National Library of Energy BETA

Sample records for low-enriched uranium billets

  1. Environmental assessment: Transfer of normal and low-enriched uranium billets to the United Kingdom, Hanford Site, Richland, Washington

    SciTech Connect (OSTI)

    1995-11-01

    Under the auspices of an agreement between the U.S. and the United Kingdom, the U.S. Department of Energy (DOE) has an opportunity to transfer approximately 710,000 kilograms (1,562,000 pounds) of unneeded normal and low-enriched uranium (LEU) to the United Kingdom; thus, reducing long-term surveillance and maintenance burdens at the Hanford Site. The material, in the form of billets, is controlled by DOE`s Defense Programs, and is presently stored as surplus material in the 300 Area of the Hanford Site. The United Kingdom has expressed a need for the billets. The surplus uranium billets are currently stored in wooden shipping containers in secured facilities in the 300 Area at the Hanford Site (the 303-B and 303-G storage facilities). There are 482 billets at an enrichment level (based on uranium-235 content) of 0.71 weight-percent. This enrichment level is normal uranium; that is, uranium having 0.711 as the percentage by weight of uranium-235 as occurring in nature. There are 3,242 billets at an enrichment level of 0.95 weight-percent (i.e., low-enriched uranium). This inventory represents a total of approximately 532 curies. The facilities are routinely monitored. The dose rate on contact of a uranium billet is approximately 8 millirem per hour. The dose rate on contact of a wooden shipping container containing 4 billets is approximately 4 millirem per hour. The dose rate at the exterior of the storage facilities is indistinguishable from background levels.

  2. EA-1123: Transfer of Normal and Low-Enriched Uranium Billets to the United Kingdom, Hanford Site, Richland, Washington

    Broader source: Energy.gov [DOE]

    This EA evaluates the environmental impacts of the proposal to transfer approximately 710,000 kilograms (1,562,000 pounds) of unneeded normal and low-enriched uranium to the United Kingdom; thus,...

  3. 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)

  4. Quadrilateral Cooperation on High-density Low-enriched Uranium...

    National Nuclear Security Administration (NNSA)

    Quadrilateral Cooperation on High-density Low-enriched Uranium Fuel Production: Fact Sheet ... Fact Sheets Quadrilateral Cooperation on High-density Low-enriched Uranium Fuel ... ...

  5. :- : DRILLING URANIUM BILLETS ON A

    Office of Legacy Management (LM)

    . . . . . . . 6 4. FIRST BILLET DRILLING TEST . .. .. . . . . . . . 10 4.1 Feedstock . . ... . . . . . . . 13 5. SECOND BILLET DRILLING TEST .. . . . . . . . . . 13 5.1 Feedstock . . ...

  6. Possibility of nuclear pumped laser experiment using low enriched uranium

    SciTech Connect (OSTI)

    Obara, Toru; Takezawa, Hiroki [Center for Research into Innovative Nuclear Energy Systems Tokyo Institute of Technology 2-12-1-N1-19, Ookayama Meguro-ku, Tokyo 152-8550 (Japan)

    2012-06-06

    Possibility to perform experiments for nuclear pumped laser oscillation by using low enriched uranium is investigated. Kinetic analyses are performed for two types of reactor design, one is using highly enriched uranium and the other is using low enriched uranium. The reactor design is based on the experiment reactor in IPPE. The results show the oscillation of nuclear pumped laser in the case of low enriched uranium reactor is also possible. The use of low enriched uranium in the experiment will make experiment easier.

  7. Simulation of transportation of low enriched uranium solutions

    SciTech Connect (OSTI)

    Hope, E.P.; Ades, M.J.

    1996-08-01

    A simulation of the transportation by truck of low enriched uranium solutions has been completed for NEPA purposes at the Savannah River Site. The analysis involves three distinct source terms, and establishes the radiological risks of shipment to three possible destinations. Additionally, loading accidents were analyzed to determine the radiological consequences of mishaps during handling and delivery. Source terms were developed from laboratory measurements of chemical samples from low enriched uranium feed materials being stored at SRS facilities, and from manufacturer data on transport containers. The transportation simulations were accomplished over the INTERNET using the DOE TRANSNET system at Sandia National Laboratory. The HIGHWAY 3.3 code was used to analyze routing scenarios, and the RADTRAN 4 code was used to analyze incident free and accident risks of transporting radiological materials. Loading accidents were assessed using the Savannah River Site AXAIR89Q and RELEASE 2 codes.

  8. The Role of COMSOL Toward a Low-Enriched Uranium Fuel Design for the High

    Office of Scientific and Technical Information (OSTI)

    Flux Isotope Reactor (Conference) | SciTech Connect Conference: The Role of COMSOL Toward a Low-Enriched 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 Isotope Reactor Design and safety analyses are underway to convert the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) from a high-enriched uranium (HEU) fuel to a low-enriched uranium

  9. Active-Interrogation Measurements of Fast Neutrons from Induced Fission in Low-Enriched Uranium

    SciTech Connect (OSTI)

    J. L. Dolan; M. J. Marcath; M. Flaska; S. A. Pozzi; D. L. Chichester; A. Tomanin; P. Peerani

    2014-02-01

    A detection system was designed with MCNPX-PoliMi to measure induced-fission neutrons from U-235 and U-238 using active interrogation. Measurements were then performed with this system at the Joint Research Centre (JRC) in Ispra, Italy on low-enriched uranium samples. Liquid scintillators measured induced fission neutron to characterize the samples in terms of their uranium mass and enrichment. Results are presented to investigate and support the use of organic liquid scintillators with active interrogation techniques to characterize uranium containing materials.

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

    SciTech Connect (OSTI)

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

    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.

  11. Calculation of parameters for inspection planning and evaluation: low enriched uranium conversion and fuel fabrication facilities

    SciTech Connect (OSTI)

    Reardon, P.T.; Mullen, M.F.; Harms, N.L.

    1981-02-01

    As part of Task C.35 (Calculation of Parameters for Inspection Planning and Evaluation) of the US Program of Technical Assistance to IAEA Safeguards, Pacific Northwest Laboratory has performed some quantitative analyses of IAEA inspection activities at low-enriched uranium (LEU) conversion and fuel fabrication facilities. This report presents the results and conclusions of those analyses. Implementation of IAEA safeguards at LEU conversion and fuel fabrication facilities must take into account a variety of practical problems and constraints. One of the key concerns is the problem of flow verification, especially product verification. The objective of this report is to help put the problem of flow verification in perspective by presenting the results of some specific calculations of inspection effort and probability of detection for various product measurement strategies. In order to provide quantitative information about the advantages and disadvantages of the various strategies, eight specific cases were examined.

  12. High Accuracy U-235 Enrichment Verification Station for Low Enriched Uranium Alloys

    SciTech Connect (OSTI)

    Lillard, C. R.; Hayward, J. P.; Williamson, M. R.

    2012-06-07

    The Y-12 National Security Complex is playing a role in the U.S. High Performance Research Reactor (USHPRR) Conversion program sponsored by the U.S. National Nuclear Security Administration's Office of Global Threat Reduction. The USHPRR program has a goal of converting remaining U.S. reactors that continue to use highly enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. The USHPRR program is currently developing a LEU Uranium-Molybdenum (U-Mo) monolithic fuel for use in the U.S. high performance research reactors.Y-12 is supporting both the fuel development and fuel fabrication efforts by fabricating low enriched U-Mo foils from its own source material for irradiation experiments and for optimizing the fabrication process in support of scaling up the process to a commercial production scale. Once the new fuel is qualified, Y-12 will produce and ship U-Mo coupons with verified 19.75% +0.2% - 0.3% U-235 enrichment to be fabricated into fuel elements for the USHPRRs. Considering this small enrichment tolerance and the transition into HEU being set strictly at 20% U-235, a characterization system with a measurement uncertainty of less than or equal to 0.1% in enrichment is desired to support customer requirements and minimize production costs. Typical uncertainty for most available characterization systems today is approximately 1-5%; therefore, a specialized system must be developed which results in a reduced measurement uncertainty. A potential system using a High-Purity Germanium (HPGe) detector has been procured, and tests have been conducted to verify its capabilities with regards to the requirements. Using four U-Mo enrichment standards fabricated with complete isotopic and chemical characterization, infinite thickness and peak-ratio enrichment measurement methods have been considered for use. As a result of inhomogeneity within the U-Mo samples, FRAM, an isotopic analysis software, has been selected for initial testing. A systematic approach towards observing effects on FRAM's enrichment analysis has been conducted with regards to count and dead time.

  13. Validity of Hansen-Roach cross sections in low-enriched uranium systems

    SciTech Connect (OSTI)

    Busch, R.D. ); O'Dell, R.D. )

    1991-01-01

    Within the nuclear criticality safety community, the Hansen-Roach 16 group cross section set has been the standard'' for use in k{sub eff} calculations over the past 30 years. Yet even with its widespread acceptance, there are still questions about its validity and adequacy, about the proper procedure for calculating the potential scattering cross section, {sigma}{sub p}, for uranium and plutonium, and about the concept of resonance self shielding and its impact on cross sections. This paper attempts to address these questions. It provides a brief background on the Hansen-Roach cross sections. Next is presented a review of resonances in cross sections, self shielding of these resonances, and the use of {sigma}{sub p} to characterize resonance self shielding. Three prescriptions for calculating {sigma}{sub p} are given. Finally, results of several calculations of k{sub eff} on low-enriched uranium systems are provided to confirm the validity of the Hansen-Roach cross sections when applied to such systems.

  14. Conversion and Blending Facility highly enriched uranium to low enriched uranium as metal. Revision 1

    SciTech Connect (OSTI)

    1995-07-05

    The mission of this Conversion and Blending Facility (CBF) will be to blend surplus HEU metal and alloy with depleted uranium metal to produce an LEU product. The primary emphasis of this blending operation will be to destroy the weapons capability of large, surplus stockpiles of HEU. The blended LEU product can only be made weapons capable again by the uranium enrichment process. The blended LEU will be produced as a waste suitable for storage or disposal.

  15. CONCEPTUAL PROCESS DESCRIPTION FOR THE MANUFACTURE OF LOW-ENRICHED URANIUM-MOLYBDENUM FUEL

    SciTech Connect (OSTI)

    Daniel M. Wachs; Curtis R. Clark; Randall J. Dunavant

    2008-02-01

    The National Nuclear Security Agency Global Threat Reduction Initiative (GTRI) is tasked with minimizing the use of high-enriched uranium (HEU) worldwide. A key component of that effort is the conversion of research reactors from HEU to low-enriched uranium (LEU) fuels. The GTRI Convert Fuel Development program, previously known as the Reduced Enrichment for Research and Test Reactors program was initiated in 1978 by the United States Department of Energy to develop the nuclear fuels necessary to enable these conversions. The program cooperates with the research reactors’ operators to achieve this goal of HEU to LEU conversion without reduction in reactor performance. The programmatic mandate is to complete the conversion of all civilian domestic research reactors by 2014. These reactors include the five domestic high-performance research reactors (HPRR), namely: the High Flux Isotope Reactor at the Oak Ridge National Laboratory, the Advanced Test Reactor at the Idaho National Laboratory, the National Bureau of Standards Reactor at the National Institute of Standards and Technology, the Missouri University Research Reactor at the University of Missouri–Columbia, and the MIT Reactor-II at the Massachusetts Institute of Technology. Characteristics for each of the HPRRs are given in Appendix A. The GTRI Convert Fuel Development program is currently engaged in the development of a novel nuclear fuel that will enable these conversions. The fuel design is based on a monolithic fuel meat (made from a uranium-molybdenum alloy) clad in Al-6061 that has shown excellent performance in irradiation testing. The unique aspects of the fuel design, however, necessitate the development and implementation of new fabrication techniques and, thus, establishment of the infrastructure to ensure adequate fuel fabrication capability. A conceptual fabrication process description and rough estimates of the total facility throughput are described in this document as a basis for establishing preconceptual fabrication facility designs.

  16. 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.

  17. Enhanced Low-Enriched Uranium Fuel Element for the Advanced Test Reactor

    SciTech Connect (OSTI)

    Pope, M. A.; DeHart, M. D.; Morrell, S. R.; Jamison, R. K.; Nef, E. C.; Nigg, D. W.

    2015-03-01

    Under the current US Department of Energy (DOE) policy and planning scenario, the Advanced Test Reactor (ATR) and its associated critical facility (ATRC) will be reconfigured to operate on low-enriched uranium (LEU) fuel. This effort has produced a conceptual design for an Enhanced LEU Fuel (ELF) element. This fuel features monolithic U-10Mo fuel foils and aluminum cladding separated by a thin zirconium barrier. As with previous iterations of the ELF design, radial power peaking is managed using different U-10Mo foil thicknesses in different plates of the element. The lead fuel element design, ELF Mk1A, features only three fuel meat thicknesses, a reduction from the previous iterations meant to simplify manufacturing. Evaluation of the ELF Mk1A fuel design against reactor performance requirements is ongoing, as are investigations of the impact of manufacturing uncertainty on safety margins. The element design has been evaluated in what are expected to be the most demanding design basis accident scenarios and has met all initial thermal-hydraulic criteria.

  18. 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.

  19. Active-Interrogation Measurements of Induced-Fission Neutrons from Low-Enriched Uranium

    SciTech Connect (OSTI)

    J. L. Dolan; M. J. Marcath; M. Flaska; S. A. Pozzi; D. L. Chichester; A. Tomanin; P. Peerani; G. Nebbia

    2012-07-01

    Protection and control of nuclear fuels is paramount for nuclear security and safeguards; therefore, it is important to develop fast and robust controlling mechanisms to ensure the safety of nuclear fuels. Through both passive- and active-interrogation methods we can use fast-neutron detection to perform real-time measurements of fission neutrons for process monitoring. Active interrogation allows us to use different ranges of incident neutron energy to probe for different isotopes of uranium. With fast-neutron detectors, such as organic liquid scintillation detectors, we can detect the induced-fission neutrons and photons and work towards quantifying a samples mass and enrichment. Using MCNPX-PoliMi, a system was designed to measure induced-fission neutrons from U-235 and U-238. Measurements were then performed in the summer of 2010 at the Joint Research Centre in Ispra, Italy. Fissions were induced with an associated particle D-T generator and an isotopic Am-Li source. The fission neutrons, as well as neutrons from (n, 2n) and (n, 3n) reactions, were measured with five 5 by 5 EJ-309 organic liquid scintillators. The D-T neutron generator was available as part of a measurement campaign in place by Padova University. The measurement and data-acquisition systems were developed at the University of Michigan utilizing a CAEN V1720 digitizer and pulse-shape discrimination algorithms to differentiate neutron and photon detections. Low-enriched uranium samples of varying mass and enrichment were interrogated. Acquired time-of-flight curves and cross-correlation curves are currently analyzed to draw relationships between detected neutrons and sample mass and enrichment. In the full paper, the promise of active-interrogation measurements and fast-neutron detection will be assessed through the example of this proof-of-concept measurement campaign. Additionally, MCNPX-PoliMi simulation results will be compared to the measured data to validate the MCNPX-PoliMi code when used for active-interrogation simulations.

  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. 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.

  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. Preliminary Evaluation of Alternate Designs for HFIR Low-Enriched Uranium Fuel

    SciTech Connect (OSTI)

    Renfro, David G; Chandler, David; Cook, David Howard; Ilas, Germina; Jain, Prashant K; Valentine, Jennifer R

    2014-11-01

    Engineering design studies of the feasibility of conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL) as part of an effort sponsored by the U.S. Department of Energy s Global Threat Reduction Initiative (GTRI)/Reduced Enrichment for Research and Test Reactors (RERTR) program. The fuel type selected by the program for the conversion of the five high-power research reactors in the U.S. that still use HEU fuel is a new U-Mo monolithic fuel. Studies by ORNL have previously indicated that HFIR can be successfully converted using the new fuel provided (1) the reactor power can be increased from 85 MW to 100 MW and (2) the fuel can be fabricated to a specific reference design. Fabrication techniques for the new fuel are under development by the program but are still immature, especially for the complex aspects of the HFIR fuel design. In FY 2012, the program underwent a major shift in focus to emphasize developing and qualifying processes for the fabrication of reliable and affordable LEU fuel. In support of this new focus and in an effort to ensure that the HFIR fuel design is as suitable for reliable fabrication as possible, ORNL undertook the present study to propose and evaluate several alternative design features. These features include (1) eliminating the fuel zone axial contouring in the previous reference design by substituting a permanent neutron absorber in the lower unfueled region of all of the fuel plates, (2) relocating the burnable neutron absorber from the fuel plates of the inner fuel element to the side plates of the inner fuel element (the fuel plates of the outer fuel element do not contain a burnable absorber), (3) relocating the fuel zone inside the fuel plate to be centered on the centerline of the depth of the plate, and (4) reshaping the radial contour of the relocated fuel zone to be symmetric about this centerline. The present studies used current analytical tools to evaluate the various alternate designs for cycle length, scientific performance (e.g., neutron scattering), and steady-state and transient thermal performance using both safety limit and nominal parameter assumptions. The studies concluded that a new reference design combining a permanent absorber in the lower unfueled region of all of the fuel plates, a burnable absorber in the inner element side plates, and a relocated and reshaped (but still radially contoured) fuel zone will allow successful conversion of HFIR. Future collaboration with the program will reveal whether the new reference design can be fabricated reliably and affordably. Following this feedback, additional studies using state-of-the-art developmental analytical tools are proposed to optimize the design of the fuel zone radial contour and the amount and location of both types of neutron absorbers to further flatten thermal peaks while maximizing the performance of the reactor.

  4. Preliminary Evaluation of Alternate Designs for HFIR Low-Enriched Uranium Fuel

    SciTech Connect (OSTI)

    Renfro, David; Chandler, David; Cook, David; Ilas, Germina; Jain, Prashant; Valentine, Jennifer

    2014-10-30

    Engineering design studies of the feasibility of conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL) as part of an effort sponsored by the U.S. Department of Energy’s Global Threat Reduction Initiative (GTRI)/Reduced Enrichment for Research and Test Reactors (RERTR) program. The fuel type selected by the program for the conversion of the five high-power research reactors in the U.S. that still use HEU fuel is a new U-Mo monolithic fuel. Studies by ORNL have previously indicated that HFIR can be successfully converted using the new fuel provided (1) the reactor power can be increased from 85 MW to 100 MW and (2) the fuel can be fabricated to a specific reference design. Fabrication techniques for the new fuel are under development by the program but are still immature, especially for the “complex” aspects of the HFIR fuel design. In FY 2012, the program underwent a major shift in focus to emphasize developing and qualifying processes for the fabrication of reliable and affordable LEU fuel. In support of this new focus and in an effort to ensure that the HFIR fuel design is as suitable for reliable fabrication as possible, ORNL undertook the present study to propose and evaluate several alternative design features. These features include (1) eliminating the fuel zone axial contouring in the previous reference design by substituting a permanent neutron absorber in the lower unfueled region of all of the fuel plates, (2) relocating the burnable neutron absorber from the fuel plates of the inner fuel element to the side plates of the inner fuel element (the fuel plates of the outer fuel element do not contain a burnable absorber), (3) relocating the fuel zone inside the fuel plate to be centered on the centerline of the depth of the plate, and (4) reshaping the radial contour of the relocated fuel zone to be symmetric about this centerline. The present studies used current analytical tools to evaluate the various alternate designs for cycle length, scientific performance (e.g., neutron scattering), and steady-state and transient thermal performance using both safety limit and nominal parameter assumptions. The studies concluded that a new reference design combining a permanent absorber in the lower unfueled region of all of the fuel plates, a burnable absorber in the inner element side plates, and a relocated and reshaped (but still radially contoured) fuel zone will allow successful conversion of HFIR. Future collaboration with the program will reveal whether the new reference design can be fabricated reliably and affordably. Following this feedback, additional studies using state-of-the-art developmental analytical tools are proposed to optimize the design of the fuel zone radial contour and the amount and location of both types of neutron absorbers to further flatten thermal peaks while maximizing the performance of the reactor.

  5. EA-1172: Sale of Surplus Natural and Low Enriched Uranium, Piketon, Ohio

    Broader source: Energy.gov [DOE]

    This EA evaluates the environmental impacts for the proposal to sell uranium for subsequent enrichment and fabrication into commercial nuclear power reactor fuel.  The uranium is currently stored...

  6. 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.

  7. Monte Carlo modeling and analyses of YALINA booster subcritical assembly, Part III : low enriched uranium conversion analyses.

    SciTech Connect (OSTI)

    Talamo, A.; Gohar, Y. (Nuclear Engineering Division) [Nuclear Engineering Division

    2011-05-12

    This study investigates the performance of the YALINA Booster subcritical assembly, located in Belarus, during operation with high (90%), medium (36%), and low (21%) enriched uranium fuels in the assembly's fast zone. The YALINA Booster is a zero-power, subcritical assembly driven by a conventional neutron generator. It was constructed for the purpose of investigating the static and dynamic neutronics properties of accelerator driven subcritical systems, and to serve as a fast neutron source for investigating the properties of nuclear reactions, in particular transmutation reactions involving minor-actinides. The first part of this study analyzes the assembly's performance with several fuel types. The MCNPX and MONK Monte Carlo codes were used to determine effective and source neutron multiplication factors, effective delayed neutron fraction, prompt neutron lifetime, neutron flux profiles and spectra, and neutron reaction rates produced from the use of three neutron sources: californium, deuterium-deuterium, and deuterium-tritium. In the latter two cases, the external neutron source operates in pulsed mode. The results discussed in the first part of this report show that the use of low enriched fuel in the fast zone of the assembly diminishes neutron multiplication. Therefore, the discussion in the second part of the report focuses on finding alternative fuel loading configurations that enhance neutron multiplication while using low enriched uranium fuel. It was found that arranging the interface absorber between the fast and the thermal zones in a circular rather than a square array is an effective method of operating the YALINA Booster subcritical assembly without downgrading neutron multiplication relative to the original value obtained with the use of the high enriched uranium fuels in the fast zone.

  8. 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.

  9. 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.

  10. Moderation control in low enriched {sup 235}U uranium hexafluoride packaging operations and transportation

    SciTech Connect (OSTI)

    Dyer, R.H.; Kovac, F.M.; Pryor, W.A.

    1993-10-01

    Moderation control is the basic parameter for ensuring nuclear criticality safety during the packaging and transport of low {sup 235}U enriched uranium hexafluoride before its conversion to nuclear power reactor fuel. Moderation control has permitted the shipment of bulk quantities in large cylinders instead of in many smaller cylinders and, therefore, has resulted in economies without compromising safety. Overall safety and uranium accountability have been enhanced through the use of the moderation control. This paper discusses moderation control and the operating procedures to ensure that moderation control is maintained during packaging operations and transportation.

  11. 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.

  12. Technical basis in support of the conversion of the University of Missouri Research Reactor (MURR) core from highly-enriched to low-enriched uranium - core neutron physics

    SciTech Connect (OSTI)

    Stillman, J.; Feldman, E.; Foyto, L; Kutikkad, K; McKibben, J C; Peters, N.; Stevens, J.

    2012-09-01

    This report contains the results of reactor design and performance for conversion of the University of Missouri Research Reactor (MURR) from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL) and the MURR Facility. The core conversion to LEU is being performed with financial support of the U. S. government.

  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, several engineering proof-of-principle tests would be required. The RERTR program is currently conducting a series of generic fuel qualification tests at the Advanced Test Reactor. A review of these tests and a review of the safety basis for the current, HEU fuel cycle led to the identification of a set of HFIR-specific fuel qualification tests. Much additional study is required to formulate a HFIR-specific fuel qualification plan from this set. However, one such test - creating a graded fuel profile across a flat foil - has been initiated with promising results.

  14. SUB-LEU-METAL-THERM-001 SUBCRITICAL MEASUREMENTS OF LOW ENRICHED TUBULAR URANIUM METAL FUEL ELEMENTS BEFORE & AFTER IRRADIATION

    SciTech Connect (OSTI)

    SCHWINKENDORF, K.N.

    2006-05-12

    With the shutdown of the Hanford PUREX (Plutonium-Uranium Extraction Plant) reprocessing plant in the 1970s, adequate storage capacity for spent Hanford N Reactor fuel elements in the K and N Reactor pools became a concern. To maximize space utilization in the pools, accounting for fuel burnup was considered. Calculations indicated that at typical fuel exposures for N Reactor, the spent-fuel critical mass would be twice the critical mass for green fuel. A decision was reached to test the calculational result with a definitive experiment. If the results proved positive, storage capacity could be increased and N Reactor operation could be prolonged. An experiment to be conducted in the N Reactor spent-fuel storage pool was designed and assembled and the services of the Battelle Northwest Laboratories (BNWL) (now Pacific Northwest National Laboratory [PNNL]) critical mass laboratory were procured for the measurements. The experiments were performed in April 1975 in the Hanford N Reactor fuel storage pool. The fuel elements were MKIA fuel assemblies, comprising two concentric tubes of low-enriched metallic uranium. Two separate sets of measurements were performed: one with ''green'' (fresh) fuel and one with spent fuel. Both the green and spent fuel, were measured in the same geometry. The spent-fuel MKIA assemblies had an average burnup of 2865 MWd (megawatt days)/t. A constraint was imposed restricting the measurements to a subcritical limit of k{sub eff} = 0.97. Subcritical count rate data was obtained with pulsed-neutron and approach-to-critical measurements. Ten (10) configurations with green fuel and nine (9) configurations with spent fuel are described and evaluated. Of these, 3 green fuel and 4 spent fuel loading configurations were considered to serve as benchmark models. However, shortcomings in experimental data failed to meet the high standards for a benchmark problem. Nevertheless, the data provided by these subcritical measurements can supply useful information to analysts evaluating spent fuel subcriticality. The original purpose of the subcritical measurements was to validate computer model predictions that spent N Reactor fuel of a particular, typical exposure (2740 MWd/t) had a critical mass equal to twice that of unexposed fuel of the same type. The motivation for performing this work was driven by the need to increase spent fuel storage limits. These subcritical measurements confirmed the computer model predictions.

  15. Report on the Effect the Low Enriched Uranium Delivered Under the Highly Enriched Uranium Agreement Between the Government of the United States and the Government of the Russian Federation has on the

    Energy Savers [EERE]

    Report on the Effect the Low Enriched Uranium Delivered Under the Highly Enriched Uranium Agreement Between the Government of the United States of America and the Government of the Russian Federation has on the Domestic Uranium Mining, Conversion, and Enrichment Industries and the Operation of the Gaseous Diffusion Plant 2008 Information Date: December 31, 2008 1 Introduction The Agreement Between the Government of the United States of America and the Government of the Russian Federation

  16. SUB-LEU-METAL-THERM-001 SUBCRITICAL MEASUREMENTS OF LOW ENRICHED TUBULAR URANIUM METAL FUEL ELEMENTS BEFORE & AFTER IRRADIATION

    SciTech Connect (OSTI)

    TOFFER, H.

    2006-07-18

    With the shutdown of the Hanford PUREX (Plutonium-Uranium Extraction Plant) reprocessing plant in the 1970s, adequate storage capacity for spent Hanford N Reactor fuel elements in the K and N Reactor pools became a concern. To maximize space utilization in the pools, accounting for fuel burnup was considered. Fuel that had experienced a neutron environment in a reactor is known as spent, exposed, or irradiated fuel. In contrast fuel that has not yet been placed in a reactor is known as green, unexposed, or unirradiated fuel. Calculations indicated that at typical fuel exposures for N Reactor, the spent-fuel critical mass would be twice the critical mass for green fuel. A decision was reached to test the calculational result with a definitive experiment. If the results proved positive, storage capacity could be increased and N Reactor operation could be prolonged. An experiment to be conducted in the N Reactor spent-fuel storage pool was designed and assembled (References 1 and 2) and the services of the Battelle Northwest Laboratories (BNWL) (now Pacific Northwest National Laboratory [PNNL]) critical mass laboratory were procured for the measurements (Reference 3). The experiments were performed in April 1975 in the Hanford N Reactor fuel storage pool. The fuel elements were MKIA fuel assemblies, comprised of two concentric tubes of low-enriched metallic uranium. Two separate sets of measurements were performed: one with unirradiated fuel and one with irradiated fuel. Both the unirradiated and irradiated fuel, were measured in the same geometry. The spent-fuel MKIA assemblies had an average burnup of 2865 MWd (megawatt days)/t. A constraint was imposed restricting the measurements to a subcritical limit of k{sub eff} = 0.97. Subcritical count rate data was obtained with pulsed-neutron and approach-to-critical measurements. Ten (10) configurations with green fuel and nine (9) configurations with spent fuel are described and evaluated. Of these, three (3) green fuel and four (4) spent fuel loading configurations were considered to serve as benchmark models. However, shortcomings in experimental data, such as the uncertainty in fuel exposure impact on reactivity and the pulse neutron data evaluation methodology, failed to meet the high standards for a benchmark problem. Nevertheless, the data provided by these subcritical measurements supply useful information to analysts evaluating spent fuel subcriticality. The original purpose of the subcritical measurements was to validate computer model predictions that spent N Reactor fuel of a particular, typical exposure (2740 MWd/t) had a critical mass equal to twice that of unexposed fuel of the same type. The motivation for performing this work was driven by the need to increase spent fuel storage limits. These subcritical measurements confirmed the computer model predictions.

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

    SciTech Connect (OSTI)

    Kuzminski, Jozef; Nesuhoff, J; Cratto, P; Pfennigwerth, G; Mikhailenko, A; Maliutina, I; Nations, J

    2009-01-01

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

  18. 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.

  19. A study of a zone approach to IAEA (International Atomic Energy Agency) safeguards: The low-enriched-uranium zone of a light-water-reactor fuel cycle

    SciTech Connect (OSTI)

    Fishbone, L.G.; Higinbotham, W.A.

    1986-06-01

    At present the IAEA designs its safeguards approach with regard to each type of nuclear facility so that the safeguards activities and effort are essentially the same for a given type and size of nuclear facility wherever it may be located. Conclusions regarding a state are derived by combining the conclusions regarding the effectiveness of safeguards for the individual facilities within a state. In this study it was convenient to define three zones in a state with a closed light-water-reactor nuclear fuel cycle. Each zone contains those facilities or parts thereof which use or process nuclear materials of the same safeguards significance: low-enriched uranium, radioactive spent fuel, or recovered plutonium. The possibility that each zone might be treated as an extended material balance area for safeguards purposes is under investigation. The approach includes defining the relevant features of the facilities in the three zones and listing the safeguards activities which are now practiced. This study has focussed on the fresh-fuel zone, the several facilities of which use or process low-enriched uranium. At one extreme, flows and inventories would be verified at each material balance area. At the other extreme, the flows into and out of the zone and the inventory of the whole zone would be verified. There are a number of possible safeguards approaches which fall between the two extremes. The intention is to develop a rational approach which will make it possible to compare the technical effectiveness and the inspection effort for the facility-oriented approach, for the approach involving the zone as a material balance area, and for some reasonable intermediate safeguards approaches.

  20. Nuclear safety analyses and core design calculations to convert the Texas A & M University Nuclear Science Center reactor to low enrichment uranium fuel. Final report

    SciTech Connect (OSTI)

    Parish, T.A.

    1995-03-02

    This project involved performing the nuclear design and safety analyses needed to modify the license issued by the Nuclear Regulatory Commission to allow operation of the Texas A& M University Nuclear Science Center Reactor (NSCR) with a core containing low enrichment uranium (LEU) fuel. The specific type of LEU fuel to be considered was the TRIGA 20-20 fuel produced by General Atomic. Computer codes for the neutronic analyses were provided by Argonne National Laboratory (ANL) and the assistance of William Woodruff of ANL in helping the NSCR staff to learn the proper use of the codes is gratefully acknowledged. The codes applied in the LEU analyses were WIMSd4/m, DIF3D, NCTRIGA and PARET. These codes allowed full three dimensional, temperature and burnup dependent calculations modelling the NSCR core to be performed for the first time. In addition, temperature coefficients of reactivity and pulsing calculations were carried out in-house, whereas in the past this modelling had been performed at General Atomic. In order to benchmark the newly acquired codes, modelling of the current NSCR core with highly enriched uranium fuel was also carried out. Calculated results were compared to both earlier licensing calculations and experimental data and the new methods were found to achieve excellent agreement with both. Therefore, even if an LEU core is never loaded at the NSCR, this project has resulted in a significant improvement in the nuclear safety analysis capabilities established and maintained at the NSCR.

  1. Field test of short-notice random inspections for inventory-change verification at a low-enriched-uranium fuel-fabrication plant: Preliminary summary

    SciTech Connect (OSTI)

    Fishbone, L.G. |; Moussalli, G.; Naegele, G.; Ikonomou, P.; Hosoya, M.; Scott, P.; Fager, J.; Sanders, C.; Colwell, D.; Joyner, C.J.

    1994-04-01

    An approach of short-notice random inspections (SNRIs) for inventory-change verification can enhance the effectiveness and efficiency of international safeguards at natural or low-enriched uranium (LEU) fuel fabrication plants. According to this approach, the plant operator declares the contents of nuclear material items before knowing if an inspection will occur to verify them. Additionally, items about which declarations are newly made should remain available for verification for an agreed time. This report details a six-month field test of the feasibility of such SNRIs which took place at the Westinghouse Electric Corporation Commercial Nuclear Fuel Division. Westinghouse personnel made daily declarations about both feed and product items, uranium hexafluoride cylinders and finished fuel assemblies, using a custom-designed computer ``mailbox``. Safeguards inspectors from the IAEA conducted eight SNRIs to verify these declarations. Items from both strata were verified during the SNRIs by means of nondestructive assay equipment. The field test demonstrated the feasibility and practicality of key elements of the SNRI approach for a large LEU fuel fabrication plant.

  2. Field test of short-notice random inspections for inventory-change verification at a low-enriched-uranium fuel-fabrication plant

    SciTech Connect (OSTI)

    Fishbone, L.G. |; Moussalli, G.; Naegele, G.

    1995-05-01

    An approach of short-notice random inspections (SNRIs) for inventory-change verification can enhance the effectiveness and efficiency of international safeguards at natural or low-enriched uranium (LEU) fuel fabrication plants. According to this approach, the plant operator declares the contents of nuclear material items before knowing if an inspection will occur to verify them. Additionally, items about which declarations are newly made should remain available for verification for an agreed time. Then a statistical inference can be made from verification results for items verified during SNRIs to the entire populations, i.e. the entire strata, even if inspectors were not present when many items were received or produced. A six-month field test of the feasibility of such SNRIs took place at the Westinghouse Electric Corporation Commercial Nuclear Fuel Division during 1993. Westinghouse personnel made daily declarations about both feed and product items, uranium hexafluoride cylinders and finished fuel assemblies, using a custom-designed computer ``mailbox``. Safeguards inspectors from the IAEA conducted eight SNRIs to verify these declarations. They arrived unannounced at the plant, in most cases immediately after travel from Canada, where the IAEA maintains a regional office. Items from both strata were verified during the SNRIs by meant of nondestructive assay equipment.

  3. In-Situ Measurements of Low Enrichment Uranium Holdup Process Gas Piping at K-25 - Paper for Waste Management Symposia 2010 East Tennessee Technology Park Oak Ridge, Tennessee

    SciTech Connect (OSTI)

    Rasmussen B.

    2010-01-01

    This document is the final version of a paper submitted to the Waste Management Symposia, Phoenix, 2010, abstract BJC/OR-3280. The primary document from which this paper was condensed is In-Situ Measurement of Low Enrichment Uranium Holdup in Process Gas Piping at K-25 Using NaI/HMS4 Gamma Detection Systems, BJC/OR-3355. This work explores the sufficiency and limitations of the Holdup Measurement System 4 (HJVIS4) software algorithms applied to measurements of low enriched uranium holdup in gaseous diffusion process gas piping. HMS4 has been used extensively during the decommissioning and demolition project of the K-25 building for U-235 holdup quantification. The HMS4 software is an integral part of one of the primary nondestructive assay (NDA) systems which was successfully tested and qualified for holdup deposit quantification in the process gas piping of the K-25 building. The initial qualification focused on the measurement of highly enriched UO{sub 2}F{sub 2} deposits. The purpose of this work was to determine if that qualification could be extended to include the quantification of holdup in UO{sub 2}F{sub 2} deposits of lower enrichment. Sample field data are presented to provide evidence in support of the theoretical foundation. The HMS4 algorithms were investigated in detail and found to sufficiently compensate for UO{sub 2}F{sub 2} source self-attenuation effects, over the range of expected enrichment (4-40%), in the North and East Wings of the K-25 building. The limitations of the HMS4 algorithms were explored for a described set of conditions with respect to area source measurements of low enriched UO{sub 2}F{sub 2} deposits when used in conjunction with a 1 inch by 1/2 inch sodium iodide (NaI) scintillation detector. The theoretical limitations of HMS4, based on the expected conditions in the process gas system of the K-25 building, are related back to the required data quality objectives (DQO) for the NBA measurement system established for the K-25 demolition project. The combined review of the HMS software algorithms and supporting field measurements lead to the conclusion that the majority of process gas pipe measurements are adequately corrected for source self-attenuation using HMS4. While there will be instances where the UO{sub 2}F{sub 2} holdup mass presents an infinitely thick deposit to the NaI-HMS4 system these situations are expected to be infrequent. This work confirms that the HMS4 system can quantify UO{sub 2}F{sub 2} holdup, in its current configuration (deposition, enrichment, and geometry), below the DQO levels for the K-25 building decommissioning and demolition project. For an area measurement of process gas pipe in the K-25 building, if an infinitely thick UO{sub 2}F{sub 2} deposit is identified in the range of enrichment of {approx}4-40%, the holdup quantity exceeds the corresponding DQO established for the K-25 building demolition project.

  4. Evaluating quantitative 3-D image analysis as a design tool for low enriched uranium fuel compacts for the transient reactor test facility: A preliminary study

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

    Kane, J. J.; van Rooyen, I. J.; Craft, A. E.; Roney, T. J.; Morrell, S. R.

    2016-02-05

    In this study, 3-D image analysis when combined with a non-destructive examination technique such as X-ray computed tomography (CT) provides a highly quantitative tool for the investigation of a material’s structure. In this investigation 3-D image analysis and X-ray CT were combined to analyze the microstructure of a preliminary subsized fuel compact for the Transient Reactor Test Facility’s low enriched uranium conversion program to assess the feasibility of the combined techniques for use in the optimization of the fuel compact fabrication process. The quantitative image analysis focused on determining the size and spatial distribution of the surrogate fuel particles andmore » the size, shape, and orientation of voids within the compact. Additionally, the maximum effect of microstructural features on heat transfer through the carbonaceous matrix of the preliminary compact was estimated. The surrogate fuel particles occupied 0.8% of the compact by volume with a log-normal distribution of particle sizes with a mean diameter of 39 μm and a standard deviation of 16 μm. Roughly 39% of the particles had a diameter greater than the specified maximum particle size of 44 μm suggesting that the particles agglomerate during fabrication. The local volume fraction of particles also varies significantly within the compact although uniformities appear to be evenly dispersed throughout the analysed volume. The voids produced during fabrication were on average plate-like in nature with their major axis oriented perpendicular to the compaction direction of the compact. Finally, the microstructure, mainly the large preferentially oriented voids, may cause a small degree of anisotropy in the thermal diffusivity within the compact. α∥/α⊥, the ratio of thermal diffusivities parallel to and perpendicular to the compaction direction are expected to be no less than 0.95 with an upper bound of 1.« less

  5. Reactor Physics Methods and Preconceptual Core Design Analyses for Conversion of the Advanced Test Reactor to Low-Enriched Uranium Fuel Annual Report for Fiscal Year 2012

    SciTech Connect (OSTI)

    David W. Nigg; Sean R. Morrell

    2012-09-01

    Under the current long-term DOE policy and planning scenario, both the ATR and the ATRC will be reconfigured at an appropriate time within the next several years to operate with low-enriched uranium (LEU) fuel. This will be accomplished under the auspices of the Reduced Enrichment Research and Test Reactor (RERTR) Program, administered by the DOE National Nuclear Security Administration (NNSA). At a minimum, the internal design and composition of the fuel element plates and support structure will change, to accommodate the need for low enrichment in a manner that maintains total core excess reactivity at a suitable level for anticipated operational needs throughout each cycle while respecting all control and shutdown margin requirements and power distribution limits. The complete engineering design and optimization of LEU cores for the ATR and the ATRC will require significant multi-year efforts in the areas of fuel design, development and testing, as well as a complete re-analysis of the relevant reactor physics parameters for a core composed of LEU fuel, with possible control system modifications. Ultimately, revalidation of the computational physics parameters per applicable national and international standards against data from experimental measurements for prototypes of the new ATR and ATRC core designs will also be required for Safety Analysis Report (SAR) changes to support routine operations with LEU. This report is focused on reactor physics analyses conducted during Fiscal Year (FY) 2012 to support the initial development of several potential preconceptual fuel element designs that are suitable candidates for further study and refinement during FY-2013 and beyond. In a separate, but related, effort in the general area of computational support for ATR operations, the Idaho National Laboratory (INL) is conducting a focused multiyear effort to introduce modern high-fidelity computational reactor physics software and associated validation protocols to replace several obsolete components of the current analytical tool set used for ATR neutronics support. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF). It will also greatly facilitate the LEU conversion effort, since the upgraded computational capabilities are now at a stage where they can be, and in fact have been, used for the required physics analysis from the beginning. In this context, extensive scoping neutronics analyses were completed for six preconceptual candidate LEU fuel element designs for the ATR (and for its companion critical facility, ATRC). Of these, four exhibited neutronics performance in what is believed to be an acceptable range. However, there are currently some concerns with regard to fabricability and mechanical performance that have emerged for one of the four latter concepts. Thus three concepts have been selected for more comprehensive conceptual design analysis during the upcoming fiscal year.

  6. 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.

  7. Disposition of DOE Excess Depleted Uranium, Natural Uranium, and

    Energy Savers [EERE]

    Low-Enriched Uranium | Department of Energy Disposition of DOE Excess Depleted Uranium, Natural Uranium, and Low-Enriched Uranium Disposition of DOE Excess Depleted Uranium, Natural Uranium, and Low-Enriched Uranium 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 depleted uranium hexafluoride (DUF6), natural uranium hexafluoride (NUF6), and

  8. Process for recovering niobium from uranium-niobium alloys

    DOE Patents [OSTI]

    Wallace, Steven A.; Creech, Edward T.; Northcutt, Walter G.

    1983-01-01

    Niobium is recovered from scrap uranium-niobium alloy by melting the scrap with tin, solidifying the billet thus formed, heating the billet to combine niobium with tin therein, placing the billet in hydrochloric acid to dissolve the uranium and leave an insoluble residue of niobium stannide, then separating the niobium stannide from the acid.

  9. Process for recovering niobium from uranium-niobium alloys

    DOE Patents [OSTI]

    Wallace, S.A.; Creech, E.T.; Northcutt, W.G.

    1982-09-27

    Niobium is recovered from scrap uranium-niobium alloy by melting the scrap with tin, solidifying the billet thus formed, heating the billet to combine niobium with tin therein, placing the billet in hydrochloric acid to dissolve the uranium and form a precipitate of niobium stannide, then separating the precipitate from the acid.

  10. EA-1123: Final Environmental Assessment

    Office of Energy Efficiency and Renewable Energy (EERE)

    Transfer of Normal and Low-Enriched Uranium Billets to the United Kingdom, Hanford Site, Richland, Washington

  11. EA-1123: Finding of No Significant Impact

    Broader source: Energy.gov [DOE]

    Transfer of Normal and Low-Enriched Uranium Billets to the United Kingdom, Hanford Site, Richland, Washington

  12. Baseline Graphite Characterization: First Billet

    SciTech Connect (OSTI)

    Mark C. Carroll; Joe Lords; David Rohrbaugh

    2010-09-01

    The Next Generation Nuclear Plant Project Graphite Research and Development program is currently establishing the safe operating envelope of graphite core components for a very high temperature reactor design. To meet this goal, the program is generating the extensive amount of quantitative data necessary for predicting the behavior and operating performance of the available nuclear graphite grades. In order determine the in-service behavior of the graphite for the latest proposed designs, two main programs are underway. The first, the Advanced Graphite Creep (AGC) program, is a set of experiments that are designed to evaluate the irradiated properties and behavior of nuclear grade graphite over a large spectrum of temperatures, neutron fluences, and compressive loads. Despite the aggressive experimental matrix that comprises the set of AGC test runs, a limited amount of data can be generated based upon the availability of space within the Advanced Test Reactor and the geometric constraints placed on the AGC specimens that will be inserted. In order to supplement the AGC data set, the Baseline Graphite Characterization program will endeavor to provide supplemental data that will characterize the inherent property variability in nuclear-grade graphite without the testing constraints of the AGC program. This variability in properties is a natural artifact of graphite due to the geologic raw materials that are utilized in its production. This variability will be quantified not only within a single billet of as-produced graphite, but also from billets within a single lot, billets from different lots of the same grade, and across different billets of the numerous grades of nuclear graphite that are presently available. The thorough understanding of this variability will provide added detail to the irradiated property data, and provide a more thorough understanding of the behavior of graphite that will be used in reactor design and licensing. This report covers the development of the Baseline Graphite Characterization program from a testing and data collection standpoint through the completion of characterization on the first billet of nuclear-grade graphite. This data set is the starting point for all future evaluations and comparisons of material properties.

  13. 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...

  14. Quadrilateral Cooperation on High-density Low-enriched Uranium...

    National Nuclear Security Administration (NNSA)

    This Site Budget IG Web Policy Privacy No Fear Act Accessibility FOIA Sitemap Federal Government The White House DOE.gov USA.gov Jobs Apply for Our Jobs Our Jobs Working at NNSA...

  15. RELAP5 model of the high flux isotope reactor with low enriched fuel thermal flux profiles

    SciTech Connect (OSTI)

    Banfield, J.; Mervin, B.; Hart, S.; Ritchie, J.; Walker, S.; Ruggles, A.; Maldonado, G. I. [Dept. of Nuclear Engineering, Univ. of Tennessee Knoxville, Knoxville, TN 37996-2300 (United States)

    2012-07-01

    The High Flux Isotope Reactor (HFIR) currently uses highly enriched uranium (HEU) fabricated into involute-shaped fuel plates. It is desired that HFIR be able to use low enriched uranium (LEU) fuel while preserving the current performance capability for its diverse missions in material irradiation studies, isotope production, and the use of neutron beam lines for basic research. Preliminary neutronics and depletion simulations of HFIR with LEU fuel have arrived to feasible fuel loadings that maintain the neutronics performance of the reactor. This article illustrates preliminary models developed for the analysis of the thermal-hydraulic characteristics of the LEU core to ensure safe operation of the reactor. The beginning of life (BOL) LEU thermal flux profile has been modeled in RELAP5 to facilitate steady state simulation of the core cooling, and of anticipated and unanticipated transients. Steady state results are presented to validate the new thermal power profile inputs. A power ramp, slow depressurization at the outlet, and flow coast down transients are also evaluated. (authors)

  16. Method for fabricating .sup.99 Mo production targets using low enriched uranium, .sup.99 Mo production targets comprising low enriched uranium

    DOE Patents [OSTI]

    Wiencek, Thomas C.; Matos, James E.; Hofman, Gerard L.

    2000-12-12

    A radioisotope production target and a method for fabricating a radioisotope production target is provided, wherein the target comprises an inner cylinder, a foil of fissionable material circumferentially contacting the outer surface of the inner cylinder, and an outer hollow cylinder adapted to receive the substantially foil-covered inner cylinder and compress tightly against the foil to provide good mechanical contact therewith. The method for fabricating a primary target for the production of fission products comprises preparing a first substrate to receive a foil of fissionable material so as to allow for later removal of the foil from the first substrate, preparing a second substrate to receive the foil so as to allow for later removal of the foil from the second substrate; attaching the first substrate to the second substrate such that the foil is sandwiched between the first substrate and second substrate to prevent foil exposure to ambient atmosphere, and compressing the exposed surfaces of the first and second substrate to assure snug mechanical contact between the foil, the first substrate and the second substrate.

  17. Method for fabricating .sup.99 Mo production targets using low enriched uranium, .sup.99 Mo production targets comprising low enriched uranium

    DOE Patents [OSTI]

    Wiencek, Thomas C.; Matos, James E.; Hofman, Gerard L.

    1997-01-01

    A radioisotope production target and a method for fabricating a radioisotope production target is provided, wherein the target comprises an inner cylinder, a foil of fissionable material circumferentially contacting the outer surface of the inner cylinder, and an outer hollow cylinder adapted to receive the substantially foil-covered inner cylinder and compress tightly against the foil to provide good mechanical contact therewith. The method for fabricating a primary target for the production of fission products comprises preparing a first substrate to receive a foil of fissionable material so as to allow for later removal of the foil from the first substrate, preparing a second substrate to receive the foil so as to allow for later removal of the foil from the second substrate; attaching the first substrate to the second substrate such that the foil is sandwiched between the first substrate and second substrate to prevent foil exposure to ambient atmosphere, and compressing the exposed surfaces of the first and second substrate to assure snug mechanical contact between the foil, the first substrate and the second substrate.

  18. Method for fabricating {sup 99}Mo production targets using low enriched uranium, {sup 99}Mo production targets comprising low enriched uranium

    DOE Patents [OSTI]

    Wiencek, T.C.; Matos, J.E.; Hofman, G.L.

    1997-03-25

    A radioisotope production target and a method for fabricating a radioisotope production target is provided, wherein the target comprises an inner cylinder, a foil of fissionable material circumferentially contacting the outer surface of the inner cylinder, and an outer hollow cylinder adapted to receive the substantially foil-covered inner cylinder and compress tightly against the foil to provide good mechanical contact therewith. The method for fabricating a primary target for the production of fission products comprises preparing a first substrate to receive a foil of fissionable material so as to allow for later removal of the foil from the first substrate, preparing a second substrate to receive the foil so as to allow for later removal of the foil from the second substrate; attaching the first substrate to the second substrate such that the foil is sandwiched between the first substrate and second substrate to prevent foil exposure to ambient atmosphere, and compressing the exposed surfaces of the first and second substrate to assure snug mechanical contact between the foil, the first substrate and the second substrate. 3 figs.

  19. FORMING TUBES AND RODS OF URANIUM METAL BY EXTRUSION

    DOE Patents [OSTI]

    Creutz, E.C.

    1959-01-27

    A method and apparatus are presented for the extrusion of uranium metal. Since uranium is very brittle if worked in the beta phase, it is desirable to extrude it in the gamma phase. However, in the gamma temperature range thc uranium will alloy with the metal of the extrusion dic, and is readily oxidized to a great degree. According to this patent, uranium extrusion in thc ganmma phase may be safely carried out by preheating a billet of uranium in an inert atmosphere to a trmperature between 780 C and 1100 C. The heated billet is then placed in an extrusion apparatus having dies which have been maintained at an elevated temperature for a sufficient length of time to produce an oxide film, and placing a copper disc between the uranium billet and the die.

  20. 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...

  1. METHOD OF HOT ROLLING URANIUM METAL

    DOE Patents [OSTI]

    Kaufmann, A.R.

    1959-03-10

    A method is given for quickly and efficiently hot rolling uranium metal in the upper part of the alpha phase temperature region to obtain sound bars and sheets possessing a good surface finish. The uranium metal billet is heated to a temperature in the range of 1000 deg F to 1220 deg F by immersion iii a molten lead bath. The heated billet is then passed through the rolls. The temperature is restored to the desired range between successive passes through the rolls, and the rolls are turned down approximately 0.050 inch between successive passes.

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

    Office of Scientific and Technical Information (OSTI)

    Several technical disciplines are required to complete this conversion including nuclear reactor physics, heat transfer, fluid dynamics, structural mechanics, fuel fabrication, and ...

  3. Method and apparatus for forming billets from metallic chip scraps

    DOE Patents [OSTI]

    Girshov, Vladimir Leonidovich (St. Petersburg, RU); Treschevskiy, Arnold Nikolayevich (St. Petersburg, RU); Kochkin, Victor Georgievich (St. Petersburg, RU); Abramov, Alexey Alexandrovich (St. Petersburg, RU); Sidenko, Natalja Semenovna (St. Petersburg, RU)

    2006-05-02

    After recycled titanium alloy chips are crushed and cleaned, they are pressed into cylindrically briquettes with a relative density of 0.6, and placed into capsules. The capsules are heated and placed into a preheated pressing rig. The pressing rig repetitively applies axial force to the capsule, resulting in a relative density of at least 0.95. The product billets are used for consumable electrodes, secondary casting alloys, forgings, extruded semi-finished products and the like.

  4. Hot rolling of thick uranium molybdenum alloys

    SciTech Connect (OSTI)

    DeMint, Amy L.; Gooch, Jack G.

    2015-11-17

    Disclosed herein are processes for hot rolling billets of uranium that have been alloyed with about ten weight percent molybdenum to produce cold-rollable sheets that are about one hundred mils thick. In certain embodiments, the billets have a thickness of about 7/8 inch or greater. Disclosed processes typically involve a rolling schedule that includes a light rolling pass and at least one medium rolling pass. Processes may also include reheating the rolling stock and using one or more heavy rolling passes, and may include an annealing step.

  5. Highly Enriched Uranium Transparency Program | National Nuclear Security

    National Nuclear Security Administration (NNSA)

    Administration Highly Enriched Uranium Transparency Program November 13, 2013 The U.S. National Nuclear Security Administration's (NNSA) Highly Enriched Uranium (HEU) Transparency Program reduces nuclear risk by monitoring the conversion of 500 metric tons (MT) of Russian HEU, enough material for 20,000 nuclear weapons, into low enriched uranium (LEU). This LEU is put into peaceful use in the United States, generating nearly 10% of all U.S. electrical power. The HEU Purchase Agreement:

  6. The effect of residuals on the presence of intergranular surface cracks on continuously cast billets

    SciTech Connect (OSTI)

    Wijngaarden, M.J.U.T. van; Visagie, G.P.

    1996-12-31

    During 1991, Iscor Vereeniging experienced a dramatic increase in the rejection rate of specialty steel bars rolled from continuously cast billets due to the presence of seams on the bars. The seams originated from tearing of the billets during the first 2 passes in the roughing mill during hot rolling. The defective billets were found to contain fine intergranular cracks on the surface. Such cracks have been described in the literature and have been attributed to the presence of high levels of residuals resulting in the well-known phenomenon of surface hot shortness which results from the enrichment of residuals at the grain boundaries after preferential oxidation of iron during scaling of the steel. The present investigation revealed that the effect of residuals on intergranular surface cracking is a complex interaction between steel composition and casting conditions such as casting speed, intensity of secondary cooling, section size, and mold type. This paper quantifies the effect of residuals on the intergranular surface cracking of continuously cast billets and quantitatively relates the incidence of these cracks to parameters which can be controlled during steelmaking and continuous casting.

  7. Assuaging Nuclear Energy Risks: The Angarsk International Uranium Enrichment Center

    SciTech Connect (OSTI)

    Myers, Astasia

    2011-06-28

    The recent nuclear renaissance has motivated many countries, especially developing nations, to plan and build nuclear power reactors. However, domestic low enriched uranium demands may trigger nations to construct indigenous enrichment facilities, which could be redirected to fabricate high enriched uranium for nuclear weapons. The potential advantages of establishing multinational uranium enrichment sites are numerous including increased low enrichment uranium access with decreased nuclear proliferation risks. While multinational nuclear initiatives have been discussed, Russia is the first nation to actualize this concept with their Angarsk International Uranium Enrichment Center (IUEC). This paper provides an overview of the historical and modern context of the multinational nuclear fuel cycle as well as the evolution of Russia's IUEC, which exemplifies how international fuel cycle cooperation is an alternative to domestic facilities.

  8. 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 Commissions Mid-Term Appraisal of the countrys 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 Indias 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.

  9. Development of uranium metal targets for {sup 99}Mo production

    SciTech Connect (OSTI)

    Wiencek, T.C.; Hofman, G.L.

    1993-10-01

    A substantial amount of high enriched uranium (HEU) is used for the production of medical-grade {sup 99}Mo. Promising methods of producing irradiation targets are being developed and may lead to the reduction or elimination of this HEU use. To substitute low enriched uranium (LEU) for HEU in the production of {sup 99}Mo, the target material may be changed to uranium metal foil. Methods of fabrication are being developed to simplify assembly and disassembly of the targets. Removal of the uranium foil after irradiation without dissolution of the cladding is a primary goal in order to reduce the amount of liquid radioactive waste material produced in the process. Proof-of-concept targets have been fabricated. Destructive testing indicates that acceptable contact between the uranium foil and the cladding can be achieved. Thermal annealing tests, which simulate the cladding/uranium diffusion conditions during irradiation, are underway. Plans are being made to irradiate test targets.

  10. Criticality concerns in cleaning large uranium hexafluoride cylinders

    SciTech Connect (OSTI)

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

    1995-06-01

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

  11. Conversion of Worcester Polytechnic Institute Reactor to low enriched uranium (LEU) fuel: Technical progress report for period August 15, 1987-February 15, 1988

    SciTech Connect (OSTI)

    Newton, T.H. Jr.

    1988-02-01

    An HEU fuel element was removed from the WPI core and tested in a Babcock-Wilcox 6M shipping container on August 27, 1987, for radiation level adequacy in shipping. Levels were found to be adequate so that use of the 6M container can be made in shipping the HEU fuel after a few weeks of decay time. A final submittal of the SAR technical specification changes relating to the fuel conversion was made on September 17, 1987. Questions regarding this submittal were received on January 25, 1988, and responses to these questions were made on February 10, 1988.

  12. 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.

  13. EIS-0240: Disposition of Surplus Highly Enriched Uranium

    Broader source: Energy.gov [DOE]

    The Department proposes to eliminate the proliferation threat of surplus highly enriched uranium (HEU) by blending it down to low enriched uranium (LEU), which is not weapons-usable. The EIS assesses the disposition of a nominal 200 metric tons of surplus HEU. The Preferred Alternative is, where practical, to blend the material for use as LEU and use overtime, in commercial nuclear reactor field to recover its economic value. Material that cannot be economically recovered would be blended to LEU for disposal as low-level radioactive waste.

  14. Report on the Effect the Low Enriched Uranium Delivered Under the Highly Enriched Uranium Agreement Between the USA and the Russian Federation has on the Domestic Uranium Mining, Conversion, and Enrichment Industries and the Ops of the Gaseous Diffusion

    Broader source: Energy.gov [DOE]

    The successful implementation of the HEU Agreement remains a high priority of the U.S. Government. The agreement also serves U.S. and Russian commercial interests. HEU Agreement deliveries are an...

  15. Safeguarding a NWS International Enrichment Center as an Enriched Uranium Store

    SciTech Connect (OSTI)

    Curtis, Michael M.

    2008-03-31

    The operational and regulatory singularities of a multilateral facility designed to provide enriched uranium to a consortium of members may engender a new sub-category of safeguard criteria for the International Atomic Energy Agency (IAEA). This paper introduces the contingency of monitoring such a facility as a uranium storage center with cylinders containing low-enriched uranium (LEU) as the principal, and perhaps only, material open to verification. Accountancy and verification techniques will be proffered together with disparate means for maintaining continuity of knowledge (CoK) on verified stock.

  16. Design and calibration of the AWCC for measuring uranium hexafluoride

    SciTech Connect (OSTI)

    Wenz, T.R.; Menlove, H.O.; WSalton, G.; Baca, J.

    1995-08-01

    An Active Well Coincidence Counter (AWCC) has been modified to measure variable enrichment uranium hexafluoride (UF{sub 6}) in storage bottles. An active assay technique was used to measure the {sup 235}U content because of the small quantity (nominal loading of 2 kg UF{sub 6}) and nonuniform distribution of UF{sub 6} in the storage bottles. A new insert was designed for the AWCC composed of graphite containing four americium-lithium sources. Monte Carlo calculations were used to design the insert and to calibrate the detector. Benchmark measurements and calculations were performed using uranium oxide resulted in assay values that agreed within 2 to 3% of destructive assay values. In addition to UF{sub 6}, the detector was also calibrated for HEU ingots, billets, and alloy scrap using the standard Mode 1 end-plug configuration.

  17. Development of three-wavelength CCD image pyrometer used for the temperature field measurements of continuous casting billets

    SciTech Connect (OSTI)

    Xie, Zhi; Bai, Haicheng [State Key Laboratory of Synthetical Automation for Process Industries, School of Information Science and Engineering, Northeastern University, Shenyang 110819 (China)] [State Key Laboratory of Synthetical Automation for Process Industries, School of Information Science and Engineering, Northeastern University, Shenyang 110819 (China)

    2014-02-15

    This paper develops an imaging based three-color pyrometer for the monitoring of temperature distribution in a continuous casting billet. A novel optical device, together with an embedded electronic system, is designed to sequentially collect a dark image and three thermal images with specified wavelengths on a same monochromatic charge-coupled-device (CCD). The three thermal images provide the basis for the determination of target temperature, while the dark image is used to online eliminate the dark noise of CCD with a differential method. This image pyrometer is not only independent of target emissivity but also overcomes the dissimilarity of measuring accuracy between the micro-sensors of CCD resulted from the non-uniformity of pixels intensity response and the vignetting of optical system. Furthermore, a precise two-color temperature field measuring model on the CCD pyrometer is established, based on which a self-adaptive light-integration mechanism is presented. Compared with the traditional fixed light-integration method, the measuring range of the pyrometer is greatly extended and its sensitivity in low temperature segment is improved. The test results in a steel factory demonstrate that the pyrometer is capable of meeting the requirement of surface temperature measurements about casting billets. Reliability and accuracy of measurement results are also discussed herein.

  18. 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.

  19. recycled_uranium.cdr

    Office of Legacy Management (LM)

    supply of natural uranium. The chemical reprocessing of spent nuclear fuel for uranium was very efficient, but trace quantities of impurities accompanied the uranium product. ...

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. Method for improving the mechanical properties of uranium-1 to 3 wt % zirconium alloy

    DOE Patents [OSTI]

    Anderson, R.C.

    1983-11-22

    A uranium-1 to 3 wt % zirconium alloy characterized by high strength, high ductility and stable microstructure is fabricated by an improved thermal mechanical process. A homogenous ingot of the alloy which has been reduced in thickness of at least 50% in the two-step forging operation, rolled into a plate with a 75% reduction and then heated in vacuum at a temperature of about 750 to 850/sup 0/C and then quenched in water, is subjected to further thermal-mechanical operation steps to increase the compressive yield strength approximately 30%, stabilize the microstructure, and decrease the variations in mechanical properties throughout the plate is provided. These thermal-mechanical steps are achieved by cold rolling the quenchd plate to reduce the thickness thereof about 8 to 12%, aging the cold rolled plate at a first temperature of about 325 to 375/sup 0/C for five to six hours and then aging the plate at a higher temperature ranging from 480 to 500/sup 0/C for five to six hours prior to cooling the billet to ambient conditions and sizing the billet or plate into articles provides the desired increase in mechanical properties and phase stability throughout the plate.

  6. 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.

  7. Research Reactor Preparations for the Air Shipment of Highly Enriched Uranium from Romania

    SciTech Connect (OSTI)

    K. J. Allen; I. Bolshinsky; L. L. Biro; M. E. Budu; N. V. Zamfir; M. Dragusin; C. Paunoiu; M. Ciocanescu

    2010-03-01

    In June 2009 two air shipments transported both unirradiated (fresh) and irradiated (spent) Russian-origin highly enriched uranium (HEU) nuclear fuel from two research reactors in Romania to the Russian Federation for conversion to low enriched uranium. The Institute for Nuclear Research at Pitesti (SCN Pitesti) shipped 30.1 kg of HEU fresh fuel pellets to Dimitrovgrad, Russia and the Horia Hulubei National Institute of Physics and Nuclear Engineering (IFIN-HH) shipped 23.7 kilograms of HEU spent fuel assemblies from the VVR S research reactor at Magurele, Romania, to Chelyabinsk, Russia. Both HEU shipments were coordinated by the Russian Research Reactor Fuel Return Program (RRRFR) as part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), were managed in Romania by the National Commission for Nuclear Activities Control (CNCAN), and were conducted in cooperation with the Russian Federation State Corporation Rosatom and the International Atomic Energy Agency. Both shipments were transported by truck to and from respective commercial airports in Romania and the Russian Federation and stored at secure nuclear facilities in Russia until the material is converted into low enriched uranium. These shipments resulted in Romania becoming the 3rd country under the RRRFR program and the 14th country under the GTRI program to remove all HEU. This paper describes the research reactor preparations and license approvals that were necessary to safely and securely complete these air shipments of nuclear fuel.

  8. 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.

  9. 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.

  10. 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.

  11. Assessment of severe accident source terms in pressurized-water reactors with a 40% mixed-oxide and 60% low-enriched uranium core using MELCOR 1.8.5.

    SciTech Connect (OSTI)

    Gauntt, Randall O.; Goldmann, Andrew S.; Wagner, Kenneth C.; Powers, Dana Auburn; Ashbaugh, Scott G.; Longmire, Pamela

    2010-04-01

    As part of a Nuclear Regulatory Commission (NRC) research program to evaluate the impact of using mixed-oxide (MOX) fuel in commercial nuclear power plants, a study was undertaken to evaluate the impact of the usage of MOX fuel on the consequences of postulated severe accidents. A series of 23 severe accident calculations was performed using MELCOR 1.8.5 for a four-loop Westinghouse reactor with an ice condenser containment. The calculations covered five basic accident classes that were identified as the risk- and consequence-dominant accident sequences in plant-specific probabilistic risk assessments for the McGuire and Catawba nuclear plants, including station blackouts and loss-of-coolant accidents of various sizes, with both early and late containment failures. Ultimately, the results of these MELCOR simulations will be used to provide a supplement to the NRC's alternative source term described in NUREG-1465. Source term magnitude and timing results are presented consistent with the NUREG-1465 format. For each of the severe accident release phases (coolant release, gap release, in-vessel release, ex-vessel release, and late in-vessel release), source term timing information (onset of release and duration) is presented. For all release phases except for the coolant release phase, magnitudes are presented for each of the NUREG-1465 radionuclide groups. MELCOR results showed variation of noble metal releases between those typical of ruthenium (Ru) and those typical of molybdenum (Mo); therefore, results for the noble metals were presented for Ru and Mo separately. The collection of the source term results can be used as the basis to develop a representative source term (across all accident types) that will be the MOX supplement to NUREG-1465.

  12. 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 1970s. 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 provide additional confidence with the results. The actual corrosion rates of UMo fuel is very likely to be lower than assumed within this report which can be confirmed with additional testing.

  13. Domestic Uranium Production Report - Quarterly

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

    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 -...

  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. DOE - Office of Legacy Management -- R K Le Blond Machine Tool...

    Office of Legacy Management (LM)

    Facility used to conduct two billet drilling tests to demonstrating the capability of ... and Development Report (NLCO-886); Drilling Uranium Billets on a LeBlond-Carlstedt ...

  16. Kr Ion Irradiation Study of the Depleted-Uranium Alloys

    SciTech Connect (OSTI)

    J. Gan; D. Keiser; B. Miller; M. Kirk; J. Rest; T. Allen; D. Wachs

    2010-12-01

    Fuel development for the Reduced Enrichment Research and Test Reactor program is tasked with the development of new low-enriched uranium nuclear fuels that can be employed to replace existing highly enriched uranium fuels currently used in some research reactors throughout the world. For dispersion-type fuels, radiation stability of the fuel/cladding interaction product has a strong impact on fuel performance. Three depleted uranium alloys are cast for the radiation stability studies of the fuel/cladding interaction product using Kr ion irradiation to investigate radiation damage from fission products. SEM analysis indicates the presence of the phases of interest: U(Si, Al)3, (U, Mo)(Si, Al)3, UMo2Al20, U6Mo4Al43, and UAl4. Irradiations of TEM disc samples were conducted with 500 keV Kr ions at 200C to ion doses up to 2.5 1015 ions/cm2 (~ 10 dpa) with an Kr ion flux of 1012 ions/cm2-sec (~ 4.0 10-3 dpa/sec). Microstructural evolution of the phases relevant to fuel-cladding interaction products was investigated using transmission electron microscopy.

  17. 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.

  18. 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.

  19. 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.

  20. Method of increasing the phase stability and the compressive yield strength of uranium-1 to 3 wt. % zirconium alloy

    DOE Patents [OSTI]

    Anderson, Robert C.

    1986-01-01

    A uranium-1 to 3 wt. % zirconium alloy characterized by high strength, high ductility and stable microstructure is fabricated by an improved thermal mechanical process. A homogenous ingot of the alloy which has been reduced in thickness of at least 50% in the two-step forging operation, rolled into a plate with a 75% reduction and then heated in vacuum at a temperature of about 750.degree. to 850.degree. C. and then quenched in water is subjected to further thermal-mechanical operation steps to increase the compressive yield strength approximately 30%, stabilize the microstructure, and decrease the variations in mechanical properties throughout the plate is provided. These thermal-mechanical steps are achieved by cold rolling the quenched plate to reduce the thickness thereof about 8 to 12%, aging the cold rolled plate at a first temperature of about 325.degree. to 375.degree. C. for five to six hours and then aging the plate at a higher temperature ranging from 480.degree. to 500.degree. C. for five to six hours prior to cooling the billet to ambient conditions and sizing the billet or plate into articles provides the desired increase in mechanical properties and phase stability throughout the plate.

  1. Method for selective recovery of PET-usable quantities of [.sup.18 F] fluoride and [.sup.13 N] nitrate/nitrite from a single irradiation of low-enriched [.sup.18 O] water

    DOE Patents [OSTI]

    Ferrieri, Richard A.; Schlyer, David J.; Shea, Colleen

    1995-06-13

    A process for simultaneously producing PET-usable quantities of [.sup.13 N]NH.sub.3 and [.sup.18 F]F.sup.- for radiotracer synthesis is disclosed. The process includes producing [.sup.13 N]NO.sub.2.sup.- /NO.sub.3.sup.- and [.sup.18 F]F.sup.- simultaneously by exposing a low-enriched (20%-30%) [.sup.18 O]H.sub.2 O target to proton irradiation, sequentially isolating the [.sup.13 N]NO.sub.2.sup.- /NO.sub.3.sup.- and [.sup.18 F]F.sup.- from the [.sup.18 O]H.sub.2 O target, and reducing the [.sup.13 N]NO.sub.2.sup.- /NO.sub.3.sup.- to [.sup.13 N]NH.sub.3. The [.sup.13 N]NH.sub.3 and [.sup.18 F]F.sup.- products are then conveyed to a laboratory for radiotracer applications. The process employs an anion exchange resin for isolation of the isotopes from the [.sup.18 O]H.sub.2 O, and sequential elution of [.sup.13 N]NO.sub.2.sup.- /NO.sub.3.sup.- and [ .sup.18 F]F.sup.- fractions. Also the apparatus is disclosed for simultaneously producing PET-usable quantities of [.sup.13 N]NH.sub.3 and [.sup.18 F]F.sup.- from a single irradiation of a single low-enriched [.sup.18 O]H.sub.2 O target.

  2. Overview of transparency issues under the US-Russian highly enriched uranium purchase agreement

    SciTech Connect (OSTI)

    Bieniawski, A.J.; Dougherty, D.R.

    1995-12-31

    The US has signed an Agreement with the Russian Federation for the purchase of 500 metric tons of highly enriched uranium (HEU) derived from dismantled Russian nuclear weapons. The BEU will be blended down to low-enriched uranium (LEU) in Russia and will be transported to the US to be used by fuel Fabricators to make fuel for commercial nuclear power plants. Both the United States and Russia have been preparing to institute transparency measures to provide confidence that the nonproliferation, physical protection, and material control and accounting requirements specified in the Agreement are met. This paper provides a background on the Agreement and subsequent on-going negotiations to develop transparency measures suited to the facilities and processes which are expected to be involved.

  3. 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.

  4. 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.

  5. Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

    4. Uranium sellers to owners and operators of U.S. civilian nuclear power reactors, 2012-14 2012 2013 2014 Advance Uranium Asset Management Ltd. (was Uranium Asset Management) American Fuel Resources, LLC Advance Uranium Asset Management Ltd. American Fuel Resources, LLC AREVA NC, Inc. AREVA / AREVA NC, Inc. AREVA NC, Inc. BHP Billiton Olympic Dam Corporation Pty Ltd ARMZ (AtomRedMetZoloto) BHP Billiton Olympic Dam Corporation Pty Ltd CAMECO BHP Billiton Olympic Dam Corporation Pty Ltd CAMECO

  6. 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.

  7. U.S. Uranium Reserves Estimates

    Gasoline and Diesel Fuel Update (EIA)

    Major U.S. Uranium Reserves

  8. 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.

  9. 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.

  10. 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.

  11. 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 of the future, but has yet to be demonstrated commercially. In the early 1980s, six countries developing gas centrifuge technology (United States, United Kingdom, Germany, the Netherlands, Japan, and Australia) along with the International Atomic Energy Agency (IAEA) and the European Atomic Energy Community (EURATOM) began developing effective safeguards techniques for GCEPs. This effort was known as the Hexapartite Safeguards Project (HSP). The HSP had the goal of maximizing safeguards effectiveness while minimizing the cost to the operator and inspectorate, and adopted several recommendations, such as the acceptance of limited-frequency unannounced access (LFUA) inspections in cascade halls, and the use of nondestructive assay (NDA) measurements and tamper-indicating seals. While only the HSP participants initially committed to implementing all the measures of the approach, it has been used as a model for the safeguards applied to GCEPs in additional states. This report provides a snapshot overview of world enrichment capacity in 2007, including profiles of the uranium enrichment programs of individual states. It is based on open-source information, which is dependent on unclassified sources and may therefore not reflect the most recent developments. In addition, it briefly describes some of the safeguards techniques being used at various enrichment plants, including implementation of HSP recommendations.

  12. 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.

  13. Method for selective recovery of PET-usable quantities of [{sup 18}F] fluoride and [{sup 13}N] nitrate/nitrite from a single irradiation of low-enriched [{sup 18}O] water

    DOE Patents [OSTI]

    Ferrieri, R.A.; Schlyer, D.J.; Shea, C.

    1995-06-13

    A process for simultaneously producing PET-usable quantities of [{sup 13}N]NH{sub 3} and [{sup 18}F]F{sup {minus}} for radiotracer synthesis is disclosed. The process includes producing [{sup 13}N]NO{sub 2}{sup {minus}}/NO{sub 3}{sup {minus}}and [{sup 18}F]F{sup {minus}} simultaneously by exposing a low-enriched (20%-30%) [{sup 18}O]H{sub 2}O target to proton irradiation, sequentially isolating the [{sup 13}N]NO{sub 2}{sup {minus}}/NO{sub 3}{sup {minus}} and [{sup 18}F]F{sup {minus}} from the [{sup 18}O]H{sub 2}O target, and reducing the [{sup 13}N]NO{sub 2}{sup {minus}}/NO{sub 3}{sup {minus}} to [{sup 13}N]NH{sub 3}. The [{sup 13}N]NH{sub 3} and [{sup 18}F]F{sup {minus}} products are then conveyed to a laboratory for radiotracer applications. The process employs an anion exchange resin for isolation of the isotopes from the [{sup 18}O]H{sub 2}O, and sequential elution of [{sup 13}N]NO{sub 2}{sup {minus}}/NO{sub 3}{sup {minus}} and [{sup 18}F]F{sup {minus}} fractions. Also the apparatus is disclosed for simultaneously producing PET-usable quantities of [{sup 13}N]NH{sub 3} and [{sup 18}F]F{sup {minus}} from a single irradiation of a single low-enriched [{sup 18}O]H{sub 2}O target. 2 figs.

  14. 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

  15. Evaluation of impact tests of solid steel billet onto concrete pads, and application to generic ISFSI storage cask for tipover and side drop

    SciTech Connect (OSTI)

    Witte, M.C.; Chen, T.F.; Murty, S.S.; Tang, D.T.; Mok, G.C.; Fischer, L.E.; Carlson, R.W.

    1997-05-01

    Twelve tests were performed at LLNL to assess loading conditions on a spent fuel casts for side drops, end drops and tipover events. The tests were performed with a 1/3-scale model concrete pad to benchmark the structural analysis code DYNA3D. The side drop and tipover test results are discussed in this report. The billet and test pad were modified with DYNA3D using material properties and techniques used in earlier tests. The peak or maximum deceleration test results were compared to the simulated analytical results. It was concluded that an analytical model based on DYNA3D code and has been adequately benchmarked for this type of application. A generic or represented cask was modified with the DYNA3D code and evaluated for ISFSI side drop and tipover events. The analytical method can be applied to similar casks to estimate impact loads on storage casks resulting from low-velocity side or tip impacts onto concrete storage pads.

  16. 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.

  17. 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.

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

    DOE Patents [OSTI]

    Hayden, Jr., Howard W. (Oakridge, TN); Horton, James A. (Livermore, CA); Elliott, Guy R. B. (Los Alamos, NM)

    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.

  19. 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 Table S3a. Foreign purchases, foreign sales, and uranium ...

  20. Domestic Uranium Production Report

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

    1993-2014 Year Exploration and development surface drilling (million feet) Exploration and development drilling expenditures 1 (million dollars) Mine production of uranium ...

  1. Domestic Uranium Production Report

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

    1. U.S. uranium drilling activities, 2003-14 Exploration drilling Development drilling Exploration and development drilling Year Number of holes Feet (thousand) Number of holes ...

  2. 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.

  3. Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    or dissolving-out from mined rock, of the soluble uranium constituents by the natural action of percolating a prepared chemical solution through mounded (heaped) rock material. ...

  4. 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

  5. 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.

  6. 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.

  7. 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.

  8. 2015 Domestic Uranium Production Report

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

    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. uranium industry, 1993-2015 Exploration and Development Surface Exploration and Development Drilling Mine Production of Uranium Uranium Concentrate Production Uranium Concentrate Shipments Employment Year Drilling (million feet) Expenditures 1 (million dollars) (million pounds U 3 O 8 ) (million pounds U 3 O 8 )

  9. 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)

  10. 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...

  11. 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 thousand ...

  12. 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 ... Source: U.S. Energy Information Administration, Form EIA-858 "Uranium Marketing Annual ...

  13. 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 Deliveries 2011 2012 2013 2014 2015 Purchases of ...

  14. 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 Delivery year Total purchased (weighted- average price) ...

  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 Purchases Weighted- average price Purchases Weighted- ...

  16. 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 ...

  17. 2015 Uranium Marketing Annual Survey

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

    U.S. Energy Information Administration, Form EIA-858 ""Uranium Marketing Annual Survey"" (2015)." " U.S. Energy Information Administration 2015 Uranium Marketing Annual Report 15

  18. 2015 Uranium Marketing Annual Report

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

    U.S. Energy Information Administration, Form EIA-858 ""Uranium Marketing Annual Survey"" (2015)." " U.S. Energy Information Administration 2015 Uranium Marketing Annual Report 25

  19. 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 ...

  20. 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 ...

  1. 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 ...

  2. 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 ...

  3. 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 ...

  4. 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 Marketing Annual Report Release Date: May 24, 2016 Next Release Date: May 2017 Purchase ...

  5. 2015 Uranium Marketing Annual Report

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

    2 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 thousand ...

  6. 2015 Uranium Marketing Annual Report

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

    Energy Information Administration, Form EIA-858 ""Uranium Marketing Annual Survey"" (2013-15)." " U.S. Energy Information Administration 2015 Uranium Marketing Annual Report 33

  7. 2015 Uranium Marketing Annual Report

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

    8 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 ...

  8. 2015 Uranium Marketing Annual Report

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

    5 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 ...

  9. 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 ...

  10. 2015 Uranium Market Annual Report

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

    U.S. Energy Information Administration, Form EIA-858 ""Uranium Marketing Annual Survey"" (2015)." "16 U.S. Energy Information Administration 2015 Uranium Marketing Annual Report

  11. 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 ...

  12. 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 ...

  13. 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 ...

  14. 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 ...

  15. 2015 Uranium Marketing Annual Report

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

    Energy Information Administration: Form EIA-858 ""Uranium Marketing Annual Survey"" (2013-15)." " U.S. Energy Information Administration 2015 Uranium Marketing Annual Report 1

  16. 2015 Uranium Marketing Annual Report

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

    Energy Information Administration: Form EIA-858 ""Uranium Marketing Annual Survey"" (2013-15)." "14 U.S. Energy Information Administration 2015 Uranium Marketing Annual Report

  17. 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.

  18. 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/.

  19. 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)

  20. 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.

  1. 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.

  2. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    8. Uranium in fuel assemblies loaded into U.S. civilian nuclear power reactors by year, 2010-14 thousand pounds U3O8 equivalent Origin of uranium 2010 2011 2012 2013 P2014 Domestic-origin uranium 4,119 4,134 4,825 3,643 3,202 Foreign-origin uranium 40,187 46,809 44,657 39,000 47,281 Total 44,306 50,943 49,483 42,642 50,483 P = Preliminary data. Final 2013 fuel assembly data reported in the 2014 survey. Notes: Includes only unirradiated uranium in new fuel assemblies loaded into reactors during

  3. Uranium dioxide electrolysis

    DOE Patents [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.

  4. Effect of short-term material balances on the projected uranium measurement uncertainties for the gas centrifuge enrichment plant

    SciTech Connect (OSTI)

    Younkin, J.M.; Rushton, J.E.

    1980-02-05

    A program is under way to design an effective International Atomic Energy Agency (IAEA) safeguards system that could be applied to the Portsmouth Gas Centrifuge Enrichment Plant (GCEP). This system would integrate nuclear material accountability with containment and surveillance. Uncertainties in material balances due to errors in the measurements of the declared uranium streams have been projected on a yearly basis for GCEP under such a system in a previous study. Because of the large uranium flows, the projected balance uncertainties were, in some cases, greater than the IAEA goal quantity of 75 kg of U-235 contained in low-enriched uranium. Therefore, it was decided to investigate the benefits of material balance periods of less than a year in order to improve the sensitivity and timeliness of the nuclear material accountability system. An analysis has been made of projected uranium measurement uncertainties for various short-term material balance periods. To simplify this analysis, only a material balance around the process area is considered and only the major UF/sub 6/ stream measurements are included. That is, storage areas are not considered and uranium waste streams are ignored. It is also assumed that variations in the cascade inventory are negligible compared to other terms in the balance so that the results obtained in this study are independent of the absolute cascade inventory. This study is intended to provide information that will serve as the basis for the future design of a dynamic materials accounting component of the IAEA safeguards system for GCEP.

  5. 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.

  6. 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.

  7. Uranium Processing Facility | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Gallery Photo Gallery Jobs Apply for Our Jobs Our Jobs Working at NNSA Blog Home Uranium Processing Facility Uranium Processing Facility Uranium Processing Facility Site...

  8. Domestic Uranium Production Report

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

    3. U.S. uranium concentrate production, shipments, and sales, 2003-14 Activity at U.S. mills and In-Situ-Leach plants 2003 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014...

  9. Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    6. Employment in the U.S. uranium production industry by category, 2003-14 person-years Year Exploration Mining Milling Processing Reclamation Total 2003 W W W W 117 321 2004 18...

  10. Domestic Uranium Production Report

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

    7. Employment in the U.S. uranium production industry by state, 2003-14 person-years State(s) 2003 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 Wyoming 134 139 181 195...

  11. Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

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

  12. 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.

  13. 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.

  14. 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.

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

    DOE Patents [OSTI]

    Kreuzmann, Alvin B. (Cincinnati, OH)

    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.

  16. 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.

  17. 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

  18. 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 demonstrated commercially. In the early 1980s, six countries developing gas centrifuge technology (United States, United Kingdom, Germany, the Netherlands, Japan, and Australia) along with the International Atomic Energy Agency and the European Atomic Energy Community began developing effective safeguards techniques for GCEPs. This effort was known as the Hexapartite Safeguards Project (HSP). The HSP had the goal of maximizing safeguards effectiveness while minimizing the cost to the operator and inspectorate, and adopted several recommendations, such as the acceptance of limited-frequency unannounced access inspections in cascade halls, and the use of nondestructive assay measurements and tamper-indicating seals. While only the HSP participants initially committed to implementing all the measures of the approach, it has been used as a model for the safeguards applied to GCEPs in additional states. Uranium enrichment capacity has continued to expand on all fronts in the last few years. GCEP capacity is expanding in anticipation of the eventual shutdown of the less-efficient GDPs, the termination of the U.S.-Russia HEU blend-down program slated for 2013, and the possible resurgence of nuclear reactor construction as part of an expected 'Nuclear Renaissance'. Overall, a clear trend in the world profile of uranium enrichment plant operation is the continued movement towards multinational projects driven by commercial and economic interests. Along this vein, the safeguards community is continuing to develop new safeguards techniques and technologies that are not overly burdensome to enrichment plant operators while delivering more effective and efficient results. This report provides a snapshot overview of world enrichment capacity in 2009, including profiles of the uranium enrichment programs of individual states. It is a revision of a 2007 report on the same topic; significant changes in world enrichment programs between the previous and current reports are emphasized. It is based entirely on open-source information, which is dependent on published sources and may therefore not be completely accurate or reflect the most recent developments. Consequently, readers should not assume that information cited here has the endorsement of either ORNL or the U.S. Department of Energy. We are merely reporting what's been reported. In addition, this report briefly describes some of the safeguards techniques being used at various enrichment plants, including implementation of HSP recommendations.

  19. 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)

  20. 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.

  1. Extrusion-formed uranium-2.4 wt. % article with decreased linear thermal expansion and method for making the same

    DOE Patents [OSTI]

    Anderson, Robert C.; Jones, Jack M.; Kollie, Thomas G.

    1982-01-01

    The present invention is directed to the fabrication of an article of uranium-2.4 wt. % niobium alloy in which the linear thermal expansion in the direction transverse to the extrusion direction is less than about 0.98% between 22.degree. C. and 600.degree. C. which corresponds to a value greater than the 1.04% provided by previous extrusion operations over the same temperature range. The article with the improved thermal expansion possesses a yield strength at 0.2% offset of at least 400 MPa, an ultimate tensile strength of 1050 MPa, a compressive yield strength of at least 0.2% offset of at least 675 MPa, and an elongation of at least 25% over 25.4 mm/sec. To provide this article with the improved thermal expansion, the uranium alloy billet is heated to 630.degree. C. and extruded in the alpha phase through a die with a reduction ratio of at least 8.4:1 at a ram speed no greater than 6.8 mm/sec. These critical extrusion parameters provide the article with the desired decrease in the linear thermal expansion while maintaining the selected mechanical properties without encountering crystal disruption in the article.

  2. 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.

  3. 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.

  4. 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 2014 2015 2014 2015 2014 2015 Weighted-average price ...

  5. 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 ... Annual, Tables 28, 29, 30 and 31. 2003-15-Form EIA-858, "Uranium Marketing Annual Survey". ...

  6. 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 ...

  7. 2015 Uranium Marketing Annual Report

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

    2015 Uranium Marketing Annual Report Release Date: May 24, 2016 Next Release Date: May ... and 16. 2003-15-Form EIA-858, "Uranium Marketing Annual Survey". million pounds U 3 O 8 ...

  8. 2015 Uranium Marketing Annual Report

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

    5 2015 Uranium Marketing Annual Report Release Date: May 24, 2016 Next Release Date: May ... and 27. 2003-15-Form EIA-858, "Uranium Marketing Annual Survey". - No data reported. 0 ...

  9. 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.

  10. 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.

  11. 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)

  12. 2015 Domestic Uranium Production Report

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

    9. Summary production statistics of the U.S. uranium industry, 1993-2015" ,"Exploration and Development Surface ","Exploration and Development Drilling","Mine Production of Uranium ","Uranium Concentrate Production ","Uranium Concentrate Shipments ","Employment " "Year","Drilling (million feet)"," Expenditures 1 (million dollars)","(million pounds U3O8)","(million pounds

  13. 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.

  14. 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.

  15. 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.

  16. 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, 2010-14 thousand pounds U3O8 equivalent; dollars per pound U3O8 equivalent Deliveries 2010 2011 2012 2013 2014 U.S.-Origin Uranium Purchases 3,687 5,205 9,807 9,484 3,316 Weighted-Average Price 45.25 52.12 59.44 56.37 48.11 Foreign-Origin Uranium Purchases 42,895 49,626 47,713 47,919 50,033 Weighted-Average Price 49.64 55.98 54.07 51.13 46.03 Total Purchases 46,582 54,831 57,520

  17. 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, 2010-14 thousand pounds U3O8 equivalent; dollars per pound U3O8 equivalent Deliveries 2010 2011 2012 2013 2014 Received U.S.-origin uranium Purchases 2,226 1,668 1,194 W 410 Weighted-average price 43.36 54.85 51.78 W 33.55 Received foreign-origin uranium Purchases 27,186 24,695 24,606 W 28,743 Weighted-average price 41.42 49.69 47.75 W 38.42 Total received by U.S. brokers and traders Purchases 29,412 26,363

  18. 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, 2010-14 thousands pounds U3O8 equivalent; dollars per pound U3O8 equivalent Deliveries to foreign suppliers and utilities 2010 2011 2012 2013 2014 U.S.-origin uranium Foreign sales 3,440 4,387 4,798 4,148 4,210 Weighted-average price 37.82 53.08 47.53 43.10 32.91 Foreign-origin uranium Foreign sales 19,708 12,297 13,185 14,717 15,794 Weighted-Average Price

  19. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    2. Inventories of natural and enriched uranium by material type as of end of year, 2010-14 thousand pounds U3O8 equivalent Inventories at the end of the year Type of uranium inventory owned by 2010 2011 2012 2013 P2014 Owners and operators of U.S. civilian nuclear power reactors inventories 86,527 89,835 97,647 113,077 116,047 Uranium concentrate (U3O8) 13,076 14,718 15,963 18,131 20,501 Natural UF6 35,767 35,883 29,084 38,332 40,972 Enriched UF6 25,392 19,596 38,428 40,841 44,605 Fabricated

  20. 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.

  1. 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)

  2. 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 mineral phases. Four case studies are presented: Water and Soil Characterization, Subsurface Stabilization of Uranium and other Toxic Metals, Reductive Precipitation (in situ bioremediation) of Uranium, and Physical Transport of Particle-bound Uranium by Erosion.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

    a. Foreign purchases, foreign sales, and uranium inventories owned by U.S. suppliers and owners and operators of U.S. civilian nuclear power reactors, 1994-2014 million pounds U3O8 equivalent Delivery year Foreign purchases by U.S. suppliers Foreign purchases by owners and operators of U.S. civilian nuclear power reactors Total foreign purchases U.S. broker and trader purchases from foreign suppliers Foreign sales U.S. supplier owned uranium inventories Owners and operators of U.S. civilian

  8. 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.

  9. Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

    5. U.S. uranium in-situ-leach plants by owner, location, capacity, and operating status at end of the year, 2011-15 In-Situ-Leach plant owner In-Situ-Leach plant name County, state (existing and planned locations) Production capacity (pounds U3O8 per year) Operating status at end of the year 2011 2012 2013 2014 2015 AUC LLC Reno Creek Campbell, Wyoming 2,000,000 - - Developing Developing Partially Permitted and Licensed Azarga Uranium Corp Dewey Burdock Project Fall River and Custer, South

  10. 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.

  11. Uranium Lease Tracts Location Map | Department of Energy

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

    Uranium Lease Tracts Location Map Uranium Lease Tracts Location Map Uranium Lease Tracts Location Map PDF icon Uranium Lease Tracts Location Map More Documents & Publications ...

  12. 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.

  13. 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)

  14. 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.

  15. 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.

  16. Domestic U.S. Reactor Conversions: Fact Sheet | National Nuclear...

    National Nuclear Security Administration (NNSA)

    the use of highly-enriched uranium (HEU) fuel to low-enriched uranium (LEU) as a key ... the use of highly-enriched uranium (HEU) fuel to low-enriched uranium (LEU) as a key ...

  17. 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.

  18. 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.

  19. 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,

  20. Air Shipment of Highly Enriched Uranium Spent Nuclear Fuel from Romania

    SciTech Connect (OSTI)

    K. J. Allen; I. Bolshinsky; L. L. Biro; M. E. Budu; N. V. Zamfir; M. Dragusin

    2010-07-01

    Romania safely air shipped 23.7 kilograms of Russian origin highly enriched uranium (HEU) spent nuclear fuel from the VVR S research reactor at Magurele, Romania, to the Russian Federation in June 2009. This was the worlds first air shipment of spent nuclear fuel transported in a Type B(U) cask under existing international laws without special exceptions for the air transport licenses. This shipment was coordinated by the Russian Research Reactor Fuel Return Program (RRRFR), part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), in cooperation with the Romania National Commission for Nuclear Activities Control (CNCAN), the Horia Hulubei National Institute of Physics and Nuclear Engineering (IFIN-HH), and the Russian Federation State Corporation Rosatom. The shipment was transported by truck to and from the respective commercial airports in Romania and the Russian Federation and stored at a secure nuclear facility in Russia where it will be converted into low enriched uranium. With this shipment, Romania became the 3rd country under the RRRFR program and the 14th country under the GTRI program to remove all HEU. This paper describes the work, equipment, and approvals that were required to complete this spent fuel air shipment.

  1. Improved Irradiation Performance of Uranium-Molybdenum/Aluminum Dispersion Fuel by Silicon Addition in Aluminum

    SciTech Connect (OSTI)

    Yeon Soo Kim; G. L. Hofman; A. B. Robinson; D. M. Wachs

    2013-10-01

    Uranium-molybdenum fuel particle dispersion in aluminum is a form of fuel under development for conversion of high-power research and test reactors from highly enriched to low-enriched uranium in the U.S. Global Threat Reduction Initiative program (also known as the Reduced Enrichment for Research and Test Reactors program). Extensive irradiation tests have been conducted to find a solution for problems caused by interaction layer growth and pore formation between U-Mo and Al. Adding a small amount of Si (up to [approximately]5 wt%) in the Al matrix was one of the proposed remedies. The effect of silicon addition in the Al matrix was examined using irradiation test results by comparing side-by-side samples with different Si additions. Interaction layer growth was progressively reduced with increasing Si addition to the matrix Al, up to 4.8 wt%. The Si addition also appeared to delay pore formation and growth between the U-Mo and Al.

  2. 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.

  3. Disposition of Uranium Oxide From Conversion of Depleted Uranium Hexafluoride

    Broader source: Energy.gov [DOE]

    This Supplemental Environmental Impact Statement (SEIS) for Disposition of Uranium Oxide Conversion Product Generated from Conversion of DOE’s Inventory of Depleted Uranium Hexafluoride [DOE/EIS-0359-S1 and DOE/EIS-0360-S1] evaluates the environmental impacts resulting from the disposition of up to 800,000 metric tons of uranium oxide resulting from the conversion of depleted uranium hexafluoride (DUF6) at the Department’s two operating DUF6 conversion facilities in Paducah, Kentucky and Portsmouth, Ohio.

  4. 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.

  5. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    3. Inventories of uranium by owner as of end of year, 2010-14 thousand pounds U3O8 equivalent Inventories at the end of the year Owner of uranium inventory 2010 2011 2012 2013 P2014 Owners and operators of U.S. civilian nuclear power reactors 86,527 89,835 97,647 113,007 116,047 U.S. brokers and traders 11,125 6,841 5,677 7,926 5,798 U.S. converter, enrichers, fabricators, and producers 13,608 15,428 17,611 13,416 12,766 Total commercial inventories 111,259 112,104 120,936 134,418 134,611 P =

  6. 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.

  7. 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.

  8. 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, 2012-14 deliveries thousand pounds U3O8 equivalent; dollars per pound U3O8 equivalent Deliveries in 2012 Deliveries in 2013 Deliveries in 2014 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

  9. 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.

  10. 2015 Uranium Marketing Annual Report

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

    2 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 2013 2014 2015 American Fuel Resources, LLC Advance Uranium Asset Management Ltd. AREVA / AREVA NC, Inc. AREVA NC, Inc. AREVA / AREVA NC, Inc. ARMZ (AtomRedMetZoloto) BHP Billiton Olympic Dam Corporation Pty Ltd ARMZ (AtomRedMetZoloto) BHP Billiton Olympic Dam Corporation Pty Ltd CAMECO BHP Billiton Olympic Dam Corporation Pty

  11. 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.

  12. 2014 Domestic Uranium Production Report

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

    Domestic Uranium Production Report 2015 Domestic Uranium Production Report Release Date: May 5, 2016 Next Release Date: May 2017 2011 2012 2013 2014 2015 AUC LLC Reno Creek Campbell, Wyoming 2,000,000 - - Developing Developing Partially Permitted and Licensed Azarga Uranium Corp Dewey Burdock Project Fall River and Custer, South Dakota 1,000,000 Undeveloped Developing Developing Partially Permitted And Licensed Partially Permitted And Licensed Cameco Crow Butte Operation Dawes, Nebraska

  13. 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....

  14. 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, ...

  15. 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...

  16. Domestic Uranium Production Report - Quarterly

    Gasoline and Diesel Fuel Update (EIA)

    or dissolving-out from mined rock, of the soluble uranium constituents by the natural action of percolating a prepared chemical solution through mounded (heaped) rock material. ...

  17. 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.

  18. 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.

  19. 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.

  20. Highly Enriched Uranium Materials Facility | National Nuclear...

    National Nuclear Security Administration (NNSA)

    Gallery Jobs Apply for Our Jobs Our Jobs Working at NNSA Blog Home Highly Enriched Uranium Materials Facility Highly Enriched Uranium Materials Facility Congressmen tour Y-12...

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

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

    for DOE's Uranium Leasing Program, under which DOE administers tracts of land in western Colorado for exploration, development, and the extraction of uranium and vanadium ores. ...

  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 Apparatus, systems, and methods for...

  3. 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...

  4. 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.

  5. 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...

  6. 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...

  7. 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.

  8. Extrusion-formed uranium-2. 4 wt % article with decreased linear thermal expansion and method for making the same. [Patent application

    DOE Patents [OSTI]

    Anderson, R.C.; Jones, J.M.; Kollie, T.G.

    1982-05-24

    The present invention is directed to the fabrication of an article of uranium-2.4 wt % niobium alloy in which the linear thermal expansion in the direction transverse to the extrusion direction is less than about 0.98% between 22 and 600/sup 0/C which corresponds to a value greater than the 1.04% provided by previous extrusion operations over the same temperature range. The article with the improved thermal expansion possesses a yield strength at 0.2% offset of at least 400 MPa, an ultimate tensile strength of 1050 MPa, a compressive yield strength of at least 0.2% offset of at least 675 MPa, and an elongation of at least 25% over 25.4 mm/s. To provide this article with the improved thermal expansion, the uranium alloy billet is heated to 630/sup 0/C and extruded in the alpha phase through a die with a reduction ratio of at least 8.4:1 at a ram speed no greater than 6.8 mm/s. These critical extrusion parameters provide the article with a desired decrease in the linear thermal expansion while maintaining the selected mechanical properties without encountering crystal disruption in the article.

  9. 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.

  10. 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/.

  11. 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.

  12. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    0. Contracted purchases of uranium from suppliers by owners and operators of U.S. civilian nuclear power reactors, in effect at the end of 2014, by delivery year, 2015-24 thousand pounds U3O8 equivalent Contracted purchases from U.S. suppliers Contracted purchases from foreign suppliers Contracted purchases from all suppliers Year of delivery Minimum Maximum Minimum Maximum Minimum Maximum 2015 8,405 8,843 31,468 34,156 39,873 42,999 2016 7,344 7,757 29,660 31,787 37,004 39,544 2017 5,980 6,561

  13. 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, 2012-14 thousand pounds U3O8 equivalent Deliveries in 2012 Deliveries in 2013 Deliveries in 2014 Origin country of feed U.S. enrichment Foreign enrichment Total U.S. enrichment Foreign enrichment Total U.S. enrichment Foreign enrichment Total Australia 3,195 3,352 6,547 2,417 2,476 4,893 910 4,467 5,377 Brazil 0 0 0 0 W W 0 W W Canada 6,741 5,007

  14. 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, 2010-14 thousand pounds U3O8 equivalent; dollars per pound U3O8 equivalent Deliveries in 2010 Deliveries in 2011 Deliveries in 2012 Deliveries in 2013 Deliveries in 2014 Origin country Purchases Weighted-average price Purchases Weighted-average price Purchases Weighted-average price Purchases Weighted-average price Purchases Weighted-average price Australia 7,112 51.35 6,001

  15. 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, 2013-14 dollars per pound U3O8 equivalent; thousand pounds U3O8 equivalent Pricing mechanisms Domestic purchases1 Foreign purchases2 Total purchases 2013 2014 2013 2014 2013 2014 Contract-specified (fixed and base-escalated) pricing Weighted-average price 54.95 41.87 55.03 49.87 54.99 45.47 Quantity with reported price 14,530 15,711 14,732

  16. 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, 2012-14 deliveries thousand pounds U3O8 equivalent; dollars per pound U3O8 equivalent Deliveries in 2012 Deliveries in 2013 Deliveries in 2014 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 7,119 38.24 7,175 34.24 6,665 30.26

  17. 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, 2014 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 8,440 38.38 20,820 47.57 29,260 44.92 Natural UF6 4,405 35.30 13,373 53.13

  18. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    9. Contracted purchases of uranium by owners and operators of U.S. civilian nuclear power reactors, signed in 2014, by delivery year, 2015-24 thousand pounds U3O8 equivalent Year of Delivery Minimum Maximum 2015 2,838 2,838 2016 3,573 3,573 2017 2,718 2,818 2018 W 2,628 2019 W W 2020 W W 2021 W W 2022 W W 2023 W W 2024 W W Total 13,991 15,591 W = Data withheld to avoid disclosure of individual company data. Note: Totals may not equal sum of components because of independent rounding

  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. Quantification of the Potential Impact on Commercial Markets...

    Office of Environmental Management (EM)

    Introduction of the Enrichment Services Component of DOE Low Enriched Uranium Inventory ... Markets of Introduction of the Enrichment Services Component of DOE Low Enriched ...

  1. 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)

  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. Uranium Downblending and Disposition Project Technology Readiness

    Energy Savers [EERE]

    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 PDF icon Uranium Downblending and Disposition Project Technology Readiness Assessment PDF icon Summary - Uranium233 Downblending and Disposition Project More Documents & Publications Compilation of TRA Summaries EA-1574: Final Environmental

  4. 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.

  5. 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)

  6. 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.

  7. 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...

  8. Highly Enriched Uranium Materials Facility | Y-12 National Security...

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

    Highly Enriched Uranium ... Highly Enriched Uranium Materials Facility HEUMF The Highly Enriched Uranium Materials Facility is our nation's central repository for highly enriched...

  9. 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.

  10. 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,...

  11. 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.

  12. 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.

  13. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    2. Maximum anticipated uranium market requirements of owners and operators of U.S. civilian nuclear power reactors, 2015-24, as of December 31, 2014 thousand pounds U3O8 equivalent Year Maximum Under Purchase Contracts Unfilled Market Requirements Maximum Anticipated Market Requirements Enrichment Feed Deliveries 2015 42,999 3,496 46,494 48,206 2016 39,544 7,384 46,929 46,529 2017 31,257 10,351 41,608 49,924 2018 26,001 18,468 44,469 51,169 2019 19,096 29,929 49,025 46,184 2020 13,308 33,521

  14. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    5. Shipments of uranium feed by owners and operators of U.S. civilian nuclear power reactors to domestic and foreign enrichment suppliers, 2015-24 thousand pounds U3O8 equivalent Amount of feed to be shipped Change from 2013 to 2014 Year of shipment As of December 31, 2013 As of December 31, 2014 Annual Cumulative 2015 45,498 48,206 2,708 2,708 2016 48,693 46,529 -2,164 544 2017 47,005 49,924 2,919 3,463 2018 52,138 51,169 -969 2,494 2019 50,041 46,184 -3,857 -1,363 2020 49,726 49,598 -128

  15. 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, 2010-14 thousand pounds U3O8 equivalent; dollars per pound U3O8 equivalent Deliveries 2010 2011 2012 2013 2014 U.S. suppliers Foreign purchases 24,985 19,318 20,196 23,233 24,199 Weighted-average price 41.30 48.80 46.80 43.25 39.13 Owners and operators of U.S. civilian nuclear power reactors Foreign purchases 30,362 35,071 36,037 34,095 34,404 Weighted-average

  16. 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...

  17. Domestic Uranium Production Report - Quarterly

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

    1. Total production of uranium concentrate in the United States, 1996 - 3rd quarter 2015 pounds U3O8 Calendar-year quarter 1st quarter 2nd quarter 3rd quarter 4th quarter...

  18. 2015 Uranium Marketing Annual Report

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

    Uranium Marketing Annual Report May 2016 Independent Statistics & Analysis www.eia.gov U.S. Department of Energy Washington, DC 20585 U.S. Energy Information Administration | 2015 Uranium Marketing Annual Report i This report was prepared by the U.S. Energy Information Administration (EIA), the statistical and analytical agency within the U.S. Department of Energy. By law, EIA's data, analyses, and forecasts are independent of approval by any other officer or employee of the United States

  19. 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.

  20. 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.

  1. 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.

  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. 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.

  4. 2015 Domestic Uranium Production Report

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

    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 Shootaring Canyon Uranium Mill Garfield, Utah 750 Standby Standby Standby Standby Standby EFR White Mesa LLC White Mesa Mill San Juan, Utah 2,000 Operating Operating Operating- Processing Alternate Feed Operating- Processing Alternate Feed Operating- Processing Alternate Feed Energy Fuels Wyoming Inc Sheep Mountain

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. Joint DOE-Rosatom Statement | Department of Energy

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

    Report on the Effect the Low Enriched Uranium Delivered Under the Highly Enriched Uranium Agreement Between the USA and the Russian Federation has on the Domestic Uranium Mining, ...

  10. EIS-0471: Final Environmental Impact Statement | Department of...

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

    EIS-0468: Final Environmental Impact Statement EA-1607: Final Environmental Assessment Disposition of DOE Excess Depleted Uranium, Natural Uranium, and Low-Enriched Uranium...

  11. 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

  12. 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.

  13. 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.

  14. 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.

  15. Inherently safe in situ uranium recovery

    DOE Patents [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. 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.

  17. Uranium Mining and Milling near Rifle, Colorado

    Broader source: Energy.gov [DOE]

    The small town of Rifle, Colorado, has an interesting history related to uranium and vanadium production. A mineral found near Rifle, called roscolite, contains both vanadium and uranium but was...

  18. 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.)

  19. 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.

  20. 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

  1. 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.

  2. 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.

  3. 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 thousand pounds U 3 O 8 equivalent 2011 2012 2013 2014 P2015 Owners and operators of U.S. civilian nuclear power reactors inventories 89,835 97,647 113,077 114,046 120,857 Uranium concentrate (U 3 O 8 ) 14,718 15,963 18,131 19,060 20,635 Natural UF 6 35,883 29,084 38,332 40,803 47,253 Enriched UF 6 19,596 38,428 40,841 43,382

  4. 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.

  5. 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)

  6. 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.

  7. 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.

  8. 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.

  9. 2015 Uranium Marketing Annual Report

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

    4. 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 / AREVA NC, Inc." "AREVA NC, Inc.","AREVA / AREVA NC, Inc.","ARMZ (AtomRedMetZoloto)" "BHP Billiton Olympic Dam Corporation Pty Ltd","ARMZ (AtomRedMetZoloto)","BHP Billiton Olympic Dam Corporation Pty Ltd"

  10. 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.

  11. Domestic Uranium Production Report - Quarterly

    Gasoline and Diesel Fuel Update (EIA)

    4. U.S. uranium in-situ-leach plants by owner, location, capacity, and operating status Operating status at the end of In-situ-leach plant owner In-situ-leach plant name County, state (existing and planned locations) Production capacity (pounds U3O8 per year) 2015 1st quarter 2016 AUC LLC Reno Creek Campbell, Wyoming 2,000,000 Partially Permitted And Licensed Partially Permitted And Licensed Azarga Uranium Corp. Dewey Burdock Project Fall River and Custer, South Dakota 1,000,000 Partially

  12. 2015 Domestic Uranium Production Report

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

    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 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

  13. 2015 Domestic Uranium Production Report

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

    Domestic Uranium Production Report 2015 Domestic Uranium Production Report Release Date: May 5, 2016 Next Release Date: May 2017 million pounds U 3 O 8 $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 Drilling W 38.2 W W 38.2 W Mines in Production W 19.2 W

  14. 2015 Domestic Uranium Production Report

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

    10. Uranium reserve estimates at the end of 2014 and 2015" "million pounds U3O8" ,"End of 2014",,,"End of 2015" "Uranium Reserve Estimates1 by Mine and Property Status, Mining Method, and State(s)","Forward Cost 2" ,"$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

  15. Secretarial Determination Pursuant to USEC Privatization Act for the Sale

    Energy Savers [EERE]

    or Transfer of Low-Enriched Uranium | Department of Energy USEC Privatization Act for the Sale or Transfer of Low-Enriched Uranium Secretarial Determination Pursuant to USEC Privatization Act for the Sale or Transfer of Low-Enriched Uranium Secretarial determination regarding the potential impacts of the transfer by DOE of up to 48 metric tons of low-enriched uranium to USEC Inc. in exchange for DOE receiving approximately 409 metric tons of uranium hexafluoride, the equivalent amount of

  16. 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 of twin boundaries formed during the martensitic transformation. Deformation actually results in a coarsening of the microstructure making EBSD more practical following a limited amount of strain. Figure 2 shows the microstructure resulting from 6% compression. Casting of U-10Mo results in considerable chemical segregation as is apparent in Figure 2a. The segregation subsists through rolling and heat treatment processes as shown in Figure 2b. EBSD was used to study the effects of homogenization time and temperature on chemical heterogeneity. It was found that times and temperatures that result in a chemically homogeneous microstructure also result in a significant increase in grain size. U-0.75Ti forms an acicular martinsite as shown in Figure 4. This microstructure prevails through cycling into the higher temperature solid uranium phases.

  17. 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.

  18. 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

  19. 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.

  20. Process for alloying uranium and niobium

    DOE Patents [OSTI]

    Holcombe, Cressie E. (Farragut, TN); Northcutt, Jr., Walter G. (Oak Ridge, TN); Masters, David R. (Knoxville, TN); Chapman, Lloyd R. (Knoxville, TN)

    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.

  1. 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.

  2. 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

  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. 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.

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

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

    Uranium Processing Facility Uranium Processing Facility UPF will be a state-of-the-art, consolidated facility for enriched uranium operations including assembly, disassembly, dismantlement, quality evaluation, and product certification. The National Nuclear Security Administration is constructing a modern Uranium Processing Facility designed specifically for processes not suitable for relocation into existing buildings at Y-12. Originally designed to house all Enriched Uranium processing

  6. Toxic Substances Control Act Uranium Enrichment Federal Facility Compliance

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

    Agreement | Department of Energy Toxic Substances Control Act Uranium Enrichment Federal Facility Compliance Agreement Toxic Substances Control Act Uranium Enrichment Federal Facility Compliance Agreement Toxic Substances Control Act (TSCA) Uranium Enrichment Federal Facility Compliance Agreement establishes a plan to bring DOE's Uranium Enrichment Plants (and support facilities) located in Portsmouth, Ohio and Paducah, Kentucky and DOE's former Uranium Enrichment Plant (and support

  7. 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.

  8. DOE - Office of Legacy Management -- Kaiser Aluminum Corp - IL...

    Office of Legacy Management (LM)

    Operations: Performed limited duration work extruding uranium billets into three CP-5 fuel elements, circa 1959. IL.19-2 Site Disposition: Eliminated - Potential for ...

  9. DOE - Office of Legacy Management -- Babcock and Wilcox Co -...

    Office of Legacy Management (LM)

    work to pierce uranium billets for extrusion to tubes. No indication that the piercing operation was conducted. PA.18-1 Site Disposition: Eliminated - No indication that...

  10. 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.

  11. 2015 Domestic Uranium Production Report

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

    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 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

  12. 2015 Domestic Uranium Production Report

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

    Domestic Uranium Production Report 2015 Domestic Uranium Production Report Release Date: May 5, 2016 Next Release Date: May 2017 Activity at U.S. Mills and In-Situ-Leach Plants 2003 2004 2005 2006 2007 2008 2009 2010 2011 2012 2013 2014 2015 Ore from Underground Mines and Stockpiles Fed to Mills 1 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 (thousand pounds U 3 O 8 ) W W W W W W W W W W W W W (thousand pounds U 3 O 8 ) W W

  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. 2015 Domestic Uranium Production Report

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

    Domestic Uranium Production Report May 2016 Independent Statistics & Analysis www.eia.gov U.S. Department of Energy Washington, DC 20585 This report was prepared by the U.S. Energy Information Administration (EIA), the statistical and analytical agency within the U.S. Department of Energy. By law, EIA's data, analyses, and forecasts are independent of approval by any other officer or employee of the United States Government. The views in this report therefore should not be construed as

  16. 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.

  17. 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.

  18. Uranium Mining, Conversion, and Enrichment Industries

    Energy Savers [EERE]

    i Analysis of Potential Impacts of Uranium Transfers on the Domestic Uranium Mining, Conversion, and Enrichment Industries May 1, 2015 ii EXECUTIVE SUMMARY: The Department of Energy ("Department" or "DOE") plans to transfer the equivalent of up to 2,100 metric tons ("MTU") of natural uranium per year (with a higher total for calendar year 2015, mainly because of transfers already executed or under way before today's determination). These transfers would include

  19. 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.

  20. Uranium Leasing Program | Department of Energy

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

    » Uranium Leasing Program Uranium Leasing Program Abandoned Mine Reclamation, Uravan Mineral Belt, Colorado Abandoned Mine Reclamation, Uravan Mineral Belt, 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

  1. 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)

  2. Uranium Weapons Components Successfully Dismantled | National Nuclear

    National Nuclear Security Administration (NNSA)

    Security Administration 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

  3. 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.

  4. Consent Order, Uranium Disposition Services, LLC - NCO-2010-01...

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

    Uranium Disposition Services, LLC - NCO-2010-01 Consent Order, Uranium Disposition Services, LLC - NCO-2010-01 March 26, 2010 Issued to Uranium Disposition Services, LLC related to ...

  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

    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.

  7. U.S. Uranium Reserves Estimates

    Gasoline and Diesel Fuel Update (EIA)

    The initial uranium property reserves estimates were based on bore hole radiometric data validated by chemical analysis of samples from cores and drill cuttings. The thickness of ...

  8. 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.

  9. 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.

  10. 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)

  11. 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.

  12. High strength and density tungsten-uranium alloys

    DOE Patents [OSTI]

    Sheinberg, Haskell (Los Alamos, NM)

    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.

  13. Uranium Nitride: Enabling New Applications for TRISO Fuel Particles...

    Office of Scientific and Technical Information (OSTI)

    Uranium Nitride: Enabling New Applications for TRISO Fuel Particles Citation Details In-Document Search Title: Uranium Nitride: Enabling New Applications for TRISO Fuel Particles ...

  14. Secretarial Determination for the Sale or Transfer of Uranium...

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

    Secretarial Determination for the Sale or Transfer of Uranium Secretarial Determination for the Sale or Transfer of Uranium Secretarial Determination for the Sale or Transfer of...

  15. Sequestering Uranium from Seawater: Binding Strength and Modes...

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

    Sequestering Uranium from Seawater: Binding Strength and Modes of Uranyl Complexes with Glutarimidedioxime Sequestering Uranium from Seawater: Binding Strength and Modes of Uranyl...

  16. Legacy Management Work Progresses on Defense-Related Uranium...

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

    defense-related legacy uranium mine sites located within 11 uranium mining districts in 6 western states. At these sites, photographs and global positioning location data were...

  17. Potential of Melastoma malabathricum as bio-accumulator for uranium...

    Office of Scientific and Technical Information (OSTI)

    > 1 for uranium in the leaf, stem and roots, indicating accumulation of uranium from soil. ... Institute of Science, Universiti Teknologi MARA, 40450 Shah Alam (Malaysia) (Malaysia) ...

  18. Mined Land Reclamation on DOE's Uranium Lease Tracts, Southwestern...

    Office of Environmental Management (EM)

    Mined Land Reclamation on DOE's Uranium Lease Tracts, Southwestern Colorado Mined Land Reclamation on DOE's Uranium Lease Tracts, Southwestern Colorado Mined Land Reclamation on...

  19. DOE Extends Contract to Operate Depleted Uranium Hexafluoride...

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

    Extends Contract to Operate Depleted Uranium Hexafluoride Conversion Plants DOE Extends Contract to Operate Depleted Uranium Hexafluoride Conversion Plants December 24, 2015 - ...

  20. DOE Seeks Contractor for Depleted Uranium Hexafluoride (DUF6...

    Office of Environmental Management (EM)

    Contractor for Depleted Uranium Hexafluoride (DUF6) Operations at Ohio and Kentucky Facilities DOE Seeks Contractor for Depleted Uranium Hexafluoride (DUF6) Operations at Ohio and ...

  1. Uranium and thorium complexes of the phosphaethynolate ion (Journal...

    Office of Scientific and Technical Information (OSTI)

    Uranium and thorium complexes of the phosphaethynolate ion Citation Details In-Document Search Title: Uranium and thorium complexes of the phosphaethynolate ion You are ...

  2. Decommissioning of U.S. Uranium Production Facilities

    Reports and Publications (EIA)

    1995-01-01

    This report analyzes the uranium production facility decommissioning process and its potential impact on uranium supply and prices. 1995 represents the most recent publication year.

  3. Highly Enriched Uranium Materials Facility, Major Design Changes...

    Energy Savers [EERE]

    Highly Enriched Uranium Materials Facility, Major Design Changes Late...Lessons Learned Report, NNSA, Dec 2010 Highly Enriched Uranium Materials Facility, Major Design Changes...

  4. Uranium Sequestration via Phosphate Infiltration/Injection Test...

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

    Uranium Sequestration via Phosphate InfiltrationInjection Test History Supporting the Preferred Alternative 1 300 Area GW Concentrations - Uranium High River Stage - GW...

  5. DOE Extends Public Comment Period for Uranium Program Environmental...

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

    Uranium Program Environmental Impact Statement DOE Extends Public Comment Period for Uranium Program Environmental Impact Statement April 18, 2013 - 1:08pm Addthis Contractor, Bob ...

  6. DOE Extends Public Comment Period for the Draft Uranium Leasing...

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

    the Draft Uranium Leasing Program Programmatic Environmental Impact Statement DOE Extends Public Comment Period for the Draft Uranium Leasing Program Programmatic Environmental ...

  7. Record of Decision for the Uranium Leasing Program Programmatic...

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

    Record of Decision for the Uranium Leasing Program Programmatic Environmental Impact Statement Record of Decision for the Uranium Leasing Program Programmatic Environmental Impact ...

  8. Toxic Substances Control Act Uranium Enrichment Federal Facility...

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

    Toxic Substances Control Act Uranium Enrichment Federal Facility Compliance Agreement Toxic Substances Control Act Uranium Enrichment Federal Facility Compliance Agreement Toxic ...

  9. Decay Heat Calculations for PWR and BWR Assemblies Fueled with Uranium and Plutonium Mixed Oxide Fuel using SCALE

    SciTech Connect (OSTI)

    Ade, Brian J; Gauld, Ian C

    2011-10-01

    In currently operating commercial nuclear power plants (NPP), there are two main types of nuclear fuel, low enriched uranium (LEU) fuel, and mixed-oxide uranium-plutonium (MOX) fuel. The LEU fuel is made of pure uranium dioxide (UO{sub 2} or UOX) and has been the fuel of choice in commercial light water reactors (LWRs) for a number of years. Naturally occurring uranium contains a mixture of different uranium isotopes, primarily, {sup 235}U and {sup 238}U. {sup 235}U is a fissile isotope, and will readily undergo a fission reaction upon interaction with a thermal neutron. {sup 235}U has an isotopic concentration of 0.71% in naturally occurring uranium. For most reactors to maintain a fission chain reaction, the natural isotopic concentration of {sup 235}U must be increased (enriched) to a level greater than 0.71%. Modern nuclear reactor fuel assemblies contain a number of fuel pins potentially having different {sup 235}U enrichments varying from {approx}2.0% to {approx}5% enriched in {sup 235}U. Currently in the United States (US), all commercial nuclear power plants use UO{sub 2} fuel. In the rest of the world, UO{sub 2} fuel is still commonly used, but MOX fuel is also used in a number of reactors. MOX fuel contains a mixture of both UO{sub 2} and PuO{sub 2}. Because the plutonium provides the fissile content of the fuel, the uranium used in MOX is either natural or depleted uranium. PuO{sub 2} is added to effectively replace the fissile content of {sup 235}U so that the level of fissile content is sufficiently high to maintain the chain reaction in an LWR. Both reactor-grade and weapons-grade plutonium contains a number of fissile and non-fissile plutonium isotopes, with the fraction of fissile and non-fissile plutonium isotopes being dependent on the source of the plutonium. While only RG plutonium is currently used in MOX, there is the possibility that WG plutonium from dismantled weapons will be used to make MOX for use in US reactors. Reactor-grade plutonium in MOX fuel is generally obtained from reprocessed irradiated nuclear fuel, whereas weapons-grade plutonium is obtained from decommissioned nuclear weapons material and thus has a different plutonium (and other actinides) concentration. Using MOX fuel instead of UOX fuel has potential impacts on the neutronic performance of the nuclear fuel and the design of the nuclear fuel must take these differences into account. Each of the plutonium sources (RG and WG) has different implications on the neutronic behavior of the fuel because each contains a different blend of plutonium nuclides. The amount of heat and the number of neutrons produced from fission of plutonium nuclides is different from fission of {sup 235}U. These differences in UOX and MOX do not end at discharge of the fuel from the reactor core - the short- and long-term storage of MOX fuel may have different requirements than UOX fuel because of the different discharged fuel decay heat characteristics. The research documented in this report compares MOX and UOX fuel during storage and disposal of the fuel by comparing decay heat rates for typical pressurized water reactor (PWR) and boiling water reactor (BWR) fuel assemblies with and without weapons-grade (WG) and reactor-grade (RG) MOX fuel.

  10. TERNARY ALLOYS OF URANIUM, COLUMBIUM, AND ZIRCONIUM

    DOE Patents [OSTI]

    Foote, F.G.

    1960-08-01

    Ternary alloys of uranium are described which are useful as neutron- reflecting materials in a fast neutron reactor. They are especially resistant to corrosion caused by oxidative processes of gascous or aqueous origin and comprise uranium as the predominant metal with zirconiunn and niobium wherein the total content of the minor alloying elements is between 2 and 8% by weight.

  11. PRODUCTION OF URANIUM AND THORIUM COMPOUNDS

    DOE Patents [OSTI]

    Arden, T.V.; Burstall, F.H.; Linstead, R.P.; Wells, R.A.

    1955-12-27

    Compounds of Th and U are extracted with an organic solvent in the presence of an adsorbent substance which has greater retentivity for impurities present than for the uranium and/or thorium. The preferred adsorbent material is noted as being cellulose. The uranium and thoriumcontaining substances treated are preferably in the form of dissolved nitrates, and the preferred organic solvent is diethyl ether.

  12. Uranium Management - Preservation of a National Asset

    SciTech Connect (OSTI)

    Jackson, J. D.; Stroud, J. C.

    2002-02-27

    The Uranium Management Group (UMG) was established at the Department of Energy's (DOE's) Oak Ridge Operations in 1999 as a mechanism to expedite the de-inventory of surplus uranium from the Fernald Environmental Management Project site. This successful initial venture has broadened into providing uranium material de-inventory and consolidation support to the Hanford site as well as retrieving uranium materials that the Department had previously provided to universities under the loan/lease program. As of December 31, 2001, {approx} 4,300 metric tons of uranium (MTU) have been consolidated into a more cost effective interim storage location at the Portsmouth site near Piketon, OH. The UMG continues to uphold its corporate support mission by promoting the Nuclear Materials Stewardship Initiative (NMSI) and the twenty-five (25) action items of the Integrated Nuclear Materials Management Plan (1). Before additional consolidation efforts may commence to remove excess inventory from Environmental Management closure sites and universities, a Programmatic Environmental Assessment (PEA) must be completed. Two (2) noteworthy efforts currently being pursued involve the investigation of re-use opportunities for surplus uranium materials and the recovery of usable uranium from the shutdown Portsmouth cascade. In summary, the UMG is available as a DOE complex-wide technical resource to promote the responsible management of surplus uranium.

  13. US-Russian collaboration in MPC & A enhancements at the Elektrostal Uranium Fuel-Fabrication Plant

    SciTech Connect (OSTI)

    Smith, H.; Murray, W.; Whiteson, R.

    1997-11-01

    Enhancement of the nuclear materials protection, control, and accounting of (MPC&A) at the Elektrostal Machine-Building Plant (ELEMASH) has proceeded in two phases. Initially, Elektrostal served as the model facility at which to test US/Russian collaboration and to demonstrate MPC&A technologies available for safeguards enhancements at Russian facilities. This phase addressed material control and accounting (MC&A) in the low-enriched uranium (LEU) fuel-fabrication processes and the physical protection (PP) of part of the (higher-enrichment) breeder-fuel process. The second phase, identified later in the broader US/Russian agreement for expanded MPC&A cooperation. includes implementation of appropriate MC&A and PP systems in the breeder-fuel fabrication processes. Within the past year, an automated physical protection system has been installed and demonstrated in building 274, and an automated MC&A system has been designed and is being installed and will be tested in the LEU process. Attention has now turned to assuring longterm sustainability for the first phase and beginning MPC&A upgrades for the second phase. Sustainability measures establish the infrastructure for operation, maintenance, and repair of the installed systems-with US support for the lifetime of the US/Russian Agreement, but evolving toward full Russian operation of the system over the long term. For phase 2, which will address higher enrichments, projects have been identified to characterize the facilities, design MPC&A systems, procure appropriate equipment, and install and test final systems. One goal in phase 2 will be to build on initial work to create shared, plant-wide MPC&A assets for operation, maintenance, and evaluation of all safeguards systems.

  14. Deep drawing of uranium metal

    SciTech Connect (OSTI)

    Jackson, R J; Lundberg, M R

    1987-01-19

    A procedure was developed to fabricate uranium forming blanks with high ''draw-ability'' so that cup shapes could be easily and uniformly deep drawn. The overall procedure involved a posttreatment to develop optimum mechanical and structural properties in the deep-drawn cups. The fabrication sequence is casting high-purity logs, pucking cast logs, cross-rolling pucks to forming blanks, annealing and outgassing forming blanks, cold deep drawing to hemispherical shapes, and stress relieving, outgassing, and annealing deep-drawn parts to restore ductility and impart dimensional stability. The fabrication development and the resulting fabrication procedure are discussed in detail. The mechanical properties and microstructural properties are discussed.

  15. 2014 Domestic Uranium Production Report

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

    5. U.S. uranium in-situ-leach plants by owner, location, capacity, and operating status at end of the year, 2011-15" "In-Situ-Leach Plant Owner","In-Situ-Leach Plant Name","County, State (existing and planned locations)","Production Capacity (pounds U3O8 per year)","Operating Status at End of the Year" ,,,,2011,2012,2013,2014,2015 "AUC LLC","Reno Creek","Campbell,

  16. 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" "Owner","Mill and Heap Leach1 Facility Name","County, State (existing and planned locations)"," Capacity","Operating Status at End of the Year" ,,,"(short tons of ore per day)",2011,2012,2013,2014,2015 "Anfield

  17. 2015 Domestic Uranium Production Report

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

    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

  18. 2015 Domestic Uranium Production Report

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

    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 Underground Mines and Stockpiles Fed to Mills 1",0,"W","W","W",0,"W","W","W","W","W","W","W",0 "Other Feed Materials

  19. Method for fabricating laminated uranium composites

    DOE Patents [OSTI]

    Chapman, L.R.

    1983-08-03

    The present invention is directed to a process for fabricating laminated composites of uranium or uranium alloys and at least one other metal or alloy. The laminated composites are fabricated by forming a casting of the molten uranium with the other metal or alloy which is selectively positioned in the casting and then hot-rolling the casting into a laminated plate in or around which the casting components are metallurgically bonded to one another to form the composite. The process of the present invention provides strong metallurgical bonds between the laminate components primarily since the bond disrupting surface oxides on the uranium or uranium alloy float to the surface of the casting to effectively remove the oxides from the bonding surfaces of the components.

  20. Scrap uranium recycling via electron beam melting

    SciTech Connect (OSTI)

    McKoon, R.

    1993-11-01

    A program is underway at the Lawrence Livermore National Laboratory (LLNL) to recycle scrap uranium metal. Currently, much of the material from forging and machining processes is considered radioactive waste and is disposed of by oxidation and encapsulation at significant cost. In the recycling process, uranium and uranium alloys in various forms will be processed by electron beam melting and continuously cast into ingots meeting applicable specifications for virgin material. Existing vacuum processing facilities at LLNL are in compliance with all current federal and state environmental, safety and health regulations for the electron beam melting and vaporization of uranium metal. One of these facilities has been retrofitted with an auxiliary electron beam gun system, water-cooled hearth, crucible and ingot puller to create an electron beam melt furnace. In this furnace, basic process R&D on uranium recycling will be performed with the goal of eventual transfer of this technology to a production facility.

  1. Uranium Pyrophoricity Phenomena and Prediction

    SciTech Connect (OSTI)

    DUNCAN, D.R.

    2000-04-20

    We have compiled a topical reference on the phenomena, experiences, experiments, and prediction of uranium pyrophoricity for the Hanford Spent Nuclear Fuel Project (SNFP) with specific applications to SNFP process and situations. The purpose of the compilation is to create a reference to integrate and preserve this knowledge. Decades ago, uranium and zirconium fires were commonplace at Atomic Energy Commission facilities, and good documentation of experiences is surprisingly sparse. Today, these phenomena are important to site remediation and analysis of packaging, transportation, and processing of unirradiated metal scrap and spent nuclear fuel. Our document, bearing the same title as this paper, will soon be available in the Hanford document system [Plys, et al., 2000]. This paper explains general content of our topical reference and provides examples useful throughout the DOE complex. Moreover, the methods described here can be applied to analysis of potentially pyrophoric plutonium, metal, or metal hydride compounds provided that kinetic data are available. A key feature of this paper is a set of straightforward equations and values that are immediately applicable to safety analysis.

  2. heu

    National Nuclear Security Administration (NNSA)

    plans to reduce the proliferation threat from stockpiles of surplus highly enriched uranium (HEU) by down-blending, or converting, it to low enriched uranium (LEU). In this...

  3. Downblend Diplomacy, Part 2 | Y-12 National Security Complex

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

    conversion of highly enriched uranium from nuclear weapons to low-enriched uranium. "Initially, it was difficult," the monitor recalled. "but for the last several years, it's been ...

  4. Ms. Maria Galanti Site Coordinator

    Office of Environmental Management (EM)

    ... uranium LEU low-enriched uranium NCP ... Office OMB Office of Management and Budget OSDC ... United States Code WAC waste acceptance criteria This page is ...

  5. NNSA Completes Conversion of the Budapest Research Reactor and...

    National Nuclear Security Administration (NNSA)

    Atomic Energy Research Institute, successfully converted the Budapest Research Reactor (BRR) from the use of highly enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. ...

  6. Friday, September 24, 2010 | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    States also disposes of its surplus highly enriched uranium by downblending it to low enriched uranium for peaceful use as commercial nuclear reactor fuel. These programs are a...

  7. IAEA 2010 General Conference Travel Blog | National Nuclear Security...

    National Nuclear Security Administration (NNSA)

    States also disposes of its surplus highly enriched uranium by downblending it to low enriched uranium for peaceful use as commercial nuclear reactor fuel. These programs are a...

  8. Technical Basis for Assessing Uranium Bioremediation Performance

    SciTech Connect (OSTI)

    PE Long; SB Yabusaki; PD Meyer; CJ Murray; AL N’Guessan

    2008-04-01

    In situ bioremediation of uranium holds significant promise for effective stabilization of U(VI) from groundwater at reduced cost compared to conventional pump and treat. This promise is unlikely to be realized unless researchers and practitioners successfully predict and demonstrate the long-term effectiveness of uranium bioremediation protocols. Field research to date has focused on both proof of principle and a mechanistic level of understanding. Current practice typically involves an engineering approach using proprietary amendments that focuses mainly on monitoring U(VI) concentration for a limited time period. Given the complexity of uranium biogeochemistry and uranium secondary minerals, and the lack of documented case studies, a systematic monitoring approach using multiple performance indicators is needed. This document provides an overview of uranium bioremediation, summarizes design considerations, and identifies and prioritizes field performance indicators for the application of uranium bioremediation. The performance indicators provided as part of this document are based on current biogeochemical understanding of uranium and will enable practitioners to monitor the performance of their system and make a strong case to clients, regulators, and the public that the future performance of the system can be assured and changes in performance addressed as needed. The performance indicators established by this document and the information gained by using these indicators do add to the cost of uranium bioremediation. However, they are vital to the long-term success of the application of uranium bioremediation and provide a significant assurance that regulatory goals will be met. The document also emphasizes the need for systematic development of key information from bench scale tests and pilot scales tests prior to full-scale implementation.

  9. uranium

    National Nuclear Security Administration (NNSA)

    a>

    NNSA Removes U.S.-Origin HEU from Jamaica, Makes the Caribbean HEU Free http:nnsa.energy.govmediaroompressreleasesnnsa-removes-u.s.-origin-heu-jamaica-mak...

  10. Uranium Leasing Program: Program Summary | Department of Energy

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

    Uranium Leasing Program » Uranium Leasing Program: Program Summary Uranium Leasing Program: Program Summary Uranium Leasing Program: Program Summary The Atomic Energy Act and other legislative actions authorized the U.S. Atomic Energy Commission (AEC), predecessor agency to the DOE, to withdraw lands from the public domain and then lease them to private industry for mineral exploration and for development and mining of uranium and vanadium ore. A total of 25,000 acres of land in southwestern

  11. Ex Parte Communications - Uranium Producers of America | Department of

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

    Energy - Uranium Producers of America Ex Parte Communications - Uranium Producers of America On Thursday, February 12, 2015, representatives from the Uranium Producers of America (UPA) met with the Department of Energy (DOE) officials to discuss the management of the federal excess uranium inventory. PDF icon Henderson - meeting summary 02 18 15 More Documents & Publications Excess Uranium Management Public Comment re Section 934 of the Energy Independence and Security Act of 2007

  12. Special Nuclear Materials: EM Manages Plutonium, Highly Enriched Uranium

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

    and Uranium-233 | Department of Energy Waste Management » Nuclear Materials & Waste » Special Nuclear Materials: EM Manages Plutonium, Highly Enriched Uranium and Uranium-233 Special Nuclear Materials: EM Manages Plutonium, Highly Enriched Uranium and Uranium-233 105-K building houses the K-Area Material Storage (KAMS) facility, designated for the consolidated storage of surplus plutonium at Savannah River Site pending disposition. The plutonium shipped to KAMS is sealed inside a

  13. Uranium at Y-12: Recovery | Y-12 National Security Complex

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

    Recovery Uranium at Y-12: Recovery Posted: July 22, 2013 - 3:44pm | Y-12 Report | Volume 10, Issue 1 | 2013 Recovery involves reclaiming uranium from numerous sources and configurations and handling uranium in almost any form, including oxides and liquids (see A Rich Resource Requires Recovery). Y-12 has the equipment and expertise to recover uranium that is present in filters, wipes, mop water and elsewhere. For many salvage materials, the uranium is extracted and then manipulated into a uranyl

  14. Method of Making Uranium Dioxide Bodies

    DOE Patents [OSTI]

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

    1973-09-25

    Sintered uranium dioxide bodies having controlled density are produced from U.sub.3 O.sub.8 and carbon by varying the mole ratio of carbon to U.sub.3 O.sub.8 in the mixture, which is compressed and sintered in a neutral or slightly oxidizing atmosphere to form dense slightly hyperstoichiometric uranium dioxide bodies. If the bodies are to be used as nuclear reactor fuel, they are subsequently heated in a hydrogen atmosphere to achieve stoichiometry. This method can also be used to produce fuel elements of uranium dioxide -- plutonium dioxide having controlled density.

  15. SEPARATION OF THORIUM FROM URANIUM BY EXTRACTION

    DOE Patents [OSTI]

    Bohlmann, E.G.

    1959-07-28

    A method is presented for the recovery and separation of uranium and thorium values contained in an aqueous nitric acid solution which is more than 3 M in nitric acid. The uranium and thorium containing solution preferable about 7 M in nitric acid is contacted with tributyl phosphatekerosene mixture. Both U and Th are extracted by the immiscible organic. After phase separation the Th is selectively back extracted by contacting with an aqueous nitric acid solution preferably between 0.1 to 1.5 M in nitric acid. The uranium which is still in the organic extractant phase may be recovered by contacting with water.

  16. Electrolytic process for preparing uranium metal

    DOE Patents [OSTI]

    Haas, Paul A.

    1990-01-01

    An electrolytic process for making uranium from uranium oxide using Cl.sub.2 anode product from an electrolytic cell to react with UO.sub.2 to form uranium chlorides. The chlorides are used in low concentrations in a melt comprising fluorides and chlorides of potassium, sodium and barium in the electrolytic cell. The electrolysis produces Cl.sub.2 at the anode that reacts with UO.sub.2 in the feed reactor to form soluble UCl.sub.4, available for a continuous process in the electrolytic cell, rather than having insoluble UO.sub.2 fouling the cell.

  17. RECOVERY OF URANIUM FROM CARBONATE LEACH LIQUORS

    DOE Patents [OSTI]

    Wilson, H.F.

    1958-07-01

    An improved process is described for the recovery of uranium from vanadifrous ores. In the prior art such ores have been digested with alkali carbonate solutions at a pH of less than 10 and then contacted with a strong base anion exchange resin to separate uranium from vanadium. It has been found that if the exchamge resin feed solution has its pH adjusted to the range 10.8 to 11.8, that vanadium adsorption on the resin is markedly decreased and the separation of uranium from the vanadium is thereby improved.

  18. METHOD OF SEPARATING URANIUM FROM ALLOYS

    DOE Patents [OSTI]

    Chiotti, P.; Shoemaker, H.E.

    1960-06-28

    Uranium can be recovered from metallic uraniumthorium mixtures containing uranium in comparatively small amounts. The method of recovery comprises adding a quantity of magnesium to a mass to obtain a content of from 48 to 85% by weight; melting and forming a magnesium-thorium alloy at a temperature of between 585 and 800 deg C; agitating the mixture, allowing the mixture to settle whereby two phases, a thorium-containing magnesium-rich liquid phase and a solid uranium-rich phase, are formed; and separating the two phases.

  19. Inherently safe in situ uranium recovery (Patent) | SciTech Connect

    Office of Scientific and Technical Information (OSTI)

    Inherently safe in situ uranium recovery Citation Details In-Document Search Title: Inherently safe in situ uranium recovery An in situ recovery of uranium operation involves...

  20. PROCESS FOR PRODUCING URANIUM HALIDES

    DOE Patents [OSTI]

    Murphree, E.V.

    1957-10-29

    A process amd associated apparatus for producing UF/sub 4/ from U/sub 3/ O/sub 8/ by a fluidized'' technique are reported. The U/sub 3/O/sub 8/ is first reduced to UO/sub 2/ by reaction with hydrogen, and the lower oxide of uranium is then reacted with gaseous HF to produce UF/sub 4/. In each case the reactant gas is used, alone or in combination with inert gases, to fluidize'' the finely divided reactant solid. The complete setup of the plant equipment including bins, reactor and the associated piping and valving, is described. An auxiliary fluorination reactor allows for the direct production of UF/sub 6/ from UF/sub 4/ and fluorine gas, or if desired, UF/sub 4/ may be collected as the product.

  1. Thermodynamic properties of uranium dioxide

    SciTech Connect (OSTI)

    Fink, J.K.; Chasanov, M.G.; Leibowitz, L.

    1981-04-01

    In order to provide reliable and consistent data on the thermophysical properties of reactor materials for reactor safety studies, this revision is prepared for the thermodynamic properties of the uranium dioxide portion of the fuel property section of the report Properties for LMFBR Safety Analysis. Since the original report was issued in 1976, there has been international agreement on a vapor pressure equation for the total pressure over UO/sub 2/, new methods have been suggested for the calculation of enthalpy and heat capacity, and a phase change at 2670 K has been proposed. In this report, an electronic term is used in place of the Frenkel defect term in the enthalpy and heat capacity equation and the phase transition is accepted.

  2. 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 2017 Deliveries to foreign suppliers and utilities 2011 2012 2013 2014 2015 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 sales 12,297 13,185 14,717 15,794 21,465 Weighted-average price 47.61 47.58 42.66 36.43 39.58 Foreign sales 16,683 17,982 18,864 20,004 25,723 Weighted-average price 49.05 47.57 42.75 35.69 39.29 Foreign sales 4,538 3,699 4,177

  3. 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 2017 thousand pounds U 3 O 8 equivalent 2011 2012 2013 2014 P2015 Owners and operators of U.S. civilian nuclear power reactors 89,835 97,647 113,007 114,046 120,857 U.S. brokers and traders 6,841 5,677 7,926 5,916 5,678 U.S. converter, enrichers, fabricators, and producers 15,428 17,611 13,416 12,766 9,388 Total commercial inventories 112,104 120,936 134,418 132,728 135,923 thousand pounds U 3 O 8 equivalent

  4. 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 2017 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 Second 7,175 41.29 6,665 35.11 6,807 36.03 Third 7,175 45.89 6,665 39.29 6,807 38.63 Fourth 7,175 49.84 6,665 43.36 6,807 41.80 Fifth 7,175 53.17 6,665 46.74 6,807 44.63 Sixth 7,175 57.24 6,665 50.65 6,807

  5. 2015 Domestic Uranium Production Report

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

    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 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 116 625

  6. 2015 Domestic Uranium Production Report

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

    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 2014 2015 (estimated contained thousand pounds U 3 O 8 ) W W W W W W W W W W W W W (estimated contained thousand pounds U 3 O 8 ) 0 0 0 0 0 0 0 0 0 0 0 0 0 (thousand pounds U 3 O 8 ) W W 2,681 4,259 W W W W W W W W W (thousand pounds U 3 O 8 ) W W W W W W W W W W W W W (thousand pounds U 3 O 8 ) E2,200 2,452 3,045 4,692

  7. 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 "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

  8. 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.

  9. Table 4.10 Uranium Reserves, 2008 (Million Pounds Uranium Oxide)

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

    0 Uranium Reserves,1 2008 (Million Pounds Uranium Oxide) State Forward-Cost 2 Category (dollars 3 per pound) $50 or Less $100 or Less Total 539 1,227 Wyoming 220 446 New Mexico 179 390 Arizona, Colorado, Utah 63 198 Texas 27 40 Others 4 50 154 1The U.S. Energy Information Administration (EIA) category of uranium reserves is equivalent to the internationally reported category of "Reasonably Assured Resources" (RAR). Notes: * Estimates are at end of year. * See "Uranium Oxide"

  10. Synthesis of uranium nitride and uranium carbide powder by carbothermic reduction

    SciTech Connect (OSTI)

    Dunwoody, J.T.; Stanek, C.R.; McClellan, K.J.; Voit, S.L.; Volz, H.M.; Hickman, R.R.

    2007-07-01

    Uranium nitride and uranium carbide are being considered as high burnup fuels in next generation nuclear reactors and accelerated driven systems for the transmutation of nuclear waste. The same characteristics that make nitrides and carbides candidates for these applications (i.e. favorable thermal properties, mutual solubility of nitrides, etc.), also make these compositions candidate fuels for space nuclear reactors. In this paper, we discuss the synthesis and characterization of depleted uranium nitride and carbide for a space nuclear reactor program. Importantly, this project emphasized that to synthesize high quality uranium nitride and carbide, it is necessary to understand the exact stoichiometry of the oxide feedstock. (authors)

  11. Manhattan Project: More Uranium Research, 1942

    Office of Scientific and Technical Information (OSTI)

    ... The production of adequate centrifuges was proving to be a very difficult task, and it looked like it might take tens of thousands of centrifuges to produce enough uranium-235 to ...

  12. JACKETED URANIUM NUCLEAR REACTOR FUEL ELEMENT

    DOE Patents [OSTI]

    Huey, W.R.

    1960-03-01

    A uranium rod encased by iwo aluminum cans internested together from opposite directions along their full lengths and with all interfaces bonded together by an aluminum - silicon alloy was developed.

  13. U.S. Uranium Reserves Estimates

    Gasoline and Diesel Fuel Update (EIA)

    1. U.S. Forward-Cost Uranium Reserves by State, Year-End 2008 State 50lb 100lb Ore (million tons) Gradea (%) U3O8 (million lbs) Ore (million tons) Gradea (%) U3O8 (million lbs)...

  14. U.S. Uranium Reserves Estimates

    Gasoline and Diesel Fuel Update (EIA)

    2. U.S. Forward-Cost Uranium Reserves by Mining Method, Year-End 2008 Mining Method 50 per pound 100 per pound Ore (million tons) Gradea (percent U3O8) U3O8 (million pounds) Ore...

  15. Uranium Marketing Annual Report - Energy Information Administration

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

    Uranium Marketing Annual Report With Data for 2015 | Release Date: May 24, 2016 | Next Release Date: May 2017 | full report Previous reports Year: 2014 2013 2012 2011 2010 2009 ...

  16. Nuclear radiation cleanup and uranium prospecting

    DOE Patents [OSTI]

    Mariella, Jr., Raymond P.; Dardenne, Yves M.

    2016-02-02

    Apparatus, systems, and methods for nuclear radiation cleanup and uranium prospecting include the steps of identifying an area; collecting samples; sample preparation; identification, assay, and analysis; and relating the samples to the area.

  17. The Uranium Resource: A Comparative Analysis

    SciTech Connect (OSTI)

    Schneider, Erich A.; Sailor, William C.

    2007-07-01

    An analogy was drawn between uranium and thirty five minerals for which the USGS maintains extensive records. The USGS mineral price data, which extends from 1900 to the present, was used to create a simple model describing long term price evolution. Making the assumption that the price of uranium, a geologically unexceptional mineral, will evolve in a manner similar to that of the USGS minerals, the model was used to project its price trend for this century. Based upon the precedent set by the USGS data, there is an 80% likelihood that the price of uranium will decline. Moreover, the most likely scenario would see the equilibrium price of uranium decline by about 40% by mid-century. (authors)

  18. High-strength ductile uranium alloy

    DOE Patents [OSTI]

    Hemperly, Vernon C.

    1976-07-13

    A novel alloy composition consisting essentially of 0.7 to 0.8 weight percent titanium and 0.2 to 0.3 weight percent vanadium with the balance being uranium.

  19. Process for reducing beta activity in uranium

    DOE Patents [OSTI]

    Briggs, G.G.; Kato, T.R.; Schonegg, E.

    1985-04-11

    This invention is a method for lowering the beta radiation hazards associated with the casting of uranium. The method reduces the beta radiation emitted from the as-cast surfaces of uranium ingots. The method also reduces the amount of beta radiation emitters retained on the interiors of the crucibles that have been used to melt the uranium charges and which undergone cleaning in a remote handling facility. The lowering of the radioactivity is done by scavenging the beta emitters from the molten uranium with a molten mixture containing the fluorides of magnesium and calcium. The method provides a means of collection and disposal of the beta emitters in a manner that reduces radiation exposure to operating personnel in the work area where the ingots are cast and processed. 5 tabs.

  20. Process for reducing beta activity in uranium

    DOE Patents [OSTI]

    Briggs, Gifford G.; Kato, Takeo R.; Schonegg, Edward

    1986-01-01

    This invention is a method for lowering the beta radiation hazards associated with the casting of uranium. The method reduces the beta radiation emitted from the as-cast surfaces of uranium ingots. The method also reduces the amount of beta radiation emitters retained on the interiors of the crucibles that have been used to melt the uranium charges and which have undergone cleaning in a remote handling facility. The lowering of the radioactivity is done by scavenging the beta emitters from the molten uranium with a molten mixture containing the fluorides of magnesium and calcium. The method provides a means of collection and disposal of the beta emitters in a manner that reduces radiation exposure to operating personnel in the work area where the ingots are cast and processed.

  1. Federal Actions to Address Impacts of Uranium

    Office of Legacy Management (LM)

    Federal Actions to Address Impacts of Uranium Contamination in the Navajo Nation 2014 Page | i TABLE OF CONTENTS Executive Summary ....................................................................................................................... 1 Introduction .................................................................................................................................... 2 Summary of Work Completed 2008-2012

  2. DIRECT INGOT PROCESS FOR PRODUCING URANIUM

    DOE Patents [OSTI]

    Leaders, W.M.; Knecht, W.S.

    1960-11-15

    A process is given in which uranium tetrafluoride is reduced to the metal with magnesium and in the same step the uranium metal formed is cast into an ingot. For this purpose a mold is arranged under and connected with the reaction bomb, and both are filled with the reaction mixture. The entire mixture is first heated to just below reaction temperature, and thereafter heating is restricted to the mixture in the mold. The reaction starts in the mold whereby heat is released which brings the rest of the mixture to reaction temperature. Pure uranium metal settles in the mold while the magnesium fluoride slag floats on top of it. After cooling, the uranium is separated from the slag by mechanical means.

  3. PROCESSES OF RECLAIMING URANIUM FROM SOLUTIONS

    DOE Patents [OSTI]

    Zumwalt, L.R.

    1959-02-10

    A process is described for reclaiming residual enriched uranium from calutron wash solutions containing Fe, Cr, Cu, Ni, and Mn as impurities. The solution is adjusted to a pH of between 2 and 4 and is contacted with a metallic reducing agent, such as iron or zinc, in order to reduce the copper to metal and thereby remove it from the solution. At the same time the uranium present is reduced to the uranous state The solution is then contacted with a precipitate of zinc hydroxide or barium carbonate in order to precipitate and carry uranium, iron, and chromium away from the nickel and manganese ions in the solution. The uranium is then recovered fronm this precipitate.

  4. Draft Environmental Impact Statement for Construction and Operation of a Depleted Uranium Hexafluoride Conversion Facility at the Paducah, Kentucky, Site

    SciTech Connect (OSTI)

    N /A

    2003-11-28

    This document is a site-specific environmental impact statement (EIS) for construction and operation of a proposed depleted uranium hexafluoride (DUF{sub 6}) conversion facility at the U.S. Department of Energy (DOE) Paducah site in northwestern Kentucky (Figure S-1). The proposed facility would convert the DUF{sub 6} stored at Paducah to a more stable chemical form suitable for use or disposal. In a Notice of Intent (NOI) published in the ''Federal Register'' (FR) on September 18, 2001 (''Federal Register'', Volume 66, page 48123 [66 FR 48123]), DOE announced its intention to prepare a single EIS for a proposal to construct, operate, maintain, and decontaminate and decommission two DUF{sub 6} conversion facilities at Portsmouth, Ohio, and Paducah, Kentucky, in accordance with the National Environmental Policy Act of 1969 (NEPA) (''United States Code'', Title 42, Section 4321 et seq. [42 USC 4321 et seq.]) and DOE's NEPA implementing procedures (''Code of Federal Regulations'', Title 10, Part 1021 [10 CFR Part 1021]). Subsequent to award of a contract to Uranium Disposition Services, LLC (hereafter referred to as UDS), Oak Ridge, Tennessee, on August 29, 2002, for design, construction, and operation of DUF{sub 6} conversion facilities at Portsmouth and Paducah, DOE reevaluated its approach to the NEPA process and decided to prepare separate site-specific EISs. This change was announced in a ''Federal Register'' Notice of Change in NEPA Compliance Approach published on April 28, 2003 (68 FR 22368); the Notice is included as Attachment B to Appendix C of this EIS. This EIS addresses the potential environmental impacts from the construction, operation, maintenance, and decontamination and decommissioning (D&D) of the proposed conversion facility at three alternative locations within the Paducah site; from the transportation of depleted uranium conversion products to a disposal facility; and from the transportation, sale, use, or disposal of the fluoride-containing conversion products (hydrogen fluoride [HF] or calcium fluoride [CaF{sub 2}]). Although not part of the proposed action, an option of shipping all cylinders (DUF{sub 6}, low-enriched UF{sub 6} [LEU-UF{sub 6}], and empty) stored at the East Tennessee Technology Park (ETTP) near Oak Ridge, Tennessee, to Paducah rather than to Portsmouth is also considered. In addition, this EIS evaluates a no action alternative, which assumes continued storage of DUF{sub 6} in cylinders at the Paducah site. A separate EIS (DOE/EIS-0360) evaluates the potential environmental impacts for the proposed Portsmouth conversion facility.

  5. Draft Environmental Impact Statement for Construction and Operation of a Depleted Uranium Hexafluoride Conversion Facility at the Portsmouth, Ohio, Site

    SciTech Connect (OSTI)

    N /A

    2003-11-28

    This document is a site-specific environmental impact statement (EIS) for construction and operation of a proposed depleted uranium hexafluoride (DUF{sub 6}) conversion facility at the U.S. Department of Energy (DOE) Portsmouth site in Ohio (Figure S-1). The proposed facility would convert the DUF{sub 6} stored at Portsmouth to a more stable chemical form suitable for use or disposal. The facility would also convert the DUF{sub 6} from the East Tennessee Technology Park (ETTP) site near Oak Ridge, Tennessee. In a Notice of Intent (NOI) published in the Federal Register on September 18, 2001 (Federal Register, Volume 66, page 48123 [66 FR 48123]), DOE announced its intention to prepare a single EIS for a proposal to construct, operate, maintain, and decontaminate and decommission two DUF{sub 6} conversion facilities at Portsmouth, Ohio, and Paducah, Kentucky, in accordance with the National Environmental Policy Act of 1969 (NEPA) (United States Code, Title 42, Section 4321 et seq. [42 USC 4321 et seq.]) and DOE's NEPA implementing procedures (Code of Federal Regulations, Title 10, Part 1021 [10 CFR Part 1021]). Subsequent to award of a contract to Uranium Disposition Services, LLC (hereafter referred to as UDS), Oak Ridge, Tennessee, on August 29, 2002, for design, construction, and operation of DUF{sub 6} conversion facilities at Portsmouth and Paducah, DOE reevaluated its approach to the NEPA process and decided to prepare separate site-specific EISs. This change was announced in a Federal Register Notice of Change in NEPA Compliance Approach published on April 28, 2003 (68 FR 22368); the Notice is included as Attachment B to Appendix C of this EIS. This EIS addresses the potential environmental impacts from the construction, operation, maintenance, and decontamination and decommissioning (D&D) of the proposed conversion facility at three alternative locations within the Portsmouth site; from the transportation of all ETTP cylinders (DUF{sub 6}, low-enriched UF6 [LEU-UF{sub 6}], and empty) to Portsmouth; from the transportation of depleted uranium conversion products to a disposal facility; and from the transportation, sale, use, or disposal of the fluoride-containing conversion products (hydrogen fluoride [HF] or calcium fluoride [CaF{sub 2}]). An option of shipping the ETTP cylinders to Paducah is also considered. In addition, this EIS evaluates a no action alternative, which assumes continued storage of DUF{sub 6} in cylinders at the Portsmouth and ETTP sites. A separate EIS (DOE/EIS-0359) evaluates potential environmental impacts for the proposed Paducah conversion facility.

  6. Depleted uranium: A DOE management guide

    SciTech Connect (OSTI)

    1995-10-01

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

  7. Uranium Leasing Program Documents | Department of Energy

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

    Documents Uranium Leasing Program Documents U.S. District Court's Order of October 18, 2011, in Colorado Environmental Coalition v. Office of Legacy Management, Civil Action No. 08-cv-01624 (D. Colo.). The Court has issued the injunctive relief described on pages 51-52 of the Order. U.S. District Court's Order of February 27, 2012, in Colorado Environmental Coalition v. Office of Legacy Management, Civil Action No. 08-cv-01624 (D. Colo.). Uranium Lease Tracts Location Map

  8. The ultimate disposition of depleted uranium

    SciTech Connect (OSTI)

    Lemons, T.R.

    1991-12-31

    Depleted uranium (DU) is produced as a by-product of the uranium enrichment process. Over 340,000 MTU of DU in the form of UF{sub 6} have been accumulated at the US government gaseous diffusion plants and the stockpile continues to grow. An overview of issues and objectives associated with the inventory management and the ultimate disposition of this material is presented.

  9. highly enriched uranium | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    highly enriched uranium NNSA deputy administrator travels to Ukraine Earlier this month, Deputy Administrator for Defense Nuclear Nonproliferation Anne Harrington traveled to Ukraine to celebrate the 20th anniversary of the Science and Technology Center in Ukraine (STCU) and visit the Neutron Source Facility at the Kharkiv Institute of Physics and Technology. The U... DOE/NNSA Successfully Establishes Uranium Lease and Takeback Program to Support Critical Medical Isotope Production In January

  10. RECOVERY OF URANIUM BY SECONDARY XANTHATE COMPLEXING

    DOE Patents [OSTI]

    Neville, O.K.

    1959-09-01

    A method is described for separating and recovering uranium values contained in an acidic aqueous solution together with thorium or protactinium values. In accordance with the invention, the acidic solution containing uranium in the uranyl form is contacted with an organic xanthate. The xanthate forms a urano-xanthate complex but is substantially non-reactive with thorium and protactinium. The urano-xanthate complex is recovered by organic solvent extraction.

  11. Domestic Uranium Production Report 2004-13

    Gasoline and Diesel Fuel Update (EIA)

    Domestic Uranium Production Report 1st Quarter 2016 May 2016 Independent Statistics & Analysis www.eia.gov U.S. Department of Energy Washington, DC 20585 U.S. Energy Information Administration | Domestic Uranium Production Report 1st Quarter 2016 ii This report was prepared by the U.S. Energy Information Administration (EIA), the statistical and analytical agency within the U.S. Department of Energy. By law, EIA's data, analyses, and forecasts are independent of approval by any other officer

  12. Electrochemical method of producing eutectic uranium alloy and apparatus

    DOE Patents [OSTI]

    Horton, James A. (Livermore, CA); Hayden, H. Wayne (Oakridge, TN)

    1995-01-01

    An apparatus and method for continuous production of liquid uranium alloys through the electrolytic reduction of uranium chlorides. The apparatus includes an electrochemical cell formed from an anode shaped to form an electrolyte reservoir, a cathode comprising a metal, such as iron, capable of forming a eutectic uranium alloy having a melting point less than the melting point of pure uranium, and molten electrolyte in the reservoir comprising a chlorine or fluorine containing salt and uranium chloride. The method of the invention produces an eutectic uranium alloy by creating an electrolyte reservoir defined by a container comprising an anode, placing an electrolyte in the reservoir, the electrolyte comprising a chlorine or fluorine containing salt and uranium chloride in molten form, positioning a cathode in the reservoir where the cathode comprises a metal capable of forming an uranium alloy having a melting point less than the melting point of pure uranium, and applying a current between the cathode and the anode.

  13. Y-12 Knows Uranium | Y-12 National Security Complex

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

    Knows Uranium Y-12 Knows Uranium Posted: July 22, 2013 - 3:45pm | Y-12 Report | Volume 10, Issue 1 | 2013 Y-12 produces many forms of uranium. They may be used in chemical processing steps on-site or shipped elsewhere to serve as raw materials for nuclear fuel or as research tools. All of uranium's uses, defense related and otherwise, are critical to the nation. Y-12's understanding of uranium, coupled with the site's work with enriched uranium metal, alloys, oxides, compounds and solutions, is

  14. URANIUM RECOVERY AND PURIFICATION PROCESS AND PRODUCTION OF HIGH PURITY URANIUM TETRAFLUORIDE

    DOE Patents [OSTI]

    Bailes, R.H.; Long, R.S.; Grinstead, R.R.

    1957-09-17

    A process is described wherein an anionic exchange technique is employed to separate uramium from a large variety of impurities. Very efficient and economical purification of contamimated uranium can be achieved by treatment of the contaminated uranium to produce a solution containing a high concentration of chloride. Under these conditions the uranium exists as an aniomic chloride complex. Then the uranium chloride complex is adsorbed from the solution on an aniomic exchange resin, whereby a portion of the impurities remain in the solution and others are retained with the uramium by the resin. The adsorbed impurities are then removed by washing the resin with pure concentrated hydrochloric acid, after which operation the uranium is eluted with pure water yielding an acidic uranyl chloride solution of high purity.

  15. Uranium mill ore dust characterization

    SciTech Connect (OSTI)

    Knuth, R.H.; George, A.C.

    1980-11-01

    Cascade impactor and general air ore dust measurements were taken in a uranium processing mill in order to characterize the airborne activity, the degree of equilibrium, the particle size distribution and the respirable fraction for the /sup 238/U chain nuclides. The sampling locations were selected to limit the possibility of cross contamination by airborne dusts originating in different process areas of the mill. The reliability of the modified impactor and measurement techniques was ascertained by duplicate sampling. The results reveal no significant deviation from secular equilibrium in both airborne and bulk ore samples for the /sup 234/U and /sup 230/Th nuclides. In total airborne dust measurements, the /sup 226/Ra and /sup 210/Pb nuclides were found to be depleted by 20 and 25%, respectively. Bulk ore samples showed depletions of 10% for the /sup 226/Ra and /sup 210/Pb nuclides. Impactor samples show disequilibrium of /sup 226/Ra as high as +-50% for different size fractions. In these samples the /sup 226/Ra ratio was generally found to increase as particle size decreased. Activity median aerodynamic diameters of the airborne dusts ranged from 5 to 30 ..mu..m with a median diameter of 11 ..mu..m. The maximum respirable fraction for the ore dusts, based on the proposed International Commission on Radiological Protection's (ICRP) definition of pulmonary deposition, was < 15% of the total airborne concentration. Ore dust parameters calculated for impactor duplicate samples were found to be in excellent agreement.

  16. 2015 Domestic Uranium Production Report

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

    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 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 43.5 33.7 319.2 2012 66.6 186.9 99.4 16.8 33.3

  17. Uranium distribution in relation to sedimentary facies, Kern Lake, California

    SciTech Connect (OSTI)

    Merifield, P.M.; Carlisle, D.; Idiz, E.; Anderhalt, R.; Reed, W.E.; Lamar, D.L.

    1980-04-01

    Kern Lake has served as a sink for drainage from the southern Sierra Nevada and, in lesser amounts, from the southern Temblor Range. Both areas contain significant uranium source rocks. The uranium content in Holocene Kern Lake sediments correlates best with the mud (silt and clay) fraction. It correlates less well with organic carbon. Biotite grains could account for much of the uranium in the sand fraction, and perhaps the silt fraction as well. The data suggest that fixation of uranium by adsorption on mineral grains is a dominant process in this lake system. Further work is required to determine the importance of cation-exchange of uranium on clays and micas and of organically complexed uranium adsorbed to mineral surfaces. These findings also raise the question of whether uranium transport down the Kern River occurs largely as uranium adsorbed to mineral surfaces.

  18. EA-1290: Disposition of Russian Federation Titled Natural Uranium

    Broader source: Energy.gov [DOE]

    This EA evaluates the potential environmental impacts of a proposal to transport up to an average of 9,000 metric tons per year of natural uranium as uranium hexafluoride (UF6) from the United...

  19. 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.

  20. DOE - Office of Legacy Management -- Abandoned Uranium Mines

    Office of Legacy Management (LM)

    Uranium Mines Report to Congress The U.S. Department of Energy (DOE) Office of Legacy Management completed a report on defense-related uranium mines in consultation with...

  1. Think Uranium. Think Y-12 | Y-12 National Security Complex

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

    | Y-12 Report | Volume 10, Issue 1 | 2013 Uranium fever: Much like the California gold rush of 1849, the uranium flurry of 1949 led Geiger counter-toting prospectors to scour...

  2. U.S. Uranium Reserves Estimates - Energy Information Administration

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

    Nuclear & Uranium Glossary › FAQS › Overview Data Status of U.S. Nuclear Outages (interactive) Summary Uranium & nuclear fuel Nuclear power plants Spent nuclear fuel International All nuclear data reports Analysis & Projections Major Topics Most popular Nuclear plants and reactors Projections Recurring Uranium All reports Browse by Tag Alphabetical Frequency Tag Cloud ‹ See all Nuclear Reports U.S. Uranium Reserves Estimates Data for: 2008 | Release Date: July 2010 | Next

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

    Office of Scientific and Technical Information (OSTI)

    Ernest Lawrence, Arthur Compton, Vannevar Bush, and James Conant discuss uranium research, Berkeley, March 29, 1940. EARLY URANIUM RESEARCH (1939-1941) Events > Early Government Support, 1939-1942 Einstein's Letter, 1939 Early Uranium Research, 1939-1941 Piles and Plutonium, 1939-1941 Reorganization and Acceleration, 1940-1941 The MAUD Report, 1941 A Tentative Decision to Build the Bomb, 1941-1942 President Franklin D. Roosevelt responded to the call for government support of uranium research

  4. Retrieval of buried depleted uranium from the T-1 trench

    SciTech Connect (OSTI)

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

    1998-07-01

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

  5. DOE Releases Excess Uranium Inventory Plan | Department of Energy

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

    Excess Uranium Inventory Plan DOE Releases Excess Uranium Inventory Plan December 16, 2008 - 8:51am Addthis WASHINGTON, D.C. - The United States Department of Energy (DOE) today issued its Excess Uranium Inventory Management Plan (the Plan), which outlines the Department's strategy for the management and disposition of its excess uranium inventories. The Plan highlights DOE's ongoing efforts to enhance national security and promote a healthy domestic nuclear infrastructure through the efficient

  6. The Uranium Processing Facility (UPF) Finite Element Meshing Discussion

    Broader source: Energy.gov [DOE]

    The Uranium Processing Facility (UPF) Finite Element Meshing Discussion Loring Wyllie Arne Halterman Degenkolb Engineers, San Francisco

  7. Reimbursements to Licensees of Active Uranium and Thorium Processing Sites,

    Energy Savers [EERE]

    Fiscal Year 2009 and 2010 Status Report | Department of Energy Reimbursements to Licensees of Active Uranium and Thorium Processing Sites, Fiscal Year 2009 and 2010 Status Report Reimbursements to Licensees of Active Uranium and Thorium Processing Sites, Fiscal Year 2009 and 2010 Status Report Reimbursements to Licensees of Active Uranium and Thorium Processing Sites, Fiscal Year 2009 and 2010 Status Report (March 2010) PDF icon Reimbursements to Licensees of Active Uranium and Thorium

  8. Secretarial Determination of No Adverse Material Impact for Uranium

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

    Transfers | Department of Energy Secretarial Determination of No Adverse Material Impact for Uranium Transfers Secretarial Determination of No Adverse Material Impact for Uranium Transfers The determination covers the Department's sales or transfers of no more than 2,705 metric tons (MTU) of natural uranium (NU) or NU equivalent in a calendar year. The proposed transfers include up to 650 MTU per year by the National Nuclear Security Administration in support of highly enriched uranium down

  9. Monitoring Uranium Transformations Determined by the Evolution of Biogeochemical Processes

    SciTech Connect (OSTI)

    Marsh, Terence L.

    2013-07-30

    Our contribution to the larger project (ANL) was the phylogenetic analysis of evolved communities capable of reducing metals including uranium.

  10. Uranium at Y-12: Inspection | Y-12 National Security Complex

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

    Inspection Uranium at Y-12: Inspection Posted: July 22, 2013 - 3:36pm | Y-12 Report | Volume 10, Issue 1 | 2013 Inspection of enriched uranium is performed by dimensional checks and radiography. Inspectors examine enriched uranium products using coordinate measuring machines, microscopy, laser inspection machines and other instruments. Technicians use X-rays to determine that the uranium metal integrity is of high quality - absent of voids. These inspections, along with impurity analyses and

  11. Sequestering Uranium from Seawater: Binding Strength and Modes of Uranyl

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

    Complexes with Glutarimidedioxime Sequestering Uranium from Seawater: Binding Strength and Modes of Uranyl Complexes with Glutarimidedioxime Sequestering Uranium from Seawater: Binding Strength and Modes of Uranyl Complexes with Glutarimidedioxime Print Sunday, 14 October 2012 00:00 The ocean is an important source of uranium if it can be extracted economically. Extraction of uranium from seawater is very challenging, not only because it is in an extremely low concentration, but also because

  12. Testing for Uranium Deuteride Initiation in Liquid Deuterium

    SciTech Connect (OSTI)

    Siekhaus, W. J.; Teslich, N. E.; Kucheyev, S. O.; Go, J.

    2015-10-29

    This report offers a description of the testing related to Uranium foil and its interaction with liquid deuterium.

  13. SOLVENT EXTRACTION PROCESS FOR URANIUM FROM CHLORIDE SOLUTIONS

    DOE Patents [OSTI]

    Blake, C.A. Jr.; Brown, K.B.; Horner, D.E.

    1960-05-24

    An improvement was made in a uranium extraction process wherein the organic extractant is a phosphine oxide. An aqueous solution containing phosphate ions or sulfate ions together with uranium is provided with a source of chloride ions during the extraction step. The presence of the chloride ions enables a phosphine oxide to extract uranium in the presence of strong uranium- complexing ions such as phosphate or sulfate ions.

  14. Depleted uranium plasma reduction system study

    SciTech Connect (OSTI)

    Rekemeyer, P.; Feizollahi, F.; Quapp, W.J.; Brown, B.W.

    1994-12-01

    A system life-cycle cost study was conducted of a preliminary design concept for a plasma reduction process for converting depleted uranium to uranium metal and anhydrous HF. The plasma-based process is expected to offer significant economic and environmental advantages over present technology. Depleted Uranium is currently stored in the form of solid UF{sub 6}, of which approximately 575,000 metric tons is stored at three locations in the U.S. The proposed system is preconceptual in nature, but includes all necessary processing equipment and facilities to perform the process. The study has identified total processing cost of approximately $3.00/kg of UF{sub 6} processed. Based on the results of this study, the development of a laboratory-scale system (1 kg/h throughput of UF6) is warranted. Further scaling of the process to pilot scale will be determined after laboratory testing is complete.

  15. Method for fluorination of uranium oxide

    DOE Patents [OSTI]

    Petit, George S. (Oak Ridge, TN)

    1987-01-01

    Highly pure uranium hexafluoride is made from uranium oxide and fluorine. The uranium oxide, which includes UO.sub.3, UO.sub.2, U.sub.3 O.sub.8 and mixtures thereof, is introduced together with a small amount of a fluorine-reactive substance, selected from alkali chlorides, silicon dioxide, silicic acid, ferric oxide, and bromine, into a constant volume reaction zone. Sufficient fluorine is charged into the zone at a temperature below approximately 0.degree. C. to provide an initial pressure of at least approximately 600 lbs/sq. in. at the ambient atmospheric temperature. The temperature is then allowed to rise in the reaction zone until reaction occurs.

  16. SEPARATION OF URANIUM FROM THORIUM AND PROTACTINIUM

    DOE Patents [OSTI]

    Musgrave, W.K.R.

    1959-06-30

    This patent relates to the separation of uranium from thorium and protactinium; such mixtures of elements usually being obtained by neutron irradiation of thorium. The method of separating the constituents has been first to dissolve the mixture of elements in concertrated nitric acid and then to remove the protactinium by absorption on manganese dioxide and the uranium by solvent extraction with ether. Prior to now, comparatively large amounts of thorium were extracted with the uranium. According to the invention this is completely prevented by adding sodium diethyldithiocarbamate to the mixture of soluble nitrate salts. The organic salt has the effect of reacting only with the uranyl nitrate to form the corresponding uranyl salt which can then be selectively extracted from the mixture with amyl acetate.

  17. SEPARATION OF URANIUM, PLUTONIUM, AND FISSION PRODUCTS

    DOE Patents [OSTI]

    Spence, R.; Lister, M.W.

    1958-12-16

    Uranium and plutonium can be separated from neutron-lrradiated uranium by a process consisting of dissolvlng the lrradiated material in nitric acid, saturating the solution with a nitrate salt such as ammonium nitrate, rendering the solution substantially neutral with a base such as ammonia, adding a reducing agent such as hydroxylamine to change plutonium to the trivalent state, treating the solution with a substantially water immiscible organic solvent such as dibutoxy diethylether to selectively extract the uranium, maklng the residual aqueous solutlon acid with nitric acid, adding an oxidizing agent such as ammonlum bromate to oxidize the plutonium to the hexavalent state, and selectlvely extracting the plutonium by means of an immlscible solvent, such as dibutoxy dlethyletber.

  18. TRACE ELEMENT ANALYSES OF URANIUM MATERIALS

    SciTech Connect (OSTI)

    Beals, D; Charles Shick, C

    2008-06-09

    The Savannah River National Laboratory (SRNL) has developed an analytical method to measure many trace elements in a variety of uranium materials at the high part-per-billion (ppb) to low part-per-million (ppm) levels using matrix removal and analysis by quadrapole ICP-MS. Over 35 elements were measured in uranium oxides, acetate, ore and metal. Replicate analyses of samples did provide precise results however none of the materials was certified for trace element content thus no measure of the accuracy could be made. The DOE New Brunswick Laboratory (NBL) does provide a Certified Reference Material (CRM) that has provisional values for a series of trace elements. The NBL CRM were purchased and analyzed to determine the accuracy of the method for the analysis of trace elements in uranium oxide. These results are presented and discussed in the following paper.

  19. Compton DIV: Using a Compton-Based Gamma-Ray Imager for Design Information Verification of Uranium Enrichment Plants

    SciTech Connect (OSTI)

    Burks, M; Verbeke, J; Dougan, A; Wang, T; Decman, D

    2009-07-04

    A feasibility study has been performed to determine the potential usefulness of Compton imaging as a tool for design information verification (DIV) of uranium enrichment plants. Compton imaging is a method of gamma-ray imaging capable of imaging with a 360-degree field of view over a broad range of energies. These systems can image a room (with a time span on the order of one hour) and return a picture of the distribution and composition of radioactive material in that room. The effectiveness of Compton imaging depends on the sensitivity and resolution of the instrument as well the strength and energy of the radioactive material to be imaged. This study combined measurements and simulations to examine the specific issue of UF{sub 6} gas flow in pipes, at various enrichment levels, as well as hold-up resulting from the accumulation of enriched material in those pipes. It was found that current generation imagers could image pipes carrying UF{sub 6} in less than one hour at moderate to high enrichment. Pipes with low enriched gas would require more time. It was also found that hold-up was more amenable to this technique and could be imaged in gram quantities in a fraction of an hour. another questions arises regarding the ability to separately image two pipes spaced closely together. This depends on the capabilities of the instrument in question. These results are described in detail. In addition, suggestions are given as to how to develop Compton imaging as a tool for DIV.

  20. Chapter 20 - Uranium Enrichment Decontamination & Decommissioning Fund

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

    0. Uranium Enrichment Decontamination and Decommissioning Fund 20-1 CHAPTER 20 URANIUM ENRICHMENT DECONTAMINATION AND DECOMMISSIONING FUND 1. INTRODUCTION. a. Purpose. To establish policies and procedures for the financial management, accounting, budget preparation, cash management of the Uranium Enrichment Decontamination and Decommissioning Fund, referred to hereafter as the Fund. b. Applicability. This chapter applies to all Departmental elements, including the National Nuclear Security

  1. President Truman Increases Production of Uranium and Plutonium | National

    National Nuclear Security Administration (NNSA)

    Nuclear Security Administration Increases Production of Uranium and Plutonium President Truman Increases Production of Uranium and Plutonium Washington, DC President Truman approves a $1.4 billion expansion of Atomic Energy Commission facilities to produce uranium and plutonium for nuclear weapons

  2. Plutonium recovery from spent reactor fuel by uranium displacement

    DOE Patents [OSTI]

    Ackerman, J.P.

    1992-03-17

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

  3. Plutonium recovery from spent reactor fuel by uranium displacement

    DOE Patents [OSTI]

    Ackerman, John P.

    1992-01-01

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

  4. Uranium in the Savannah River Site environment

    SciTech Connect (OSTI)

    Evans, A.G.; Bauer, L.R.; Haselow, J.S.; Hayes, D.W.; Martin, H.L.; McDowell, W.L.; Pickett, J.B.

    1992-12-09

    The purpose of this report is to consolidate the history of environmental uranium studies conducted by SRS and to describe the status of uranium in the environment. The report is intended to be a living document'' that will be updated periodically. This draft issue, February 1992, documents studies that occurred from 1954 to 1989. Data in this report are taken primarily from annual and semiannual environmental reports for SRS. Semiannual reports were published from 1954 through 1962. Annual reports have been published since 1963. Occasionally unpublished data are included in this report for completeness.

  5. Uranium in the Savannah River Site environment

    SciTech Connect (OSTI)

    Evans, A.G.; Bauer, L.R.; Haselow, J.S.; Hayes, D.W.; Martin, H.L.; McDowell, W.L.; Pickett, J.B.

    1992-12-09

    The purpose of this report is to consolidate the history of environmental uranium studies conducted by SRS and to describe the status of uranium in the environment. The report is intended to be a ``living document`` that will be updated periodically. This draft issue, February 1992, documents studies that occurred from 1954 to 1989. Data in this report are taken primarily from annual and semiannual environmental reports for SRS. Semiannual reports were published from 1954 through 1962. Annual reports have been published since 1963. Occasionally unpublished data are included in this report for completeness.

  6. Evaluation of an automatic uranium titration system

    SciTech Connect (OSTI)

    Lewis, K.

    1980-01-01

    The titration system utilizes the constant current coulometric titration of Goldbeck and Lerner. U(VI) is reduced to U(IV) by Fe(II). V(V) is generated to titrate the U(IV), and the titration is followed potentiometrically. The evaluation shows that the recovery of uranium is 100% at the 40-mg level. The accuracy is generally +-0.10% or better. The smallest sample weight at which reliable results were obtained was 40 mg of uranium. Time for one analysis is 15 minutes. Advantages and disadvantages of the automated titrator are listed. (DLC)

  7. Aseismic design criteria for uranium enrichment plants

    SciTech Connect (OSTI)

    Beavers, J.E.

    1980-01-01

    In this paper technological, economical, and safety issues of aseismic design of uranium enrichment plants are presented. The role of management in the decision making process surrounding these issues is also discussed. The resolution of the issues and the decisions made by management are controlling factors in developing aseismic design criteria for any facility. Based on past experience in developing aseismic design criteria for the GCEP various recommendations are made for future enrichment facilities, and since uranium enrichment plants are members of the nuclear fuel cycle the discussion and recommendations presented herein are applicable to other nonreactor nuclear facilities.

  8. PROCESS FOR THE PRODUCTION OF AMMONIUM URANIUM FLUORIDE

    DOE Patents [OSTI]

    Ellis, A.S.; Mooney, R.B.

    1953-08-25

    This patent relates to the preparation of ammonium uranium fluoride. The process comprises adding a water soluble fluoride to an aqueous solution of a uranous compound containing an ammonium salt, and isolating the resulting precipitate. This patent relates to the manufacture of uranium tetnafluoride from ammonium uranium fluoride, NH/sub 4/UF/sub 5/. Uranium tetrafluoride is prepared by heating the ammonium uranium fluoride to a temperature at which dissociation occurs with liberation of ammonium fluoride. Preferably the process is carried out under reduced pressure, or in a current of an inert gas.

  9. Steady State Sputtering Yields and Surface Compositions of Depleted Uranium and Uranium Carbide bombarded by 30 keV Gallium or 16 keV Cesium Ions.

    SciTech Connect (OSTI)

    Siekhaus, W. J.; Teslich, N. E.; Weber, P. K.

    2014-10-23

    Depleted uranium that included carbide inclusions was sputtered with 30-keV gallium ions or 16-kev cesium ions to depths much greater than the ions range, i.e. using steady-state sputtering. The recession of both the uraniums and uranium carbides surfaces and the ion corresponding fluences were used to determine the steady-state target sputtering yields of both uranium and uranium carbide, i.e. 6.3 atoms of uranium and 2.4 units of uranium carbide eroded per gallium ion, and 9.9 uranium atoms and 3.65 units of uranium carbide eroded by cesium ions. The steady state surface composition resulting from the simultaneous gallium or cesium implantation and sputter-erosion of uranium and uranium carbide were calculated to be U??Ga??, (UC)??Ga?? and U??Cs?, (UC)??Cs??, respectively.

  10. PROCESSING OF NEUTRON-IRRADIATED URANIUM

    DOE Patents [OSTI]

    Hopkins, H.H. Jr.

    1960-09-01

    An improved "Purex" process for separating uranium, plutonium, and fission products from nitric acid solutions of neutron-irradiated uranium is offered. Uranium is first extracted into tributyl phosphate (TBP) away from plutonium and fission products after adjustment of the acidity from 0.3 to 0.5 M and heating from 60 to 70 deg C. Coextracted plutonium, ruthenium, and fission products are fractionally removed from the TBP by three scrubbing steps with a 0.5 M nitric acid solution of ferrous sulfamate (FSA), from 3.5 to 5 M nitric acid, and water, respectively, and the purified uranium is finally recovered from the TBP by precipitation with an aqueous solution of oxalic acid. The plutonium in the 0.3 to 0.5 M acid solution is oxidized to the tetravalent state with sodium nitrite and extracted into TBP containing a small amount of dibutyl phosphate (DBP). Plutonium is then back-extracted from the TBP-DBP mixture with a nitric acid solution of FSA, reoxidized with sodium nitrite in the aqueous strip solution obtained, and once more extracted with TBP alone. Finally the plutonium is stripped from the TBP with dilute acid, and a portion of the strip solution thus obtained is recycled into the TBPDBP for further purification.

  11. Statistical design of a uranium corrosion experiment

    SciTech Connect (OSTI)

    Wendelberger, Joanne R; Moore, Leslie M

    2009-01-01

    This work supports an experiment being conducted by Roland Schulze and Mary Ann Hill to study hydride formation, one of the most important forms of corrosion observed in uranium and uranium alloys. The study goals and objectives are described in Schulze and Hill (2008), and the work described here focuses on development of a statistical experiment plan being used for the study. The results of this study will contribute to the development of a uranium hydriding model for use in lifetime prediction models. A parametric study of the effect of hydrogen pressure, gap size and abrasion on hydride initiation and growth is being planned where results can be analyzed statistically to determine individual effects as well as multi-variable interactions. Input to ESC from this experiment will include expected hydride nucleation, size, distribution, and volume on various uranium surface situations (geometry) as a function of age. This study will also address the effect of hydrogen threshold pressure on corrosion nucleation and the effect of oxide abrasion/breach on hydriding processes. Statistical experiment plans provide for efficient collection of data that aids in understanding the impact of specific experiment factors on initiation and growth of corrosion. The experiment planning methods used here also allow for robust data collection accommodating other sources of variation such as the density of inclusions, assumed to vary linearly along the cast rods from which samples are obtained.

  12. Uranium Battery Development Project Final Report

    SciTech Connect (OSTI)

    Dunbar, Paul D; Lee-Desautels, Rhonda

    2007-06-01

    This report summarizes the research funded by the Department of Energy, Oak Ridge National Labs, and the Kentucky Science and Engineering Foundation. This report briefly presents the theory behind our experimental methods and the most important experiments that were performed. This research focused on the reuse of uranium materials in lithium ion batteries. The majority of experiments involved lithium salts and organic solvents.

  13. IRON COATED URANIUM AND ITS PRODUCTION

    DOE Patents [OSTI]

    Gray, A.G.

    1960-03-15

    A method of applying a protective coating to a metallic uranium article is given. The method comprises etching the surface of the article with an etchant solution containlng chloride ions, such as a solution of phosphoric acid and hydrochloric acid, cleaning the etched surface, electroplating iron thereon from a ferrous ammonium sulfate electroplating bath, and soldering an aluminum sheath to the resultant iron layer.

  14. PROCESS OF RECOVERING URANIUM FROM ITS ORES

    DOE Patents [OSTI]

    Galvanek, P. Jr.

    1959-02-24

    A process is presented for recovering uranium from its ores. The crushed ore is mixed with 5 to 10% of sulfuric acid and added water to about 5 to 30% of the weight of the ore. This pugged material is cured for 2 to 3 hours at 100 to 110 deg C and then cooled. The cooled mass is nitrate-conditioned by mixing with a solution equivalent to 35 pounds of ammunium nitrate and 300 pounds of water per ton of ore. The resulting pulp containing 70% or more solids is treated by upflow percolation with a 5% solution of tributyl phosphate in kerosene at a rate equivalent to a residence time of about one hour to extract the solubilized uranium. The uranium is recovered from the pregnant organic liquid by counter-current washing with water. The organic extractant may be recycled. The uranium is removed from the water solution by treating with ammonia to precipitate ammonium diuranate. The filtrate from the last step may be recycled for the nitrate-conditioning treatment.

  15. URANIUM EXTRACTION PROCESS USING SYNERGISTIC REAGENTS

    DOE Patents [OSTI]

    Schmitt, J.M.; Blake, C.A. Jr.; Brown, K.B.; Coleman, C.F.

    1958-11-01

    Improved methods are presented for recovering uranium values from aqueous solutions by organic solvent extraction. The improvement lies in the use, in combination, of two classes of organic compounds so that their extracting properties are enhanced synergistically. The two classes of organic compounds are dialkylphosphoric acid and certain neutral organophosphorus compounds such as trialkylphosphates, trialkylphosphonates, trlalkylphosphinates and trialkylphosphine oxides.

  16. Fermentation and Hydrogen Metabolism Affect Uranium Reduction by Clostridia

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

    Gao, Weimin; Francis, Arokiasamy J.

    2013-01-01

    Previously, it has been shown that not only is uranium reduction under fermentation condition common among clostridia species, but also the strains differed in the extent of their capability and the pH of the culture significantly affected uranium(VI) reduction. In this study, using HPLC and GC techniques, metabolic properties of those clostridial strains active in uranium reduction under fermentation conditions have been characterized and their effects on capability variance of uranium reduction discussed. Then, the relationship between hydrogen metabolism and uranium reduction has been further explored and the important role played by hydrogenase in uranium(VI) and iron(III) reduction bymore » clostridia demonstrated. When hydrogen was provided as the headspace gas, uranium(VI) reduction occurred in the presence of whole cells of clostridia. This is in contrast to that of nitrogen as the headspace gas. Without clostridia cells, hydrogen alone could not result in uranium(VI) reduction. In alignment with this observation, it was also found that either copper(II) addition or iron depletion in the medium could compromise uranium reduction by clostridia. In the end, a comprehensive model was proposed to explain uranium reduction by clostridia and its relationship to the overall metabolism especially hydrogen (H 2 ) production.« less

  17. Uranium-Loaded Water Treatment Resins: 'Equivalent Feed' at NRC and Agreement State-Licensed Uranium Recovery Facilities - 12094

    SciTech Connect (OSTI)

    Camper, Larry W.; Michalak, Paul; Cohen, Stephen; Carter, Ted

    2012-07-01

    Community Water Systems (CWSs) are required to remove uranium from drinking water to meet EPA standards. Similarly, mining operations are required to remove uranium from their dewatering discharges to meet permitted surface water discharge limits. Ion exchange (IX) is the primary treatment strategy used by these operations, which loads uranium onto resin beads. Presently, uranium-loaded resin from CWSs and mining operations can be disposed as a waste product or processed by NRC- or Agreement State-licensed uranium recovery facilities if that licensed facility has applied for and received permission to process 'alternate feed'. The disposal of uranium-loaded resin is costly and the cost to amend a uranium recovery license to accept alternate feed can be a strong disincentive to commercial uranium recovery facilities. In response to this issue, the NRC issued a Regulatory Issue Summary (RIS) to clarify the agency's policy that uranium-loaded resin from CWSs and mining operations can be processed by NRC- or Agreement State-licensed uranium recovery facilities without the need for an alternate feed license amendment when these resins are essentially the same, chemically and physically, to resins that licensed uranium recovery facilities currently use (i.e., equivalent feed). NRC staff is clarifying its current alternate feed policy to declare IX resins as equivalent feed. This clarification is necessary to alleviate a regulatory and financial burden on facilities that filter uranium using IX resin, such as CWSs and mine dewatering operations. Disposing of those resins in a licensed facility could be 40 to 50 percent of the total operations and maintenance (O and M) cost for a CWS. Allowing uranium recovery facilities to treat these resins without requiring a license amendment lowers O and M costs and captures a valuable natural resource. (authors)

  18. 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.

  19. Assessment of Preferred Depleted Uranium Disposal Forms

    SciTech Connect (OSTI)

    Croff, A.G.; Hightower, J.R.; Lee, D.W.; Michaels, G.E.; Ranek, N.L.; Trabalka, J.R.

    2000-06-01

    The Department of Energy (DOE) is in the process of converting about 700,000 metric tons (MT) of depleted uranium hexafluoride (DUF6) containing 475,000 MT of depleted uranium (DU) to a stable form more suitable for long-term storage or disposal. Potential conversion forms include the tetrafluoride (DUF4), oxide (DUO2 or DU3O8), or metal. If worthwhile beneficial uses cannot be found for the DU product form, it will be sent to an appropriate site for disposal. The DU products are considered to be low-level waste (LLW) under both DOE orders and Nuclear Regulatory Commission (NRC) regulations. The objective of this study was to assess the acceptability of the potential DU conversion products at potential LLW disposal sites to provide a basis for DOE decisions on the preferred DU product form and a path forward that will ensure reliable and efficient disposal.

  20. Uranium enrichment management review: summary of analysis

    SciTech Connect (OSTI)

    Not Available

    1981-01-01

    In May 1980, the Assistant Secretary for Resource Applications within the Department of Energy requested that a group of experienced business executives be assembled to review the operation, financing, and management of the uranium enrichment enterprise as a basis for advising the Secretary of Energy. After extensive investigation, analysis, and discussion, the review group presented its findings and recommendations in a report on December 2, 1980. The following pages contain background material on which that final report was based. This report is arranged in chapters that parallel those of the uranium enrichment management review final report - chapters that contain summaries of the review group's discussion and analyses in six areas: management of operations and construction; long-range planning; marketing of enrichment services; financial management; research and development; and general management. Further information, in-depth analysis, and discussion of suggested alternative management practices are provided in five appendices.

  1. PROCESS OF PREPARING A FLUORIDE OF TETRAVLENT URANIUM

    DOE Patents [OSTI]

    Wheelwright, E.J.

    1959-02-17

    A method is described for producing a fluoride salt pf tetravalent uranium suitable for bomb reduction to metallic uranium. An aqueous solution of uranyl nitrate is treated with acetic acid and a nitrite-suppressor and then contacted with metallic lead whereby uranium is reduced from the hexavalent to the tetravalent state and soluble lead acetate is formed. Sulfate ions are then added to the solution to precipitate and remove the lead values. Hydrofluoric acid and alkali metal ions are then added causing the formation of an alkali metal uranium double-fluoride in which the uranium is in the tetravalent state. After recovery, this precipitate is suitable for using in the limited production of metallic uranium.

  2. 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.

  3. PROCESSING OF URANIUM-METAL-CONTAINING FUEL ELEMENTS

    DOE Patents [OSTI]

    Moore, R.H.

    1962-10-01

    A process is given for recovering uranium from neutronbombarded uranium- aluminum alloys. The alloy is dissolved in an aluminum halide--alkali metal halide mixture in which the halide is a mixture of chloride and bromide, the aluminum halide is present in about stoichiometric quantity as to uranium and fission products and the alkali metal halide in a predominant quantity; the uranium- and electropositive fission-products-containing salt phase is separated from the electronegative-containing metal phase; more aluminum halide is added to the salt phase to obtain equimolarity as to the alkali metal halide; adding an excess of aluminum metal whereby uranium metal is formed and alloyed with the excess aluminum; and separating the uranium-aluminum alloy from the fission- productscontaining salt phase. (AEC)

  4. 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.

  5. The uranium cylinder assay system for enrichment plant safeguards

    SciTech Connect (OSTI)

    Miller, Karen A; Swinhoe, Martyn T; Marlow, Johnna B; Menlove, Howard O; Rael, Carlos D; Iwamoto, Tomonori; Tamura, Takayuki; Aiuchi, Syun

    2010-01-01

    Safeguarding sensitive fuel cycle technology such as uranium enrichment is a critical component in preventing the spread of nuclear weapons. A useful tool for the nuclear materials accountancy of such a plant would be an instrument that measured the uranium content of UF{sub 6} cylinders. The Uranium Cylinder Assay System (UCAS) was designed for Japan Nuclear Fuel Limited (JNFL) for use in the Rokkasho Enrichment Plant in Japan for this purpose. It uses total neutron counting to determine uranium mass in UF{sub 6} cylinders given a known enrichment. This paper describes the design of UCAS, which includes features to allow for unattended operation. It can be used on 30B and 48Y cylinders to measure depleted, natural, and enriched uranium. It can also be used to assess the amount of uranium in decommissioned equipment and waste containers. Experimental measurements have been carried out in the laboratory and these are in good agreement with the Monte Carlo modeling results.

  6. Engineering assessment of inactive uranium mill tailings

    SciTech Connect (OSTI)

    Not Available

    1981-07-01

    The Grand Junction site has been reevaluated in order to revise the October 1977 engineering assessment of the problems resulting from the existence of radioactive uranium mill tailings at Grand Junction, Colorado. This engineering assessment has included the preparation of topographic maps, the performance of core drillings and radiometric measurements sufficient to determine areas and volumes of tailings and radiation exposures of individuals and nearby populations, the investigations of site hydrology and meteorology, and the evaluation and costing of alternative corrective actions. Radon gas released from the 1.9 million tons of tailings at the Grand Junction site constitutes the most significant environmental impact, although windblown tailings and external gamma radiation are also factors. The eight alternative actions presented herein range from millsite and off-site decontamination with the addition of 3 m of stabilization cover material (Option I), to removal of the tailings to remote disposal sites and decontamination of the tailings site (Options II through VIII). Cost estimates for the eight options range from about $10,200,000 for stabilization in-place to about $39,500,000 for disposal in the DeBeque area, at a distance of about 35 mi, using transportation by rail. If transportation to DeBeque were by truck, the cost estimated to be about $41,900,000. Three principal alternatives for the reprocessing of the Grand Junction tailings were examined: (a) heap leaching; (b) treatment at an existing mill; and (c) reprocessing at a new conventional mill constructed for tailings reprocessing. The cost of the uranium recovered would be about $200/lb by heap leach and $150/lb by conventional plant processes. The spot market price for uranium was $25/lb early in 1981. Therefore, reprocessing the tailings for uranium recovery appears not to be economically attractive.

  7. HIGH ENERGY RATE EXTRUSION OF URANIUM

    DOE Patents [OSTI]

    Lewis, L.

    1963-07-23

    A method of extruding uranium at a high energy rate is described. Conditions during the extrusion are such that the temperature of the metal during extrusion reaches a point above the normal alpha to beta transition, but the metal nevertheless remains in the alpha phase in accordance with the Clausius- Clapeyron equation. Upon exiting from the die, the metal automatically enters the beta phase, after which the metal is permitted to cool. (AEC)

  8. PROCESS OF SEPARATING PLUTONIUM FROM URANIUM

    DOE Patents [OSTI]

    Brown, H.S.; Hill, O.F.

    1958-09-01

    A process is presented for recovering plutonium values from aqueous solutions. It comprises forming a uranous hydroxide precipitate in such a plutonium bearing solution, at a pH of at least 5. The plutonium values are precipitated with and carried by the uranium hydroxide. The carrier precipitate is then redissolved in acid solution and the pH is adjusted to about 2.5, causing precipitation of the uranous hydroxide but leaving the still soluble plutonium values in solution.

  9. Colloids generation from metallic uranium fuel

    SciTech Connect (OSTI)

    Metz, C.; Fortner, J.; Goldberg, M.; Shelton-Davis, C.

    2000-07-20

    The possibility of colloid generation from spent fuel in an unsaturated environment has significant implications for storage of these fuels in the proposed repository at Yucca Mountain. Because colloids can act as a transport medium for sparingly soluble radionuclides, it might be possible for colloid-associated radionuclides to migrate large distances underground and present a human health concern. This study examines the nature of colloidal materials produced during corrosion of metallic uranium fuel in simulated groundwater at elevated temperature in an unsaturated environment. Colloidal analyses of the leachates from these corrosion tests were performed using dynamic light scattering and transmission electron microscopy. Results from both techniques indicate a bimodal distribution of small discrete particles and aggregates of the small particles. The average diameters of the small, discrete colloids are {approximately}3--12 nm, and the large aggregates have average diameters of {approximately}100--200 nm. X-ray diffraction of the solids from these tests indicates a mineral composition of uranium oxide or uranium oxy-hydroxide.

  10. LM Issues Final Programmatic Environmental Impact Statement on the Uranium

    Energy Savers [EERE]

    Leasing Program | Department of Energy Issues Final Programmatic Environmental Impact Statement on the Uranium Leasing Program LM Issues Final Programmatic Environmental Impact Statement on the Uranium Leasing Program April 8, 2014 - 6:26pm Addthis What does this project do? Goal 4. Optimize the use of land and assets The U.S. Department of Energy (DOE) has released the Final Uranium Leasing Program Programmatic Environmental Impact Statement (PEIS) to the public. The document can be found

  11. In Situ Biological Uranium Remediation within a Highly Contaminated Aquifer

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

    In Situ Biological Uranium Remediation within a Highly Contaminated Aquifer Matthew Ginder-Vogel1, Wei-Min Wu1, Jack Carley2, Phillip Jardine2, Scott Fendorf1 and Craig Criddle1 1Stanford University, Stanford, CA 2Oak Ridge National Laboratory, Oak Ridge, TN Microbial Respiration Figure 1. Uranium(VI) reduction is driven by microbial respiration resulting in the precipitation of uraninite. Uranium contamination of ground and surface waters has been detected at numerous sites throughout the

  12. Multiple Mechanisms of Uranium Immobilization by Cellulomonas sp. Strain

    Office of Scientific and Technical Information (OSTI)

    ES6 (Journal Article) | SciTech Connect Journal Article: Multiple Mechanisms of Uranium Immobilization by Cellulomonas sp. Strain ES6 Citation Details In-Document Search Title: Multiple Mechanisms of Uranium Immobilization by Cellulomonas sp. Strain ES6 Removal of hexavalent uranium (U(VI)) from aqueous solution was studied using a Gram-positive facultative anaerobe, Cellulomonas sp. strain ES6, under anaerobic, non-growth conditions in bicarbonate and PIPES buffers. Inorganic phosphate was

  13. Excess Uranium Inventory Management Plan | Department of Energy

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

    Plan Excess Uranium Inventory Management Plan 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 environmental cleanup and national security missions. The Plan is not a commitment to specific activities beyond those that have already been contracted nor is it a restriction on actions that the Department may undertake in the future as a result of changing

  14. Nuclear & Uranium - U.S. Energy Information Administration (EIA)

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

    Nuclear & Uranium Glossary › FAQS › Overview Data Status of U.S. Nuclear Outages (interactive) Summary Uranium & nuclear fuel Nuclear power plants Spent nuclear fuel International All nuclear data reports Analysis & Projections Major Topics Most popular Nuclear plants and reactors Projections Recurring Uranium All reports Browse by Tag Alphabetical Frequency Tag Cloud Current Issues & Trends See more › Updated EIA survey provides data on spent nuclear fuel in the United

  15. The US uranium industry: Regulatory and policy impediments

    SciTech Connect (OSTI)

    Drennen, T.E.; Glicken, J.

    1995-06-01

    The Energy Policy Act of 1992 required the DOE to develop recommendations and implement government programs to assist the domestic uranium industry in increasing export opportunities. In 1993, as part of that effort, the Office of Nuclear Energy identified several key factors that could (or have) significantly impact(ed) export opportunities for domestic uranium. This report addresses one of these factors: regulatory and policy impediments to the flow of uranium products between the US and other countries. It speaks primarily to the uranium market for civil nuclear power. Changes in the world political and economic order have changed US national security requirements, and the US uranium industry has found itself without the protected market it once enjoyed. An unlevel playing field for US uranium producers has resulted from a combination of geology, history, and a general US political philosophy of nonintervention that precludes the type of industrial policy practiced in other uranium-exporting countries. The US has also been hampered in its efforts to support the domestic uranium-producing industry by its own commitment to free and open global markets and by international agreements such as GATT and NAFTA. Several US policies, including the imposition of NRC fees and licensing costs and Harbor Maintenance fees, directly harm the competitiveness of the domestic uranium industry. Finally, requirements under US law, such as those in the 1979 Nuclear Nonproliferation Act, place very strict limits on the use of US-origin uranium, limitations not imposed by other uranium-producing countries. Export promotion and coordination are two areas in which the US can help the domestic uranium industry without violating existing trade agreements or other legal or policy constraints.

  16. Uranium Mill Tailings Radiation Control Act Sites Fact Sheet

    Energy Savers [EERE]

    This fact sheet provides information about the Uranium Mill Tailings Radiation Control Act Title I and II disposal and processing sites. The sites are managed by the U.S. Department of Energy Office of Legacy Management. Introduction The Uranium Mill Tailings Radiation Control Act (UMTRCA) of 1978 (Public Law 95-604) is a federal law that provides for the safe and environmentally sound disposal, long-term stabilization, and control of uranium mill tailings in a manner that minimizes or

  17. DOE Announces Policy for Managing Excess Uranium Inventory | Department of

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

    Energy Policy for Managing Excess Uranium Inventory DOE Announces Policy for Managing Excess Uranium Inventory March 12, 2008 - 10:52am Addthis WASHINGTON, DC - U.S. Secretary of Energy Samuel W. Bodman today released a Policy Statement on the management of the Department of Energy's (DOE) excess uranium inventory, providing the framework within which DOE will make decisions concerning future use and disposition of its inventory. During the coming year, DOE will continue its ongoing program

  18. DOE Announces Preferred Alternatives For Moab, Utah, Uranium Mill Tailings

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

    | Department of Energy Preferred Alternatives For Moab, Utah, Uranium Mill Tailings DOE Announces Preferred Alternatives For Moab, Utah, Uranium Mill Tailings April 6, 2005 - 11:33am Addthis WASHINGTON, DC - The U.S. Department of Energy today announced the department's preferred alternatives for remediation of the Moab, Utah, Uranium Mill Tailings Remedial Action Project Site: active groundwater remediation, and offsite disposal of the tailings pile and other contaminated materials to the

  19. DOE Extends Contract to Operate Depleted Uranium Hexafluoride Conversion

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

    Plants | Department of Energy Extends Contract to Operate Depleted Uranium Hexafluoride Conversion Plants DOE Extends Contract to Operate Depleted Uranium Hexafluoride Conversion Plants December 24, 2015 - 11:06am Addthis LEXINGTON, Ky. (Dec. 24, 2015) - The U.S. Department of Energy's Office of Environmental Management (EM) today announced it is extending its contract for Operations of Depleted Uranium Hexafluoride (DUF6) Conversion Facilities at Paducah, Kentucky and Portsmouth, Ohio for a

  20. DOE Extends Public Comment Period for Uranium Program Environmental Impact

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

    Statement | Department of Energy Uranium Program Environmental Impact Statement DOE Extends Public Comment Period for Uranium Program Environmental Impact Statement April 18, 2013 - 1:08pm Addthis Contractor, Bob Darr, S.M. Stoller Corporation Public Affairs, (720) 377-9672, ULinfo@lm.doe.gov GRAND JUNCTION, Colo. - The U.S. Department of Energy (DOE) today announced that the public comment period for the Draft Uranium Leasing Program Programmatic Environmental Impact Statement (ULP PEIS)