organic solvent prior to filtration. Tbe result is that uranium values contained in the residue are extracted along with the uranium values contained th the solution in one step.

  2. URANIUM SEPARATION PROCESS

    DOE Patents [OSTI]

    Hyde, E.K.; Katzin, L.I.; Wolf, M.J.

    1959-07-14

    The separation of uranium from a mixture of uranium and thorium by organic solvent extraction from an aqueous solution is described. The uranium is separrted from an aqueous mixture of uranium and thorium nitrates 3 N in nitric acid and containing salting out agents such as ammonium nitrate, so as to bring ihe total nitrate ion concentration to a maximum of about 8 N by contacting the mixture with an immiscible aliphatic oxygen containing organic solvent such as diethyl carbinol, hexone, n-amyl acetate and the like. The uranium values may be recovered from the organic phase by back extraction with water.

  3. PROCESS OF RECOVERING URANIUM

    DOE Patents [OSTI]

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

    1958-10-01

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

  4. PRODUCTION OF URANIUM

    DOE Patents [OSTI]

    Spedding, F.H.; Wilhelm, H.A.; Keller, W.H.

    1958-04-15

    The production of uranium metal by the reduction of uranium tetrafluoride is described. Massive uranium metal of high purily is produced by reacting uranium tetrafluoride with 2 to 20% stoichiometric excess of magnesium at a temperature sufficient to promote the reaction and then mantaining the reaction mass in a sealed vessel at temperature in the range of 1150 to 2000 d C, under a superatomospheric pressure of magnesium for a period of time sufficient 10 allow separation of liquid uranium and liquid magnesium fluoride into separate layers.

  5. Method for converting uranium oxides to uranium metal

    DOE Patents [OSTI]

    Duerksen, Walter K.

    1988-01-01

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

  6. Isotope geochemistry

    SciTech Connect (OSTI)

    Cole, D.R.; Curtis, D.B.; DePaolo, D.J.; Gerlach, T.M.; Laul, J.C.; Shaw, H.; Smith, B.M.; Sturchio, N.C.

    1990-09-01

    This document represents the consensus of members of the ad hoc Committee on Isotope Geochemistry in the US Department of Energy; the committee is composed of researchers in isotope geochemistry from seven of the national laboratories. Information included in this document was presented at workshops at Lawrence Berkeley Laboratory (April 1989) and at Los Alamos National Laboratory (August 1989).

  7. Depleted and Recyclable Uranium in the United States: Inventories and Options

    SciTech Connect (OSTI)

    Schneider, Erich; Scopatza, Anthony; Deinert, Mark

    2007-07-01

    International consumption of uranium currently outpaces production by nearly a factor of two. Secondary supplies from dismantled nuclear weapons, along with civilian and governmental stockpiles, are being used to make up the difference but supplies are limited. Large amounts of {sup 235}U are contained in spent nuclear fuel as well as in the tails left over from past uranium enrichment. The usability of these inhomogeneous uranium supplies depends on their isotopics. We present data on the {sup 235}U content of spent nuclear fuel and depleted uranium tails in the US and discuss the factors that affect its marketability and alternative uses. (authors)

  8. About the Uranium Mine Team | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Uranium Mine Team About the Uranium Mine Team Text coming

  9. Atomic vapor laser isotope separation

    SciTech Connect (OSTI)

    Stern, R.C.; Paisner, J.A.

    1985-11-08

    Atomic vapor laser isotope separation (AVLIS) is a general and powerful technique. A major present application to the enrichment of uranium for light-water power reactor fuel has been under development for over 10 years. In June 1985 the Department of Energy announced the selection of AVLIS as the technology to meet the nation's future need for the internationally competitive production of uranium separative work. The economic basis for this decision is considered, with an indicated of the constraints placed on the process figures of merit and the process laser system. We then trace an atom through a generic AVLIS separator and give examples of the physical steps encountered, the models used to describe the process physics, the fundamental parameters involved, and the role of diagnostic laser measurements.

  10. Isotopes Products

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Isotopes Products Isotopes Products Isotopes produced at Los Alamos National Laboratory are saving lives, advancing cutting-edge research and keeping the U.S. safe. Products stress and rest Stress and rest Rb-82 PET images in a patient with dipyridamole stress-inducible lateral wall and apical ischemia. (http://www.fac.org.ar/scvc/llave/image/machac/machaci.htm#f2,3,4) Strontium-82 is supplied to our customers for use in Sr-82/Rb-82 generator technologies. The generators in turn are supplied to

  11. Radionuclide inventories : ORIGEN2.2 isotopic depletion calculation for high burnup low-enriched uranium and weapons-grade mixed-oxide pressurized-water reactor fuel assemblies.

    SciTech Connect (OSTI)

    Gauntt, Randall O.; Ross, Kyle W.; Smith, James Dean; Longmire, Pamela

    2010-04-01

    The Oak Ridge National Laboratory computer code, ORIGEN2.2 (CCC-371, 2002), was used to obtain the elemental composition of irradiated low-enriched uranium (LEU)/mixed-oxide (MOX) pressurized-water reactor fuel assemblies. Described in this report are the input parameters for the ORIGEN2.2 calculations. The rationale for performing the ORIGEN2.2 calculation was to generate inventories to be used to populate MELCOR radionuclide classes. Therefore the ORIGEN2.2 output was subsequently manipulated. The procedures performed in this data reduction process are also described herein. A listing of the ORIGEN2.2 input deck for two-cycle MOX is provided in the appendix. The final output from this data reduction process was three tables containing the radionuclide inventories for LEU/MOX in elemental form. Masses, thermal powers, and activities were reported for each category.

  12. Method of isotope separation by chemi-ionization

    DOE Patents [OSTI]

    Wexler, Sol; Young, Charles E.

    1977-05-17

    A method for separating specific isotopes present in an isotopic mixture by aerodynamically accelerating a gaseous compound to form a jet of molecules, and passing the jet through a stream of electron donor atoms whereby an electron transfer takes place, thus forming negative ions of the molecules. The molecular ions are then passed through a radiofrequency quadrupole mass filter to separate the specific isotopes. This method may be used for any compounds having a sufficiently high electron affinity to permit negative ion formation, and is especially useful for the separation of plutonium and uranium isotopes.

  13. Preparation of uranium compounds

    DOE Patents [OSTI]

    Kiplinger, Jaqueline L; Montreal, Marisa J; Thomson, Robert K; Cantat, Thibault; Travia, Nicholas E

    2013-02-19

    UI.sub.3(1,4-dioxane).sub.1.5 and UI.sub.4(1,4-dioxane).sub.2, were synthesized in high yield by reacting turnings of elemental uranium with iodine dissolved in 1,4-dioxane under mild conditions. These molecular compounds of uranium are thermally stable and excellent precursor materials for synthesizing other molecular compounds of uranium including alkoxide, amide, organometallic, and halide compounds.

  14. Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

    S2. Uranium feed deliveries, enrichment services, and uranium loaded by owners and operators of U.S. civilian nuclear power reactors, 1994-2015 million pounds U3O8 equivalent million separative work units (SWU) Year Feed deliveries by owners and operators of U.S. civilian nuclear power reactors Uranium in fuel assemblies loaded into U.S. civilian nuclear power reactors U.S.-origin enrichment services purchased Foreign-origin enrichment services purchased Total purchased enrichment services

  15. Process for continuous production of metallic uranium and uranium alloys

    DOE Patents [OSTI]

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

    1995-06-06

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

  16. Process for continuous production of metallic uranium and uranium alloys

    DOE Patents [OSTI]

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

    1995-01-01

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

  17. Uranium Dispersion & Dosimetry Model.

    Energy Science and Technology Software Center (OSTI)

    2002-03-22

    The Uranium Dispersion and Dosimetry (UDAD) program provides estimates of potential radiation exposure to individuals and to the general population in the vicinity of a uranium processing facility such as a uranium mine or mill. Only transport through the air is considered. Exposure results from inhalation, external irradiation from airborne and ground-deposited activity, and ingestion of foodstuffs. Individual dose commitments, population dose commitments, and environmental dose commitments are computed. The program was developed for applicationmore » to uranium mining and milling; however, it may be applied to dispersion of any other pollutant.« less

  18. Uranium Purchases Report

    Reports and Publications (EIA)

    1996-01-01

    Final issue. This report details natural and enriched uranium purchases as reported by owners and operators of commercial nuclear power plants. 1996 represents the most recent publication year.

  19. URANIUM RECOVERY PROCESS

    DOE Patents [OSTI]

    Kaufman, D.

    1958-04-15

    A process of recovering uranium from very low-grade ore residues is described. These low-grade uraniumcontaining hydroxide precipitates, which also contain hydrated silica and iron and aluminum hydroxides, are subjected to multiple leachings with aqueous solutions of sodium carbonate at a pH of at least 9. This leaching serves to selectively extract the uranium from the precipitate, but to leave the greater part of the silica, iron, and aluminum with the residue. The uranium is then separated from the leach liquor by the addition of an acid in sufficient amount to destroy the carbonate followed by the addition of ammonia to precipitate uranium as ammonium diuranate.

  20. PRODUCTION OF URANIUM

    DOE Patents [OSTI]

    Ruehle, A.E.; Stevenson, J.W.

    1957-11-12

    An improved process is described for the magnesium reduction of UF/sub 4/ to produce uranium metal. In the past, there have been undesirable premature reactions between the Mg and the bomb liner or the UF/sub 4/ before the actual ignition of the bomb reaction. Since these premature reactions impair the yield of uranium metal, they have been inhibited by forming a protective film upon the particles of Mg by reacting it with hydrated uranium tetrafluoride, sodium bifluoride, uranyl fluoride, or uranium trioxide. This may be accomplished by adding about 0.5 to 2% of the additive to the bomb charge.

  1. COATING URANIUM FROM CARBONYLS

    DOE Patents [OSTI]

    Gurinsky, D.H.; Storrs, S.S.

    1959-07-14

    Methods are described for making adherent corrosion resistant coatings on uranium metal. According to the invention, the uranium metal is heated in the presence of an organometallic compound such as the carbonyls of nickel, molybdenum, chromium, niobium, and tungsten at a temperature sufficient to decompose the metal carbonyl and dry plate the resultant free metal on the surface of the uranium metal body. The metal coated body is then further heated at a higher temperature to thermally diffuse the coating metal within the uranium bcdy.

  2. highly enriched uranium

    National Nuclear Security Administration (NNSA)

    and radioisotope supply capabilities of MURR and Nordion with General Atomics' selective gas extraction technology-which allows their low-enriched uranium (LEU) targets to remain...

  3. METHOD OF ROLLING URANIUM

    DOE Patents [OSTI]

    Smith, C.S.

    1959-08-01

    A method is described for rolling uranium metal at relatively low temperatures and under non-oxidizing conditions. The method involves the steps of heating the uranium to 200 deg C in an oil bath, withdrawing the uranium and permitting the oil to drain so that only a thin protective coating remains and rolling the oil coated uranium at a temperature of 200 deg C to give about a 15% reduction in thickness at each pass. The operation may be repeated to accomplish about a 90% reduction without edge cracking, checking or any appreciable increase in brittleness.

  4. Domestic Uranium Production Report

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    Resources, Inc., dba Cameco Resources Smith Ranch-Highland Operation Converse, Wyoming ... Uranium is first processed at the Nichols Ranch plant and then transported to the Smith ...

  5. Performance Assessment Transport Modeling of Uranium at the Area 5 Radioactive Waste Management Site at the Nevada National Security Site

    SciTech Connect (OSTI)

    NSTec Radioactive Waste

    2010-10-12

    Following is a brief summary of the assumptions that are pertinent to the radioactive isotope transport in the GoldSim Performance Assessment model of the Area 5 Radioactive Waste Management Site, with special emphasis on the water-phase reactive transport of uranium, which includes depleted uranium products.

  6. U.S.Uranium Reserves

    Gasoline and Diesel Fuel Update (EIA)

    Uranium Reserves Data for: 2003 Release Date: June 2004 Next Release: Not determined Uranium Reserves Estimates The Energy Information Administration (EIA) has reported the...

  7. 2015 Uranium Marketing Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    1 2015 Uranium Marketing Annual Report Release Date: May 24, 2016 Next Release Date: May ... Source: U.S. Energy Information Administration, Form EIA-858 "Uranium Marketing Annual ...

  8. 2015 Uranium Marketing Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    U.S. Energy Information Administration 2015 Uranium Marketing Annual Report 2015 Uranium Marketing Annual Report Release Date: May 24, 2016 Next Release Date: May 2017 Minimum ...

  9. 2015 Uranium Marketing Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    U.S. Energy Information Administration 2015 Uranium Marketing Annual Report 2015 Uranium Marketing Annual Report Release Date: May 24, 2016 Next Release Date: May 2017 Origin of ...

  10. 2015 Uranium Marketing Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    7 2015 Uranium Marketing Annual Report Release Date: May 24, 2016 Next Release Date: May ... Source: U.S. Energy Information Administration, Form EIA-858 "Uranium Marketing Annual ...

  11. 2015 Uranium Marketing Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    3 2015 Uranium Marketing Annual Report Release Date: May 24, 2016 Next Release Date: May ... Source: U.S. Energy Information Administration: Form EIA-858 "Uranium Marketing Annual ...

  12. 2015 Uranium Marketing Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    9 2015 Uranium Marketing Annual Report Release Date: May 24, 2016 Next Release Date: May ... Source: U.S. Energy Information Administration, Form EIA-858 "Uranium Marketing Annual ...

  13. PROCESS FOR MAKING URANIUM HEXAFLUORIDE

    DOE Patents [OSTI]

    Rosen, R.

    1959-07-14

    A process is described for producing uranium hexafluoride by reacting uranium hexachloride with hydrogen fluoride at a temperature below about 150 deg C, under anhydrous conditions.

  14. 2015 Uranium Market Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    U.S. Energy Information Administration 2015 Uranium Marketing Annual Report 2015 Uranium ... received in 2015 Weighted-average price Number of purchase contracts for ...

  15. 2015 Uranium Marketing Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    Uranium Marketing Annual Report 2015 Uranium Marketing Annual Report Release Date: May 24, 2016 Next Release Date: May 2017 Number of purchasers Quantity with reported price ...

  16. 2015 Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    data set of uranium reserves that were published in the July 2010 report U.S. Uranium Reserves Estimates at http:www.eia.govcneafnuclearpagereservesures.html. ...

  17. 2015 Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    Domestic Uranium Production Report 2015 Domestic Uranium Production Report Release Date: May 5, 2016 Next Release Date: May 2017 Number of Holes Feet (thousand) Number of Holes ...

  18. URANIUM LEACHING AND RECOVERY PROCESS

    DOE Patents [OSTI]

    McClaine, L.A.

    1959-08-18

    A process is described for recovering uranium from carbonate leach solutions by precipitating uranium as a mixed oxidation state compound. Uranium is recovered by adding a quadrivalent uranium carbon;te solution to the carbonate solution, adjusting the pH to 13 or greater, and precipitating the uranium as a filterable mixed oxidation state compound. In the event vanadium occurs with the uranium, the vanadium is unaffected by the uranium precipitation step and remains in the carbonate solution. The uranium-free solution is electrolyzed in the cathode compartment of a mercury cathode diaphragm cell to reduce and precipitate the vanadium.

  19. Disposition of DOE Excess Depleted Uranium, Natural Uranium, and Low-Enriched Uranium

    Broader source: Energy.gov [DOE]

    The U.S. Department of Energy (DOE) owns and manages an inventory of depleted uranium (DU), natural uranium (NU), and low-enriched uranium (LEU) that is currently stored in large cylinders as...

  20. Atomic vapor laser isotope separation using resonance ionization

    SciTech Connect (OSTI)

    Comaskey, B.; Crane, J.; Erbert, G.; Haynam, C.; Johnson, M.; Morris, J.; Paisner, J.; Solarz, R.; Worden, E.

    1986-09-01

    Atomic vapor laser isotope separation (AVLIS) is a general and powerful technique. A major present application to the enrichment of uranium for light-water power-reactor fuel has been under development for over 10 years. In June 1985, the Department of Energy announced the selection of AVLIS as the technology to meet the nation's future need for enriched uranium. Resonance photoionization is the heart of the AVLIS process. We discuss those fundamental atomic parameters that are necessary for describing isotope-selective resonant multistep photoionization along with the measurement techniques that we use. We illustrate the methodology adopted with examples of other elements that are under study in our program.

  1. Laser Isotope Enrichment for Medical and Industrial Applications

    SciTech Connect (OSTI)

    Leonard Bond

    2006-07-01

    Laser Isotope Enrichment for Medical and Industrial Applications by Jeff Eerkens (University of Missouri), Jay Kunze (Idaho State University), and Leonard Bond (Idaho National Laboratory) The principal isotope enrichment business in the world is the enrichment of uranium for commercial power reactor fuels. However, there are a number of other needs for separated isotopes. Some examples are: 1) Pure isotopic targets for irradiation to produce medical radioisotopes. 2) Pure isotopes for semiconductors. 3) Low neutron capture isotopes for various uses in nuclear reactors. 4) Isotopes for industrial tracer/identification applications. Examples of interest to medicine are targets to produce radio-isotopes such as S-33, Mo-98, Mo-100, W-186, Sn-112; while for MRI diagnostics, the non-radioactive Xe-129 isotope is wanted. For super-semiconductor applications some desired industrial isotopes are Si-28, Ga-69, Ge-74, Se-80, Te-128, etc. An example of a low cross section isotope for use in reactors is Zn-68 as a corrosion inhibitor material in nuclear reactor primary systems. Neutron activation of Ar isotopes is of interest in industrial tracer and diagnostic applications (e.g. oil-logging). . In the past few years there has been a sufficient supply of isotopes in common demand, because of huge Russian stockpiles produced with old electromagnetic and centrifuge separators previously used for uranium enrichment. Production of specialized isotopes in the USA has been largely accomplished using old ”calutrons” (electromagnetic separators) at Oak Ridge National Laboratory. These methods of separating isotopes are rather energy inefficient. Use of lasers for isotope separation has been considered for many decades. None of the proposed methods have attained sufficient proof of principal status to be economically attractive to pursue commercially. Some of the authors have succeeded in separating sulfur isotopes using a rather new and different method, known as condensation

  2. DECONTAMINATION OF URANIUM

    DOE Patents [OSTI]

    Spedding, F.H.; Butler, T.A.

    1962-05-15

    A process is given for separating fission products from uranium by extracting the former into molten aluminum. Phase isolation can be accomplished by selectively hydriding the uranium at between 200 and 300 deg C and separating the hydride powder from coarse particles of fissionproduct-containing aluminum. (AEC)

  3. URANIUM SEPARATION PROCESS

    DOE Patents [OSTI]

    McVey, W.H.; Reas, W.H.

    1959-03-10

    The separation of uranium from an aqueous solution containing a water soluble uranyl salt is described. The process involves adding an alkali thiocyanate to the aqueous solution, contacting the resulting solution with methyl isobutyl ketons and separating the resulting aqueous and organic phase. The uranium is extracted in the organic phase as UO/sub 2/(SCN)/sub/.

  4. Radiochronological Age of a Uranium Metal Sample from an Abandoned Facility

    SciTech Connect (OSTI)

    Meyers, L A; Williams, R W; Glover, S E; LaMont, S P; Stalcup, A M; Spitz, H B

    2012-03-16

    A piece of scrap uranium metal bar buried in the dirt floor of an old, abandoned metal rolling mill was analyzed using multi-collector inductively coupled plasma mass spectroscopy (MC-ICP-MS). The mill rolled uranium rods in the 1940s and 1950s. Samples of the contaminated dirt in which the bar was buried were also analyzed. The isotopic composition of uranium in the bar and dirt samples were both the same as natural uranium, though a few samples of dirt also contained recycled uranium; likely a result of contamination with other material rolled at the mill. The time elapsed since the uranium metal bar was last purified can be determined by the in-growth of the isotope {sup 230}Th from the decay of {sup 234}U, assuming that only uranium isotopes were present in the bar after purification. The age of the metal bar was determined to be 61 years at the time of this analysis and corresponds to a purification date of July 1950 {+-} 1.5 years.

  5. Uranium dioxide electrolysis

    SciTech Connect (OSTI)

    Willit, James L.; Ackerman, John P.; Williamson, Mark A.

    2009-12-29

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

  6. URANIUM PRECIPITATION PROCESS

    DOE Patents [OSTI]

    Thunaes, A.; Brown, E.A.; Smith, H.W.; Simard, R.

    1957-12-01

    A method for the recovery of uranium from sulfuric acid solutions is described. In the present process, sulfuric acid is added to the uranium bearing solution to bring the pH to between 1 and 1.8, preferably to about 1.4, and aluminum metal is then used as a reducing agent to convert hexavalent uranium to the tetravalent state. As the reaction proceeds, the pH rises amd a selective precipitation of uranium occurs resulting in a high grade precipitate. This process is an improvement over the process using metallic iron, in that metallic aluminum reacts less readily than metallic iron with sulfuric acid, thus avoiding consumption of the reducing agent and a raising of the pH without accomplishing the desired reduction of the hexavalent uranium in the solution. Another disadvantage to the use of iron is that positive ferric ions will precipitate with negative phosphate and arsenate ions at the pH range employed.

  7. Laser-isotope-separation technology. [Review; economics

    SciTech Connect (OSTI)

    Jensen, R.J.; Blair, L.S.

    1981-01-01

    The Molecular Laser Isotope Separation (MLIS) process currently under development is discussed as an operative example of the use of lasers for material processing. The MLIS process, which uses infrared and ultraviolet lasers to process uranium hexafluoride (UF/sub 6/) resulting in enriched uranium fuel to be used in electrical-power-producing nuclear reactor, is reviewed. The economics of the MLIS enrichment process is compared with conventional enrichment technique, and the projected availability of MLIS enrichment capability is related to estimated demands for U.S. enrichment service. The lasers required in the Los Alamos MLIS program are discussed in detail, and their performance and operational characteristics are summarized. Finally, the timely development of low-cost, highly efficient ultraviolet and infrared lasers is shownd to be the critical element controlling the ultimate deployment of MLIS uranium enrichment. 8 figures, 7 tables.

  8. HEU Minimization and the Reliable Supply of Medical Isotopes Nuclear

    National Nuclear Security Administration (NNSA)

    Security Summit: Fact Sheet | National Nuclear Security Administration | (NNSA) HEU Minimization and the Reliable Supply of Medical Isotopes Nuclear Security Summit: Fact Sheet March 26, 2012 Molybdenum-99 (Mo-99) is used to produce technetium-99m (Tc-99m), a medical isotope that is used in about 100,000 diagnostic medical procedures globally every day. Today, Mo-99 is produced at aging facilities in Europe, Canada and South Africa primarily using highly-enriched uranium (HEU) - a

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

    SciTech Connect (OSTI)

    Young, Carl; Weismann, Joseph; Caputo, Daniel [Cabrera Services, Inc., East Hartford, Connecticut (United States)

    2007-07-01

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

  10. US developments in technology for uranium enrichment

    SciTech Connect (OSTI)

    Wilcox, W.J. Jr.; McGill, R.M.

    1982-01-01

    The purpose of this paper is to review recent progress and the status of the work in the United States on that part of the fuel cycle concerned with uranium enrichment. The United States has one enrichment process, gaseous diffusion, which has been continuously operated in large-scale production for the past 37 years; another process, gas centrifugation, which is now in the construction phase; and three new processes, molecular laser isotope separation, atomic vapor laser isotope separation, plasma separation process, in which the US has also invested sizable research and development efforts over the last few years. The emphasis in this paper is on the technical aspects of the various processes, but the important economic factors which will define the technological mix which may be applied in the next two decades are also discussed.

  11. ISOTOPE SEPARATORS

    DOE Patents [OSTI]

    Bacon, C.G.

    1958-08-26

    An improvement is presented in the structure of an isotope separation apparatus and, in particular, is concerned with a magnetically operated shutter associated with a window which is provided for the purpose of enabling the operator to view the processes going on within the interior of the apparatus. The shutier is mounted to close under the force of gravity in the absence of any other force. By closing an electrical circuit to a coil mouated on the shutter the magnetic field of the isotope separating apparatus coacts with the magnetic field of the coil to force the shutter to the open position.

  12. I ISOTOPES

    Office of Legacy Management (LM)

    fl6-6 ' , WTELEEYNE I ISOTOPES i - ' 50<77 /,' y. 6 IWL-5025-473 SUBSURFACE URASIUM OJ: THE GROUNDS OF NL BEARINGS, ALBAh'Y Heyitt Iv. Jeter Douglas M. Eagleson Fred J. Frullo TELEDYNE ISOTOPES 50 VAK BUREN A\!EMJE WESTKOOD, NEK JERSEY 07675 7 Dcccmhcr 1953 Prepnrcd for NL f%carings/NL Tndustrics, Inc. 1130 CCVltrill AXr~lMIC Allmy, New York 12205 TABLE OF CONTEhTS 1.0 INTRODUCTION 2.0 METHODS 2.1 Soil Sampling 2.2 Sample Preparation 2.3 Analysis of Samples 3.0 RESULTS 4.0 SUMMARY REFERENCES

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

    SciTech Connect (OSTI)

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

    1995-11-30

    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.

  14. PROCESS OF RECOVERING URANIUM

    DOE Patents [OSTI]

    Kilner, S.B.

    1959-12-29

    A method is presented for separating and recovering uranium from a complex mixure of impurities. The uranium is dissolved to produce an aqueous acidic solution including various impurities. In accordance with one method, with the uranium in the uranyl state, hydrogen cyanide is introduced into the solution to complex the impurities. Subsequently, ammonia is added to the solution to precipitate the uraniunn as ammonium diuranate away from the impurities in the solution. Alternatively, the uranium is precipitated by adding an alkaline metal hydroxide. In accordance with the second method, the uranium is reduced to the uranous state in the solution. The reduced solution is then treated with solid alkali metal cyanide sufficient to render the solution about 0.1 to 1.0 N in cyanide ions whereat cyanide complex ions of the metal impurities are produced and the uranium is simultaneously precipituted as uranous hydroxide. Alternatively, hydrogen cyanide may be added to the reduced solution and the uranium precipitated subsequently by adding ammonium hydroxide or an alkali metal hydroxide. Other refinements of the method are also disclosed.

  15. India's Worsening Uranium Shortage

    SciTech Connect (OSTI)

    Curtis, Michael M.

    2007-01-15

    As a result of NSG restrictions, India cannot import the natural uranium required to fuel its Pressurized Heavy Water Reactors (PHWRs); consequently, it is forced to rely on the expediency of domestic uranium production. However, domestic production from mines and byproduct sources has not kept pace with demand from commercial reactors. This shortage has been officially confirmed by the Indian Planning Commission’s Mid-Term Appraisal of the country’s current Five Year Plan. The report stresses that as a result of the uranium shortage, Indian PHWR load factors have been continually decreasing. The Uranium Corporation of India Ltd (UCIL) operates a number of underground mines in the Singhbhum Shear Zone of Jharkhand, and it is all processed at a single mill in Jaduguda. UCIL is attempting to aggrandize operations by establishing new mines and mills in other states, but the requisite permit-gathering and development time will defer production until at least 2009. A significant portion of India’s uranium comes from byproduct sources, but a number of these are derived from accumulated stores that are nearing exhaustion. A current maximum estimate of indigenous uranium production is 430t/yr (230t from mines and 200t from byproduct sources); whereas, the current uranium requirement for Indian PHWRs is 455t/yr (depending on plant capacity factor). This deficit is exacerbated by the additional requirements of the Indian weapons program. Present power generation capacity of Indian nuclear plants is 4350 MWe. The power generation target set by the Indian Department of Atomic Energy (DAE) is 20,000 MWe by the year 2020. It is expected that around half of this total will be provided by PHWRs using indigenously supplied uranium with the bulk of the remainder provided by breeder reactors or pressurized water reactors using imported low-enriched uranium.

  16. Depleted uranium management alternatives

    SciTech Connect (OSTI)

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

    1994-08-01

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

  17. The IMCA: A field instrument for uranium enrichment measurements

    SciTech Connect (OSTI)

    Gardner, G.H.; Koskelo, M.; Moeslinger, M.; Mayer, R.L. II; McGinnis, B.R.; Wishard, B.

    1996-12-31

    The IMCA (Inspection Multi-Channel Analyzer) is a portable gamma-ray spectrometer designed to measure the enrichment of uranium either in a laboratory or in the field. The IMCA consists of a Canberra InSpector Multi-Channel Analyzer, sodium iodide or a planar germanium detector, and special application software. The system possesses a high degree of automation. The IMCA uses the uranium enrichment meter principle, and is designed to meet the International Atomic Energy Agency (IAEA) requirements for the verification of enriched uranium materials. The IMCA is available with MGA plutonium isotopic analysis software or MGAU uranium analysis software as well. In this paper, the authors present a detailed description of the hardware and software of the IMCA system, as well as results from preliminary measurements testing compliance of IMCA with IAEA requirements using uranium standards and UF6 cylinders. Measurements performed on UF6 cylinders in the field under variable environmental conditions (temperatures ranging from 0 to 35 C) have shown that good results can be achieved. The enrichment of UF6 contained in the cylinder is determined by using calibration constants generated from an instrument calibration, using traceable uranium oxide standards, performed in the laboratory under controlled environmental conditions. The IMCA software is designed to make the necessary matrix and container corrections to ensure that accurate results are achieved in the field.

  18. URANIUM RECOVERY PROCESS

    DOE Patents [OSTI]

    Stevenson, J.W.; Werkema, R.G.

    1959-07-28

    The recovery of uranium from magnesium fluoride slag obtained as a by- product in the production of uranium metal by the bomb reduction prccess is presented. Generally the recovery is accomplished by finely grinding the slag, roasting ihe ground slag air, and leaching the roasted slag with a hot, aqueous solution containing an excess of the sodium bicarbonate stoichiometrically required to form soluble uranium carbonate complex. The roasting is preferably carried out at between 425 and 485 deg C for about three hours. The leaching is preferably done at 70 to 90 deg C and under pressure. After leaching and filtration the uranium may be recovered from the clear leach liquor by any desired method.

  19. Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    9. Summary production statistics of the U.S. uranium industry, 1993-2015 Year Exploration and development surface drilling (million feet) Exploration and development drilling expenditures 1 (million dollars) Mine production of uranium (million pounds U3O8) Uranium concentrate production (million pounds U3O8) Uranium concentrate shipments (million pounds U3O8) Employment (person-years) 1993 1.1 5.7 2.1 3.1 3.4 871 1994 0.7 1.1 2.5 3.4 6.3 980 1995 1.3 2.6 3.5 6.0 5.5 1,107 1996 3.0 7.2 4.7 6.3

  20. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    4. Deliveries of uranium feed for enrichment by owners and operators of U.S. civilian nuclear power reactors by origin country and delivery year, 2013-15 thousand pounds U3O8 ...

  1. Uranium Marketing Annual Report -

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    9. Contracted purchases of uranium by owners and operators of U.S. civilian nuclear power reactors, signed in 2015, by delivery year, 2016-25 thousand pounds U3O8 equivalent Year ...

  2. PURIFICATION OF URANIUM FUELS

    DOE Patents [OSTI]

    Niedrach, L.W.; Glamm, A.C.

    1959-09-01

    An electrolytic process of refining or decontaminating uranium is presented. The impure uranium is made the anode of an electrolytic cell. The molten salt electrolyte of this cell comprises a uranium halide such as UF/sub 4/ or UCl/sub 3/ and an alkaline earth metal halide such as CaCl/sub 2/, BaF/sub 2/, or BaCl/sub 2/. The cathode of the cell is a metal such as Mn, Cr, Co, Fe, or Ni which forms a low melting eutectic with U. The cell is operated at a temperature below the melting point of U. In operation the electrodeposited uranium becomes alloyed with the metal of the cathode, and the low melting alloy thus formed drips from the cathode.

  3. ANODIC TREATMENT OF URANIUM

    DOE Patents [OSTI]

    Kolodney, M.

    1959-02-01

    A method is presented for effecting eloctrolytic dissolution of a metallic uranium article at a uniform rate. The uranium is made the anode in an aqueous phosphoric acid solution containing nitrate ions furnished by either ammonium nitrate, lithium nitrate, sodium nitrate, or potassium nitrate. A stainless steel cathode is employed and electrolysls carried out at a current density of about 0.1 to 1 ampere per square inch.

  4. URANIUM EXTRACTION PROCESS

    DOE Patents [OSTI]

    Baldwin, W.H.; Higgins, C.E.

    1958-12-16

    A process is described for recovering uranium values from acidic aqueous solutions containing hexavalent uranium by contacting the solution with an organic solution comprised of a substantially water-immiscible organlc diluent and an organic phosphate to extract the uranlum values into the organic phase. Carbon tetrachloride and a petroleum hydrocarbon fraction, such as kerosene, are sultable diluents to be used in combination with organlc phosphates such as dibutyl butylphosphonate, trlbutyl phosphine oxide, and tributyl phosphate.

  5. Method for the recovery of uranium values from uranium tetrafluoride

    DOE Patents [OSTI]

    Kreuzmann, Alvin B.

    1983-01-01

    The invention is a novel method for the recovery of uranium from dry, particulate uranium tetrafluoride. In one aspect, the invention comprises reacting particulate uranium tetrafluoride and calcium oxide in the presence of gaseous oxygen to effect formation of the corresponding alkaline earth metal uranate and alkaline earth metal fluoride. The product uranate is highly soluble in various acidic solutions wherein the product fluoride is virtually insoluble therein. The product mixture of uranate and alkaline earth metal fluoride is contacted with a suitable acid to provide a uranium-containing solution, from which the uranium is recovered. The invention can achieve quantitative recovery of uranium in highly pure form.

  6. Method for the recovery of uranium values from uranium tetrafluoride

    DOE Patents [OSTI]

    Kreuzmann, A.B.

    1982-10-27

    The invention is a novel method for the recovery of uranium from dry, particulate uranium tetrafluoride. In one aspect, the invention comprises reacting particulate uranium tetrafluoride and calcium oxide in the presence of gaseous oxygen to effect formation of the corresponding alkaline earth metal uranate and alkaline earth metal fluoride. The product uranate is highly soluble in various acidic solutions whereas the product fluoride is virtually insoluble therein. The product mixture of uranate and alkaline earth metal fluoride is contacted with a suitable acid to provide a uranium-containing solution, from which the uranium is recovered. The invention can achieve quantitative recovery of uranium in highly pure form.

  7. 230Th-234U Age-Dating Uranium by Mass Spectrometry

    SciTech Connect (OSTI)

    Williams, R W; Gaffney, A M

    2012-04-18

    This is the standard operating procedure used by the Isotope Ratio Mass Spectrometry Group of the Chemical Sciences Division at LLNL for the preparation of a sample of uranium oxide or uranium metal for {sup 230}Th-{sup 234}U age-dating. The method described here includes the dissolution of a sample of uranium oxide or uranium metal, preparation of a secondary dilution, spiking of separate aliquots for uranium and thorium isotope dilution measurements, and purification of uranium and thorium aliquots for mass spectrometry. This SOP may be applied to uranium samples of unknown purity as in a nuclear forensic investigation, and also to well-characterized samples such as, for example, U{sub 3}O{sub 8} and U-metal certified reference materials. The sample of uranium is transferred to a quartz or PFA vial, concentrated nitric acid is added and the sample is heated on a hotplate at approximately 100 C for several hours until it dissolves. The sample solution is diluted with water to make the solution approximately 4 M HNO{sub 3} and hydrofluoric acid is added to make it 0.05 M HF. A secondary dilution of the primary uranium solution is prepared. Separate aliquots for uranium and thorium isotope dilution measurements are taken and spiked with {sup 233}U and {sup 229}Th, respectively. The spiked aliquot for uranium isotope dilution analysis is purified using EiChrom UTEVA resin. The spiked aliquot for thorium isotope dilution analysis is purified by, first, a 1.8 mL AG1x8 resin bed in 9 M HCl on which U adsorbs and Th passes through; second, adsorbing Th on a 1 mL AG1x8 resin bed in 8 M HNO{sub 3} and then eluting it with 9 M HCl followed by 0.1 M HCl + 0.005 M HF; and third, by passing the Th through a final 1.0 mL AG1x8 resin bed in 9 M HCl. The mass spectrometry is performed using the procedure 'Th and U Mass Spectrometry for {sup 230}Th-{sup 234}U Age Dating'.

  8. Perimeter safeguards techniques for uranium-enrichment plants

    SciTech Connect (OSTI)

    Fehlau, P.E.; Chamber, W.H.

    1981-09-01

    In 1972, a working group of the International Atomic Energy Agency identified a goal to develop and evaluate perimeter safeguards for uranium isotope enrichment plants. As part of the United State's response to that goal, Los Alamos Detection and Verification personnel studied gamma-ray and neutron emissions from uranium hexafluoride. They developed instruments that use the emissions to verify uranium enrichment and to monitor perimeter personnel and shipping portals. Unattended perimeter monitors and hand-held verification instruments were evaluated in field measurements and, when possible, were loaned to enrichment facilities for trials. None of the seven package monitoring techniques that were investigated proved entirely satisfactory for an unattended monitor. They either revealed proprietary information about centrifuge design or were subject to interference by shielding materials that could be present in a package. Further evaluation in a centrifuge facility may help in developing an acceptable attended package monitor. 34 figures, 9 tables.

  9. Depleted uranium as a backfill for nuclear fuel waste package

    DOE Patents [OSTI]

    Forsberg, C.W.

    1998-11-03

    A method is described for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package. 6 figs.

  10. Depleted uranium as a backfill for nuclear fuel waste package

    DOE Patents [OSTI]

    Forsberg, Charles W.

    1998-01-01

    A method for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package.

  11. Nuclear Fuel Facts: Uranium | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Facts: Uranium Nuclear Fuel Facts: Uranium Nuclear Fuel Facts: Uranium Uranium is a silvery-white metallic chemical element in the periodic table, with atomic number 92. It is assigned the chemical symbol U. A uranium atom has 92 protons and 92 electrons, of which 6 are valence electrons. Uranium has the highest atomic weight (19 kg m) of all naturally occurring elements. Uranium occurs naturally in low concentrations in soil, rock and water, and is commercially extracted from uranium-bearing

  12. Uranium Processing Facility Team Signs Partnering Agreement ...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Processing Facility ... Uranium Processing Facility Team Signs Partnering Agreement ... Nuclear Security, LLC; John Eschenberg, Uranium Processing Facility Project Office; Brian ...

  13. Influence of uranium hydride oxidation on uranium metal behaviour

    SciTech Connect (OSTI)

    Patel, N.; Hambley, D.; Clarke, S.A.; 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)

  14. Process for electrolytically preparing uranium metal

    DOE Patents [OSTI]

    Haas, Paul A.

    1989-08-01

    A process for making uranium metal from uranium oxide by first fluorinating uranium oxide to form uranium tetrafluoride and next electrolytically reducing the uranium tetrafluoride with a carbon anode to form uranium metal and CF.sub.4. The CF.sub.4 is reused in the fluorination reaction rather than being disposed of as a hazardous waste.

  15. Process for electrolytically preparing uranium metal

    DOE Patents [OSTI]

    Haas, Paul A.

    1989-01-01

    A process for making uranium metal from uranium oxide by first fluorinating uranium oxide to form uranium tetrafluoride and next electrolytically reducing the uranium tetrafluoride with a carbon anode to form uranium metal and CF.sub.4. The CF.sub.4 is reused in the fluorination reaction rather than being disposed of as a hazardous waste.

  16. PRODUCTION OF URANIUM HEXAFLUORIDE

    DOE Patents [OSTI]

    Fowler, R.D.

    1957-08-27

    A process for the production of uranium hexafluoride from the oxides of uranium is reported. In accordance with the method, the higher oxides of uranium may be reduced to uranium dioxide (UO/sub 2/), the latter converted into uranium tetrafluoride by reaction with hydrogen fluoride, and the UF/sub 4/ converted to UF/sub 6/ by reaction with a fluorinating agent, such as CoF/sub 3/. The UO/sub 3/ or U/sub 3/O/sub 8/ is placed in a reac tion chamber in a copper boat or tray enclosed in a copper oven, and heated to 500 to 650 deg C while hydrogen gas is passed through the oven. After nitrogen gas is used to sweep out the hydrogen and the water vapor formed, and while continuing to inaintain the temperature between 400 deg C and 600 deg C, anhydrous hydrogen fluoride is passed through. After completion of the conversion of UO/sub 2/ to UF/sub 4/ the temperature of the reaction chamber is lowered to about 400 deg C or less, the UF/sub 4/ is mixed with the requisite quantity of CoF/sub 3/, and after evacuating the chamber, the mixture is heated to 300 to 400 deg C, and the resulting UF/sub 6/ is led off and delivered to a condenser.

  17. Uranium hexafluoride handling. Proceedings

    SciTech Connect (OSTI)

    Not Available

    1991-12-31

    The United States Department of Energy, Oak Ridge Field Office, and Martin Marietta Energy Systems, Inc., are co-sponsoring this Second International Conference on Uranium Hexafluoride Handling. The conference is offered as a forum for the exchange of information and concepts regarding the technical and regulatory issues and the safety aspects which relate to the handling of uranium hexafluoride. Through the papers presented here, we attempt not only to share technological advances and lessons learned, but also to demonstrate that we are concerned about the health and safety of our workers and the public, and are good stewards of the environment in which we all work and live. These proceedings are a compilation of the work of many experts in that phase of world-wide industry which comprises the nuclear fuel cycle. Their experience spans the entire range over which uranium hexafluoride is involved in the fuel cycle, from the production of UF{sub 6} from the naturally-occurring oxide to its re-conversion to oxide for reactor fuels. The papers furnish insights into the chemical, physical, and nuclear properties of uranium hexafluoride as they influence its transport, storage, and the design and operation of plant-scale facilities for production, processing, and conversion to oxide. The papers demonstrate, in an industry often cited for its excellent safety record, continuing efforts to further improve safety in all areas of handling uranium hexafluoride. Selected papers were processed separately for inclusion in the Energy Science and Technology Database.

  18. METHOD OF RECOVERING URANIUM COMPOUNDS

    DOE Patents [OSTI]

    Poirier, R.H.

    1957-10-29

    S>The recovery of uranium compounds which have been adsorbed on anion exchange resins is discussed. The uranium and thorium-containing residues from monazite processed by alkali hydroxide are separated from solution, and leached with an alkali metal carbonate solution, whereby the uranium and thorium hydrorides are dissolved. The carbonate solution is then passed over an anion exchange resin causing the uranium to be adsorbed while the thorium remains in solution. The uranium may be recovered by contacting the uranium-holding resin with an aqueous ammonium carbonate solution whereby the uranium values are eluted from the resin and then heating the eluate whereby carbon dioxide and ammonia are given off, the pH value of the solution is lowered, and the uranium is precipitated.

  19. 2015 Uranium Marketing Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    2015 Uranium Marketing Annual Report 2015 Uranium Marketing Annual Report Release Date: May 24, 2016 Next Release Date: May 2017 thousand pounds U 3 O 8 equivalent Year Maximum ...

  20. 2015 Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

    Uranium Marketing Annual Report 2015 Uranium Marketing Annual Report Release Date: May 24, 2016 Next Release Date: May 2017 2014 2015 2014 2015 2014 2015 Weighted-average price ...

  1. 2015 Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

    Figure 3. Uranium purchased by owners and operators of U.S. civilian nuclear power reactors by origin and delivery year, 2011-15 Figure 4. Weighted-average price of uranium ...

  2. 2015 Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    Domestic Uranium Production Report 2015 Domestic Uranium Production Report Release Date: May 5, 2016 Next Release Date: May 2017 Table 9. Summary production statistics of the U.S. ...

  3. 2015 Domestic Uranium Production Report

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    Domestic Uranium Production Report 2015 Domestic Uranium Production Report Release Date: May 5, 2016 Next Release Date: May 2017 State(s) 2003 2004 2005 2006 2007 2008 2009 2010 ...

  4. Uranium-titanium-niobium alloy

    DOE Patents [OSTI]

    Ludtka, Gail M.; Ludtka, Gerard M.

    1990-01-01

    A uranium alloy having small additions of Ti and Nb shows improved strength and ductility in cross section of greater than one inch over prior uranium alloy having only Ti as an alloying element.

  5. METHOD OF SINTERING URANIUM DIOXIDE

    DOE Patents [OSTI]

    Henderson, C.M.; Stavrolakis, J.A.

    1963-04-30

    This patent relates to a method of sintering uranium dioxide. Uranium dioxide bodies are heated to above 1200 nif- C in hydrogen, sintered in steam, and then cooled in hydrogen. (AEC)

  6. Excess Uranium Inventory Management Plan

    Office of Energy Efficiency and Renewable Energy (EERE)

    The 2013 Excess Uranium Inventory Management Plan describes a framework for the effective management of the Energy Department’s surplus uranium inventory in support of meeting its critical...

  7. uranium | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    uranium Klotz visits Y-12 to see progress on new projects and ongoing work on NNSA's national security missions Last week, NNSA Administrator Lt. Gen. Frank Klotz (Ret.) visited the Y-12 National Security Complex to check on the status of ongoing projects like the Uranium Processing Facility as well as the site's continuing uranium operations. He also met with the Region 2 volunteers of the Radiogical... NNSA Announces Arrival of Plutonium and Uranium from Japan's Fast Critical Assembly at

  8. EXTRACTION OF URANIUM

    DOE Patents [OSTI]

    Kesler, R.D.; Rabb, D.D.

    1959-07-28

    An improved process is presented for recovering uranium from a carnotite ore. In the improved process U/sub 2/O/sub 5/ is added to the comminuted ore along with the usual amount of NaCl prior to roasting. The amount of U/sub 2/O/ sub 5/ is dependent on the amount of free calcium oxide and the uranium in the ore. Specifically, the desirable amount of U/sub 2/O/sub 5/ is 3.2% for each 1% of CaO, and 5 to 6% for each 1% of uranium. The mixture is roasted at about 1560 deg C for about 30 min and then leached with a 3 to 9% aqueous solution of sodium carbonate.

  9. Process for recovering uranium

    DOE Patents [OSTI]

    MacWood, G. E.; Wilder, C. D.; Altman, D.

    1959-03-24

    A process useful in recovering uranium from deposits on stainless steel liner surfaces of calutrons is presented. The deposit is removed from the stainless steel surface by washing with aqueous nitric acid. The solution obtained containing uranium, chromium, nickel, copper, and iron is treated with an excess of ammonium hydroxide to precipitnte the uranium, iron, and chromium and convert the nickel and copper to soluble ammonio complexions. The precipitated material is removed, dried and treated with carbon tetrachloride at an elevated temperature of about 500 to 600 deg C to form a vapor mixture of UCl/ sub 4/, UCl/sub 5/, FeCl/sub 3/, and CrCl/sub 4/. The UCl/sub 4/ is separated from this vapor mixture by selective fractional condensation at a temperature of about 500 to 400 deg C.

  10. Uranium industry annual, 1987

    SciTech Connect (OSTI)

    Not Available

    1988-09-29

    This report provides current statistical data on the US uranium industry for the Congress, federal and state agencies, the uranium and utility industries, and the public. It utilizes data from the mandatory ''Uranium Industry Annual Survey,'' Form EIA-858; historical data collected by the Energy Information Administration (EIA) and by the Grand Junction (Colorado) Project Office of the Idaho Operations Office of the US Department of Energy (DOE); and other data from federal agencies that preceded the DOE. The data provide a comprehensive statistical characterization of the industry's annual activities and include some information about industry plans and commitments over the next several years. Where these data are presented in aggregate form, care has been taken to protect the confidentiality of company-specific data while still conveying an accurate and complete statistical representation of the industry data.

  11. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    . Uranium purchased by owners and operators of U.S. civilian nuclear power reactors by supplier and delivery year, 2011-15 thousand pounds U3O8 equivalent, dollars per pound U3O8 equivalent Deliveries 2011 2012 2013 2014 2015 Purchased from U.S. producers Purchases of U.S.-origin and foreign-origin uranium 550 W W W 1,455 Weighted-average price 58.12 W W W 52.35 Purchased from U.S. brokers and traders Purchases of U.S.-origin and foreign-origin uranium 14,778 11,545 12,835 17,111 13,852

  12. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    . Uranium purchased by owners and operators of U.S. civilian nuclear power reactors by origin and delivery year, 2011-15 thousand pounds U3O8 equivalent; dollars per pound U3O8 equivalent Deliveries 2011 2012 2013 2014 2015 U.S.-Origin Uranium Purchases 5,205 9,807 9,484 3,316 3,419 Weighted-Average Price 52.12 59.44 56.37 48.11 43.86 Foreign-Origin Uranium Purchases 49,626 47,713 47,919 50,033 53,106 Weighted-Average Price 55.98 54.07 51.13 46.03 44.14 Total Purchases 54,831 57,520 57,403

  13. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    0. U.S. broker and trader purchases of uranium by origin, supplier, and delivery year, 2011-15 thousand pounds U3O8 equivalent; dollars per pound U3O8 equivalent Deliveries 2011 2012 2013 2014 2015 Received U.S.-origin uranium Purchases 1,668 1,194 W 410 2,702 Weighted-average price 54.85 51.78 W 33.55 35.04 Received foreign-origin uranium Purchases 24,695 24,606 W 28,743 33,014 Weighted-average price 49.69 47.75 W 38.42 39.58 Total received by U.S. brokers and traders Purchases 26,363 25,800

  14. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    1. Foreign sales of uranium from U.S. suppliers and owners and operators of U.S. civilian nuclear power reactors by origin and delivery year, 2011-15 thousands pounds U3O8 equivalent; dollars per pound U3O8 equivalent Deliveries to foreign suppliers and utilities 2011 2012 2013 2014 2015 U.S.-origin uranium Foreign sales 4,387 4,798 4,148 4,210 4,258 Weighted-average price 53.08 47.53 43.10 32.91 37.85 Foreign-origin uranium Foreign sales 12,297 13,185 14,717 15,794 21,465 Weighted-Average Price

  15. Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    3. U.S. uranium concentrate production, shipments, and sales, 2003-15 Activity at U.S. mills and In-Situ-Leach plants 2003 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 2015 Estimated contained U3O8 (thousand pounds) Ore from Mines and Stockpiles Fed to Mills1 0 W W W 0 W W W W W W W 0 Other Feed Materials 2 W W W W W W W W W W W W W Total Mill Feed W W W W W W W W W W W W W Uranium Concentrate Produced at U.S. Mills (thousand pounds U3O8) W W W W W W W W W W W W W Uranium Concentrate

  16. PROCESS FOR RECOVERING URANIUM

    DOE Patents [OSTI]

    MacWood, G.E.; Wilder, C.D.; Altman, D.

    1959-03-24

    A process is described for recovering uranium from deposits on stainless steel liner surfaces of calutrons. The deposit is removed from the stainless steel surface by washing with aqueous nitric acid. The solution obtained containing uranium, chromium, nickels copper, and iron is treated with excess of ammonium hydroxide to precipitatc the uranium, irons and chromium and convert thc nickel and copper to soluble ammonia complexions. The precipitated material is removed, dried, and treated with carbon tetrachloride at an elevated temperature of about 500 to 600 deg C to form a vapor mixture of UCl/sub 4/, UCl/sub 5/, FeCl/ sub 3/, and CrCl/sub 4/. The UCl/sub 4/ is separated from this vapor mixture by selective fractional condensation at a temprrature of about 300 to400 deg C.

  17. Uranium immobilization and nuclear waste

    SciTech Connect (OSTI)

    Duffy, C.J.; Ogard, A.E.

    1982-02-01

    Considerable information useful in nuclear waste storage can be gained by studying the conditions of uranium ore deposit formation. Further information can be gained by comparing the chemistry of uranium to nuclear fission products and other radionuclides of concern to nuclear waste disposal. Redox state appears to be the most important variable in controlling uranium solubility, especially at near neutral pH, which is characteristic of most ground water. This is probably also true of neptunium, plutonium, and technetium. Further, redox conditions that immobilize uranium should immobilize these elements. The mechanisms that have produced uranium ore bodies in the Earth's crust are somewhat less clear. At the temperatures of hydrothermal uranium deposits, equilibrium models are probably adequate, aqueous uranium (VI) being reduced and precipitated by interaction with ferrous-iron-bearing oxides and silicates. In lower temperature roll-type uranium deposits, overall equilibrium may not have been achieved. The involvement of sulfate-reducing bacteria in ore-body formation has been postulated, but is uncertain. Reduced sulfur species do, however, appear to be involved in much of the low temperature uranium precipitation. Assessment of the possibility of uranium transport in natural ground water is complicated because the system is generally not in overall equilibrium. For this reason, Eh measurements are of limited value. If a ground water is to be capable of reducing uranium, it must contain ions capable of reducing uranium both thermodynamically and kinetically. At present, the best candidates are reduced sulfur species.

  18. PROCESS OF PREPARING URANIUM CARBIDE

    DOE Patents [OSTI]

    Miller, W.E.; Stethers, H.L.; Johnson, T.R.

    1964-03-24

    A process of preparing uranium monocarbide is de scribed. Uranium metal is dissolved in cadmium, zinc, cadmium-- zinc, or magnesium-- zinc alloy and a small quantity of alkali metal is added. Addition of stoichiometric amounts of carbon at 500 to 820 deg C then precipitates uranium monocarbide. (AEC)

  19. Uranium Transport Modeling

    SciTech Connect (OSTI)

    Bostick, William D.

    2008-01-15

    Uranium contamination is prevalent at many of the U.S. DOE facilities and at several civilian sites that have supported the nuclear fuel cycle. The potential off-site mobility of uranium depends on the partitioning of uranium between aqueous and solid (soil and sediment) phases. Hexavalent U (as uranyl, UO{sub 2}{sup 2+}) is relatively mobile, forming strong complexes with ubiquitous carbonate ion which renders it appreciably soluble even under mild reducing conditions. In the presence of carbonate, partition of uranyl to ferri-hydrate and select other mineral phases is usually maximum in the near-neutral pH range {approx} 5-8. The surface complexation reaction of uranyl with iron-containing minerals has been used as one means to model subsurface migration, used in conjunction with information on the site water chemistry and hydrology. Partitioning of uranium is often studied by short-term batch 'equilibrium' or long-term soil column testing ; MCLinc has performed both of these methodologies, with selection of method depending upon the requirements of the client or regulatory authority. Speciation of uranium in soil may be determined directly by instrumental techniques (e.g., x-ray photoelectron spectroscopy, XPS; x-ray diffraction, XRD; etc.) or by inference drawn from operational estimates. Often, the technique of choice for evaluating low-level radionuclide partitioning in soils and sediments is the sequential extraction approach. This methodology applies operationally-defined chemical treatments to selectively dissolve specific classes of macro-scale soil or sediment components. These methods recognize that total soil metal inventory is of limited use in understanding bioavailability or metal mobility, and that it is useful to estimate the amount of metal present in different solid-phase forms. Despite some drawbacks, the sequential extraction method can provide a valuable tool to distinguish among trace element fractions of different solubility related to

  20. METHOD OF ELECTROPOLISHING URANIUM

    DOE Patents [OSTI]

    Walker, D.E.; Noland, R.A.

    1959-07-14

    A method of electropolishing the surface of uranium articles is presented. The process of this invention is carried out by immersing the uranium anticle into an electrolyte which contains from 35 to 65% by volume sulfuric acid, 1 to 20% by volume glycerine and 25 to 50% by volume of water. The article is made the anode in the cell and polished by electrolyzing at a voltage of from 10 to 15 volts. Discontinuing the electrolysis by intermittently withdrawing the anode from the electrolyte and removing any polarized film formed therein results in an especially bright surface.

  1. TREATMENT OF URANIUM SURFACES

    DOE Patents [OSTI]

    Slunder, C.J.

    1959-02-01

    An improved process is presented for prcparation of uranium surfaces prior to electroplating. The surfacc of the uranium to be electroplated is anodized in a bath comprising a solution of approximately 20 to 602 by weight of phosphoric acid which contains about 20 cc per liter of concentrated hydrochloric acid. Anodization is carried out for approximately 20 minutes at a current density of about 0.5 amperes per square inch at a temperature of about 35 to 45 C. The oxidic film produced by anodization is removed by dipping in strong nitric acid, followed by rinsing with water just prior to electroplating.

  2. PREPARATION OF URANIUM TRIOXIDE

    DOE Patents [OSTI]

    Buckingham, J.S.

    1959-09-01

    The production of uranium trioxide from aqueous solutions of uranyl nitrate is discussed. The uranium trioxide is produced by adding sulfur or a sulfur-containing compound, such as thiourea, sulfamic acid, sulfuric acid, and ammonium sulfate, to the uranyl solution in an amount of about 0.5% by weight of the uranyl nitrate hexahydrate, evaporating the solution to dryness, and calcining the dry residue. The trioxide obtained by this method furnished a dioxide with a considerably higher reactivity with hydrogen fluoride than a trioxide prepared without the sulfur additive.

  3. Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

    b. Weighted-average price of uranium purchased by owners and operators of U.S. civilian nuclear power reactors, 1994-2015 dollars per pound U3O8 equivalent Delivery year Total purchased (weighted-average price) Purchased from U.S. producers Purchased from U.S. brokers and traders Purchased from other owners and operators of U.S. civilian nuclear power reactors, other U.S. suppliers, (and U.S. government for 2007)1 Purchased from foreign suppliers U.S.-origin uranium (weighted-average price)

  4. Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    4. U.S. uranium mills by owner, location, capacity, and operating status at end of the year, 2011-15 Owner Mill and Heap Leach1 Facility name County, state (existing and planned locations) Capacity (short tons of ore per day) Operating status at end of the year 2011 2012 2013 2014 2015 Anfield Resources Shootaring Canyon Uranium Mill Garfield, Utah 750 Standby Standby Standby Standby Standby EPR White Mesa LLC White Mesa Mill San Juan, Utah 2,000 Operating Operating Operating- Processing

  5. Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    10. Uranium reserve estimates at the end of 2014 and 2015 million pounds U3O8 End of 2014 End of 2015 Forward Cost2 Uranium Reserve Estimates1 by Mine and Property Status, Mining Method, and State(s) $0 to $30 per pound $0 to $50 per pound $0 to $100 per pound $0 to $30 per pound $0 to $50 per pound $0 to $100 per pound Properties with Exploration Completed, Exploration Continuing, and Only Assessment Work W W 154.6 24.3 W 151.6 Properties Under Development for Production and Development

  6. PROCESS OF RECOVERING URANIUM

    DOE Patents [OSTI]

    Price, T.D.; Jeung, N.M.

    1958-06-17

    An improved precipitation method is described for the recovery of uranium from aqueous solutions. After removal of all but small amounts of Ni or Cu, and after complexing any iron present, the uranium is separated as the peroxide by adding H/sub 2/O/sub 2/. The improvement lies in the fact that the addition of H/sub 2/O/sub 2/ and consequent precipitation are carried out at a temperature below the freezing; point of the solution, so that minute crystals of solvent are present as seed crystals for the precipitation.

  7. Corrosion-resistant uranium

    DOE Patents [OSTI]

    Hovis, V.M. Jr.; Pullen, W.C.; Kollie, T.G.; Bell, R.T.

    1981-10-21

    The present invention is directed to the protecting of uranium and uranium alloy articles from corrosion by providing the surfaces of the articles with a layer of an ion-plated metal selected from aluminum and zinc to a thickness of at least 60 microinches and then converting at least the outer surface of the ion-plated layer of aluminum or zinc to aluminum chromate or zinc chromate. This conversion of the aluminum or zinc to the chromate form considerably enhances the corrosion resistance of the ion plating so as to effectively protect the coated article from corrosion.

  8. Corrosion-resistant uranium

    DOE Patents [OSTI]

    Hovis, Jr., Victor M.; Pullen, William C.; Kollie, Thomas G.; Bell, Richard T.

    1983-01-01

    The present invention is directed to the protecting of uranium and uranium alloy articles from corrosion by providing the surfaces of the articles with a layer of an ion-plated metal selected from aluminum and zinc to a thickness of at least 60 microinches and then converting at least the outer surface of the ion-plated layer of aluminum or zinc to aluminum chromate or zinc chromate. This conversion of the aluminum or zinc to the chromate form considerably enhances the corrosion resistance of the ion plating so as to effectively protect the coated article from corrosion.

  9. Isotope Enrichment Detection by Laser Ablation - Dual Tunable Diode Laser Absorption Spectrometry

    SciTech Connect (OSTI)

    Anheier, Norman C.; Bushaw, Bruce A.

    2009-07-01

    The rapid global expansion of nuclear energy is motivating the expedited development of new safeguards technology to mitigate potential proliferation threats arising from monitoring gaps within the uranium enrichment process. Current onsite enrichment level monitoring methods are limited by poor sensitivity and accuracy performance. Offsite analysis has better performance, but this approach requires onsite hand sampling followed by time-consuming and costly post analysis. These limitations make it extremely difficult to implement comprehensive safeguards accounting measures that can effectively counter enrichment facility misuse. In addition, uranium enrichment by modern centrifugation leads to a significant proliferation threat, since the centrifuge cascades can quickly produce a significant quantity of highly enriched uranium (HEU). The Pacific Northwest National Laboratory is developing an engineered safeguards approach having continuous aerosol particulate collection and uranium isotope analysis to provide timely detection of HEU production in a low enriched uranium facility. This approach is based on laser vaporization of aerosol particulate samples, followed by wavelength tuned laser diode spectroscopy, to characterize the 235U/238U isotopic ratio by subtle differences in atomic absorption wavelengths arising from differences in each isotopes nuclear mass, volume, and spin (hyperfine structure for 235U). Environmental sampling media is introduced into a small, reduced pressure chamber, where a focused pulsed laser vaporizes a 10 to 20-m sample diameter. The ejected plasma forms a plume of atomic vapor. A plume for a sample containing uranium has atoms of the 235U and 238U isotopes present. Tunable diode lasers are directed through the plume to selectively excite each isotope and their presence is detected by monitoring absorbance signals on a shot-to-shot basis. Single-shot detection sensitivity approaching the femtogram range and abundance uncertainty less

  10. Uranium Lease and Take-Back | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Uranium Lease and Take-Back

  11. High loading uranium fuel plate

    DOE Patents [OSTI]

    Wiencek, Thomas C.; Domagala, Robert F.; Thresh, Henry R.

    1990-01-01

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

  12. RECOVERY OF URANIUM FROM PITCHBLENDE

    DOE Patents [OSTI]

    Ruehle, A.E.

    1958-06-24

    The decontamination of uranium from molybdenum is described. When acid solutions containing uranyl nitrate are contacted with ether for the purpose of extracting the uranium values, complex molybdenum compounds are coextracted with the uranium and also again back-extracted from the ether with the uranium. This invention provides a process for extracting uranium in which coextraction of molybdenum is avoided. It has been found that polyhydric alcohols form complexes with molybdenum which are preferentially water-soluble are taken up by the ether extractant to only a very minor degree. The preferred embodiment of the process uses mannitol, sorbitol or a mixture of the two as the complexing agent.

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

    SciTech Connect (OSTI)

    Forsberg, C.W.

    1996-01-20

    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.

  14. Nuclear forensic analysis of uranium oxide powders interdicted in Victoria, Australia

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Kristo, Michael Joseph; Keegan, Elizabeth; Colella, Michael; Williams, Ross; Lindvall, Rachel; Eppich, Gary; Roberts, Sarah; Borg, Lars; Gaffney, Amy; Plaue, Jonathan; et al

    2015-04-13

    Nuclear forensic analysis was conducted on two uranium samples confiscated during a police investigation in Victoria, Australia. The first sample, designated NSR-F-270409-1, was a depleted uranium powder of moderate purity (~1000 μg/g total elemental impurities). The chemical form of the uranium was a compound similar to K2(UO2)3O4·4H2O. While aliquoting NSR-F-270409-1 for analysis, the body and head of a Tineid moth was discovered in the sample. The second sample, designated NSR-F-270409-2, was also a depleted uranium powder. It was of reasonably high purity (~380 μg/g total elemental impurities). The chemical form of the uranium was primarily UO3·2H2O, with minor phases ofmore » U3O8 and UO2. While aliquoting NSR-F-270409-2 for analysis, a metal staple of unknown origin was discovered in the sample. The presence of 236U and 232U in both samples indicates that the uranium feed stocks for these samples experienced a neutron flux at some point in their history. The reactor burn-up calculated from the isotopic composition of the uranium is consistent with that of spent fuel from natural uranium (NU) fueled Pu production. These nuclear forensic conclusions allow us to categorically exclude Australia as the origin of the material and greatly reduce the number of candidate sources.« less

  15. STRIPPING OF URANIUM FROM ORGANIC EXTRACTANTS

    DOE Patents [OSTI]

    Crouse, D.J. Jr.

    1962-09-01

    A liquid-liquid extraction method is given for recovering uranium values from uranium-containing solutions. Uranium is removed from a uranium-containing organic solution by contacting said organic solution with an aqueous ammonium carbonate solution substantially saturated in uranium values. A uranium- containing precipitate is thereby formed which is separated from the organic and aqueous phases. Uranium values are recovered from this separated precipitate. (AE C)

  16. Evaluation of Uranium Measurements in Water by Various Methods - 13571

    SciTech Connect (OSTI)

    Tucker, Brian J.; Workman, Stephen M.

    2013-07-01

    In December 2000, EPA amended its drinking water regulations for radionuclides by adding a Maximum Contaminant Level (MCL) for uranium (so called MCL Rule)[1] of 30 micrograms per liter (?g/L). The MCL Rule also included MCL goals of zero for uranium and other radionuclides. Many radioactively contaminated sites must test uranium in wastewater and groundwater to comply with the MCL rule as well as local publicly owned treatment works discharge limitations. This paper addresses the relative sensitivity, accuracy, precision, cost and comparability of two EPA-approved methods for detection of total uranium: inductively plasma/mass spectrometry (ICP-MS) and alpha spectrometry. Both methods are capable of measuring the individual uranium isotopes U-234, U- 235, and U-238 and both methods have been deemed acceptable by EPA. However, the U-238 is by far the primary contributor to the mass-based ICP-MS measurement, especially for naturally-occurring uranium, which contains 99.2745% U-238. An evaluation shall be performed relative to the regulatory requirement promulgated by EPA in December 2000. Data will be garnered from various client sample results measured by ALS Laboratory in Fort Collins, CO. Data shall include method detection limits (MDL), minimum detectable activities (MDA), means and trends in laboratory control sample results, performance evaluation data for all methods, and replicate results. In addition, a comparison will be made of sample analyses results obtained from both alpha spectrometry and the screening method Kinetic Phosphorescence Analysis (KPA) performed at the U.S. Army Corps of Engineers (USACE) FUSRAP Maywood Laboratory (UFML). Many uranium measurements occur in laboratories that only perform radiological analysis. This work is important because it shows that uranium can be measured in radiological as well as stable chemistry laboratories and it provides several criteria as a basis for comparison of two uranium test methods. This data will

  17. URANIUM RECOVERY PROCESS

    DOE Patents [OSTI]

    Hyman, H.H.; Dreher, J.L.

    1959-07-01

    The recovery of uranium from the acidic aqueous metal waste solutions resulting from the bismuth phosphate carrier precipitation of plutonium from solutions of neutron irradiated uranium is described. The waste solutions consist of phosphoric acid, sulfuric acid, and uranium as a uranyl salt, together with salts of the fission products normally associated with neutron irradiated uranium. Generally, the process of the invention involves the partial neutralization of the waste solution with sodium hydroxide, followed by conversion of the solution to a pH 11 by mixing therewith sufficient sodium carbonate. The resultant carbonate-complexed waste is contacted with a titanated silica gel and the adsorbent separated from the aqueous medium. The aqueous solution is then mixed with sufficient acetic acid to bring the pH of the aqueous medium to between 4 and 5, whereby sodium uranyl acetate is precipitated. The precipitate is dissolved in nitric acid and the resulting solution preferably provided with salting out agents. Uranyl nitrate is recovered from the solution by extraction with an ether such as diethyl ether.

  18. Uranium Reduction by Clostridia

    SciTech Connect (OSTI)

    Francis, A.J.; Dodge, Cleveland J.; Gillow, Jeffrey B.

    2006-04-05

    The FRC groundwater and sediment contain significant concentrations of U and Tc and are dominated by low pH, and high nitrate and Al concentrations where dissimilatory metal reducing bacterial activity may be limited. The presence of Clostridia in Area 3 at the FRC site has been confirmed and their ability to reduce uranium under site conditions will be determined. Although the phenomenon of uranium reduction by Clostridia has been firmly established, the molecular mechanisms underlying such a reaction are not very clear. The authors are exploring the hypothesis that U(VI) reduction occurs through hydrogenases and other enzymes (Matin and Francis). Fundamental knowledge of metal reduction using Clostridia will allow us to exploit naturally occurring processes to attenuate radionuclide and metal contaminants in situ in the subsurface. The outline for this report are as follows: (1) Growth of Clostridium sp. under normal culture conditions; (2) Fate of metals and radionuclides in the presence of Clostridia; (3) Bioreduction of uranium associated with nitrate, citrate, and lepidocrocite; and (4) Utilization of Clostridium sp. for immobilization of uranium at the FRC Area 3 site.

  19. Small cell experiments for electrolytic reduction of uranium oxides to uranium metal using fluoride salts

    SciTech Connect (OSTI)

    Haas, P.A.; Adcock, P.W.; Coroneos, A.C.; Hendrix, D.E. )

    1994-08-01

    Electrolytic reduction of uranium oxide was proposed for the preparation of uranium metal feed for the atomic vapor laser isotope separation (AVLIS) process. A laboratory cell of 25-cm ID was operated to obtain additional information in areas important to design and operation of a pilot plant cell. Reproducible test results and useful operating and control procedures were demonstrated. About 20 kg of uranium metal of acceptable purity were prepared. A good supply of dissolved UO[sub 2] feed at the anode is the most important controlling requirement for efficient cell operation. A large fraction of the cell current is nonproductive in that it does not produce a metal product nor consume carbon anodes. All useful test conditions gave some reduction of UF[sub 4] to produce CF[sub 4] in addition to the reduction of UO[sub 2], but the fraction of metal from the reduction of UF[sub 4] can be decreased by increasing the concentration of dissolved UO[sub 2]. Operation of large continuous cells would probably be limited to current efficiencies of less than 60 pct, and more than 20 pct of the metal would result from the reduction of UF[sub 4].

  20. Laser Isotope Separation Employing Condensation Repression

    SciTech Connect (OSTI)

    Eerkens, Jeff W.; Miller, William H.

    2004-09-15

    Molecular laser isotope separation (MLIS) techniques using condensation repression (CR) harvesting are reviewed and compared with atomic vapor laser isotope separation (AVLIS), gaseous diffusion (DIF), ultracentrifuges (UCF), and electromagnetic separations (EMS). Two different CR-MLIS or CRISLA (Condensation Repression Isotope Separation by Laser Activation) approaches have been under investigation at the University of Missouri (MU), one involving supersonic super-cooled free jets and dimer formation, and the other subsonic cold-wall condensation. Both employ mixtures of an isotopomer (e.g. {sup i}QF{sub 6}) and a carrier gas, operated at low temperatures and pressures. Present theories of VT relaxation, dimerization, and condensation are found to be unsatisfactory to explain/predict experimental CRISLA results. They were replaced by fundamentally new models that allow ab-initio calculation of isotope enrichments and predictions of condensation parameters for laser-excited and non-excited vapors which are in good agreement with experiment. Because of supersonic speeds, throughputs for free-jet CRISLA are a thousand times higher than cold-wall CRISLA schemes, and thus preferred for large-quantity Uranium enrichments. For small-quantity separations of (radioactive) medical isotopes, the simpler coldwall CRISLA method may be adequate.

  1. Preserving Ultra-Pure Uranium-233

    SciTech Connect (OSTI)

    Krichinsky, Alan M; Goldberg, Dr. Steven A.; Hutcheon, Dr. Ian D.

    2011-10-01

    Uranium-233 ({sup 233}U) is a synthetic isotope of uranium formed under reactor conditions during neutron capture by natural thorium ({sup 232}Th). At high purities, this synthetic isotope serves as a crucial reference material for accurately quantifying and characterizing uranium-bearing materials assays and isotopic distributions for domestic and international nuclear safeguards. Separated, high purity {sup 233}U is stored in vaults at Oak Ridge National Laboratory (ORNL). These materials represent a broad spectrum of {sup 233}U from the standpoint of isotopic purity - the purest being crucial for precise analyses in safeguarding uranium. All {sup 233}U at ORNL is currently scheduled to be disposed of by down-blending with depleted uranium beginning in 2015. This will reduce safety concerns and security costs associated with storage. Down-blending this material will permanently destroy its potential value as a certified reference material for use in uranium analyses. Furthermore, no credible options exist for replacing {sup 233}U due to the lack of operating production capability and the high cost of restarting currently shut down capabilities. A study was commissioned to determine the need for preserving high-purity {sup 233}U. This study looked at the current supply and the historical and continuing domestic need for this crucial isotope. It examined the gap in supplies and uses to meet domestic needs and extrapolated them in the context of international safeguards and security activities - superimposed on the recognition that existing supplies are being depleted while candidate replacement material is being prepared for disposal. This study found that the total worldwide need by this projection is at least 850 g of certified {sup 233}U reference material over the next 50 years. This amount also includes a strategic reserve. To meet this need, 18 individual items totaling 959 g of {sup 233}U were identified as candidates for establishing a lasting supply of

  2. Corrosion Evaluation of RERTR Uranium Molybdenum Fuel

    SciTech Connect (OSTI)

    A K Wertsching

    2012-09-01

    As part of the National Nuclear Security Agency (NNSA) mandate to replace the use of highly enriched uranium (HEU) fuel for low enriched uranium (LEU) fuel, research into the development of LEU fuel for research reactors has been active since the late 1970’s. Originally referred to as the Reduced Enrichment for Research and Test Reactor (RERTR) program the new effort named Global Threat Reduction Initiative (GTRI) is nearing the goal of replacing the standard aluminum clad dispersion highly enriched uranium aluminide fuel with a new LEU fuel. The five domestic high performance research reactors undergoing this conversion are High Flux Isotope reactor (HFIR), Advanced Test Reactor (ATR), National Institute of Standards and Technology (NIST) Reactor, Missouri University Research Reactor (MURR) and the Massachusetts Institute of Technology Reactor II (MITR-II). The design of these reactors requires a higher neutron flux than other international research reactors, which to this point has posed unique challenges in the design and development of the new mandated LEU fuel. The new design utilizes a monolithic fuel configuration in order to obtain sufficient 235U within the LEU stoichoimetry to maintain the fission reaction within the domestic test reactors. The change from uranium aluminide dispersion fuel type to uranium molybdenum (UMo) monolithic configuration requires examination of possible corrosion issues associated with the new fuel meat. A focused analysis of the UMo fuel under potential corrosion conditions, within the ATR and under aqueous storage indicates a slow and predictable corrosion rate. Additional corrosion testing is recommended for the highest burn-up fuels to confirm observed corrosion rate trends. This corrosion analysis will focus only on the UMo fuel and will address corrosion of ancillary components such as cladding only in terms of how it affects the fuel. The calculations and corrosion scenarios are weighted with a conservative bias to

  3. Isotope separation and advanced manufacturing technology. Volume 2, No. 2, Semiannual report, April--September 1993

    SciTech Connect (OSTI)

    Kan, Tehmanu; Carpenter, J.

    1993-12-31

    This is the second issue of a semiannual report for the Isotope Separation and Advanced Manufacturing (ISAM) Technology Program at Lawrence Livermore National Laboratory. Primary objectives of the ISAM Program include: the Uranium Atomic Vapor Laser Isotope Separation (U-AVLIS) process, and advanced manufacturing technologies which include industrial laser materials processing and new manufacturing technologies for uranium, plutonium, and other strategically important materials in support of DOE and other national applications. Topics included in this issue are: production plant product system conceptual design, development and operation of a solid-state switch for thyratron replacement, high-performance optical components for high average power laser systems, use of diode laser absorption spectroscopy for control of uranium vaporization rates, a two-dimensional time dependent hydrodynamical ion extraction model, and design of a formaldehyde photodissociation process for carbon and oxygen isotope separation.

  4. Method for separating isotopes

    DOE Patents [OSTI]

    Jepson, B.E.

    1975-10-21

    Isotopes are separated by contacting a feed solution containing the isotopes with a cyclic polyether wherein a complex of one isotope is formed with the cyclic polyether, the cyclic polyether complex is extracted from the feed solution, and the isotope is thereafter separated from the cyclic polyether.

  5. Stable isotope studies

    SciTech Connect (OSTI)

    Ishida, T.

    1992-01-01

    The research has been in four general areas: (1) correlation of isotope effects with molecular forces and molecular structures, (2) correlation of zero-point energy and its isotope effects with molecular structure and molecular forces, (3) vapor pressure isotope effects, and (4) fractionation of stable isotopes. 73 refs, 38 figs, 29 tabs.

  6. PRODUCTION OF URANIUM METAL BY CARBON REDUCTION

    DOE Patents [OSTI]

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

    1959-09-22

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

  7. Method of preparation of uranium nitride

    DOE Patents [OSTI]

    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.

  8. Isotope separation by photochromatography

    DOE Patents [OSTI]

    Suslick, K.S.

    1975-10-03

    A photochromatographic method for isotope separation is described. An isotopically mixed molecular species is adsorbed on an adsorptive surface, and the adsorbed molecules are irradiated with radiation of a predetermined wavelength which will selectively excite desired isotopic species. Sufficient energy is transferred to the excited molecules to desorb them from the surface and thus separate them from the undesired isotopic species. The method is particularly applicable to the separation of hydrogen isotopes. (BLM)

  9. Isotope separation by photochromatography

    DOE Patents [OSTI]

    Suslick, Kenneth S.

    1977-01-01

    An isotope separation method which comprises physically adsorbing an isotopically mixed molecular species on an adsorptive surface and irradiating the adsorbed molecules with radiation of a predetermined wavelength which will selectively excite a desired isotopic species. Sufficient energy is transferred to the excited molecules to desorb them from the surface and thereby separate them from the unexcited undesired isotopic species. The method is particularly applicable to the separation of hydrogen isotopes.

  10. file://\\fs-f1\shared\uranium\uranium.html

    U.S. Energy Information Administration (EIA) Indexed Site

    Glossary Home > Nuclear > U.S. Uranium Reserves Estimates U.S. Uranium Reserves Estimates Data for: 2008 Report Released: July 2010 Next Release Date: 2012 Summary The U.S. Energy Information Administration (EIA) has updated its estimates of uranium reserves for year-end 2008. This represents the first revision of the estimates since 2004. The update is based on analysis of company annual reports, any additional information reported by companies at conferences and in news releases,

  11. Method of preparing uranium nitride or uranium carbonitride bodies

    DOE Patents [OSTI]

    Wilhelm, Harley A.; McClusky, James K.

    1976-04-27

    Sintered uranium nitride or uranium carbonitride bodies having a controlled final carbon-to-uranium ratio are prepared, in an essentially continuous process, from U.sub.3 O.sub.8 and carbon by varying the weight ratio of carbon to U.sub.3 O.sub.8 in the feed mixture, which is compressed into a green body and sintered in a continuous heating process under various controlled atmospheric conditions to prepare the sintered bodies.

  12. RECOVERY OF URANIUM FROM ZIRCONIUM-URANIUM NUCLEAR FUELS

    DOE Patents [OSTI]

    Gens, T.A.

    1962-07-10

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

  13. PREPARATION OF DENSE URANIUM DIOXIDE PARTICLES FROM URANIUM HEXAFLUORI...

    Office of Scientific and Technical Information (OSTI)

    Visit OSTI to utilize additional information resources in energy science and technology. A ... A fluid-bed method was developed for the direct preparation from uranium hexafluoride of ...

  14. Method for fabricating uranium foils and uranium alloy foils

    DOE Patents [OSTI]

    Hofman, Gerard L.; Meyer, Mitchell K.; Knighton, Gaven C.; Clark, Curtis R.

    2006-09-05

    A method of producing thin foils of uranium or an alloy. The uranium or alloy is cast as a plate or sheet having a thickness less than about 5 mm and thereafter cold rolled in one or more passes at substantially ambient temperatures until the uranium or alloy thereof is in the shape of a foil having a thickness less than about 1.0 mm. The uranium alloy includes one or more of Zr, Nb, Mo, Cr, Fe, Si, Ni, Cu or Al.

  15. METHOD OF PRODUCING URANIUM

    DOE Patents [OSTI]

    Foster, L.S.; Magel, T.T.

    1958-05-13

    A modified process is described for the production of uranium metal by means of a bomb reduction of UF/sub 4/. Difficulty is sometimes experienced in obtaining complete separation of the uranium from the slag when the process is carried out on a snnall scale, i.e., for the production of 10 grams of U or less. Complete separation may be obtained by incorporating in the reaction mixture a quantity of MnCl/sub 2/, so that this compound is reduced along with the UF/sub 4/ . As a result a U--Mn alloy is formed which has a melting point lower than that of pure U, and consequently the metal remains molten for a longer period allowing more complete separation from the slag.

  16. ELECTROLYSIS OF THORIUM AND URANIUM

    DOE Patents [OSTI]

    Hansen, W.N.

    1960-09-01

    An electrolytic method is given for obtaining pure thorium, uranium, and thorium-uranium alloys. The electrolytic cell comprises a cathode composed of a metal selected from the class consisting of zinc, cadmium, tin, lead, antimony, and bismuth, an anode composed of at least one of the metals selected from the group consisting of thorium and uranium in an impure state, and an electrolyte composed of a fused salt containing at least one of the salts of the metals selected from the class consisting of thorium, uranium. zinc, cadmium, tin, lead, antimony, and bismuth. Electrolysis of the fused salt while the cathode is maintained in the molten condition deposits thorium, uranium, or thorium-uranium alloys in pure form in the molten cathode which thereafter may be separated from the molten cathode product by distillation.

  17. PROCESS FOR PRODUCING URANIUM TETRAFLUORIDE

    DOE Patents [OSTI]

    Harvey, B.G.

    1954-09-14

    >This patent relates to improvements in the method for producing uranium tetrafluoride by treating an aqueous solutlon of a uranyl salt at an elevated temperature with a reducing agent effective in acld solutlon in the presence of hydrofluoric acid. Uranium tetrafluoride produced this way frequentiy contains impurities in the raw material serving as the source of uranium. Uranium tetrafluoride much less contaminated with impurities than when prepared by the above method can be prepared from materials containing such impurities by first adding a small proportion of reducing agent so as to cause a small fraction, for example 1 to 5% of the uranium tetrafluoride to be precipitated, rejecting such precipitate, and then precipitating and recovering the remainder of the uranium tetrafluoride.

  18. WELDED JACKETED URANIUM BODY

    DOE Patents [OSTI]

    Gurinsky, D.H.

    1958-08-26

    A fuel element is presented for a neutronic reactor and is comprised of a uranium body, a non-fissionable jacket surrounding sald body, thu jacket including a portion sealed by a weld, and an inclusion in said sealed jacket at said weld of a fiux having a low neutron capture cross-section. The flux is provided by combining chlorine gas and hydrogen in the intense heat of-the arc, in a "Heliarc" welding muthod, to form dry hydrochloric acid gas.

  19. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    b. Uranium purchased by owners and operators of U.S. civilian nuclear power reactors ranked by price and distributed by purchaser, 2013-15 deliveries thousand pounds U3O8 equivalent; dollars per pound U3O8 equivalent Deliveries in 2013 Deliveries in 2014 Deliveries in 2015 Distribution of purchasers Number of purchasers Quantity with reported price Weighted-average price Number of purchasers Quantity with reported price Weighted-average price Number of purchasers Quantity with reported price

  20. METHOD OF DISSOLVING URANIUM METAL

    DOE Patents [OSTI]

    Slotin, L.A.

    1958-02-18

    This patent relates to an economicai means of dissolving metallic uranium. It has been found that the addition of a small amount of perchloric acid to the concentrated nitric acid in which the uranium is being dissolved greatly shortens the time necessary for dissolution of the metal. Thus the use of about 1 or 2 percent of perchioric acid based on the weight of the nitric acid used, reduces the time of dissolution of uranium by a factor of about 100.

  1. PROCESS FOR PREPARING URANIUM METAL

    DOE Patents [OSTI]

    Prescott, C.H. Jr.; Reynolds, F.L.

    1959-01-13

    A process is presented for producing oxygen-free uranium metal comprising contacting iodine vapor with crude uranium in a reaction zone maintained at 400 to 800 C to produce a vaporous mixture of UI/sub 4/ and iodine. Also disposed within the maction zone is a tungsten filament which is heated to about 1600 C. The UI/sub 4/, upon contacting the hot filament, is decomposed to molten uranium substantially free of oxygen.

  2. Method of and apparatus for measuring the mean concentration of thoron and/or radon in a gas mixture

    DOE Patents [OSTI]

    Lucas, Henry

    1990-01-01

    A method of and an apparatus for detecting and accurately measuring the mean concentrations of .sup.222 Rn and .sup.220 Tn in a gas mixture, such as the ambient atmosphere in a mine, is provided. The apparatus includes an alpha target member which defines at least one operative target surface and which is preferably fabricated from a single piece of an alpha particle sensitive material. At least one portion of the operative target surface is covered with an alpha particle filter. The uncovered and filter covered operative surface is exposed to the gas mixture containing the .sup.222 Rn and .sup.220 Tn. In the radioactive decay series of these isotopes the maximum kinetic energy emitted by the alpha decay of .sup.222 Rn is about 1.1 MeV less than the maximum kinetic energy emitted by the alpha decay of a .sup.220 Tn. The alpha particle filter has a predetermined mass per unit area of the covered portion of the operative target surface that prevents penetration of alpha particles which originate from .sup.222 Rn decay, but which allows passage therethrough of the maximum kinetic energy alpha particles from .sup.220 Tn decay. Thus, a count of the alpha particle tracks in the uncovered portion of the target member is proportional to the mean concentration of sum of .sup.222 Rn and .sup.220 Tn in the gas mixture, while the count of alpha tracks in the target member under the filter is proportional to the concentration of only the .sup.220 Tn in the gas mixture.

  3. VANE Uranium One JV | Open Energy Information

    Open Energy Info (EERE)

    VANE Uranium One JV Jump to: navigation, search Name: VANE-Uranium One JV Place: London, England, United Kingdom Zip: EC4V 6DX Product: JV between VANE Minerals Plc & Uranium One....

  4. SEPARATION OF THORIUM FROM URANIUM

    DOE Patents [OSTI]

    Bane, R.W.

    1959-09-01

    A description is given for the separation of thorium from uranium by forming an aqueous acidic solution containing ionic species of thorium, uranyl uranium, and hydroxylamine, flowing the solution through a column containing the phenol-formaldehyde type cation exchange resin to selectively adsorb substantially all the thorium values and a portion of the uranium values, flowing a dilute solution of hydrochloric acid through the column to desorb the uranium values, and then flowing a dilute aqueous acidic solution containing an ion, such as bisulfate, which has a complexing effect upon thortum through the column to desorb substantially all of the thorium.

  5. Highly Enriched Uranium Materials Facility

    National Nuclear Security Administration (NNSA)

    Appropriations Subcommittee, is shown some of the technology in the Highly Enriched Uranium Materials Facility by Warehousing and Transportation Operations Manager Byron...

  6. 2014 Uranium Marketing Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    By law, EIA's data, analyses, and forecasts are independent ... on information reported on Form EIA-858, "Uranium Marketing ... nuclear power reactors by contract type and material type, ...

  7. THERMAL DECOMPOSITION OF URANIUM COMPOUNDS

    DOE Patents [OSTI]

    Magel, T.T.; Brewer, L.

    1959-02-10

    A method is presented of preparing uranium metal of high purity consisting contacting impure U metal with halogen vapor at between 450 and 550 C to form uranium halide vapor, contacting the uranium halide vapor in the presence of H/sub 2/ with a refractory surface at about 1400 C to thermally decompose the uranium halides and deposit molten U on the refractory surface and collecting the molten U dripping from the surface. The entire operation is carried on at a sub-atmospheric pressure of below 1 mm mercury.

  8. 2015 Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

    3 2015 Uranium Marketing Annual Report Release Date: May 24, 2016 Next Release Date: May 2017 Quantity with reported price Weighted-average price Quantity with reported price ...

  9. 2015 Uranium Market Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    received in 2015","Weighted-average price","Number of purchase contracts for ... Administration, Form EIA-858 ""Uranium Marketing Annual Survey"" (2015)." "16 ...

  10. 2015 Uranium Marketing Annual Survey

    U.S. Energy Information Administration (EIA) Indexed Site

    5 2015 Uranium Marketing Annual Report Release Date: May 24, 2016 Next Release Date: May 2017 Quantity with reported price Weighted-average price Quantity with reported price ...

  11. ELECTROLYTIC PRODUCTION OF URANIUM TETRAFLUORIDE

    DOE Patents [OSTI]

    Lofthouse, E.

    1954-08-31

    This patent relates to electrolytic methods for the production of uranium tetrafluoride. According to the present invention a process for the production of uranium tetrafluoride comprises submitting to electrolysis an aqueous solution of uranyl fluoride containing free hydrofluoric acid. Advantageously the aqueous solution of uranyl fluoride is obtained by dissolving uranium hexafluoride in water. On electrolysis, the uranyl ions are reduced to uranous tons at the cathode and immediately combine with the fluoride ions in solution to form the insoluble uranium tetrafluoride which is precipitated.

  12. 2014 Domestic Uranium Production Report

    U.S. Energy Information Administration (EIA) Indexed Site

    Power Resources Inc., dba Cameco Resources Smith Ranch-Highland Operation Converse, ... Uranium is first processed at the Nichols Ranch plant and then transported to the Smith ...

  13. Domestic Uranium Production Report - Quarterly

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    Resources, Inc. dba Cameco Resources Smith Ranch-Highland Operation Converse, Wyoming ... Uranium is first processed at the Nichols Ranch plant and then transported to the Smith ...

  14. 2014 Domestic Uranium Production Report

    U.S. Energy Information Administration (EIA) Indexed Site

    Resources Inc., dba Cameco Resources","Smith Ranch-Highland Operation","Converse, ... Uranium is first processed at the Nichols Ranch plant and then transported to the Smith ...

  15. Profile of World Uranium Enrichment Programs - 2007

    SciTech Connect (OSTI)

    Laughter, Mark D

    2007-11-01

    It is generally agreed that the most difficult step in building a nuclear weapon is acquiring weapons grade fissile material, either plutonium or highly enriched uranium (HEU). Plutonium is produced in a nuclear reactor, while HEU is produced using a uranium enrichment process. Enrichment is also an important step in the civil nuclear fuel cycle, in producing low enriched uranium (LEU) for use in fuel for nuclear reactors. However, the same equipment used to produce LEU for nuclear fuel can also be used to produce HEU for weapons. Safeguards at an enrichment plant are the array of assurances and verification techniques that ensure uranium is only enriched to LEU, no undeclared LEU is produced, and no uranium is enriched to HEU or secretly diverted. There are several techniques for enriching uranium. The two most prevalent are gaseous diffusion, which uses older technology and requires a lot of energy, and gas centrifuge separation, which uses more advanced technology and is more energy efficient. Gaseous diffusion plants (GDPs) provide about 40% of current world enrichment capacity, but are being phased out as newer gas centrifuge enrichment plants (GCEPs) are constructed. Estimates of current and future enrichment capacity are always approximate, due to the constant upgrades, expansions, and shutdowns occurring at enrichment plants, largely determined by economic interests. Currently, the world enrichment capacity is approximately 53 million kg-separative work units (SWU) per year, with 22 million in gaseous diffusion and 31 million in gas centrifuge plants. Another 23 million SWU/year of capacity are under construction or planned for the near future, almost entirely using gas centrifuge separation. Other less-efficient techniques have also been used in the past, including electromagnetic and aerodynamic separations, but these are considered obsolete, at least from a commercial perspective. Laser isotope separation shows promise as a possible enrichment technique

  16. METHOD FOR RECOVERING URANIUM FROM OILS

    DOE Patents [OSTI]

    Gooch, L.H.

    1959-07-14

    A method is presented for recovering uranium from hydrocarbon oils, wherein the uranium is principally present as UF/sub 4/. According to the invention, substantially complete removal of the uranium from the hydrocarbon oil may be effected by intimately mixing one part of acetone to about 2 to 12 parts of the hydrocarbon oil containing uranium and separating the resulting cake of uranium from the resulting mixture. The uranium in the cake may be readily recovered by burning to the oxide.

  17. Uranium in granites from the Southwestern United States: actinide parent-daughter systems, sites and mobilization. First year report

    SciTech Connect (OSTI)

    Silver, L T; Williams, I S; Woodhead, J A

    1980-10-01

    Some of the principal findings of the study on the Lawler Peak Granite are: the granite is dated precisely by this work at 1411 +- 3 m.y., confirming its synchroneity with a great regional terrane of granites. Uranium is presently 8-10 times crustal abundance and thorium 2-3 times in this granite. Uranium is found to be enriched in at least eight, possibly ten, primary igneous mineral species over the whole-rock values. Individual mineral species show distinct levels in, and characteristics ranges of, uranium concentration. It appears that in a uraniferous granite such as this, conventional accuracy mineral suites probably cannot account for most of the uranium in the rock, and more rare, high U-concentration phases also are present and are significant uranium hosts. It appears that at least two different geological episodes have contributed to the disturbance of the U-Th-Pb isotope systems. Studies of various sites for transient dispersal of uranium, thorium, and radiogenic lead isotopes indicate a non-uniform dispersal of these components. It appears that the bulk rock has lost at least 24 percent of its original uranium endowment, accepting limited or no radiogenic lead or thorium migration from the sample.

  18. Isotopic Generation and Confirmation of the PWR Application Model 

    SciTech Connect (OSTI)

    L.B. Wimmer

    2003-11-10

    The objective of this calculation is to establish an isotopic database to represent commercial spent nuclear fuel (CSNF) from pressurized water reactors (PWRs) in criticality analyses performed for the proposed Monitored Geologic Repository at Yucca Mountain, Nevada. Confirmation of the conservatism with respect to criticality in the isotopic concentration values represented by this isotopic database is performed as described in Section 3.5.3.1.2 of the ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2000). The isotopic database consists of the set of 14 actinides and 15 fission products presented in Section 3.5.2.1.1 of YMP 2000 for use in CSNF burnup credit. This set of 29 isotopes is referred to as the principal isotopes. The oxygen isotope from the UO{sub 2} fuel is also included in the database. The isotopic database covers enrichments of {sup 235}U ranging from 1.5 to 5.5 weight percent (wt%) and burnups ranging from approximately zero to 75 GWd per metric ton of uranium (mtU). The choice of fuel assembly and operating history values used in generating the isotopic database are provided is Section 5. Tables of isotopic concentrations for the 29 principal isotopes (plus oxygen) as a function of enrichment and burnup are provided in Section 6.1. Results of the confirmation of the conservatism with respect to criticality in the isotopic concentration values are provided in Section 6.2.

  19. Uranium Enrichment Decontamination and Decommissioning Fund's...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Uranium Enrichment Decontamination and Decommissioning Fund's Fiscal Year 2008 and 2007 Financial Statement Audit, OAS-FS-10-05 Uranium Enrichment Decontamination and...

  20. Calculating Atomic Number Densities for Uranium

    Energy Science and Technology Software Center (OSTI)

    1993-01-01

    Provides method to calculate atomic number densities of selected uranium compounds and hydrogenous moderators for use in nuclear criticality safety analyses at gaseous diffusion uranium enrichment facilities.

  1. Nuclear radiation cleanup and uranium prospecting (Patent) |...

    Office of Scientific and Technical Information (OSTI)

    Nuclear radiation cleanup and uranium prospecting Citation Details In-Document Search Title: Nuclear radiation cleanup and uranium prospecting Apparatus, systems, and methods for...

  2. Nuclear radiation cleanup and uranium prospecting (Patent) |...

    Office of Scientific and Technical Information (OSTI)

    Nuclear radiation cleanup and uranium prospecting Citation Details In-Document Search Title: Nuclear radiation cleanup and uranium prospecting You are accessing a document from...

  3. Uranium Resources Inc URI | Open Energy Information

    Open Energy Info (EERE)

    exploring, developing and mining uranium properties using the in situ recovery (ISR) or solution mining process. References: Uranium Resources, Inc. (URI)1 This article...

  4. Uranium Biomineralization By Natural Microbial Phosphatase Activities...

    Office of Scientific and Technical Information (OSTI)

    Technical Report: Uranium Biomineralization By Natural Microbial Phosphatase Activities in the Subsurface Citation Details In-Document Search Title: Uranium Biomineralization By ...

  5. PROCESS OF PRODUCING REFRACTORY URANIUM OXIDE ARTICLES

    DOE Patents [OSTI]

    Hamilton, N.E.

    1957-12-01

    A method is presented for fabricating uranium oxide into a shaped refractory article by introducing a uranium halide fluxing reagent into the uranium oxide, and then mixing and compressing the materials into a shaped composite mass. The shaped mass of uranium oxide and uranium halide is then fired at an elevated temperature so as to form a refractory sintered article. It was found in the present invention that the introduction of a uraninm halide fluxing agent afforded a fluxing action with the uranium oxide particles and that excellent cohesion between these oxide particles was obtained. Approximately 90% of uranium dioxide and 10% of uranium tetrafluoride represent a preferred composition.

  6. Structural Sequestration of Uranium in Bacteriogenic Manganese...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Sequestration of Uranium in Bacteriogenic Manganese Oxides Samuel M. Webb (Stanford ... Uranium is a key contaminant of concern at US DOE sites and shuttered mining and ore ...

  7. Uranium Processing Facility team signs partnering agreement ...

    National Nuclear Security Administration (NNSA)

    Uranium Processing Facility team signs partnering agreement Thursday, July 24, 2014 - 9:40am Officials from NNSA's Uranium Processing Facility Project Office and Consolidated ...

  8. Conversion of depleted uranium hexafluoride to a solid uranium compound

    DOE Patents [OSTI]

    Rothman, Alan B.; Graczyk, Donald G.; Essling, Alice M.; Horwitz, E. Philip

    2001-01-01

    A process for converting UF.sub.6 to a solid uranium compound such as UO.sub.2 and CaF. The UF.sub.6 vapor form is contacted with an aqueous solution of NH.sub.4 OH at a pH greater than 7 to precipitate at least some solid uranium values as a solid leaving an aqueous solution containing NH.sub.4 OH and NH.sub.4 F and remaining uranium values. The solid uranium values are separated from the aqueous solution of NH.sub.4 OH and NH.sub.4 F and remaining uranium values which is then diluted with additional water precipitating more uranium values as a solid leaving trace quantities of uranium in a dilute aqueous solution. The dilute aqueous solution is contacted with an ion-exchange resin to remove substantially all the uranium values from the dilute aqueous solution. The dilute solution being contacted with Ca(OH).sub.2 to precipitate CaF.sub.2 leaving dilute NH.sub.4 OH.

  9. Advanced uranium enrichment technologies

    SciTech Connect (OSTI)

    Merriman, R.

    1983-03-10

    The Advanced Gas Centrifuge and Atomic Vapor Laser Isotope Separation methods are described. The status and potential of the technologies are summarized, the programs outlined, and the economic incentives are noted. How the advanced technologies, once demonstrated, might be deployed so that SWV costs in the 1990s can be significantly reduced is described.

  10. Manus Water Isotope Investigation

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    9 Manus Water Isotope Investigation Field Campaign Report JL Conroy D Noone KM Cobb March ... DOESC-ARM-15-079 Manus Water Isotope Investigation Field Campaign Report JL Conroy, ...

  11. Manus Water Isotope Investigation

    Office of Scientific and Technical Information (OSTI)

    ENERGY Office of Science DOESC-ARM-15-079 Manus Water Isotope Investigation Field ... DOESC-ARM-15-079 Manus Water Isotope Investigation Field Campaign Report JL Conroy, ...

  12. SOLVENT EXTRACTION OF URANIUM VALUES

    DOE Patents [OSTI]

    Feder, H.M.; Ader, M.; Ross, L.E.

    1959-02-01

    A process is presented for extracting uranium salt from aqueous acidic solutions by organic solvent extraction. It consists in contacting the uranium bearing solution with a water immiscible dialkylacetamide having at least 8 carbon atoms in the molecule. Mentioned as a preferred extractant is dibutylacetamide. The organic solvent is usually used with a diluent such as kerosene or CCl/sub 4/.

  13. PLUTONIUM-URANIUM-TITANIUM ALLOYS

    DOE Patents [OSTI]

    Coffinberry, A.S.

    1959-07-28

    A plutonium-uranium alloy suitable for use as the fuel element in a fast breeder reactor is described. The alloy contains from 15 to 60 at.% titanium with the remainder uranium and plutonium in a specific ratio, thereby limiting the undesirable zeta phase and rendering the alloy relatively resistant to corrosion and giving it the essential characteristic of good mechanical workability.

  14. ELECTRODEPOSITION OF NICKEL ON URANIUM

    DOE Patents [OSTI]

    Gray, A.G.

    1958-08-26

    A method is described for preparing uranium objects prior to nickel electroplating. The process consiats in treating the surface of the uranium with molten ferric chloride hexahydrate, at a slightiy elevated temperature. This treatment etches the metal surface providing a structure suitable for the application of adherent electrodeposits and at the same time plates the surface with a thin protective film of iron.

  15. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    3. Uranium purchased by owners and operators of U.S. civilian nuclear power reactors by origin country and delivery year, 2011-15 thousand pounds U3O8 equivalent; dollars per pound U3O8 equivalent Deliveries in 2011 Deliveries in 2012 Deliveries in 2013 Deliveries in 2014 Deliveries in 2015 Origin country Purchases Weighted-average price Purchases Weighted-average price Purchases Weighted-average price Purchases Weighted-average price Purchases Weighted-average price Australia 6,001 57.47 6,724

  16. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    5. Average price and quantity for uranium purchased by owners and operators of U.S. civilian nuclear power reactors by pricing mechanisms and delivery year, 2014-15 dollars per pound U3O8 equivalent; thousand pounds U3O8 equivalent Pricing mechanisms Domestic purchases1 Foreign purchases2 Total purchases 2014 2015 2014 2015 2014 2015 Contract-specified (fixed and base-escalated) pricing Weighted-average price 41.87 40.34 49.87 44.93 45.47 42.88 Quantity with reported price 15,711 13,862 12,815

  17. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    a. Uranium purchased by owners and operators of U.S. civilian nuclear power reactors ranked by price and distributed by quantity, 2013-15 deliveries thousand pounds U3O8 equivalent; dollars per pound U3O8 equivalent Deliveries in 2013 Deliveries in 2014 Deliveries in 2015 Quantity 1 distribution Quantity with reported price Weighted-average price Quantity with reported price Weighted-average price Quantity with reported price Weighted-average price First 7,175 34.34 6,665 30.26 6,807 29.68

  18. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    7. Uranium purchased by owners and operators of U.S. civilian nuclear power reactors by contract type and material type, 2015 deliveries thousand pounds U3O8 equivalent; dollars per pound U3O8 equivalent Spot 1 Contracts Long-Term Contracts 2 Total Material Type Quantity with reported price Weighted-average price Quantity with reported price Weighted-average price Quantity with reported price Weighted-average price U3O8 6,175 36.40 24,107 45.76 30,282 43.85 Natural UF6 3,879 38.52 12,292 48.13

  19. METHOD OF ELECTROPLATING ON URANIUM

    DOE Patents [OSTI]

    Rebol, E.W.; Wehrmann, R.F.

    1959-04-28

    This patent relates to a preparation of metallic uranium surfaces for receiving coatings, particularly in order to secure adherent electroplated coatings upon uranium metal. In accordance with the invention the uranium surface is pretreated by degreasing in trichloroethylene, followed by immersion in 25 to 50% nitric acid for several minutes, and then rinsed with running water, prior to pickling in trichloroacetic acid. The last treatment is best accomplished by making the uranium the anode in an aqueous solution of 50 per cent by weight trichloroacetic acid until work-distorted crystals or oxide present on the metal surface have been removed and the basic crystalline structure of the base metal has been exposed. Following these initial steps the metallic uranium is rinsed in dilute nitric acid and then electroplated with nickel. Adnerent firmly-bonded coatings of nickel are obtained.

  20. ISOTOPE CONVERSION DEVICE

    DOE Patents [OSTI]

    Wigner, E.P.; Young, G.J.; Ohlinger, L.A.

    1957-12-01

    This patent relates to nuclear reactors of tbe type utilizing a liquid fuel and designed to convert a non-thermally fissionable isotope to a thermally fissionable isotope by neutron absorption. A tank containing a reactive composition of a thermally fissionable isotope dispersed in a liquid moderator is disposed within an outer tank containing a slurry of a non-thermally fissionable isotope convertible to a thermally fissionable isotope by neutron absorption. A control rod is used to control the chain reaction in the reactive composition and means are provided for circulating and cooling the reactive composition and slurry in separate circuits.

  1. THE RECOVERY OF URANIUM FROM GAS MIXTURE

    DOE Patents [OSTI]

    Jury, S.H.

    1964-03-17

    A method of separating uranium from a mixture of uranium hexafluoride and other gases is described that comprises bringing the mixture into contact with anhydrous calcium sulfate to preferentially absorb the uranium hexafluoride on the sulfate. The calcium sulfate is then leached with a selective solvent for the adsorbed uranium. (AEC)

  2. Process for removing carbon from uranium

    DOE Patents [OSTI]

    Powell, George L.; Holcombe, Jr., Cressie E.

    1976-01-01

    Carbon contamination is removed from uranium and uranium alloys by heating in inert atmosphere to 700.degree.-1900.degree.C in effective contact with yttrium to cause carbon in the uranium to react with the yttrium. The yttrium is either in direct contact with the contaminated uranium or in indirect contact by means of an intermediate transport medium.

  3. PREPARATION OF URANIUM-ALUMINUM ALLOYS

    DOE Patents [OSTI]

    Moore, R.H.

    1962-09-01

    A process is given for preparing uranium--aluminum alloys from a solution of uranium halide in an about equimolar molten alkali metal halide-- aluminum halide mixture and excess aluminum. The uranium halide is reduced and the uranium is alloyed with the excess aluminum. The alloy and salt are separated from each other. (AEC)

  4. Evaluation of Background Concentrations of Contaminants in an Unusual Desert Arroyo Near a Uranium Mill Tailings Disposal Cell - 12260

    SciTech Connect (OSTI)

    Bush, Richard P.; Morrison, Stan J.

    2012-07-01

    The U.S. Department of Energy (DOE) Office of Legacy Management (LM) manages 27 sites that have groundwater containing uranium concentrations above background levels. The distal portions of the plumes merge into background groundwater that can have 50 μg/L or more uranium. Distinguishing background from site-related uranium is often problematic, but it is critical to determining if remediation is warranted, establishing appropriate remediation goals, and evaluating disposal cell performance. In particular, groundwater at disposal cells located on the upper Cretaceous Mancos Shale may have relatively high background concentrations of uranium. Elevated concentrations of nitrate, selenium, and sulfate accompany the uranium. LM used geologic analogs and uranium isotopic signatures to distinguish background groundwater from groundwater contaminated by a former uranium processing site. The same suite of contaminants is present in groundwater near former uranium processing sites and in groundwater seeps emanating from the Mancos Shale over a broad area. The concentrations of these contaminants in Many Devils Wash, located near LM's Shiprock disposal cell, are similar to those in samples collected from many Mancos seeps, including two analog sites that are 8 to 11 km from the disposal cell. Samples collected from Many Devils Wash and the analog sites have high AR values (about 2.0)-in contrast, groundwater samples collected near the tailings disposal cell have AR values near 1.0. These chemical signatures raise questions about the origin of the contamination seeping into Many Devils Wash. (authors)

  5. ELUTION OF URANIUM FROM RESIN

    DOE Patents [OSTI]

    McLEan, D.C.

    1959-03-10

    A method is described for eluting uranium from anion exchange resins so as to decrease vanadium and iron contamination and permit recycle of the major portion of the eluats after recovery of the uranium. Diminution of vanadium and iron contamination of the major portion of the uranium is accomplished by treating the anion exchange resin, which is saturated with uranium complex by adsorption from a sulfuric acid leach liquor from an ore bearing uranium, vanadium and iron, with one column volume of eluant prepared by passing chlorine into ammonium hydroxide until the chloride content is about 1 N and the pH is about 1. The resin is then eluted with 8 to 9 column volumes of 0.9 N ammonium chloride--0.1 N hydrochloric acid solution. The eluants are collected separately and treated with ammonia to precipitate ammonium diuranate which is filtered therefrom. The uranium salt from the first eluant is contaminated with the major portion of ths vanadium and iron and is reworked, while the uranium recovered from the second eluant is relatively free of the undesirable vanadium and irons. The filtrate from the first eluant portion is discarded. The filtrate from the second eluant portion may be recycled after adding hydrochloric acid to increase the chloride ion concentration and adjust the pH to about 1.

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

    SciTech Connect (OSTI)

    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.

  7. SEPARATION OF URANIUM FROM THORIUM

    DOE Patents [OSTI]

    Hellman, N.N.

    1959-07-01

    A process is presented for separating uranium from thorium wherein the ratio of thorium to uranium is between 100 to 10,000. According to the invention the thoriumuranium mixture is dissolved in nitric acid, and the solution is prepared so as to obtain the desired concentration within a critical range of from 4 to 8 N with regard to the total nitrate due to thorium nitrate, with or without nitric acid or any nitrate salting out agent. The solution is then contacted with an ether, such as diethyl ether, whereby uranium is extracted into ihe organic phase while thorium remains in the aqueous phase.

  8. URANIUM RECOVERY FROM NUCLEAR FUEL

    DOE Patents [OSTI]

    Vogel, R.C.; Rodger, W.A.

    1962-04-24

    A process of recovering uranium from a UF/sub 4/-NaFZrF/sub 4/ mixture by spraying the molten mixture at about 200 deg C in nitrogen of super- atmospheric pressure into droplets not larger than 100 microns, and contacting the molten droplets with fluorine at about 200 deg C for 0.01 to 10 seconds in a container the walls of which have a temperature below the melting point of the mixture is described. Uranium hexafluoride is formed and volatilized and the uranium-free salt is solidified. (AEC)

  9. Excess Uranium Inventory Management Plan | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Excess Uranium Inventory Management Plan Excess Uranium Inventory Management Plan The 2013 Excess Uranium Inventory Management Plan describes a framework for the effective...

  10. FLUX COMPOSITION AND METHOD FOR TREATING URANIUM-CONTAINING METAL

    DOE Patents [OSTI]

    Foote, F.

    1958-08-26

    A flux composition is preseated for use with molten uranium and uranium alloys. It consists of about 60% calcium fluoride, 30% calcium chloride and 10% uranium tetrafluoride.

  11. Uranium Processing Facility | Y-12 National Security Complex

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Y-12 Uranium Processing Facility Uranium Processing Facility UPF will be a state-of-the-art, consolidated facility for enriched uranium operations including assembly,...

  12. DOE Uranium Leasing Program 2015 Mitigation Action Plan Activity...

    Energy Savers [EERE]

    DOE Uranium Leasing Program 2015 Mitigation Action Plan Activity Summary Report DOE Uranium Leasing Program 2015 Mitigation Action Plan Activity Summary Report DOE Uranium Leasing ...

  13. Researchers use light to create rare uranium molecule

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Rare uranium molecule Researchers use light to create rare uranium molecule Uranium nitride materials show promise as advanced nuclear fuels due to their high density, high ...

  14. URANIUM PURIFICATION PROCESS

    DOE Patents [OSTI]

    Ruhoff, J.R.; Winters, C.E.

    1957-11-12

    A process is described for the purification of uranyl nitrate by an extraction process. A solution is formed consisting of uranyl nitrate, together with the associated impurities arising from the HNO/sub 3/ leaching of the ore, in an organic solvent such as ether. If this were back extracted with water to remove the impurities, large quantities of uranyl nitrate will also be extracted and lost. To prevent this, the impure organic solution is extracted with small amounts of saturated aqueous solutions of uranyl nitrate thereby effectively accomplishing the removal of impurities while not allowing any further extraction of the uranyl nitrate from the organic solvent. After the impurities have been removed, the uranium values are extracted with large quantities of water.

  15. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    3. Deliveries of uranium feed by owners and operators of U.S. civilian nuclear power reactors by enrichment country and delivery year, 2013-15 thousand pounds U3O8 equivalent Feed deliveries in 2013 Feed deliveries in 2014 Feed deliveries in 2015 Enrichment country U.S.-origin Foreign-origin Total U.S.-origin Foreign-origin Total U.S.-origin Foreign-origin Total China 0 W W W W W 0 W W France 0 1,606 1,606 0 3,055 3,055 W W 3,299 Germany W W W W W 2,140 W W W Netherlands 1,058 2,773 3,831 0

  16. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    9. Foreign purchases of uranium by U.S. suppliers and owners and operators of U.S. civilian nuclear power reactors by delivery year, 2011-15 thousand pounds U3O8 equivalent; dollars per pound U3O8 equivalent Deliveries 2011 2012 2013 2014 2015 U.S. suppliers Foreign purchases 19,318 20,196 23,233 24,199 27,233 Weighted-average price 48.80 46.80 43.25 39.13 40.68 Owners and operators of U.S. civilian nuclear power reactors Foreign purchases 35,071 36,037 34,095 34,404 36,912 Weighted-average

  17. Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    1. U.S. uranium drilling activities, 2003-15 Exploration drilling Development drilling Exploration and development drilling Year Number of holes Feet (thousand) Number of holes Feet (thousand) Number of holes Feet (thousand) 2003 NA NA NA NA W W 2004 W W W W 2,185 1,249 2005 W W W W 3,143 1,668 2006 1,473 821 3,430 1,892 4,903 2,713 2007 4,351 2,200 4,996 2,946 9,347 5,146 2008 5,198 2,543 4,157 2,551 9,355 5,093 2009 1,790 1,051 3,889 2,691 5,679 3,742 2010 2,439 1,460 4,770 3,444 7,209 4,904

  18. Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    6. Employment in the U.S. uranium production industry by category, 2003-15 person-years Year Exploration Mining Milling Processing Reclamation Total 2003 W W W W 117 321 2004 18 108 W W 121 420 2005 79 149 142 154 124 648 2006 188 121 W W 155 755 2007 375 378 107 216 155 1,231 2008 457 558 W W 154 1,563 2009 175 441 W W 162 1,096 2010 211 400 W W 125 1,073 2011 208 462 W W 102 1,191 2012 161 462 W W 179 1,196 2013 149 392 W W 199 1,156 2014 86 246 W W 161 787 2015 58 251 W W 116

  19. Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    7. Employment in the U.S. uranium production industry by state, 2003-15 person-years State(s) 2003 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 2015 Wyoming 134 139 181 195 245 301 308 348 424 512 531 416 343 Colorado and Texas 48 140 269 263 557 696 340 292 331 248 198 105 79 Nebraska and New Mexico 92 102 123 160 149 160 159 134 127 W W W W Arizona, Utah, and Washington 47 40 75 120 245 360 273 281 W W W W W Alaska, Michigan, Nevada, and South Dakota 0 0 0 16 25 30 W W W W W 0 0

  20. Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    2. U.S. uranium mine production and number of mines and sources, 2003-15 Production / Mining method 2003 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 2015 Underground (estimated contained thousand pounds U3O8) W W W W W W W W W W W W W Open Pit (estimated contained thousand pounds U3O8) 0 0 0 0 0 0 0 0 0 0 0 0 0 In-Situ Leaching (thousand pounds U3O8) W W 2,681 4,259 W W W W W W W W W Other1 (thousand pounds U3O8) W W W W W W W W W W W W W Total Mine Production (thousand pounds U3O8)

  1. Uranium hexafluoride bibliography

    SciTech Connect (OSTI)

    Burnham, S.L.

    1988-01-01

    This bibliography is a compilation of reports written about the transportation, handling, safety, and processing of uranium hexafluoride. An on-line literature search was executed using the DOE Energy files and the Nuclear Science Abstracts file to identify pertinent reports. The DOE Energy files contain unclassified information that is processed at the Office of Scientific and Technical Information of the US Department of Energy. The reports selected from these files were published between 1974 and 1983. Nuclear Science Abstracts contains unclassified international nuclear science and technology literature published from 1948 to 1976. In addition, scientific and technical reports published by the US Atomic Energy Commission and the US Energy Research and Development Administration, as well as those published by other agencies, universities, and industrial and research organizations, are included in the Nuclear Science Abstracts file. An alphabetical listing of the acronyms used to denote the corporate sponsors follows the bibliography.

  2. 2015 Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    Activity at U.S. Mills and In-Situ-Leach Plants 2003 2004 2005 2006 2007 2008 2009 2010 ... Total Uranium Concentrate Shipped from U.S. Mills and In-Situ-Leach Plants Table 3. U.S. ...

  3. 2015 Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

    The natural UF 6 and enriched UF 6 weighted-average price represent only the U 3 O 8 equivalent uranium-component price specified in the contract for each delivery of natural UF 6 ...

  4. 2015 Domestic Uranium Production Report

    U.S. Energy Information Administration (EIA) Indexed Site

    Uranium Concentrate Sales by U.S. Producers 3" "Deliveries (thousand pounds U3O8)","W","W","W",3786,3602,3656,2044,2684,2870,3630,4447,4746,3634 "Weighted-Average Price ...

  5. 2015 Uranium Marketing Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    Uranium Marketing Annual Report Release Date: May 24, 2016 Next Release Date: May 2017 Deliveries 2011 2012 2013 2014 2015 Purchases 1,668 1,194 W 410 2,702 Weighted-average price ...

  6. 2015 Uranium Marketing Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    7 2015 Uranium Marketing Annual Report Release Date: May 24, 2016 Next Release Date: May 2017 Table S3b. Weighted-average price of foreign purchases and foreign sales by U.S. ...

  7. 2015 Uranium Marketing Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    b. Uranium purchased by owners and operators of U.S. civilian nuclear power reactors ranked by price and distributed by purchaser, 2013-15 deliveries" "thousand pounds U3O8 ...

  8. 2015 Domestic Uranium Production Report

    U.S. Energy Information Administration (EIA) Indexed Site

    Forward costs are neither the full costs of production nor the market price at which the uranium, when produced, might be sold." "Note: Totals may not equal sum of components ...

  9. 2015 Uranium Marketing Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    6a. Uranium purchased by owners and operators of U.S. civilian nuclear power reactors ranked by price and distributed by quantity, 2013-15 deliveries" "thousand pounds U3O8 ...

  10. 2015 Uranium Marketing Annual Survey

    U.S. Energy Information Administration (EIA) Indexed Site

    and enriched UF6 weighted-average price represent only the U3O8 equivalent uranium-component price specified in the contract for each delivery of natural UF6 and enriched UF6, ...

  11. 2015 Uranium Marketing Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    2013-15" 2013,2014,2015 "American Fuel Resources, LLC","Advance Uranium Asset Management Ltd.","AREVA AREVA NC, Inc." "AREVA NC, Inc.","AREVA AREVA NC, Inc.","ARMZ ...

  12. 2015 Uranium Marketing Annual Report

    U.S. Energy Information Administration (EIA) Indexed Site

    Next Release Date: May 2017 2013 2014 2015 American Fuel Resources, LLC Advance Uranium Asset Management Ltd. AREVA AREVA NC, Inc. AREVA NC, Inc. AREVA AREVA NC, Inc. ARMZ ...

  13. 2015 Uranium Marketing Annual Report

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    Uranium purchased by owners and operators of U.S. civilian nuclear power reactors, ... Purchased from other owners and operators of U.S. civilian nuclear power reactors, other ...

  14. 2015 Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    5 2015 Domestic Uranium Production Report Release Date: May 5, 2016 Next Release Date: May 2017 Production Mining Method 2003 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 ...

  15. 2015 Uranium Marketing Annual Report

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    owners and operators of U.S. civilian nuclear power reactors, 1994-2015 Year Feed deliveries by owners and operators of U.S. civilian nuclear power reactors Uranium in fuel ...

  16. 2015 Domestic Uranium Production Report

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    7 2015 Domestic Uranium Production Report Release Date: May 5, 2016 Next Release Date: May 2017 Capacity (short tons of ore per day) 2011 2012 2013 2014 2015 Anfield Resources ...

  17. 2015 Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    9 2015 Domestic Uranium Production Report Release Date: May 5, 2016 Next Release Date: May 2017 Year Exploration Mining Milling Processing Reclamation Total 2003 W W W W 117 321 ...

  18. Y-12 and uranium history

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    did happen six days after he was given the assignment. The history of uranium at Y-12 began with that decision, which will be commemorated on September 19, 2012, at...

  19. 2015 Domestic Uranium Production Report

    U.S. Energy Information Administration (EIA) Indexed Site

    May 5, 2016" "Next Release Date: May 2017" "Table 4. U.S. uranium mills and heap leach facilities by owner, location, capacity, and operating status at end of the year, 2011-15" ...

  20. 2015 Domestic Uranium Production Report

    U.S. Energy Information Administration (EIA) Indexed Site

    7. Employment in the U.S. uranium production industry by state, 2003-15" "person-years" "State(s)",2003,2004,2005,2006,2007,2008,2009,2010,2011,2012,2013,2014,2015 ...

  1. Laser induced phosphorescence uranium analysis

    DOE Patents [OSTI]

    Bushaw, B.A.

    1983-06-10

    A method is described for measuring the uranium content of aqueous solutions wherein a uranyl phosphate complex is irradiated with a 5 nanosecond pulse of 425 nanometer laser light and resultant 520 nanometer emissions are observed for a period of 50 to 400 microseconds after the pulse. Plotting the natural logarithm of emission intensity as a function of time yields an intercept value which is proportional to uranium concentration.

  2. LIQUID METAL COMPOSITIONS CONTAINING URANIUM

    DOE Patents [OSTI]

    Teitel, R.J.

    1959-04-21

    Liquid metal compositions containing a solid uranium compound dispersed therein is described. Uranium combines with tin to form the intermetallic compound USn/sub 3/. It has been found that this compound may be incorporated into a liquid bath containing bismuth and lead-bismuth components, if a relatively small percentage of tin is also included in the bath. The composition has a low thermal neutron cross section which makes it suitable for use in a liquid metal fueled nuclear reactor.

  3. MELTING AND PURIFICATION OF URANIUM

    DOE Patents [OSTI]

    Spedding, F.H.; Gray, C.F.

    1958-09-16

    A process is described for treating uranium ingots having inner metal portions and an outer oxide skin. The method consists in partially supporting such an ingot on the surface of a grid or pierced plate. A sufficient weight of uranium is provided so that when the mass becomes molten, the oxide skin bursts at the unsupported portions of its bottom surface, allowing molten urantum to flow through the burst skin and into a container provided below.

  4. SURFACE TREATMENT OF METALLIC URANIUM

    DOE Patents [OSTI]

    Gray, A.G.; Schweikher, E.W.

    1958-05-27

    The treatment of metallic uranium to provide a surface to which adherent electroplates can be applied is described. Metallic uranium is subjected to an etchant treatment in aqueous concentrated hydrochloric acid, and the etched metal is then treated to dissolve the resulting black oxide and/or chloride film without destroying the etched metal surface. The oxide or chloride removal is effected by means of moderately concentrated nitric acid in 3 to 20 seconds.

  5. Laser induced phosphorescence uranium analysis

    DOE Patents [OSTI]

    Bushaw, Bruce A.

    1986-01-01

    A method is described for measuring the uranium content of aqueous solutions wherein a uranyl phosphate complex is irradiated with a 5 nanosecond pulse of 425 nanometer laser light and resultant 520 nanometer emissions are observed for a period of 50 to 400 microseconds after the pulse. Plotting the natural logarithm of emission intensity as a function of time yields an intercept value which is proportional to uranium concentration.

  6. A record of uranium-series transport at Nopal I, Sierra Pena Blanca, Mexico: Implications for natural uranium deposits and radioactive waste repositories

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Denton, J. S.; Goldstein, S. J.; Paviet, P.; Nunn, A. J.; Amato, R. S.; Hinrichs, K. A.

    2016-04-10

    Studies of uranium-series (U-series) disequilibria within and around ore deposits provide valuable information on the extent and timing of actinide mobility, via mineral-fluid interaction, over a range of spatial and temporal scales. Such information is useful in studies of analogs of high-level nuclear-waste repositories, as well as for mining and mineral extraction sites, locations of previous nuclear weapons testing, and legacy nuclear waste contamination. In this study we present isotope dilution mass spectrometry U-series measurements for fracture-fill materials (hematite, goethite, kaolinite, calcite, dolomite and quartz) from one such analog; the Nopal I uranium ore deposit situated at Peña Blanca inmore » the Chihuahua region of northern Mexico. The ore deposit is located in fractured, unsaturated volcanic tuff and fracture-fill materials from surface fractures as well as fractures in a vertical drill core have been analyzed. High uranium concentrations in the fracture-fill materials (between 12 and 7700 ppm) indicate uranium mobility and transport from the deposit. Furthermore, uranium concentrations generally decrease with horizontal distance away from the deposit but in this deposit there is no trend with depth below the surface.« less

  7. PROCESS FOR PRODUCING URANIUM HEXAFLUORIDE

    DOE Patents [OSTI]

    Fowler, R.D.

    1957-10-22

    A process for the production of uranium hexafluoride from the oxides of uranium is reported. In accordance with the method the higher oxides of uranium may be reduced to uranium dioxide (UO/sub 2/), the latter converted into uranium tetrafluoride by reaction with hydrogen fluoride, and the UF/sub 4/ convented to UF/sub 6/ by reaction with a fluorinating agent. The UO/sub 3/ or U/sub 3/O/sub 8/ is placed in a reaction chamber in a copper boat or tray enclosed in a copper oven, and heated to 500 to 650 deg C while hydrogen gas is passed through the oven. The oven is then swept clean of hydrogen and the water vapor formed by means of nitrogen and then while continuing to maintain the temperature between 400 and 600 deg C, anhydrous hydrogen fluoride is passed through. After completion of the conversion to uranium tetrafluoride, the temperature of the reaction chamber is lowered to ahout 400 deg C, and elemental fluorine is used as the fluorinating agent for the conversion of UF/sub 4/ into UF/sub 6/. The fluorine gas is passed into the chamber, and the UF/sub 6/ formed passes out and is delivered to a condenser.

  8. Nuclear forensic analysis of an unknown uranium ore concentrate sample seized in a criminal investigation in Australia

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Keegan, Elizabeth; Kristo, Michael J.; Colella, Michael; Robel, Martin; Williams, Ross; Lindvall, Rachel; Eppich, Gary; Roberts, Sarah; Borg, Lars; Gaffney, Amy; et al

    2014-04-13

    In early 2009, a state policing agency raided a clandestine drug laboratory in a suburb of a major city in Australia. While searching the laboratory, they discovered a small glass jar labelled “Gamma Source” and containing a green powder. The powder was radioactive. This paper documents the detailed nuclear forensic analysis undertaken to characterize and identify the material and determine its provenance. Isotopic and impurity content, phase composition, microstructure and other characteristics were measured on the seized sample, and the results were compared with similar material obtained from the suspected source (ore and ore concentrate material). While an extensive rangemore » of parameters were measured, the key ‘nuclear forensic signatures’ used to identify the material were the U isotopic composition, Pb and Sr isotope ratios, and the rare earth element pattern. These measurements, in combination with statistical analysis of the elemental and isotopic content of the material against a database of uranium ore concentrates sourced from mines located worldwide, led to the conclusion that the seized material (a uranium ore concentrate of natural isotopic abundance) most likely originated from Mary Kathleen, a former Australian uranium mine.« less

  9. Future of the Department of Energy's uranium enrichment enterprise

    SciTech Connect (OSTI)

    Sewell, P.G.

    1991-11-01

    The national energy strategy (NES) developed at President Bush's direction provides a focus for the US Department of Energy (DOE) future policy and funding initiatives including those of the uranium enrichment enterprise. The NES identifies an important and continuing role for nuclear energy as part of a balanced array of energy sources for meeting US energy needs, especially the growing demand for electricity. For many years, growth in US electricity demand has exhibited a strong correlation with growth in gross national product. NEW projections indicate that the US will need between 190 and 275 GW of additional system capacity by 2010. In order to unable nuclear power to help meet this need, the NEW establishes basic objectives for nuclear power. These objectives are to have a first order of a new nuclear power plant by 1995 and to have such a plant operational by 2000. The expansion of nuclear power anticipated in the NEW affirms a continuing need for a strong domestic uranium enrichment services supply capability. In terms of the future outlook for uranium enrichment, the atomic vapor laser isotope separation (AVLIS) technology continues to hold great promise for commercial application. If AVLIS efforts are successful, significant financial benefits from the commercial use of AVLIS will be realized by customers and the AVLIS deployment entity by approximately the year 2000 and thereafter.

  10. Uranium enrichment: investment options for the long term

    SciTech Connect (OSTI)

    Not Available

    1983-01-01

    The US government supplies a major portion of the enriched uranium used to fuel most of the nuclear power plants that furnish electricity in the free world. As manager of the US uranium enrichment concern, the Department of Energy (DOE) is investigating a number of technological choices to improve enrichment service and remain a significant world supplier. The Congress will ultimately select a strategy for federal investment in the uranium enrichment enterprise. A fundamental policy choice between possible future roles - that of the free world's main supplier of enrichment services, and that of a mainly domestic supplier - will underlie any investment decision the Congress makes. The technological choices are gaseous diffusion, gas centrifuge, and atomic vapor laser isotope separation (AVLIS). A base plan and four alternatives were examined by DOE and the Congressional Budget Office. In terms of total enterprise costs, Option IV, ultimately relying on advanced gas centrifuges for enrichment services, would offer the most economic approach, with costs over the full projection period totaling $123.5 billion. Option III, ultimately relying on AVLIS without gas centrifuge enrichment or gaseous diffusion, falls next in the sequence, with costs of $128.2 billion. Options I and II, involving combinations of the gas centrifuge and AVLIS technologies, follow closely with costs of $128.7 and $129.6 billion. The base plan has costs of $136.8 billion over the projection period. 1 figure, 22 tables.

  11. HYDROGEN ISOTOPE TARGETS

    DOE Patents [OSTI]

    Ashley, R.W.

    1958-08-12

    The design of targets for use in the investigation of nuclear reactions of hydrogen isotopes by bombardment with accelerated particles is described. The target con struction eomprises a backing disc of a metal selected from the group consisting of molybdenunn and tungsten, a eoating of condensed titaniunn on the dise, and a hydrogen isotope selected from the group consisting of deuterium and tritium absorbed in the coatiag. The proeess for preparing these hydrogen isotope targets is described.

  12. ARM - Measurement - Isotope ratio

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    govMeasurementsIsotope ratio ARM Data Discovery Browse Data Comments? We would love to hear from you! Send us a note below or call us at 1-888-ARM-DATA. Send Measurement : Isotope ratio Ratio of stable isotope concentrations. Categories Atmospheric State, Atmospheric Carbon Instruments The above measurement is considered scientifically relevant for the following instruments. Refer to the datastream (netcdf) file headers of each instrument for a list of all available measurements, including those

  13. Hybrid isotope separation scheme

    DOE Patents [OSTI]

    Maya, J.

    1991-06-18

    A method is described for yielding selectively a desired enrichment in a specific isotope including the steps of inputting into a spinning chamber a gas from which a scavenger, radiating the gas with a wave length or frequency characteristic of the absorption of a particular isotope of the atomic or molecular gas, thereby inducing a photochemical reaction between the scavenger, and collecting the specific isotope-containing chemical by using a recombination surface or by a scooping apparatus. 2 figures.

  14. Hybrid isotope separation scheme

    DOE Patents [OSTI]

    Maya, Jakob

    1991-01-01

    A method of yielding selectively a desired enrichment in a specific isotope including the steps of inputting into a spinning chamber a gas from which a scavenger, radiating the gas with a wave length or frequency characteristic of the absorption of a particular isotope of the atomic or molecular gas, thereby inducing a photochemical reaction between the scavenger, and collecting the specific isotope-containing chemical by using a recombination surface or by a scooping apparatus.

  15. Nuclear forensic analysis of uranium oxide powders interdicted in Victoria, Australia

    SciTech Connect (OSTI)

    Kristo, Michael Joseph; Keegan, Elizabeth; Colella, Michael; Williams, Ross; Lindvall, Rachel; Eppich, Gary; Roberts, Sarah; Borg, Lars; Gaffney, Amy; Plaue, Jonathan; Knight, Kim; Loi, Elaine; Hotchkis, Michael; Moody, Kenton; Singleton, Michael; Robel, Martin; Hutcheon, Ian

    2015-04-13

    Nuclear forensic analysis was conducted on two uranium samples confiscated during a police investigation in Victoria, Australia. The first sample, designated NSR-F-270409-1, was a depleted uranium powder of moderate purity (~1000 μg/g total elemental impurities). The chemical form of the uranium was a compound similar to K2(UO2)3O4·4H2O. While aliquoting NSR-F-270409-1 for analysis, the body and head of a Tineid moth was discovered in the sample. The second sample, designated NSR-F-270409-2, was also a depleted uranium powder. It was of reasonably high purity (~380 μg/g total elemental impurities). The chemical form of the uranium was primarily UO3·2H2O, with minor phases of U3O8 and UO2. While aliquoting NSR-F-270409-2 for analysis, a metal staple of unknown origin was discovered in the sample. The presence of 236U and 232U in both samples indicates that the uranium feed stocks for these samples experienced a neutron flux at some point in their history. The reactor burn-up calculated from the isotopic composition of the uranium is consistent with that of spent fuel from natural uranium (NU) fueled Pu production. These nuclear forensic conclusions allow us to categorically exclude Australia as the origin of the material and greatly reduce the number of candidate sources.

  16. Stable isotope enrichment

    ScienceCinema (OSTI)

    Egle, Brian

    2014-07-15

    Brian Egle is working to increase the nation's capacity to produce stable isotopes for use including medicine, industry and national security.

  17. Stable isotope enrichment

    SciTech Connect (OSTI)

    Egle, Brian

    2014-07-14

    Brian Egle is working to increase the nation's capacity to produce stable isotopes for use including medicine, industry and national security.

  18. PRINCIPAL ISOTOPE SELECTION REPORT

    SciTech Connect (OSTI)

    K. D. Wright

    1998-08-28

    Utilizing nuclear fuel to produce power in commercial reactors results in the production of hundreds of fission product and transuranic isotopes in the spent nuclear fuel (SNF). When the SNF is disposed of in a repository, the criticality analyses could consider all of the isotopes, some principal isotopes affecting criticality, or none of the isotopes, other than the initial loading. The selected set of principal isotopes will be the ones used in criticality analyses of the SNF to evaluate the reactivity of the fuel/waste package composition and configuration. This technical document discusses the process used to select the principal isotopes and the possible affect that these isotopes could have on criticality in the SNF. The objective of this technical document is to discuss the process used to select the principal isotopes for disposal criticality evaluations with commercial SNF. The principal isotopes will be used as supporting information in the ''Disposal Criticality Analysis Methodology Topical Report'' which will be presented to the United States Nuclear Regulatory Commission (NRC) when approved by the United States Department of Energy (DOE) Office of Civilian Radioactive Waste Management (OCRWM).

  19. METHOD OF APPLYING NICKEL COATINGS ON URANIUM

    DOE Patents [OSTI]

    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.

  20. SOLVENT EXTRACTION PROCESS FOR URANIUM RECOVERY

    DOE Patents [OSTI]

    Clark, H.M.; Duffey, D.

    1958-06-17

    A process is described for extracting uranium from uranium ore, wherein the uranium is substantially free from molybdenum contamination. In a solvent extraction process for recovering uranium, uranium and molybdenum ions are extracted from the ore with ether under high acidity conditions. The ether phase is then stripped with water at a lower controiled acidity, resaturated with salting materials such as sodium nitrate, and reextracted with the separation of the molybdenum from the uranium without interference from other metals that have been previously extracted.

  1. Round-robin 230Th–234U age dating of bulk uranium for nuclear forensics

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Gaffney, Amy M.; Hubert, Amélie; Kinman, William S.; Magara, Masaaki; Okubo, Ayako; Pointurier, Fabien; Schorzman, Kerri C.; Steiner, Robert E.; Williams, Ross W.

    2015-07-30

    In an inter-laboratory measurement comparison study, four laboratories determined 230Th–234U model ages of uranium certified reference material NBL U050 using isotope dilution mass spectrometry. The model dates determined by the participating laboratories range from 9 March 1956 to 19 October 1957, and are indistinguishable given the associated measurement uncertainties. As a result, these model ages are concordant with to slightly older than the known production age of NBL U050.

  2. Price Quotes and Isotope Ordering

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Ordering Price Quotes and Isotope Ordering Isotopes produced at Los Alamos National Laboratory are saving lives, advancing cutting-edge research and keeping the U.S. safe. Isotope...

  3. Profile of World Uranium Enrichment Programs-2009

    SciTech Connect (OSTI)

    Laughter, Mark D

    2009-04-01

    It is generally agreed that the most difficult step in building a nuclear weapon is acquiring fissile material, either plutonium or highly enriched uranium (HEU). Plutonium is produced in a nuclear reactor, whereas HEU is produced using a uranium enrichment process. Enrichment is also an important step in the civil nuclear fuel cycle, in producing low enriched uranium (LEU) for use as fuel for nuclear reactors to generate electricity. However, the same equipment used to produce LEU for nuclear reactor fuel can also be used to produce HEU for weapons. Safeguards at an enrichment plant are the array of assurances and verification techniques that ensure uranium is not diverted or enriched to HEU. There are several techniques for enriching uranium. The two most prevalent are gaseous diffusion, which uses older technology and requires a lot of energy, and gas centrifuge separation, which uses more advanced technology and is more energy efficient. Gaseous diffusion plants (GDPs) provide about 40% of current world enrichment capacity but are being phased out as newer gas centrifuge enrichment plants (GCEPs) are constructed. Estimates of current and future enrichment capacity are always approximate, due to the constant upgrades, expansions, and shutdowns occurring at enrichment plants, largely determined by economic interests. Currently, the world enrichment capacity is approximately 56 million kilogram separative work units (SWU) per year, with 22.5 million in gaseous diffusion and more than 33 million in gas centrifuge plants. Another 34 million SWU/year of capacity is under construction or planned for the near future, almost entirely using gas centrifuge separation. Other less-efficient techniques have also been used in the past, including electromagnetic and aerodynamic separations, but these are considered obsolete, at least from a commercial perspective. Laser isotope separation shows promise as a possible enrichment technique of the future but has yet to be

  4. Uranium reference materials

    SciTech Connect (OSTI)

    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.

  5. Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    8. U.S. uranium expenditures, 2003-15 million dollars Year Drilling1 Production2 Land and other 3 Total expenditures Total land and other Land Exploration Reclamation 2003 W W 31.3 NA NA NA W 2004 10.6 27.8 48.4 NA NA NA 86.9 2005 18.1 58.2 59.7 NA NA NA 136.0 2006 40.1 65.9 115.2 41.0 23.3 50.9 221.2 2007 67.5 90.4 178.2 77.7 50.3 50.2 336.2 2008 81.9 221.2 164.4 65.2 50.2 49.1 467.6 2009 35.4 141.0 104.0 17.3 24.2 62.4 280.5 2010 44.6 133.3 99.5 20.2 34.5 44.7 277.3 2011 53.6 168.8 96.8 19.6

  6. SIMS Analyses of Aerodynamic Fallout from a Uranium-Fueled Test

    SciTech Connect (OSTI)

    Lewis, L. A.; Knight, K. B.; Matzel, J. E.; Prussin, S. G.; Ryerson, F. J.; Kinman, W. S.; Zimmer, M. M.; Hutcheon, I. D.

    2014-09-09

    Five silicate fallout glass spherules produced in a uranium-fueled, near-surface nuclear test were characterized by secondary ion mass spectrometry, electron probe microanalysis, autoradiography, scanning electron microscopy, and energy-dispersive x ray spectroscopy. Several samples display distinctive compositional heterogeneity suggestive of incomplete mixing, and exhibit heterogeneity in U isotopes with 0.02 < 235U/ 238U < 11.8 among all five samples and 0.02 < 235U/ 238U < 7.81 within a single sample. In two samples, the 235U/ 238U ratio is correlated with major element composition, consistent with the agglomeration of chemically and isotopically distinct molten precursors. Two samples are quasi-homogeneous with respect to composition and uranium isotopic composition, suggesting extensive mixing possibly due longer residence time in the fireball. Correlated variations between 234U, 235U, 236U and 238U abundances point to mixing of end-members corresponding to uranium derived from the device and natural U ( 238U/ 235U = 0.00725) found in soil.

  7. Uranium Leasing Program Environmental Documents | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Environmental Documents Uranium Leasing Program Environmental Documents Uranium Leasing Program 2015 Mitigation Action Plan Activity Summary Report (March 2016) The DOE Uranium Leasing Program's 2015 Mitigation Action Plan Activity Summary fulfills the mitigation plan's requirement to annually notify the public of mitigation activities completed by Uranium Leasing Program lessees. Uranium Leasing Program Mitigation Action Plan for the Final Uranium Leasing Program Programmatic Environmental

  8. Evaluation of kinetic phosphorescence analysis for the determination of uranium

    SciTech Connect (OSTI)

    Croatto, P.V.; Frank, I.W.; Johnson, K.D.; Mason, P.B.; Smith, M.M.

    1997-12-01

    In the past, New Brunswick Laboratory (NBL) has used a fluorometric method for the determination of sub-microgram quantities of uranium. In its continuing effort to upgrade and improve measurement technology, NBL has evaluated the commercially-available KPA-11 kinetic phosphorescence analyzer (Chemchek, Richland, WA). The Chemchek KPA-11 is a bench-top instrument which performs single-measurement, quench-corrected analyses for trace uranium. It incorporates patented kinetic phosphorimetry techniques to measure and analyze sample phosphorescence as a function of time. With laser excitation and time-corrected photon counting, the KPA-11 has a lower detection limit than conventional fluorometric methods. Operated with a personal computer, the state-of-the-art KPA-11 offers extensive time resolution and phosphorescence lifetime capabilities for additional specificity. Interferences are thereby avoided while obtaining precise measurements. Routine analyses can be easily and effectively accomplished, with the accuracy and precision equivalent to the pulsed-laser fluorometric method presently performed at NBL, without the need for internal standards. Applications of kinetic phosphorimetry at NBL include the measurement of trace level uranium in retention tank, waste samples, and low-level samples. It has also been used to support other experimental activities at NBL by the measuring of nanogram amounts of uranium contamination (in blanks) in isotopic sample preparations, and the determining of elution curves of different ion exchange resins used for uranium purification. In many cases, no pretreatment of samples was necessary except to fume them with nitric acid, and then to redissolve and dilute them to an appropriate concentration with 1 M HNO{sub 3} before measurement. Concentrations were determined on a mass basis ({micro}g U/g of solution), but no density corrections were needed since all the samples (including the samples used for calibration) were in the same

  9. Reducing emissions from uranium dissolving

    SciTech Connect (OSTI)

    Griffith, W.L.; Compere, A.L.; Huxtable, W.P.; Googin, J.M.

    1992-10-01

    This study was designed to assess the feasibility of decreasing NO[sub x] emissions from the current uranium alloy scrap tray dissolving facility. In the current process, uranium scrap is dissolved in boiling nitric acid in shallow stainless-steel trays. As scrap dissolves, more metal and more nitric acid are added to the tray by operating personnel. Safe geometry is assured by keeping liquid level at or below 5 cm, the depth of a safe infinite slab. The accountability batch control system provides additional protection against criticality. Both uranium and uranium alloys are dissolved. Nitric acid is recovered from the vapors for reuse. Metal nitrates are sent to uranium recovery. Brown NO[sub x] fumes evolved during dissolving have occasionally resulted in a visible plume from the trays. The fuming is most noticeable during startup and after addition of fresh acid to a tray. Present environmental regulations are expected to require control of brown NO[sub x] emissions. A detailed review of the literature, indicated the feasibility of slightly altering process chemistry to favor the production of NO[sub 2] which can be scrubbed and recycled as nitric acid. Methods for controlling the process to manage offgas product distribution and to minimize chemical reaction hazards were also considered.

  10. Reducing emissions from uranium dissolving

    SciTech Connect (OSTI)

    Griffith, W.L.; Compere, A.L.; Huxtable, W.P.; Googin, J.M.

    1992-10-01

    This study was designed to assess the feasibility of decreasing NO{sub x} emissions from the current uranium alloy scrap tray dissolving facility. In the current process, uranium scrap is dissolved in boiling nitric acid in shallow stainless-steel trays. As scrap dissolves, more metal and more nitric acid are added to the tray by operating personnel. Safe geometry is assured by keeping liquid level at or below 5 cm, the depth of a safe infinite slab. The accountability batch control system provides additional protection against criticality. Both uranium and uranium alloys are dissolved. Nitric acid is recovered from the vapors for reuse. Metal nitrates are sent to uranium recovery. Brown NO{sub x} fumes evolved during dissolving have occasionally resulted in a visible plume from the trays. The fuming is most noticeable during startup and after addition of fresh acid to a tray. Present environmental regulations are expected to require control of brown NO{sub x} emissions. A detailed review of the literature, indicated the feasibility of slightly altering process chemistry to favor the production of NO{sub 2} which can be scrubbed and recycled as nitric acid. Methods for controlling the process to manage offgas product distribution and to minimize chemical reaction hazards were also considered.

  11. Uranium Biomineralization By Natural Microbial Phosphatase Activities...

    Office of Scientific and Technical Information (OSTI)

    Finally, the minerals produced during this process are stable in low pH conditions or ... strategy to uranium bioreduction in low pH uranium-contaminated environments. ...

  12. The Electrolytic Production of Metallic Uranium

    DOE Patents [OSTI]

    Rosen, R.

    1950-08-22

    This patent covers a process for producing metallic uranium by electrolyzing uranium tetrafluoride at an elevated temperature in a fused bath consisting essentially of mixed alkali and alkaline earth halides.

  13. Absorption of Thermal Neutrons in Uranium

    DOE R&D Accomplishments [OSTI]

    Creutz, E. C.; Wilson, R. R.; Wigner, E. P.

    1941-09-26

    A knowledge of the absorption processes for neutrons in uranium is important for planning a chain reaction experiment. The absorption of thermal neutrons in uranium and uranium oxide has been studied. Neutrons from the cyclotron were slowed down by passage through a graphite block. A uranium or uranium oxide sphere was placed at various positions in the block. The neutron intensity at different points in the sphere and in the graphite was measured by observing the activity induced in detectors or uranium oxide or manganese. It was found that both the fission activity in the uranium oxide and the activity induced in manganese was affected by non-thermal neutrons. An experimental correction for such effects was made by making measurements with the detectors surrounded by cadmium. After such corrections the results from three methods of procedure with the uranium oxide detectors and from the manganese detectors were consistent to within a few per cent.

  14. Uranium Marketing Annual Report - Energy Information Administration

    Gasoline and Diesel Fuel Update (EIA)

    Uranium purchases and prices Owners and operators of U.S. civilian nuclear power reactors ... Uranium in fuel assemblies loaded into U.S. civilian nuclear power reactors during 2015 ...

  15. Inherently safe in situ uranium recovery

    SciTech Connect (OSTI)

    Krumhansl, James L; Brady, Patrick V

    2014-04-29

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

  16. RECOVERY OF URANIUM VALUES FROM URANIUM BEARING RAW MATERIALS

    DOE Patents [OSTI]

    Michal, E.J.; Porter, R.R.

    1959-06-16

    Uranium leaching from ground uranium-bearing raw materials using MnO/sub 2/ in H/sub 2/SO/sub 4/ is described. The MnO/sub 2/ oxidizes U to the leachable hexavalent state. The MnO/sub 2/ does not replace Fe normally added, because the Fe complexes P and catalyzes the MnO/sub 2/ reaction. Three examples of continuous processes are given, but batch operation is also possible. The use of MnO/sub 2/ makes possible recovery of very low U values. (T.R.H.)

  17. PROCESS FOR THE RECOVERY OF URANIUM

    DOE Patents [OSTI]

    Morris, G.O.

    1955-06-21

    This patent relates to a process for the recovery of uranium from impure uranium tetrafluoride. The process consists essentially of the steps of dissolving the impure uranium tetrafluoride in excess dilute sulfuric acid in the presence of excess hydrogen peroxide, precipitating ammonium uranate from the solution so formed by adding an excess of aqueous ammonia, dissolving the precipitate in sulfuric acid and adding hydrogen peroxide to precipitate uranium peroxdde.

  18. METHOD OF APPLYING COPPER COATINGS TO URANIUM

    DOE Patents [OSTI]

    Gray, A.G.

    1959-07-14

    A method is presented for protecting metallic uranium, which comprises anodic etching of the uranium in an aqueous phosphoric acid solution containing chloride ions, cleaning the etched uranium in aqueous nitric acid solution, promptly electro-plating the cleaned uranium in a copper electro-plating bath, and then electro-plating thereupon lead, tin, zinc, cadmium, chromium or nickel from an aqueous electro-plating bath.

  19. Statistical data of the uranium industry

    SciTech Connect (OSTI)

    1981-01-01

    Data are presented on US uranium reserves, potential resources, exploration, mining, drilling, milling, and other activities of the uranium industry through 1980. The compendium reflects the basic programs of the Grand Junction Office. Statistics are based primarily on information provided by the uranium exploration, mining, and milling companies. Data on commercial U/sub 3/O/sub 8/ sales and purchases are included. Data on non-US uranium production and resources are presented in the appendix. (DMC)

  20. Excess Uranium Management | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Excess Uranium Management Excess Uranium Management Request for Information - July 2016 On July 19, 2016, the Department issued a Request for Information on the effects of DOE transfers of excess uranium on domestic uranium mining, conversion, and enrichment industries. The Request for Information established an August 18, 2016 deadline for the submission of written comments. The Request for Information is available here. On August 1, 2016, the Department extended the comment period to September

  1. Uranium Management and Policy | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Uranium Management and Policy Uranium Management and Policy The Paducah Gaseous Diffusion Plant is located 3 miles south of the Ohio River and is 12 miles west of Paducah, Kentucky. Paducah remains the only operating gaseous diffusion uranium enrichment plant in the United States. The Paducah Gaseous Diffusion Plant is located 3 miles south of the Ohio River and is 12 miles west of Paducah, Kentucky. Paducah remains the only operating gaseous diffusion uranium enrichment plant in the United

  2. URANIUM BISMUTHIDE DISPERSION IN MOLTEN METAL

    DOE Patents [OSTI]

    Teitel, R.J.

    1959-10-27

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

  3. Uranium Downblending and Disposition Project Technology Readiness

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Assessment | Department of Energy Uranium Downblending and Disposition Project Technology Readiness Assessment Uranium Downblending and Disposition Project Technology Readiness Assessment Full Document and Summary Versions are available for download Uranium Downblending and Disposition Project Technology Readiness Assessment (1.11 MB) Summary - Uranium233 Downblending and Disposition Project (146.5 KB) More Documents & Publications Compilation of TRA Summaries EA-1574: Final

  4. Alternative applications of atomic vapor laser isotope separation technology

    SciTech Connect (OSTI)

    Not Available

    1991-01-01

    This report was commissioned by the Secretary of Energy. It summarizes the main features of atomic vapor laser isotope separation (AVLIS) technology and subsystems; evaluates applications, beyond those of uranium enrichment, suggested by Lawrence Livermore National Laboratory (LLNL) and a wide range of US industries and individuals; recommends further work on several applications; recommends the provision of facilities for evaluating potential new applications; and recommends the full involvement of end users from the very beginning in the development of any application. Specifically excluded from this report is an evaluation of the main AVLIS missions, uranium enrichment and purification of plutonium for weapons. In evaluating many of the alternative applications, it became clear that industry should play a greater and earlier role in the definition and development of technologies with the Department of Energy (DOE) if the nation is to derive significant commercial benefit. Applications of AVLIS to the separation of alternate (nonuranium) isotopes were considered. The use of {sup 157}Gd as burnable poison in the nuclear fuel cycle, the use {sup 12}C for isotopically pure diamond, and the use of plutonium isotopes for several nonweapons applications are examples of commercially useful products that might be produced at a cost less than the product value. Separations of other isotopes such as the elemental constituents of semiconductors were suggested; it is recommended that proposed applications be tested by using existing supplies to establish their value before more efficient enrichment processes are developed. Some applications are clear, but their production costs are too high, the window of opportunity in the market has passed, or societal constraints (e.g., on reprocessing of reactor fuel) discourage implementation.

  5. Domestic Uranium Production Report - Quarterly

    Gasoline and Diesel Fuel Update (EIA)

    2. Number of uranium mills and plants producing uranium concentrate in the United States Uranium concentrate processing facilities End of Mills - conventional milling 1 Mills - other operations 2 In-situ-leach plants 3 Byproduct recovery plants 4 Total 1996 0 2 5 2 9 1997 0 3 6 2 11 1998 0 2 6 1 9 1999 1 2 4 0 7 2000 1 2 3 0 6 2001 0 1 3 0 4 2002 0 1 2 0 3 2003 0 0 2 0 2 2004 0 0 3 0 3 2005 0 1 3 0 4 2006 0 1 5 0 6 2007 0 1 5 0 6 2008 1 0 6 0 7 2009 0 1 3 0 4 2010 1 0 4 0 5 2011 1 0 5 0 6 2012 1

  6. Uranium distribution and geology in the Fish Lake surficial uranium deposit, Esmeralda County, Nevada

    SciTech Connect (OSTI)

    Macke, D.L.; Schumann, R.R.; Otton, J.K.

    1990-01-01

    This paper reports on approximately 675 acres of uranium-enriched lacustrine and marsh sediments in Fish Lake Valley, in southern Nevada and California. Uranium concentrations from 253 samples averaged 64.3 ppm uranium, with a range from 6 to 800 ppm. Uranium was supplied to the marsh sediments by ground water derived from Tertiary volcanic rocks of the Silver Peak Range. Reconnaissance sampling in the surrounding areas shows minor enrichment of uranium in other wetland areas.

  7. PROCESS FOR SEPARATING URANIUM FISSION PRODUCTS

    DOE Patents [OSTI]

    Spedding, F.H.; Butler, T.A.; Johns, I.B.

    1959-03-10

    The removal of fission products such as strontium, barium, cesium, rubidium, or iodine from neutronirradiated uranium is described. Uranium halide or elemental halogen is added to melted irradiated uranium to convert the fission products to either more volatile compositions which vaporize from the melt or to higher melting point compositions which separate as solids.

  8. CATALYZED OXIDATION OF URANIUM IN CARBONATE SOLUTIONS

    DOE Patents [OSTI]

    Clifford, W.E.

    1962-05-29

    A process is given wherein carbonate solutions are employed to leach uranium from ores and the like containing lower valent uranium species by utilizing catalytic amounts of copper in the presence of ammonia therein and simultaneously supplying an oxidizing agent thereto. The catalysis accelerates rate of dissolution and increases recovery of uranium from the ore. (AEC)

  9. High strength uranium-tungsten alloy process

    DOE Patents [OSTI]

    Dunn, Paul S.; Sheinberg, Haskell; Hogan, Billy M.; Lewis, Homer D.; Dickinson, James M.

    1990-01-01

    Alloys of uranium and tungsten and a method for making the alloys. The amount of tungsten present in the alloys is from about 4 wt % to about 35 wt %. Tungsten particles are dispersed throughout the uranium and a small amount of tungsten is dissolved in the uranium.

  10. High strength uranium-tungsten alloys

    DOE Patents [OSTI]

    Dunn, Paul S.; Sheinberg, Haskell; Hogan, Billy M.; Lewis, Homer D.; Dickinson, James M.

    1991-01-01

    Alloys of uranium and tungsten and a method for making the alloys. The amount of tungsten present in the alloys is from about 4 wt % to about 35 wt %. Tungsten particles are dispersed throughout the uranium and a small amount of tungsten is dissolved in the uranium.

  11. Medical Isotope Production Analyses In KIPT Neutron Source Facility

    SciTech Connect (OSTI)

    Talamo, Alberto; Gohar, Yousry

    2016-01-01

    Medical isotope production analyses in Kharkov Institute of Physics and Technology (KIPT) neutron source facility were performed to include the details of the irradiation cassette and the self-shielding effect. An updated detailed model of the facility was used for the analyses. The facility consists of an accelerator-driven system (ADS), which has a subcritical assembly using low-enriched uranium fuel elements with a beryllium-graphite reflector. The beryllium assemblies of the reflector have the same outer geometry as the fuel elements, which permits loading the subcritical assembly with different number of fuel elements without impacting the reflector performance. The subcritical assembly is driven by an external neutron source generated from the interaction of 100-kW electron beam with a tungsten target. The facility construction was completed at the end of 2015, and it is planned to start the operation during the year of 2016. It is the first ADS in the world, which has a coolant system for removing the generated fission power. Argonne National Laboratory has developed the design concept and performed extensive design analyses for the facility including its utilization for the production of different radioactive medical isotopes. 99Mo is the parent isotope of 99mTc, which is the most commonly used medical radioactive isotope. Detailed analyses were performed to define the optimal sample irradiation location and the generated activity, for several radioactive medical isotopes, as a function of the irradiation time.

  12. HIGHLY ENRICHED URANIUM BLEND DOWN PROGRAM AT THE SAVANNAH RIVER SITE PRESENT AND FUTURE

    SciTech Connect (OSTI)

    Magoulas, V; Charles Goergen, C; Ronald Oprea, R

    2008-06-05

    The Department of Energy (DOE) and Tennessee Valley Authority (TVA) entered into an Interagency Agreement to transfer approximately 40 metric tons of highly enriched uranium (HEU) to TVA for conversion to fuel for the Browns Ferry Nuclear Power Plant. Savannah River Site (SRS) inventories included a significant amount of this material, which resulted from processing spent fuel and surplus materials. The HEU is blended with natural uranium (NU) to low enriched uranium (LEU) with a 4.95% 235U isotopic content and shipped as solution to the TVA vendor. The HEU Blend Down Project provided the upgrades needed to achieve the product throughput and purity required and provided loading facilities. The first blending to low enriched uranium (LEU) took place in March 2003 with the initial shipment to the TVA vendor in July 2003. The SRS Shipments have continued on a regular schedule without any major issues for the past 5 years and are due to complete in September 2008. The HEU Blend program is now looking to continue its success by dispositioning an additional approximately 21 MTU of HEU material as part of the SRS Enriched Uranium Disposition Project.

  13. The Next Generation Safeguards Initiative s High-Purity Uranium-233 Preservation Effort

    SciTech Connect (OSTI)

    Krichinsky, Alan M; Bostick, Debra A; Giaquinto, Joseph; Bayne, Charles; Goldberg, Dr. Steven A.; Humphrey, Dr. Marc; Hutcheon, Dr. Ian D.; Sobolev, Taissa

    2012-01-01

    High-purity 233U serves as a crucial reference material for accurately quantifying and characterizing uranium. The most accurate analytical results which can be obtained only with high-purity 233U certified reference material (CRM) are required when used to confirm compliance with international safeguards obligations and international nonproliferation agreements. The U.S. supply of 233U CRM is almost depleted, and existing domestic stocks of this synthetic isotope are scheduled to be down-blended for disposition with depleted uranium beginning in 2015. Down blending batches of high-purity 233U will permanently eliminate the value of this material as a CRM. Furthermore, no replacement 233U stocks are expected to be produced in the future due to a lack of operating production capability and the high cost of replacing such capability. Therefore, preserving select batches of high-purity 233U is of great value and will assist in retaining current analytical capabilities for uranium-bearing samples. Any organization placing a priority on accurate results of uranium analyses, or on the confirmation of trace uranium in environmental samples, has a vested interest in preserving this material. This paper describes the need for high-purity 233U, the consequences organizations and agencies face if this material is not preserved, and the progress and future plans for preserving select batches of the purest 233U materials from disposition. This work is supported by the Next Generation Safeguards Initiative, Office of Nonproliferation and International Security, National Nuclear Security Administration.

  14. Fate of soluble uranium in the I{sub 2}/KI leaching process for mercury removal

    SciTech Connect (OSTI)

    Bostick, W.D.; Davis, W.H.; Jarabek, R.J.

    1997-09-01

    General Electric Corporation has developed an extraction and recovery system for mercury, based upon the use of iodine (oxidant) and iodide ion (complexing agent). This system has been proposed for application to select mercury-contaminated mixed waste (i.e., waste containing radionuclides as well as other hazardous constituents), which have been generated by historic activities in support of US Department of Energy (DOE) missions. This system is compared to a system utilizing hypochlorite and chloride ions for removal of mercury and uranium from a sample of authentic mixed waste sludge. Relative to the hypochlorite (bleach) system, the iodine system mobilized more mercury and less uranium from the sludge. An engineering flowsheet has been developed to treat spent iodine-containing extraction medium, allowing the system to be recycled. The fate of soluble uranium in this series of treatment unit operations was monitored by tracing isotopically-enriched uranyl ion into simulated spent extraction medium. Treatment with use of elemental iron is shown to remove > 85% of the traced uranium while concurrently reducing excess iodine to the iodide ion. The next unit operation, adjustment of the solution pH to a value near 12 by the addition of lime slurry to form a metal-laden sludge phase (an operation referred to as lime-softening), removed an additional 57% of soluble uranium activity, for an over-all removal efficiency of {approximately} 96%. However, the precipitated solids did not settle well, and some iodide reagent is held up in the wet filtercake.

  15. Domestic Uranium Production Report - Quarterly

    Gasoline and Diesel Fuel Update (EIA)

    3. U.S. uranium mills and heap leach facilities by owner, location, capacity, and operating status Operating status at the end of Owner Mill and Heap Leach1 Facility name County, state (existing and planned locations) Capacity (short tons of ore per day) 2015 1st Quarter 2016 2nd quarter 2016 Anfield Resources Inc. Shootaring Canyon Uranium Mill Garfield, Utah 750 Standby Standby Standby EFR White Mesa LLC White Mesa Mill San Juan, Utah 2,000 Operating-Processing Alternate Feed

  16. PROCESS FOR PRODUCTION OF URANIUM

    DOE Patents [OSTI]

    Crawford, J.W.C.

    1959-09-29

    A process is described for the production of uranium by the autothermic reduction of an anhydrous uranium halide with an alkaline earth metal, preferably magnesium One feature is the initial reduction step which is brought about by locally bringing to reaction temperature a portion of a mixture of the reactants in an open reaction vessel having in contact with the mixture a lining of substantial thickness composed of calcium fluoride. The lining is prepared by coating the interior surface with a plastic mixture of calcium fluoride and water and subsequently heating the coating in situ until at last the exposed surface is substantially anhydrous.

  17. METHOD OF PROTECTIVELY COATING URANIUM

    DOE Patents [OSTI]

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

    1959-02-01

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

  18. Performance and safety parameters for the high flux isotope reactor

    SciTech Connect (OSTI)

    Ilas, G. [Oak Ridge National Laboratory, 1 Bethel Valley Road, Oak Ridge, TN 37831-6172 (United States); Primm III, T. [Oak Ridge National Laboratory, 1 Bethel Valley Road, Oak Ridge, TN 37831-6172 (United States); Primm Consulting, LLC, 945 Laurel Hill Road, Knoxville, TN 37923 (United States)

    2012-07-01

    A Monte Carlo depletion model for the High Flux Isotope Reactor (HFIR) Cycle 400 and its use in calculating parameters of relevance to the reactor performance and safety during the reactor cycle are presented in this paper. This depletion model was developed to serve as a reference for the design of a low-enriched uranium (LEU) fuel for an ongoing study to convert HFIR from high-enriched uranium (HEU) to LEU fuel; both HEU and LEU depletion models use the same methodology and ENDF/B-VII nuclear data as discussed in this paper. The calculated HFIR Cycle 400 parameters, which are compared with measurement data from critical experiments performed at HFIR, data included in the HFIR Safety Analysis Report (SAR), or data reported by previous calculations, provide a basis for verification or updating of the corresponding SAR data. (authors)

  19. Performance and Safety Parameters for the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Ilas, Germina [ORNL; Primm, Trent [Primm Consulting, LLC

    2012-01-01

    A Monte Carlo depletion model for the High Flux Isotope Reactor (HFIR) Cycle 400 and its use in calculating parameters of relevance to the reactor performance and safety during the reactor cycle are presented in this paper. This depletion model was developed to serve as a reference for the design of a low-enriched uranium (LEU) fuel for an ongoing study to convert HFIR from high-enriched uranium (HEU) to LEU fuel; both HEU and LEU depletion models use the same methodology and ENDV/B-VII nuclear data as discussed in this paper. The calculated HFIR Cycle 400 parameters, which are compared when available with measurement data from critical experiments performed at HFIR, data included in the HFIR Safety Analysis Report (SAR), or data reported by previous calculations, provide a basis for verification or updating of the corresponding SAR data.

  20. Science on Tap - Isotopes

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Science on Tap - Isotopes Science on Tap - Isotopes WHEN: Jun 16, 2016 5:30 PM - 7:00 PM WHERE: UnQuarked Wine Room 145 Central Park Square, Los Alamos, New Mexico 87544 USA CONTACT: Linda Anderman (505) 665-9196 CATEGORY: Bradbury INTERNAL: Calendar Login Event Description Short presentation followed by lively interaction on the topic at hand. While isotopes are chemical elements (think periodic table), their varying numbers of neutrons mean they can be used in lots of different way. Join us

  1. Electron Backscatter Diffraction (EBSD) Characterization of Uranium and Uranium Alloys

    SciTech Connect (OSTI)

    McCabe, Rodney J.; Kelly, Ann Marie; Clarke, Amy J.; Field, Robert D.; Wenk, H. R.

    2012-07-25

    Electron backscatter diffraction (EBSD) was used to examine the microstructures of unalloyed uranium, U-6Nb, U-10Mo, and U-0.75Ti. For unalloyed uranium, we used EBSD to examine the effects of various processes on microstructures including casting, rolling and forming, recrystallization, welding, and quasi-static and shock deformation. For U-6Nb we used EBSD to examine the microstructural evolution during shape memory loading. EBSD was used to study chemical homogenization in U-10Mo, and for U-0.75Ti, we used EBSD to study the microstructure and texture evolution during thermal cycling and deformation. The studied uranium alloys have significant microstructural and chemical differences and each of these alloys presents unique preparation challenges. Each of the alloys is prepared by a sequence of mechanical grinding and polishing followed by electropolishing with subtle differences between the alloys. U-6Nb and U-0.75Ti both have martensitic microstructures and both require special care in order to avoid mechanical polishing artifacts. Unalloyed uranium has a tendency to rapidly oxidize when exposed to air and a two-step electropolish is employed, the first step to remove the damaged surface layer resulting from the mechanical preparation and the second step to passivate the surface. All of the alloying additions provide a level of surface passivation and different one and two step electropolishes are employed to create good EBSD surfaces. Because of its low symmetry crystal structure, uranium exhibits complex deformation behavior including operation of multiple deformation twinning modes. EBSD was used to observe and quantify twinning contributions to deformation and to examine the fracture behavior. Figure 1 shows a cross section of two mating fracture surfaces in cast uranium showing the propensity of deformation twinning and intergranular fracture largely between dissimilarly oriented grains. Deformation of U-6Nb in the shape memory regime occurs by the motion

  2. Qualification and initial characterization of a high-purity 233U spike for use in uranium analyses

    SciTech Connect (OSTI)

    Mathew, K. J.; Canaan, R. D.; Hexel, C.; Giaquinto, J.; Krichinsky, A. M.

    2015-08-20

    Several high-purity 233U items potentially useful as isotope dilution mass spectrometry standards for safeguards, non-proliferation, and nuclear forensics measurements are identified and rescued from downblending. By preserving the supply of 233U materials of different pedigree for use as source materials for certified reference materials (CRMs), it is ensured that the safeguards community has high quality uranium isotopic standards required for calibration of the analytical instruments. One of the items identified as a source material for a high-purity CRM is characterized for the uranium isotope-amount ratios using thermal ionization mass spectrometry (TIMS). Additional verification measurements on this material using quadrupole inductively coupled plasma mass spectrometry (ICPMS) are also performed. As a result, the comparison of the ICPMS uranium isotope-amount ratios with the TIMS data, with much smaller uncertainties, validated the ICPMS measurement practices. ICPMS is proposed for the initial screening of the purity of items in the rescue campaign.

  3. Qualification and initial characterization of a high-purity 233U spike for use in uranium analyses

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Mathew, K. J.; Canaan, R. D.; Hexel, C.; Giaquinto, J.; Krichinsky, A. M.

    2015-08-20

    Several high-purity 233U items potentially useful as isotope dilution mass spectrometry standards for safeguards, non-proliferation, and nuclear forensics measurements are identified and rescued from downblending. By preserving the supply of 233U materials of different pedigree for use as source materials for certified reference materials (CRMs), it is ensured that the safeguards community has high quality uranium isotopic standards required for calibration of the analytical instruments. One of the items identified as a source material for a high-purity CRM is characterized for the uranium isotope-amount ratios using thermal ionization mass spectrometry (TIMS). Additional verification measurements on this material using quadrupole inductivelymore » coupled plasma mass spectrometry (ICPMS) are also performed. As a result, the comparison of the ICPMS uranium isotope-amount ratios with the TIMS data, with much smaller uncertainties, validated the ICPMS measurement practices. ICPMS is proposed for the initial screening of the purity of items in the rescue campaign.« less

  4. Atomic vapor laser isotope separation at Lawrence Livermore National Laboratory: a status report

    SciTech Connect (OSTI)

    Davis, J.I.

    1980-01-01

    The field of laser induced chemistry began in earnest early in the 1970's with the initiation of major efforts in laser isotope separation (LIS) of uranium. Though many specialized, small-scale photochemical and diagnostic applications have been identified and evaluated experimentally, and continue to show promise, currently the only high payoff, large-scale applications remain LIS of special elements. Aspects of the physical scaling, technology status and economic basis of uranium LIS are examined with special emphasis on the effort at LLNL.

  5. Domestic Uranium Production Report - Energy Information Administration

    U.S. Energy Information Administration (EIA) Indexed Site

    Domestic Uranium Production Report - Annual With Data for 2015 | Release Date: May 5, 2016 | Next Release Date: May 2017 | full report Previous domestic uranium production reports Year: 2014 2013 2012 2011 2010 2009 2008 2007 2006 2005 2004 Go Drilling Total uranium drilling was 1,518 holes covering 0.9 million feet, 13% fewer holes than in 2015. Expenditures for uranium drilling in the United States were $29 million in 2015, an increase of 2% compared with 2014. Figure 1. U.S. Uranium drilling

  6. Development of pulsed neutron uranium logging instrument

    SciTech Connect (OSTI)

    Wang, Xin-guang; Liu, Dan; Zhang, Feng

    2015-03-15

    This article introduces a development of pulsed neutron uranium logging instrument. By analyzing the temporal distribution of epithermal neutrons generated from the thermal fission of {sup 235}U, we propose a new method with a uranium-bearing index to calculate the uranium content in the formation. An instrument employing a D-T neutron generator and two epithermal neutron detectors has been developed. The logging response is studied using Monte Carlo simulation and experiments in calibration wells. The simulation and experimental results show that the uranium-bearing index is linearly correlated with the uranium content, and the porosity and thermal neutron lifetime of the formation can be acquired simultaneously.

  7. Process for alloying uranium and niobium

    DOE Patents [OSTI]

    Holcombe, Cressie E.; Northcutt, Jr., Walter G.; Masters, David R.; Chapman, Lloyd R.

    1991-01-01

    Alloys such as U-6Nb are prepared by forming a stacked sandwich array of uraniun sheets and niobium powder disposed in layers between the sheets, heating the array in a vacuum induction melting furnace to a temperature such as to melt the uranium, holding the resulting mixture at a temperature above the melting point of uranium until the niobium dissolves in the uranium, and casting the uranium-niobium solution. Compositional uniformity in the alloy product is enabled by use of the sandwich structure of uranium sheets and niobium powder.

  8. Removal of uranium from aqueous HF solutions

    DOE Patents [OSTI]

    Pulley, Howard; Seltzer, Steven F.

    1980-01-01

    This invention is a simple and effective method for removing uranium from aqueous HF solutions containing trace quantities of the same. The method comprises contacting the solution with particulate calcium fluoride to form uranium-bearing particulates, permitting the particulates to settle, and separting the solution from the settled particulates. The CaF.sub.2 is selected to have a nitrogen surface area in a selected range and is employed in an amount providing a calcium fluoride/uranium weight ratio in a selected range. As applied to dilute HF solutions containing 120 ppm uranium, the method removes at least 92% of the uranium, without introducing contaminants to the product solution.

  9. Method for producing uranium atomic beam source

    DOE Patents [OSTI]

    Krikorian, Oscar H.

    1976-06-15

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

  10. Benchmark of SCALE (SAS2H) isotopic predictions of depletion analyses for San Onofre PWR MOX fuel

    SciTech Connect (OSTI)

    Hermann, O.W.

    2000-02-01

    The isotopic composition of mixed-oxide (MOX) fuel, fabricated with both uranium and plutonium, after discharge from reactors is of significant interest to the Fissile Materials Disposition Program. The validation of the SCALE (SAS2H) depletion code for use in the prediction of isotopic compositions of MOX fuel, similar to previous validation studies on uranium-only fueled reactors, has corresponding significance. The EEI-Westinghouse Plutonium Recycle Demonstration Program examined the use of MOX fuel in the San Onofre PWR, Unit 1, during cycles 2 and 3. Isotopic analyses of the MOX spent fuel were conducted on 13 actinides and {sup 148}Nd by either mass or alpha spectrometry. Six fuel pellet samples were taken from four different fuel pins of an irradiated MOX assembly. The measured actinide inventories from those samples has been used to benchmark SAS2H for MOX fuel applications. The average percentage differences in the code results compared with the measurement were {minus}0.9% for {sup 235}U and 5.2% for {sup 239}Pu. The differences for most of the isotopes were significantly larger than in the cases for uranium-only fueled reactors. In general, comparisons of code results with alpha spectrometer data had extreme differences, although the differences in the calculations compared with mass spectrometer analyses were not extremely larger than that of uranium-only fueled reactors. This benchmark study should be useful in estimating uncertainties of inventory, criticality and dose calculations of MOX spent fuel.

  11. EA-1929: NorthStar Medical Technologies LLC, Commercial Domestic Production of the Medical Isotope Molybdenum-99

    Broader source: Energy.gov [DOE]

    This EA evaluates the potential environmental impacts of a proposal to use federal funds to support and accelerate Northstar Medical Radioisotopes' project to develop domestic, commercial production capability for the medical isotope Molybdenum-99 without the use of highly enriched uranium.

  12. Plasma isotope separation methods

    SciTech Connect (OSTI)

    Grossman, M.W. ); Shepp, T.A. )

    1991-12-01

    Isotope separation has many important industrial, medical, and research applications. Large-scale processes have typically utilized complex cascade systems; for example, the gas centrifuge. Alternatively, high single-stage enrichment processes (as in the case of the calutron) are very energy intensive. Plasma-based methods being developed for the past 15 to 20 years have attempted to overcome these two drawbacks. In this review, six major types of isotope separation methods which involve plasma phenomena are discussed. These methods are: plasma centrifuge, AVLIS (atomic vapor laser isotope separation), ion wave, ICR (ion-cyclotron resonance), calutron, and gas discharge. The emphasis of this paper is to describe the plasma phenomena in these major categories. An attempt was made to include enough references so that more detailed study or evaluation of a particular method could readily be pursued. A brief discussion of isotope separation using mass balance concepts is also carried out.

  13. METHOD OF PURIFYING URANIUM METAL

    DOE Patents [OSTI]

    Blanco, R.E.; Morrison, B.H.

    1958-12-23

    The removal of lmpurities from uranlum metal can be done by a process conslstlng of contacting the metal with liquid mercury at 300 icient laborato C, separating the impunitycontalnlng slag formed, cooling the slag-free liquld substantlally below the point at which uranlum mercurlde sollds form, removlng the mercury from the solids, and recovering metallic uranium by heating the solids.

  14. 2015 Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

    Deliveries 2011 2012 2013 2014 2015 Purchases of U.S.-origin and foreign-origin uranium 550 W W W 1,455 Weighted-average price 58.12 W W W 52.35 Purchases of U.S.-origin and ...

  15. 2015 Uranium Marketing Annual Report

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    7 2015 Uranium Marketing Annual Report Release Date: May 24, 2016 Next Release Date: May ... Foreign purchases 19,318 20,196 23,233 24,199 27,233 Weighted-average price 48.80 46.80 ...

  16. 2015 Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

    9 2015 Uranium Marketing Annual Report Release Date: May 24, 2016 Next Release Date: May ... Foreign sales 4,387 4,798 4,148 4,210 4,258 Weighted-average price 53.08 47.53 43.10 32.91 ...

  17. SEPARATION OF PLUTONIUM FROM URANIUM

    DOE Patents [OSTI]

    Feder, H.M.; Nuttall, R.L.

    1959-12-15

    A process is described for extracting plutonium from powdered neutron- irradiated urarium metal by contacting the latter, while maintaining it in the solid form, with molten magnesium which takes up the plutonium and separating the molten magnesium from the solid uranium.

  18. GRAIN REFINEMENT OF URANIUM BILLETS

    DOE Patents [OSTI]

    Lewis, L.

    1964-02-25

    A method of refining the grain structure of massive uranium billets without resort to forging is described. The method consists in the steps of beta- quenching the billets, annealing the quenched billets in the upper alpha temperature range, and extrusion upset of the billets to an extent sufficient to increase the cross sectional area by at least 5 per cent. (AEC)

  19. 2015 Domestic Uranium Production Report

    Annual Energy Outlook [U.S. Energy Information Administration (EIA)]

    11 2015 Domestic Uranium Production Report Release Date: May 5, 2016 Next Release Date: May 2017 Total Land and Other 2003 W W 31.3 NA NA NA W 2004 10.6 27.8 48.4 NA NA NA 86.9 ...

  20. Sandia National Laboratories Medical Isotope Reactor concept.

    SciTech Connect (OSTI)

    Coats, Richard Lee; Dahl, James J.; Parma, Edward J., Jr.

    2010-04-01

    This report describes the Sandia National Laboratories Medical Isotope Reactor and hot cell facility concepts. The reactor proposed is designed to be capable of producing 100% of the U.S. demand for the medical isotope {sup 99}Mo. The concept is novel in that the fuel for the reactor and the targets for the {sup 99}Mo production are the same. There is no driver core required. The fuel pins that are in the reactor core are processed on a 7 to 21 day irradiation cycle. The fuel is low enriched uranium oxide enriched to less than 20% {sup 235}U. The fuel pins are approximately 1 cm in diameter and 30 to 40 cm in height, clad with Zircaloy (zirconium alloy). Approximately 90 to 150 fuel pins are arranged in the core in a water pool {approx}30 ft deep. The reactor power level is 1 to 2 MW. The reactor concept is a simple design that is passively safe and maintains negative reactivity coefficients. The total radionuclide inventory in the reactor core is minimized since the fuel/target pins are removed and processed after 7 to 21 days. The fuel fabrication, reactor design and operation, and {sup 99}Mo production processing use well-developed technologies that minimize the technological and licensing risks. There are no impediments that prevent this type of reactor, along with its collocated hot cell facility, from being designed, fabricated, and licensed today.

  1. Reducing Emissions from Uranium Dissolving

    SciTech Connect (OSTI)

    Griffith, W.L.

    1992-01-01

    This study was designed to assess the feasibility of decreasing NO{sub x} emissions from the current uranium alloy scrap tray dissolving facility. In the current process, uranium scrap is dissolved in boiling nitric acid in shallow stainless-steel trays. As scrap dissolves, more metal and more nitric acid are added to the tray by operating personnel. Safe geometry is assured by keeping liquid level at or below 5 cm, the depth of a safe infinite slab. The accountability batch control system provides additional protection against criticality. The trays are steam coil heated. The process has operated satisfactorily, with few difficulties, for decades. Both uranium and uranium alloys are dissolved. Nitric acid is recovered from the vapors for reuse. Metal nitrates are sent to uranium recovery. Brown NO{sub x} fumes evolved during dissolving have occasionally resulted in a visible plume from the trays. The fuming is most noticeable during startup and after addition of fresh acid to a tray. Present environmental regulations are expected to require control of brown NO{sub x} emissions. Because NO{sub x} is hazardous, fumes should be suppressed whenever the electric blower system is inoperable. Because the tray dissolving process has worked well for decades, as much of the current capital equipment and operating procedures as possible were preserved. A detailed review of the literature, indicated the feasibility of slightly altering process chemistry to favor the production of NO{sub 2}, which can be scrubbed and recycled as nitric acid. Methods for controlling the process to manage offgas product distribution and to minimize chemical reaction hazards were also considered.

  2. Comments on proposed legislation to restructure DOE's uranium enrichment program

    SciTech Connect (OSTI)

    Not Available

    1991-04-01

    This book focuses on H.R.145, H.R.788, and S.210. Each of the proposed bills would restructure DOE's enrichment program as a government corporation with private financing and would encourage the eventual sale of the corporation to the private sector. In doing so, the bills would, among other things, allow the corporation to set prices to maximize long-term returns; establish a fund to meet the costs of decontamination, decommissioning, and other environmental cleanup costs associated with uranium enrichment activities; transfer interest in DOE's new atomic vapor laser isotope separation (AVLIS) process to the new corporation; and, except for H.R. 145, require the government to pay its share of the costs to clean up mill tailings (mining wastes) generated under government contracts.

  3. Inherently safe in situ uranium recovery.

    SciTech Connect (OSTI)

    Krumhansl, James Lee; Beauheim, Richard Louis; Brady, Patrick Vane; Arnold, Bill Walter; Kanney, Joseph F.; McKenna, Sean Andrew

    2009-05-01

    Expansion of uranium mining in the United States is a concern to some environmental groups and sovereign Native American Nations. An approach which may alleviate some problems is to develop inherently safe in situ uranium recovery ('ISR') technologies. Current ISR technology relies on chemical extraction of trace levels of uranium from aquifers that, once mined, can still contain dissolved uranium and other trace metals that are a health concern. Existing ISR operations are few in number; however, high uranium prices are driving the industry to consider expanding operations nation-wide. Environmental concerns and enforcement of the new 30 ppb uranium drinking water standard may make opening new mining operations more difficult and costly. Here we propose a technological fix: the development of inherently safe in situ recovery (ISISR) methods. The four central features of an ISISR approach are: (1) New 'green' leachants that break down predictably in the subsurface, leaving uranium, and associated trace metals, in an immobile form; (2) Post-leachant uranium/metals-immobilizing washes that provide a backup decontamination process; (3) An optimized well-field design that increases uranium recovery efficiency and minimizes excursions of contaminated water; and (4) A combined hydrologic/geochemical protocol for designing low-cost post-extraction long-term monitoring. ISISR would bring larger amounts of uranium to the surface, leave fewer toxic metals in the aquifer, and cost less to monitor safely - thus providing a 'win-win-win' solution to all stakeholders.

  4. Reports on investigations of uranium anomalies. National Uranium Resource Evaluation

    SciTech Connect (OSTI)

    Goodknight, C.S.; Burger, J.A.

    1982-10-01

    During the National Uranium Resource Evaluation (NURE) program, conducted for the US Department of Energy (DOE) by Bendix Field Engineering Corporation (BFEC), radiometric and geochemical surveys and geologic investigations detected anomalies indicative of possible uranium enrichment. Data from the Aerial Radiometric and Magnetic Survey (ARMS) and the Hydrogeochemical and Stream-Sediment Reconnaissance (HSSR), both of which were conducted on a national scale, yielded numerous anomalies that may signal areas favorable for the occurrence of uranium deposits. Results from geologic evaluations of individual 1/sup 0/ x 2/sup 0/ quadrangles for the NURE program also yielded anomalies, which could not be adequately checked during scheduled field work. Included in this volume are individual reports of field investigations for the following six areas which were shown on the basis of ARMS, HSSR, and (or) geologic data to be anomalous: (1) Hylas zone and northern Richmond basin, Virginia; (2) Sischu Creek area, Alaska; (3) Goodman-Dunbar area, Wisconsin; (4) McCaslin syncline, Wisconsin; (5) Mt. Withington Cauldron, Socorro County, New Mexico; (6) Lake Tecopa, Inyo County, California. Field checks were conducted in each case to verify an indicated anomalous condition and to determine the nature of materials causing the anomaly. The ultimate objective of work is to determine whether favorable conditions exist for the occurrence of uranium deposits in areas that either had not been previously evaluated or were evaluated before data from recent surveys were available. Most field checks were of short duration (2 to 5 days). The work was done by various investigators using different procedures, which accounts for variations in format in their reports. All papers have been abstracted and indexed.

  5. Transportation of medical isotopes

    SciTech Connect (OSTI)

    Nielsen, D.L.

    1997-11-19

    A Draft Technical Information Document (HNF-1855) is being prepared to evaluate proposed interim tritium and medical isotope production at the Fast Flux Test Facility (FFTF). This assessment examines the potential health and safety impacts of transportation operations associated with the production of medical isotopes. Incident-free and accidental impacts are assessed using bounding source terms for the shipment of nonradiological target materials to the Hanford Site, the shipment of irradiated targets from the FFTF to the 325 Building, and the shipment of medical isotope products from the 325 Building to medical distributors. The health and safety consequences to workers and the public from the incident-free transportation of targets and isotope products would be within acceptable levels. For transportation accidents, risks to works and the public also would be within acceptable levels. This assessment is based on best information available at this time. As the medical isotope program matures, this analysis will be revised, if necessary, to support development of a final revision to the Technical Information Document.

  6. Separation of sulfur isotopes

    DOE Patents [OSTI]

    DeWitt, Robert; Jepson, Bernhart E.; Schwind, Roger A.

    1976-06-22

    Sulfur isotopes are continuously separated and enriched using a closed loop reflux system wherein sulfur dioxide (SO.sub.2) is reacted with sodium hydroxide (NaOH) or the like to form sodium hydrogen sulfite (NaHSO.sub.3). Heavier sulfur isotopes are preferentially attracted to the NaHSO.sub.3, and subsequently reacted with sulfuric acid (H.sub.2 SO.sub.4) forming sodium hydrogen sulfate (NaHSO.sub.4) and SO.sub.2 gas which contains increased concentrations of the heavier sulfur isotopes. This heavy isotope enriched SO.sub.2 gas is subsequently separated and the NaHSO.sub.4 is reacted with NaOH to form sodium sulfate (Na.sub.2 SO.sub.4) which is subsequently decomposed in an electrodialysis unit to form the NaOH and H.sub.2 SO.sub.4 components which are used in the aforesaid reactions thereby effecting sulfur isotope separation and enrichment without objectionable loss of feed materials.

  7. Determination of initial fuel state and number of reactor shutdowns in archived low-burnup uranium targets

    SciTech Connect (OSTI)

    Byerly, Benjamin; Tandon, Lav; Hayes-Sterbenz, Anna; Martinez, Patrick; Keller, Russ; Stanley, Floyd; Spencer, Khalil; Thomas, Mariam; Xu, Ning; Schappert, Michael; Fulwyler, James

    2015-10-26

    This article presents a method for destructive analysis of irradiated uranium (U) targets, with a focus on collection and measurement of long-lived (t1/2 > ~10 years) and stable fission product isotopes of ruthenium and cesium. Long-lived and stable isotopes of these elements can provide information on reactor conditions (e.g. flux, irradiation time, cooling time) in old samples (> 5–10 years) whose short-lived fission products have decayed away. The separation and analytical procedures were tested on archived U reactor targets at Los Alamos National Laboratory as part of an effort to evaluate reactor models at low-burnup.

  8. Determination of initial fuel state and number of reactor shutdowns in archived low-burnup uranium targets

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Byerly, Benjamin; Tandon, Lav; Hayes-Sterbenz, Anna; Martinez, Patrick; Keller, Russ; Stanley, Floyd; Spencer, Khalil; Thomas, Mariam; Xu, Ning; Schappert, Michael; et al

    2015-10-26

    This article presents a method for destructive analysis of irradiated uranium (U) targets, with a focus on collection and measurement of long-lived (t1/2 > ~10 years) and stable fission product isotopes of ruthenium and cesium. Long-lived and stable isotopes of these elements can provide information on reactor conditions (e.g. flux, irradiation time, cooling time) in old samples (> 5–10 years) whose short-lived fission products have decayed away. The separation and analytical procedures were tested on archived U reactor targets at Los Alamos National Laboratory as part of an effort to evaluate reactor models at low-burnup.

  9. Hybrid interferometric/dispersive atomic spectroscopy of laser-induced uranium plasma

    SciTech Connect (OSTI)

    Morgan, Phyllis K.; Scott, Jill R.; Jovanovic, Igor

    2015-12-19

    An established optical emission spectroscopy technique, laser-induced breakdown spectroscopy (LIBS), holds promise for detection and rapid analysis of elements relevant for nuclear safeguards, nonproliferation, and nuclear power, including the measurement of isotope ratios. One such important application of LIBS is the measurement of uranium enrichment (235U/238U), which requires high spectral resolution (e.g., 25 pm for the 424.4 nm U II line). High-resolution dispersive spectrometers necessary for such measurements are typically bulky and expensive. We demonstrate the use of an alternative measurement approach, which is based on an inexpensive and compact Fabry–Perot etalon integrated with a low to moderate resolution Czerny–Turner spectrometer, to achieve the resolution needed for isotope selectivity of LIBS of uranium in ambient air. Furthermore, spectral line widths of ~ 10 pm have been measured at a center wavelength 424.437 nm, clearly discriminating the natural from the highly enriched uranium.

  10. Hybrid interferometric/dispersive atomic spectroscopy of laser-induced uranium plasma

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Morgan, Phyllis K.; Scott, Jill R.; Jovanovic, Igor

    2015-12-19

    An established optical emission spectroscopy technique, laser-induced breakdown spectroscopy (LIBS), holds promise for detection and rapid analysis of elements relevant for nuclear safeguards, nonproliferation, and nuclear power, including the measurement of isotope ratios. One such important application of LIBS is the measurement of uranium enrichment (235U/238U), which requires high spectral resolution (e.g., 25 pm for the 424.4 nm U II line). High-resolution dispersive spectrometers necessary for such measurements are typically bulky and expensive. We demonstrate the use of an alternative measurement approach, which is based on an inexpensive and compact Fabry–Perot etalon integrated with a low to moderate resolution Czerny–Turnermore » spectrometer, to achieve the resolution needed for isotope selectivity of LIBS of uranium in ambient air. Furthermore, spectral line widths of ~ 10 pm have been measured at a center wavelength 424.437 nm, clearly discriminating the natural from the highly enriched uranium.« less

  11. 230Th-234U Model-Ages of Some Uranium Standard Reference Materials

    SciTech Connect (OSTI)

    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.

  12. Depleted uranium disposal options evaluation

    SciTech Connect (OSTI)

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

    1994-05-01

    The Department of Energy (DOE), Office of Environmental Restoration and Waste Management, has chartered a study to evaluate alternative management strategies for depleted uranium (DU) currently stored throughout the DOE complex. Historically, DU has been maintained as a strategic resource because of uses for DU metal and potential uses for further enrichment or for uranium oxide as breeder reactor blanket fuel. This study has focused on evaluating the disposal options for DU if it were considered a waste. This report is in no way declaring these DU reserves a ``waste,`` but is intended to provide baseline data for comparison with other management options for use of DU. To PICS considered in this report include: Retrievable disposal; permanent disposal; health hazards; radiation toxicity and chemical toxicity.

  13. Uranium Metal Analysis via Selective Dissolution

    SciTech Connect (OSTI)

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

    2008-09-10

    Uranium metal, which is present in sludge held in the Hanford Site K West Basin, can create hazardous hydrogen atmospheres during sludge handling, immobilization, or subsequent transport and storage operations by its oxidation/corrosion in water. A thorough knowledge of the uranium metal concentration in sludge therefore is essential to successful sludge management and waste process design. The goal of this work was to establish a rapid routine analytical method to determine uranium metal concentrations as low as 0.03 wt% in sludge even in the presence of up to 1000-fold higher total uranium concentrations (i.e., up to 30 wt% and more uranium) for samples to be taken during the upcoming sludge characterization campaign and in future analyses for sludge handling and processing. This report describes the experiments and results obtained in developing the selective dissolution technique to determine uranium metal concentration in K Basin sludge.

  14. PRETREATING URANIUM FOR METAL PLATING

    DOE Patents [OSTI]

    Wehrmann, R.F.

    1961-05-01

    A process is given for anodically treating the surface of uranium articles, prior to metal plating. The metal is electrolyzed in an aqueous solution of about 10% polycarboxylic acid, preferably oxalic acid, from 1 to 5% by weight of glycerine and from 1 to 5% by weight of hydrochloric acid at from 20 to 75 deg C for from 30 seconds to 15 minutes. A current density of from 60 to 100 amperes per square foot is used.

  15. :- : DRILLING URANIUM BILLETS ON A

    Office of Legacy Management (LM)

    'Xxy";^ ...... ' '. .- -- Metals, Ceramics, and Materials. : . - ,.. ; - . _ : , , ' z . , -, .- . >. ; . .. :- : DRILLING URANIUM BILLETS ON A .-... r .. .. i ' LEBLOND-CARLSTEDT RAPID BORER 4 r . _.i'- ' ...... ' -'".. :-'' ,' :... : , '.- ' ;BY R.' J. ' ANSEN .AEC RESEARCH AND DEVELOPMENT REPORT PERSONAL PROPERTY OF J. F. Schlltz .:- DECLASSIFIED - PER AUTHORITY OF (DAlE) (NhTI L (DATE)UE) FEED MATERIALS PRODUCTION CENTER NATIONAL LFE A COMPANY OF OHIO 26 1 3967 3035406 NLCO -

  16. Isotope separation apparatus

    DOE Patents [OSTI]

    Arnush, Donald; MacKenzie, Kenneth R.; Wuerker, Ralph F.

    1980-01-01

    Isotope separation apparatus consisting of a plurality of cells disposed adjacent to each other in an evacuated container. A common magnetic field is established extending through all of the cells. A source of energetic electrons at one end of the container generates electrons which pass through the cells along the magnetic field lines. Each cell includes an array of collector plates arranged in parallel or in tandem within a common magnetic field. Sets of collector plates are disposed adjacent to each other in each cell. Means are provided for differentially energizing ions of a desired isotope by applying energy at the cyclotron resonant frequency of the desired isotope. As a result, the energized desired ions are preferentially collected by the collector plates.

  17. REMOVAL OF URANIUM FROM ORGANIC LIQUIDS

    DOE Patents [OSTI]

    Vavalides, S.P.

    1959-08-25

    A process is described for recovering small quantities of uranium from organic liquids such as hydrocarbon oils. halogen-substituted hydrocarbons, and alcohols. The organic liquid is contacted with a comminuted alkaline earth hydroxide, calcium hydroxide particularly, and the resulting uranium-bearing solid is separated from the liquid by filtration. Uranium may then be recovered from the solid by means of dissolution in nitric acid and conventional extraction with an organic solvent such as tributyl phosphate.

  18. Uranium Weapons Components Successfully Dismantled | National Nuclear

    National Nuclear Security Administration (NNSA)

    Security Administration | (NNSA) Uranium Weapons Components Successfully Dismantled Uranium Weapons Components Successfully Dismantled Oak Ridge, TN Continuing its efforts to reduce the size of the U.S. nuclear weapons stockpile, the National Nuclear Security Administration announced that uranium components from two major nuclear weapons systems formerly deployed on U.S. Air Force missiles and aircraft have been dismantled at the Y-12 National Security Complex in Oak Ridge, TN. Y-12 workers

  19. Uranium Leasing Program | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    » Uranium Leasing Program Uranium Leasing Program Reclaimed C-CM-25 Mine Site, Montrose County, Colorado Reclaimed C-CM-25 Mine Site, Montrose County, Colorado LM currently manages the Uranium Leasing Program and continues to administer 31 lease tracts, all located within the Uravan Mineral Belt in southwestern Colorado. Twenty-nine of these lease tracts are actively held under lease and two tracts have been placed in inactive status indefinitely. Administrative duties include ongoing

  20. METHOD FOR THE REDUCTION OF URANIUM COMPOUNDS

    DOE Patents [OSTI]

    Cooke, W.H.; Crawford, J.W.C.

    1959-05-12

    An improved technique of preparing massive metallic uranium by the reaction at elevated temperature between an excess of alkali in alkaline earth metal and a uranium halide, such ss uranium tetrafluoride is presented. The improvement comprises employing a reducing atmosphere of hydrogen or the like, such as coal gas, in the vessel during the reduction stage and then replacing the reducing atmosphere with argon gas prior to cooling to ambient temperature.

  1. ELECTROCHEMICAL DECONTAMINATION AND RECOVERY OF URANIUM VALUES

    DOE Patents [OSTI]

    McLaren, J.A.; Goode, J.H.

    1958-05-13

    An electrochemical process is described for separating uranium from fission products. The method comprises subjecting the mass of uranium to anodic dissolution in an electrolytic cell containing aqueous alkali bicarbonate solution as its electrolyte, thereby promoting a settling from the solution of a solid sludge from about the electrodes and separating the resulting electrolyte solution containing the anodically dissolved uranium from the sludge which contains the rare earth fission products.

  2. ELECTROLYTIC CLADDING OF ZIRCONIUM ON URANIUM

    DOE Patents [OSTI]

    Wick, J.J.

    1959-09-22

    A method is presented for coating uranium with zircoalum by rendering the uranium surface smooth and oxidefree, immersing it in a molten electrolytic bath in NaCI, K/sub 2/ZrF/sub 6/, KF, and ZrO/sub 2/, and before the article reaches temperature equilibrium with the bath, applying an electrolyzing current of 60 amperes per square dectmeter at approximately 3 volts to form a layer of zirconium metal on the uranium.

  3. METHOD OF PRODUCING URANIUM METAL BY ELECTROLYSIS

    DOE Patents [OSTI]

    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)

  4. Think Uranium. Think Y-12 | Y-12 National Security Complex

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Think Uranium. Think Y-12 Think Uranium. Think Y-12 Posted: July 22, 2013 - 3:12pm | Y-12 Report | Volume 10, Issue 1 | 2013 Uranium fever: Much like the California gold rush of ...

  5. Uranium metal reactions with hydrogen and water vapour and the reactivity of the uranium hydride produced

    SciTech Connect (OSTI)

    Godfrey, H.; Broan, C.; Goddard, D.; Hodge, N.; Woodhouse, G.; Diggle, A.; Orr, R.

    2013-07-01

    Within the nuclear industry, metallic uranium has been used as a fuel. If this metal is stored in a hydrogen rich environment then the uranium metal can react with the hydrogen to form uranium hydride which can be pyrophoric when exposed to air. The UK National Nuclear Laboratory has been carrying out a programme of research for Sellafield Limited to investigate the conditions required for the formation and persistence of uranium hydride and the reactivity of the material formed. The experimental results presented here have described new results characterising uranium hydride formed from bulk uranium at 50 and 160 C. degrees and measurements of the hydrolysis kinetics of these materials in liquid water. It has been shown that there is an increase in the proportion of alpha-uranium hydride in material formed at lower temperatures and that there is an increase in the rate of reaction with water of uranium hydride formed at lower temperatures. This may at least in part be attributable to a difference in the reaction rate between alpha and beta-uranium hydride. A striking observation is the strong dependence of the hydrolysis reaction rate on the temperature of preparation of the uranium hydride. For example, the reaction rate of uranium hydride prepared at 50 C. degrees was over ten times higher than that prepared at 160 C. degrees at 20% extent of reaction. The decrease in reaction rate with the extent of reaction also depended on the temperature of uranium hydride preparation.

  6. SEPARATION OF URANIUM, PLUTONIUM AND FISSION PRODUCTS FROM NEUTRON- BOMBARDED URANIUM

    DOE Patents [OSTI]

    Martin, A.E.; Johnson, I.; Burris, L. Jr.; Winsch, I.O.; Feder, H.M.

    1962-11-13

    A process is given for removing plutonium and/or fission products from uranium fuel. The fuel is dissolved in molten zinc--magnesium (10 to 18% Mg) alloy, more magnesium is added to obtain eutectic composition whereby uranium precipitates, and the uranium are separated from the Plutoniumand fission-product- containing eutectic. (AEC)

  7. Potentiometric determination of uranium in organic extracts

    SciTech Connect (OSTI)

    Bodnar, L.Z.

    1980-05-01

    The potentimetric determination of uranium in organic extracts was studied. A mixture of 30% TBP, (tributylphosphate), in carbon tetrachloride was used, with the NBL (New Brunswick Laboratory) titrimetric procedure. Results include a comparative analysis performed on organic extracts of fissium alloys vs those performed on aqueous samples of the same alloys which had been treated to remove interfering elements. Also comparative analyses were performed on sample solutions from a typical scrap recovery operation common in the uranium processing industry. A limited number of residue type materials, calciner products, and presscakes were subjected to analysis by organic extraction. The uranium extraction was not hindered by 30% TBP/CCl/sub 4/. To fully demonstrate the capabilities of the extraction technique and its compatibility with the NBL potentiometric uranium determination, a series of uranium standards was subjected to uranium extraction with 30% TBP/CCl/sub 4/. The uranium was then stripped out of the organic phase with 40 mL of H/sub 3/PO/sub 4/, 15 mL of H/sub 2/0, and 1 mL of 1M FeSO/sub 4/ solution. The uranium was then determined in the aqueous phosphoric phase by the regular NBL potentiometric method, omitting only the addition of another 40 mL of H/sub 3/PO/sub 4/. Uranium determinations ranging from approximately 20 to 150 mg of U were successfully made with the same accuracy and precision normally achieved. 8 tables. (DP)

  8. Colorimetric detection of uranium in water

    DOE Patents [OSTI]

    DeVol, Timothy A.; Hixon, Amy E.; DiPrete, David P.

    2012-03-13

    Disclosed are methods, materials and systems that can be used to determine qualitatively or quantitatively the level of uranium contamination in water samples. Beneficially, disclosed systems are relatively simple and cost-effective. For example, disclosed systems can be utilized by consumers having little or no training in chemical analysis techniques. Methods generally include a concentration step and a complexation step. Uranium concentration can be carried out according to an extraction chromatographic process and complexation can chemically bind uranium with a detectable substance such that the formed substance is visually detectable. Methods can detect uranium contamination down to levels even below the MCL as established by the EPA.

  9. Assessment of radionuclides (uranium and thorium) atmospheric...

    Office of Scientific and Technical Information (OSTI)

    Title: Assessment of radionuclides (uranium and thorium) atmospheric pollution around Manjung district, Perak using moss as bio-indicator Bio-monitoring method using mosses have ...

  10. SEPARATION OF URANIUM, PLUTONIUM AND FISSION PRODUCTS

    DOE Patents [OSTI]

    Nicholls, C.M.; Wells, I.; Spence, R.

    1959-10-13

    The separation of uranium and plutonium from neutronirradiated uranium is described. The neutron-irradiated uranium is dissolved in nitric acid to provide an aqueous solution 3N in nitric acid. The fission products of the solution are extruded by treating the solution with dibutyl carbitol substantially 1.8N in nitric acid. The organic solvent phase is separated and neutralized with ammonium hydroxide and the plutonium reduced with hydroxylamine base to the trivalent state. Treatment of the mixture with saturated ammonium nitrate extracts the reduced plutonium and leaves the uranium in the organic solvent.

  11. Oxidation and crystal field effects in uranium

    SciTech Connect (OSTI)

    Tobin, J. G.; Booth, C. H.; Shuh, D. K.; van der Laan, G.; Sokaras, D.; Weng, T. -C.; Yu, S. W.; Bagus, P. S.; Tyliszczak, T.; Nordlund, D.

    2015-07-06

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

  12. SEPARATION OF URANIUM FROM OTHER METALS

    DOE Patents [OSTI]

    Hyman, H.H.

    1959-07-01

    The separation of uranium from other elements, such as ruthenium, zirconium, niobium, cerium, and other rare earth metals is described. According to the invention, this is accomplished by adding hydrazine to an acid aqueous solution containing salts of uranium, preferably hexavalent uranium, and then treating the mixture with a substantially water immiscible ketone, such as hexone. A reaction takes place between the ketone and the hydrazine whereby a complex, a ketazine, is formed; this complex has a greater power of extraction for uranium than the ketone by itself. When contaminating elements are present, they substantially remain in ihe aqueous solution.

  13. Plutonium Uranium Extraction Plant (PUREX) - Hanford Site

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    About Us Projects & Facilities Plutonium Uranium Extraction Plant (PUREX) About Us About ... and 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage ...

  14. High strength and density tungsten-uranium alloys

    DOE Patents [OSTI]

    Sheinberg, Haskell

    1993-01-01

    Alloys of tungsten and uranium and a method for making the alloys. The amount of tungsten present in the alloys is from about 55 vol % to about 85 vol %. A porous preform is made by sintering consolidated tungsten powder. The preform is impregnated with molten uranium such that (1) uranium fills the pores of the preform to form uranium in a tungsten matrix or (2) uranium dissolves portions of the preform to form a continuous uranium phase containing tungsten particles.

  15. Depleted uranium oxides and silicates as spent nuclear fuel waste package fill materials

    SciTech Connect (OSTI)

    Forsberg, C.W.

    1996-09-10

    A new repository waste package (WP) concept for spent nuclear fuel (SNF) is being investigated that uses depleted uranium (DU) to improve performance and reduce the uncertainties of geological disposal of SNF. The WP would be filled with SNF and then filled with depleted uranium (DU) ({approximately}0.2 wt % {sup 235}U) dioxide (UO{sub 2}) or DU silicate-glass beads. Fission products and actinides can not escape the SNF UO{sub 2} crystals until the UO{sub 2} dissolves or is transformed into other chemical species. After WP failure, the DU fill material slows dissolution by three mechanisms: (1) saturation of AT groundwater with DU and suppression of SNF dissolution, (2) maintenance of chemically reducing conditions in the WP that minimize SNF solubility by sacrificial oxidation of DU from the +4 valence state, and (3) evolution of DU to lower-density hydrated uranium silicates. The fill expansion seals the WP from water flow. The DU also isotopically exchanges with SNF uranium as the SNF degrades to reduce long-term nuclear-criticality concerns.

  16. Recovery and Blend-Down Uranium for Beneficial use in Commercial Reactors - 13373

    SciTech Connect (OSTI)

    Magoulas, Virginia [Savannah River National Laboratory, Savannah River Site, Aiken, SC 29808 (United States)] [Savannah River National Laboratory, Savannah River Site, Aiken, SC 29808 (United States)

    2013-07-01

    In April 2001 the Department of Energy (DOE) and the Tennessee Valley Authority (TVA) signed an Interagency Agreement to transfer approximately 33 MT of off-specification (off-spec) highly enriched uranium (HEU) from DOE to TVA for conversion to commercial reactor fuel. Since that time additional surplus off-spec HEU material has been added to the program, making the total approximately 46 MT off-spec HEU. The disposition path for approximately half (23 MT) of this 46 MT of surplus HEU material, was down blending through the H-canyon facility at the Savannah River Site (SRS). The HEU is purified through the H-canyon processes, and then blended with natural uranium (NU) to form low enriched uranium (LEU) solution with a 4.95% U-235 isotopic content. This material was then transported to a TVA subcontractor who converted the solution to uranium oxide and then fabricated into commercial light water reactor (LWR) fuel. This fuel is now powering TVA reactors and supplying electricity to approximately 1 million households in the TVA region. There is still in excess of approximately 10 to 14 MT of off-spec HEU throughout the DOE complex or future foreign and domestic research reactor returns that could be recovered and down blended for use in either currently designed light water reactors, ?5% enriched LEU, or be made available for use in subsequent advanced 'fast' reactor fuel designs, ?19% LEU. (authors)

  17. Spatially-Resolved Analyses of Aerodynamic Fallout from a Uranium-Fueled Nuclear Test

    SciTech Connect (OSTI)

    Lewis, L. A.; Knight, K. B.; Matzel, J. E.; Prussin, S. G.; Zimmer, M. M.; Kinman, W S; Ryerson, F. J.; Hutcheon, I. D.

    2015-07-28

    The fiive silicate fallout glass spherules produced in a uranium-fueled, near-surface nuclear test were characterized by secondary ion mass spectrometry, electron probe microanalysis, autoradiography, scanning electron microscopy, and energy-dispersive x-ray spectroscopy. Several samples display compositional heterogeneity suggestive of incomplete mixing between major elements and natural U (238U/235U = 0.00725) and enriched U. Samples exhibit extreme spatial heterogeneity in U isotopic composition with 0.02 < 235U/238U < 11.84 among all five spherules and 0.02 < 235U/238U < 7.41 within a single spherule. Moreover, in two spherules, the 235U/238U ratio is correlated with changes in major element composition, suggesting the agglomeration of chemically and isotopically distinct molten precursors. Two samples are nearly homogenous with respect to major element and uranium isotopic composition, suggesting extensive mixing possibly due to experiencing higher temperatures or residing longer in the fireball. Linear correlations between 234U/238U, 235U/238U, and 236U/238U ratios are consistent with a two-component mixing model, which is used to illustrate the extent of mixing between natural and enriched U end members.

  18. Spatially-Resolved Analyses of Aerodynamic Fallout from a Uranium-Fueled Nuclear Test

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Lewis, L. A.; Knight, K. B.; Matzel, J. E.; Prussin, S. G.; Zimmer, M. M.; Kinman, W S; Ryerson, F. J.; Hutcheon, I. D.

    2015-07-28

    The fiive silicate fallout glass spherules produced in a uranium-fueled, near-surface nuclear test were characterized by secondary ion mass spectrometry, electron probe microanalysis, autoradiography, scanning electron microscopy, and energy-dispersive x-ray spectroscopy. Several samples display compositional heterogeneity suggestive of incomplete mixing between major elements and natural U (238U/235U = 0.00725) and enriched U. Samples exhibit extreme spatial heterogeneity in U isotopic composition with 0.02 < 235U/238U < 11.84 among all five spherules and 0.02 < 235U/238U < 7.41 within a single spherule. Moreover, in two spherules, the 235U/238U ratio is correlated with changes in major element composition, suggesting the agglomeration ofmore » chemically and isotopically distinct molten precursors. Two samples are nearly homogenous with respect to major element and uranium isotopic composition, suggesting extensive mixing possibly due to experiencing higher temperatures or residing longer in the fireball. Linear correlations between 234U/238U, 235U/238U, and 236U/238U ratios are consistent with a two-component mixing model, which is used to illustrate the extent of mixing between natural and enriched U end members.« less

  19. DEEP WATER ISOTOPIC CURRENT ANALYZER

    DOE Patents [OSTI]

    Johnston, W.H.

    1964-04-21

    A deepwater isotopic current analyzer, which employs radioactive isotopes for measurement of ocean currents at various levels beneath the sea, is described. The apparatus, which can determine the direction and velocity of liquid currents, comprises a shaft having a plurality of radiation detectors extending equidistant radially therefrom, means for releasing radioactive isotopes from the shaft, and means for determining the time required for the isotope to reach a particular detector. (AEC)

  20. Method for separating boron isotopes

    DOE Patents [OSTI]

    Rockwood, Stephen D.

    1978-01-01

    A method of separating boron isotopes .sup.10 B and .sup.11 B by laser-induced selective excitation and photodissociation of BCl.sub.3 molecules containing a particular boron isotope. The photodissociation products react with an appropriate chemical scavenger and the reaction products may readily be separated from undissociated BCl.sub.3, thus effecting the desired separation of the boron isotopes.