National Library of Energy BETA

Sample records for highly enriched uranium

  1. 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...

  2. 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...

  3. 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...

  4. 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...

  5. 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 ... ...

  6. 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...

  7. 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

  8. 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

  9. NNSA Authorizes Start-Up of Highly Enriched Uranium Materials...

    National Nuclear Security Administration (NNSA)

    Authorizes Start-Up of Highly Enriched Uranium Materials Facility at Y-12 | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the...

  10. German Pebble Bed Research Reactor Highly Enriched Uranium (HEU) Fuel

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

    German Pebble Bed Research Reactor Highly Enriched Uranium (HEU) Fuel Environmental Assessment Maxcine Maxted, DOE-SR Used Nuclear Fuel Program Manager June 24, 2014 Public Scoping Meeting

  11. German Pebble Bed Research Reactor Highly Enriched Uranium (HEU...

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

    Potential Acceptance and Disposition of German Pebble Bed Research Reactor Highly Enriched Uranium (HEU) Fuel Environmental Assessment Maxcine Maxted, DOE-SR Used Nuclear Fuel...

  12. 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:

  13. Belgium Highly Enriched Uranium and Plutonium Removals | National Nuclear

    National Nuclear Security Administration (NNSA)

    Security Administration Highly Enriched Uranium and Plutonium Removals March 24, 2014 Belgium has been a global leader in nonproliferation, working with the United States since 2006 to minimize highly enriched uranium (HEU) and plutonium inventories in Belgium through the return of a significant amount of HEU and plutonium to the United States. At the 2014 Nuclear Security Summit, the United States and Belgium announced the successful removal of all excess fresh HEU and plutonium from

  14. Italy Highly Enriched Uranium and Plutonium Removals | National Nuclear

    National Nuclear Security Administration (NNSA)

    Security Administration Highly Enriched Uranium and Plutonium Removals March 24, 2014 Italy has been a global leader in nuclear nonproliferation, working with the United States since 1997 to eliminate more than 100 kilograms of highly enriched uranium (HEU) and separated plutonium. At the 2014 Nuclear Security Summit, the United States and Italy announced the successful removal of all eligible fresh HEU and plutonium from Italy. These shipments were completed via a joint effort between the

  15. 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.

  16. 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.

  17. 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.

  18. Nickel container of highly-enriched uranium bodies and sodium

    DOE Patents [OSTI]

    Zinn, Walter H.

    1976-01-01

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

  19. 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.

  20. GTRI's Convert Program: Minimizing the Use of Highly Enriched Uranium |

    National Nuclear Security Administration (NNSA)

    National Nuclear Security Administration GTRI's Convert Program: Minimizing the Use of Highly Enriched Uranium May 29, 2014 Mission In 2004 NNSA established the Global Threat Reduction Initiative (GTRI) in the Office of Defense Nuclear Nonproliferation to, as quickly as possible, identify, secure, remove and/or facilitate the disposition of high risk vulnerable nuclear and radiological materials around the world that pose a threat to the United States and the international community. GTRI's

  1. 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

  2. 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.

  3. Realities of verifying the absence of highly enriched uranium (HEU) in gas centrifuge enrichment plants

    SciTech Connect (OSTI)

    Swindle, D.W.

    1990-03-01

    Over a two and one-half year period beginning in 1981, representatives of six countries (United States, United Kingdom, Federal Republic of Germany, Australia, The Netherlands, and Japan) and the inspectorate organizations of the International Atomic Energy Agency and EURATOM developed and agreed to a technically sound approach for verifying the absence of highly enriched uranium (HEU) in gas centrifuge enrichment plants. This effort, known as the Hexapartite Safeguards Project (HSP), led to the first international concensus on techniques and requirements for effective verification of the absence of weapons-grade nuclear materials production. Since that agreement, research and development has continued on the radiation detection technology-based technique that technically confirms the HSP goal is achievable. However, the realities of achieving the HSP goal of effective technical verification have not yet been fully attained. Issues such as design and operating conditions unique to each gas centrifuge plant, concern about the potential for sensitive technology disclosures, and on-site support requirements have hindered full implementation and operator support of the HSP agreement. In future arms control treaties that may limit or monitor fissile material production, the negotiators must recognize and account for the realities and practicalities in verifying the absence of HEU production. This paper will describe the experiences and realities of trying to achieve the goal of developing and implementing an effective approach for verifying the absence of HEU production. 3 figs.

  4. 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.

  5. Initial report on characterization of excess highly enriched uranium

    SciTech Connect (OSTI)

    1996-07-01

    DOE`s Office of Fissile Materials Disposition assigned to this Y-12 division the task of preparing a report on the 174.4 metric tons of excess highly enriched U. Characterization included identification by category, gathering existing data (assay), defining the likely needed processing steps for prepping for transfer to a blending site, and developing a range of preliminary cost estimates for those steps. Focus is on making commercial reactor fuel as a final disposition path.

  6. Utilization of non-weapons-grade plutonium and highly enriched uranium with

    Office of Scientific and Technical Information (OSTI)

    breeding of the {sup 233}U isotope in the VVER reactors using thorium and heavy water (Journal Article) | SciTech Connect Utilization of non-weapons-grade plutonium and highly enriched uranium with breeding of the {sup 233}U isotope in the VVER reactors using thorium and heavy water Citation Details In-Document Search Title: Utilization of non-weapons-grade plutonium and highly enriched uranium with breeding of the {sup 233}U isotope in the VVER reactors using thorium and heavy water A

  7. DOE to Remove 200 Metric Tons of Highly Enriched Uranium from U.S. Nuclear

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

    Weapons Stockpile | Department of Energy to Remove 200 Metric Tons of Highly Enriched Uranium from U.S. Nuclear Weapons Stockpile DOE to Remove 200 Metric Tons of Highly Enriched Uranium from U.S. Nuclear Weapons Stockpile November 7, 2005 - 12:38pm Addthis Will Be Redirected to Naval Reactors, Down-blended or Used for Space Programs WASHINGTON, DC - Secretary of Energy Samuel W. Bodman today announced that the Department of Energy's (DOE) National Nuclear Security Administration (NNSA) will

  8. Compact reaction cell for homogenizing and down-blanding highly enriched uranium metal

    DOE Patents [OSTI]

    McLean, II, William; Miller, Philip E.; Horton, James A.

    1995-01-01

    The invention is a specialized reaction cell for converting uranium metal to uranium oxide. In a preferred form, the reaction cell comprises a reaction chamber with increasing diameter along its length (e.g. a cylindrical chamber having a diameter of about 2 inches in a lower portion and having a diameter of from about 4 to about 12 inches in an upper portion). Such dimensions are important to achieve the necessary conversion while at the same time affording criticality control and transportability of the cell and product. The reaction chamber further comprises an upper port and a lower port, the lower port allowing for the entry of reactant gasses into the reaction chamber, the upper port allowing for the exit of gasses from the reaction chamber. A diffuser plate is attached to the lower port of the reaction chamber and serves to shape the flow of gas into the reaction chamber. The reaction cell further comprises means for introducing gasses into the reaction chamber and a heating means capable of heating the contents of the reaction chamber. The present invention also relates to a method for converting uranium metal to uranium oxide in the reaction cell of the present invention. The invention is useful for down-blending highly enriched uranium metal by the simultaneous conversion of highly enriched uranium metal and natural or depleted uranium metal to uranium oxide within the reaction cell.

  9. Compact reaction cell for homogenizing and down-blending highly enriched uranium metal

    DOE Patents [OSTI]

    McLean, W. II; Miller, P.E.; Horton, J.A.

    1995-05-02

    The invention is a specialized reaction cell for converting uranium metal to uranium oxide. In a preferred form, the reaction cell comprises a reaction chamber with increasing diameter along its length (e.g. a cylindrical chamber having a diameter of about 2 inches in a lower portion and having a diameter of from about 4 to about 12 inches in an upper portion). Such dimensions are important to achieve the necessary conversion while at the same time affording criticality control and transportability of the cell and product. The reaction chamber further comprises an upper port and a lower port, the lower port allowing for the entry of reactant gases into the reaction chamber, the upper port allowing for the exit of gases from the reaction chamber. A diffuser plate is attached to the lower port of the reaction chamber and serves to shape the flow of gas into the reaction chamber. The reaction cell further comprises means for introducing gases into the reaction chamber and a heating means capable of heating the contents of the reaction chamber. The present invention also relates to a method for converting uranium metal to uranium oxide in the reaction cell of the present invention. The invention is useful for down-blending highly enriched uranium metal by the simultaneous conversion of highly enriched uranium metal and natural or depleted uranium metal to uranium oxide within the reaction cell. 4 figs.

  10. 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.

  11. 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

  12. 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.

  13. 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.

  14. 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.

  15. 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.

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

    SciTech Connect (OSTI)

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

    2010-02-01

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

  18. 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...

  19. 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.

  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. Reactor Physics Measurements and Benchmark Specifications for Oak Ridge Highly Enriched Uranium Sphere (ORSphere)

    SciTech Connect (OSTI)

    Marshall, Margaret A.

    2014-11-04

    In the early 1970s Dr. John T. Mihalczo (team leader), J.J. Lynn, and J.R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) in an effort to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s. The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with the GODIVA I experiments. Additionally, various material reactivity worths, the surface material worth coefficient, the delayed neutron fraction, the prompt neutron decay constant, relative fission density, and relative neutron importance were all measured. The critical assembly, material reactivity worths, the surface material worth coefficient, and the delayed neutron fraction were all evaluated as benchmark experiment measurements. The reactor physics measurements are the focus of this paper; although for clarity the critical assembly benchmark specifications are briefly discussed.

  2. Reactor Physics Measurements and Benchmark Specifications for Oak Ridge Highly Enriched Uranium Sphere (ORSphere)

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

    Marshall, Margaret A.

    2014-11-04

    In the early 1970s Dr. John T. Mihalczo (team leader), J.J. Lynn, and J.R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) in an effort to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s. The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with themore » GODIVA I experiments. Additionally, various material reactivity worths, the surface material worth coefficient, the delayed neutron fraction, the prompt neutron decay constant, relative fission density, and relative neutron importance were all measured. The critical assembly, material reactivity worths, the surface material worth coefficient, and the delayed neutron fraction were all evaluated as benchmark experiment measurements. The reactor physics measurements are the focus of this paper; although for clarity the critical assembly benchmark specifications are briefly discussed.« less

  3. 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.

  4. 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.

  5. Why is weapons grade plutonium more hazardous to work with than highly enriched uranium?

    SciTech Connect (OSTI)

    Cournoyer, Michael E.; Costigan, Stephen A.; Schake, Bradley S.

    2015-08-01

    Highly Enriched Uranium and Weapons grade plutonium have assumed positions of dominant importance among the actinide elements because of their successful uses as explosive ingredients in nuclear weapons and the place they hold as key materials in the development of industrial use of nuclear power. While most chemists are familiar with the practical interest concerning HEU and WG Pu, fewer know the subtleties among their hazards. In this study, a primer is provided regarding the hazards associated with working with HEU and WG Pu metals and oxides. The care that must be taken to safely handle these materials is emphasized and the extent of the hazards is described. The controls needed to work with HEU and WG Pu metals and oxides are differentiated. Given the choice, one would rather work with HEU metal and oxides than WG Pu metal and oxides.

  6. Why is weapons grade plutonium more hazardous to work with than highly enriched uranium?

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

    Cournoyer, Michael E.; Costigan, Stephen A.; Schake, Bradley S.

    2015-08-01

    Highly Enriched Uranium and Weapons grade plutonium have assumed positions of dominant importance among the actinide elements because of their successful uses as explosive ingredients in nuclear weapons and the place they hold as key materials in the development of industrial use of nuclear power. While most chemists are familiar with the practical interest concerning HEU and WG Pu, fewer know the subtleties among their hazards. In this study, a primer is provided regarding the hazards associated with working with HEU and WG Pu metals and oxides. The care that must be taken to safely handle these materials is emphasizedmore » and the extent of the hazards is described. The controls needed to work with HEU and WG Pu metals and oxides are differentiated. Given the choice, one would rather work with HEU metal and oxides than WG Pu metal and oxides.« less

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

    SciTech Connect (OSTI)

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

    2010-09-30

    Over a period of 30 years, more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited for nuclear data validation and to form the basis for criticality safety benchmarks. A number of the Argonne ZPR/ZPPR critical assemblies have been evaluated as ICSBEP and IRPhEP benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. ZPR-3 Assembly 6 consisted of six phases, A through F. In each phase a critical configuration was constructed to simulate a very simple shape such as a slab, cylinder or sphere that could be analyzed with the limited analytical tools available in the 1950s. In each case the configuration consisted of a core region of metal plates surrounded by a thick depleted uranium metal reflector. The average compositions of the core configurations were essentially identical in phases A - F. ZPR-3 Assembly 6F (ZPR-3/6F), the final phase of the Assembly 6 program, simulated a spherical core with a thick depleted uranium reflector. ZPR-3/6F was designed as a fast reactor physics benchmark experiment with an average core {sup 235}U enrichment of approximately 47 at.%. Approximately 81.4% of the total fissions in this assembly occur above 100 keV, approximately 18.6% occur below 100 keV, and essentially none below 0.625 eV - thus the classification as a 'fast' assembly. This assembly is Fast Reactor Benchmark No. 7 in the Cross Section Evaluation Working Group (CSEWG) Benchmark Specifications and has historically been used as a data validation benchmark assembly. Loading of ZPR-3/6F began in late December 1956, and the experimental measurements were performed in January 1957. The core consisted of highly enriched uranium (HEU) plates, depleted uranium plates, perforated aluminum plates and stainless steel plates loaded into aluminum drawers, which were inserted into the central square stainless steel tubes of a 31 x 31 matrix on a split table machine. The core unit cell consisted of three columns of 0.125 in.-wide (3.175 mm) HEU plates, three columns of 0.125 in.-wide depleted uranium plates, nine columns of 0.125 in.-wide perforated aluminum plates and one column of stainless steel plates. The maximum length of each column of core material in a drawer was 9 in. (228.6 mm). Because of the goal to produce an approximately spherical core, core fuel and diluent column lengths generally varied between adjacent drawers and frequently within an individual drawer. The axial reflector consisted of depleted uranium plates and blocks loaded in the available space in the front (core) drawers, with the remainder loaded into back drawers behind the front drawers. The radial reflector consisted of blocks of depleted uranium loaded directly into the matrix tubes. The assembly geometry approximated a reflected sphere as closely as the square matrix tubes, the drawers and the shapes of fuel and diluent plates allowed. According to the logbook and loading records for ZPR-3/6F, the reference critical configuration was loading 5 which was critical on January 4, 1957. The subsequent loadings were very similar but were less clean for criticality because there were modifications made to accommodate reactor physics measurements other than criticality. Accordingly, ZPR-3/6F loading 5 was selected as the only configuration for this benchmark. As documented below, it was determined to be acceptable as a criticality safety benchmark experiment. A very accurate transformation to a simplified model is needed to make any ZPR assembly a practical criticality-safety benchmark. There is simply too much geometric detail in an exact (as-built) model of a ZPR assembly. This is especially true of ZPR-3/6F because of the complex core loading required to approximate a sphere with rectangular plates in a square matrix.

  8. 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.

  9. ES-3100: A New Generation Shipping Container for Bulk Highly Enriched Uranium and Other Fissile Materials

    SciTech Connect (OSTI)

    Arbital, J.G.; Byington, G.A.; Tousley, D.R.

    2004-07-01

    The U.S. Department of Energy (DOE) National Nuclear Security Administration (NNSA) is shipping bulk quantities of surplus fissile materials, primarily highly enriched uranium (HEU), over the next 15 to 20 years for disposition purposes. The U.S. Department of Transportation (DOT) specification 6M container is the package of choice for most of these shipments. However, the 6M does not conform to the Type B packaging requirements in the ''Code of Federal Regulations'' (10CFR71) and, for that reason, is being phased out for use in the secure transportation system of DOE. BWXT Y-12 is currently developing a package to replace the DOT 6M container for HEU disposition shipping campaigns. The new package is based on state-of-the-art, proven, and patented insulation technologies that have been successfully applied in the design of other packages. The new package, designated the ES-3100, will have a 50% greater capacity for HEU than the 6M and will be easier to use. Engineering analysis on the new package includes detailed dynamic impact finite element analysis (FEA). This analysis gives the ES-3100 a high probability of complying with regulatory requirements.

  10. REMOVAL OF SOLIDS FROM HIGHLY ENRICHED URANIUM SOLUTIONS USING THE H-CANYON CENTRIFUGE

    SciTech Connect (OSTI)

    Rudisill, T; Fernando Fondeur, F

    2009-01-15

    Prior to the dissolution of Pu-containing materials in HB-Line, highly enriched uranium (HEU) solutions stored in Tanks 11.1 and 12.2 of H-Canyon must be transferred to provide storage space. The proposed plan is to centrifuge the solutions to remove solids which may present downstream criticality concerns or cause operational problems with the 1st Cycle solvent extraction due to the formation of stable emulsions. An evaluation of the efficiency of the H-Canyon centrifuge concluded that a sufficient amount (> 90%) of the solids in the Tank 11.1 and 12.2 solutions will be removed to prevent any problems. We based this conclusion on the particle size distribution of the solids isolated from samples of the solutions and the calculation of particle settling times in the centrifuge. The particle size distributions were calculated from images generated by scanning electron microscopy (SEM). The mean particle diameters for the distributions were 1-3 {micro}m. A significant fraction (30-50%) of the particles had diameters which were < 1 {micro}m; however, the mass of these solids is insignificant (< 1% of the total solids mass) when compared to particles with larger diameters. It is also probable that the number of submicron particles was overestimated by the software used to generate the particle distribution due to the morphology of the filter paper used to isolate the solids. The settling times calculated for the H-Canyon centrifuge showed that particles with diameters less than 1 to 0.5 {micro}m will not have sufficient time to settle. For this reason, we recommend the use of a gelatin strike to coagulate the submicron particles and facilitate their removal from the solution; although we have no experimental basis to estimate the level of improvement. Incomplete removal of particles with diameters < 1 {micro}m should not cause problems during purification of the HEU in the 1st Cycle solvent extraction. Particles with diameters > 1 {micro}m account for > 99% of the solid mass and will be efficiently removed by the centrifuge; therefore, the formation of emulsions during solvent extraction operations is not an issue. Under the current processing plan, the solutions from Tanks 11.1 and 12.2 will be transferred to the enriched uranium storage (EUS) tank following centrifugation. The solution from Tanks 11.1 and 12.2 may remain in the EUS tank for an extended time prior to purification. The effects of extended storage on the solution were not evaluated as part of this study.

  11. ZPR-3 Assembly 12 : A cylindrical assembly of highly enriched uranium, depleted uranium and graphite with an average {sup 235}U enrichment of 21 atom %.

    SciTech Connect (OSTI)

    Lell, R. M.; McKnight, R. D.; Perel, R. L.; Wagschal, J. J.; Nuclear Engineering Division; Racah Inst. of Physics

    2010-09-30

    Over a period of 30 years, more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited for nuclear data validation and to form the basis for criticality safety benchmarks. A number of the Argonne ZPR/ZPPR critical assemblies have been evaluated as ICSBEP and IRPhEP benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. ZPR-3 Assembly 12 (ZPR-3/12) was designed as a fast reactor physics benchmark experiment with an average core {sup 235}U enrichment of approximately 21 at.%. Approximately 68.9% of the total fissions in this assembly occur above 100 keV, approximately 31.1% occur below 100 keV, and essentially none below 0.625 eV - thus the classification as a 'fast' assembly. This assembly is Fast Reactor Benchmark No. 9 in the Cross Section Evaluation Working Group (CSEWG) Benchmark Specifications and has historically been used as a data validation benchmark assembly. Loading of ZPR-3 Assembly 12 began in late Jan. 1958, and the Assembly 12 program ended in Feb. 1958. The core consisted of highly enriched uranium (HEU) plates, depleted uranium plates and graphite plates loaded into stainless steel drawers which were inserted into the central square stainless steel tubes of a 31 x 31 matrix on a split table machine. The core unit cell consisted of two columns of 0.125 in.-wide (3.175 mm) HEU plates, seven columns of 0.125 in.-wide depleted uranium plates and seven columns of 0.125 in.-wide graphite plates. The length of each column was 9 in. (228.6 mm) in each half of the core. The graphite plates were included to produce a softer neutron spectrum that would be more characteristic of a large power reactor. The axial blanket consisted of 12 in. (304.8 mm) of depleted uranium behind the core. The thickness of the radial blanket was approximately 12 in. and the length of the radial blanket in each half of the matrix was 21 in. (533.4 mm). The assembly geometry approximated a right circular cylinder as closely as the square matrix tubes allowed. According to the logbook and loading records for ZPR-3/12, the reference critical configuration was loading 10 which was critical on Feb. 5, 1958. The subsequent loadings were very similar but less clean for criticality because there were modifications made to accommodate reactor physics measurements other than criticality. Accordingly, ZPR-3/12 loading 10 was selected as the only configuration for this benchmark. As documented below, it was determined to be acceptable as a criticality safety benchmark experiment. An accurate transformation to a simplified model is needed to make any ZPR assembly a practical criticality-safety benchmark. There is simply too much geometric detail in an exact (as-built) model of a ZPR assembly, even a clean core such as ZPR-3/12 loading 10. The transformation must reduce the detail to a practical level without masking any of the important features of the critical experiment. And it must d

  12. 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.

  13. EVALUATION OF FLOWSHEET CHANGES FOR THE HIGHLY ENRICHED URANIUM BLENDDOWN PROGRAM

    SciTech Connect (OSTI)

    Crowder, M.; Rudisill, T.; Laurinat, J.; Mickalonis, J.

    2007-10-22

    H Canyon is considering a flowsheet change for Plutonium (Pu) Contaminated Scrap (PuCS) material. The proposed change is to route dissolved PuCS material directly to a uranium (U) storage tank. As a result, the PuCS solution will bypass Head End and First U Cycle, and will be purified by solvent extraction in Second U Cycle. The PuCS solution contains appreciable amounts of boron (B) and fluoride (F{sup -}), which are currently at trace levels in the U storage tank. Though unlikely, if the B concentration in the U storage tank were to reach 1.8 g B/g U, the entire contents of the U storage tank would likely require a second pass through Second U Cycle to provide sufficient decontamination to meet the Tennessee Valley Authority (TVA) Blend Grade Highly Enriched Uranium (HEU) specification for B, which is 30 {micro}g/g U. In addition, Second U Cycle is expected to provide sufficient decontamination of F{sup -} and Pu regardless of the amount of PuCS solution sent to the storage tank. Though aluminum (Al) is not present in the PuCS solution, B can be credited as a complexant of F{sup -}. Both stability constants from the literature and Savannah River National Laboratory (SRNL) corrosion studies were documented to demonstrate that B complexation of F{sup -} in nitric acid solutions is sufficient to prevent excessive corrosion. Though B and Al complex F{sup -} to a similar degree, neither completely eliminates the presence of free F{sup -} in solution. Therefore, a limited amount of corrosion is expected even with complexed F{sup -} solutions. Tanks maintained at ambient temperature are not expected to experience significant corrosion. However, the Low Activity Waste (LAW) evaporators may be subjected to a corrosion rate of about 25 mils per year (mpy) as they reach their highest F{sup -} concentrations. The feed adjustment evaporator would only be subjected to the corrosion rate of about 25 mpy in the latter stages of the PuCS campaign. An issue that must be addressed as part of the proposed PuCS flowsheet change is that B has limited solubility in concentrated nitric acid solutions. As the proposed PuCS campaign progresses, the B concentration will increase in the U storage tank, in Second U Cycle feed, and in the 1DW stream sent to the LAW evaporators. Limitations on the B concentration in the LAW evaporators will be needed to prevent formation of boron-containing solids.

  14. Transmutation Analysis of Enriched Uranium and Deep Burn High Temperature Reactors

    SciTech Connect (OSTI)

    Michael A. Pope

    2012-07-01

    High temperature reactors (HTRs) have been under consideration for production of electricity, process heat, and for destruction of transuranics for decades. As part of the transmutation analysis efforts within the Fuel Cycle Research and Development (FCR&D) campaign, a need was identified for detailed discharge isotopics from HTRs for use in the VISION code. A conventional HTR using enriched uranium in UCO fuel was modeled having discharge burnup of 120 GWd/MTiHM. Also, a deep burn HTR (DB-HTR) was modeled burning transuranic (TRU)-only TRU-O2 fuel to a discharge burnup of 648 GWd/MTiHM. For each of these cases, unit cell depletion calculations were performed with SCALE/TRITON. Unit cells were used to perform this analysis using SCALE 6.1. Because of the long mean free paths (and migration lengths) of neutrons in HTRs, using a unit cell to represent a whole core can be non-trivial. The sizes of these cells were first set by using Serpent calculations to match a spectral index between unit cell and whole core domains. In the case of the DB-HTR, the unit cell which was arrived at in this way conserved the ratio of fuel to moderator found in a single block of fuel. In the conventional HTR case, a larger moderator-to-fuel ratio than that of a single block was needed to simulate the whole core spectrum. Discharge isotopics (for 500 nuclides) and one-group cross-sections (for 1022 nuclides) were delivered to the transmutation analysis team. This report provides documentation for these calculations. In addition to the discharge isotopics, one-group cross-sections were provided for the full list of 1022 nuclides tracked in the transmutation library.

  15. AIR SHIPMENT OF HIGHLY ENRICHED URANIUM SPENT NUCLEAR FUEL FROM ROMANIA AND LIBYA

    SciTech Connect (OSTI)

    Christopher Landers; Igor Bolshinsky; Ken Allen; Stanley Moses

    2010-07-01

    In June 2009 Romania successfully completed the worlds first air shipment of highly enriched uranium (HEU) spent nuclear fuel transported in Type B(U) casks under existing international laws and without special exceptions for the air transport licenses. Special 20-foot ISO shipping containers and cask tiedown supports were designed to transport Russian TUK 19 shipping casks for the Romanian air shipment and the equipment was certified for all modes of transport, including road, rail, water, and air. In December 2009 Libya successfully used this same equipment for a second air shipment of HEU spent nuclear fuel. Both spent fuel shipments were transported by truck from the originating nuclear facilities to nearby commercial airports, were flown by commercial cargo aircraft to a commercial airport in Yekaterinburg, Russia, and then transported by truck to their final destinations at the Production Association Mayak facility in Chelyabinsk, Russia. Both air shipments were performed under the Russian Research Reactor Fuel Return Program (RRRFR) as part of the U.S. National Nuclear Security Administration (NNSA) Global Threat Reduction Initiative (GTRI). The Romania air shipment of 23.7 kg of HEU spent fuel from the VVR S research reactor was the last of three HEU fresh and spent fuel shipments under RRRFR that resulted in Romania becoming the 3rd RRRFR participating country to remove all HEU. Libya had previously completed two RRRFR shipments of HEU fresh fuel so the 5.2 kg of HEU spent fuel air shipped from the IRT 1 research reactor in December made Libya the 4th RRRFR participating country to remove all HEU. This paper describes the equipment, preparations, and license approvals required to safely and securely complete these two air shipments of spent nuclear fuel.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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

  1. 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.

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

    SciTech Connect (OSTI)

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

    2010-09-30

    Over a period of 30 years, more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited for nuclear data validation and to form the basis for criticality safety benchmarks. A number of the Argonne ZPR/ZPPR critical assemblies have been evaluated as ICSBEP and IRPhEP benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. ZPR-3 Assembly 11 (ZPR-3/11) was designed as a fast reactor physics benchmark experiment with an average core {sup 235}U enrichment of approximately 12 at.% and a depleted uranium reflector. Approximately 79.7% of the total fissions in this assembly occur above 100 keV, approximately 20.3% occur below 100 keV, and essentially none below 0.625 eV - thus the classification as a 'fast' assembly. This assembly is Fast Reactor Benchmark No. 8 in the Cross Section Evaluation Working Group (CSEWG) Benchmark Specificationsa and has historically been used as a data validation benchmark assembly. Loading of ZPR-3 Assembly 11 began in early January 1958, and the Assembly 11 program ended in late January 1958. The core consisted of highly enriched uranium (HEU) plates and depleted uranium plates loaded into stainless steel drawers, which were inserted into the central square stainless steel tubes of a 31 x 31 matrix on a split table machine. The core unit cell consisted of two columns of 0.125 in.-wide (3.175 mm) HEU plates, six columns of 0.125 in.-wide (3.175 mm) depleted uranium plates and one column of 1.0 in.-wide (25.4 mm) depleted uranium plates. The length of each column was 10 in. (254.0 mm) in each half of the core. The axial blanket consisted of 12 in. (304.8 mm) of depleted uranium behind the core. The thickness of the depleted uranium radial blanket was approximately 14 in. (355.6 mm), and the length of the radial blanket in each half of the matrix was 22 in. (558.8 mm). The assembly geometry approximated a right circular cylinder as closely as the square matrix tubes allowed. According to the logbook and loading records for ZPR-3/11, the reference critical configuration was loading 10 which was critical on January 21, 1958. Subsequent loadings were very similar but less clean for criticality because there were modifications made to accommodate reactor physics measurements other than criticality. Accordingly, ZPR-3/11 loading 10 was selected as the only configuration for this benchmark. As documented below, it was determined to be acceptable as a criticality safety benchmark experiment. A very accurate transformation to a simplified model is needed to make any ZPR assembly a practical criticality-safety benchmark. There is simply too much geometric detail in an exact (as-built) model of a ZPR assembly, even a clean core such as ZPR-3/11 loading 10. The transformation must reduce the detail to a practical level without masking any of the important features of the critical experiment. And it must do this without increasing the total uncertain

  3. 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 ...

  4. 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.

  5. Gamma/neutron time-correlation for special nuclear material detection Active stimulation of highly enriched uranium

    SciTech Connect (OSTI)

    Paff, Marc G.; Monterial, Mateusz; Marleau, Peter; Kiff, Scott; Nowack, Aaron; Clarke, Shaun D.; Pozzi, Sara A.

    2014-06-21

    A series of simulations and experiments were undertaken to explore and evaluate the potential for a novel new technique for fissile material detection and characterization, the timecorrelated pulse-height (TCPH) method, to be used concurrent with active stimulation of potential nuclear materials. In previous work TCPH has been established as a highly sensitive method for the detection and characterization of configurations of fissile material containing Plutonium in passive measurements. By actively stimulating fission with the introduction of an external radiation source, we have shown that TCPH is also an effective method of detecting and characterizing configurations of fissile material containing Highly Enriched Uranium (HEU). The TCPH method is shown to be robust in the presence of the proper choice of external radiation source. An evaluation of potential interrogation sources is presented.

  6. Parametric Evaluation of Active Neutron Interrogation for the Detection of Shielded Highly-Enriched Uranium in the Field

    SciTech Connect (OSTI)

    D. L. Chcihester; E. H. Seabury; S. J. Thompson; R. R. C. Clement

    2011-10-01

    Parametric studies using numerical simulations are being performed to assess the performance capabilities and limits of active neutron interrogation for detecting shielded highly enriched uranium (HEU). Varying the shield material, HEU mass, HEU depth inside the shield, and interrogating neutron source energy, the simulations account for both neutron and photon emission signatures from the HEU with resolution in both energy and time. The results are processed to represent different irradiation timing schemes and several different classes of radiation detectors, and evaluated using a statistical approach considering signal intensity over background. This paper describes the details of the modeling campaign and some preliminary results, weighing the strengths of alternative measurement approaches for the different irradiation scenarios.

  7. Gamma/neutron time-correlation for special nuclear material detection – Active stimulation of highly enriched uranium

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

    Paff, Marc G.; Monterial, Mateusz; Marleau, Peter; Kiff, Scott; Nowack, Aaron; Clarke, Shaun D.; Pozzi, Sara A.

    2014-06-21

    A series of simulations and experiments were undertaken to explore and evaluate the potential for a novel new technique for fissile material detection and characterization, the timecorrelated pulse-height (TCPH) method, to be used concurrent with active stimulation of potential nuclear materials. In previous work TCPH has been established as a highly sensitive method for the detection and characterization of configurations of fissile material containing Plutonium in passive measurements. By actively stimulating fission with the introduction of an external radiation source, we have shown that TCPH is also an effective method of detecting and characterizing configurations of fissile material containing Highlymore » Enriched Uranium (HEU). The TCPH method is shown to be robust in the presence of the proper choice of external radiation source. An evaluation of potential interrogation sources is presented.« less

  8. 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

  9. Verification of the MCU precision code and ROSFOND neutron data in application to the calculations of criticality of fast reactors with highly enriched uranium

    SciTech Connect (OSTI)

    Alekseev, N. I.; Kalugin, M. A.; Kulakov, A. S.; Novoseltsev, A. P.; Sergeev, G. S.; Shkarovskiy, D. A.; Yudkevich, M. S.

    2014-12-15

    Calculation of 335 critical assemblies (benchmark experiments) with the core of highly enriched uranium and reflectors of various materials is performed. The statistical analysis of the results shows that, for all 16 materials studied, the absolute value of the most probable deviation of the calculated value of K{sub eff} from the experimental one does not exceed 0.005.

  10. 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.

  11. The Complete Burning of Weapons Grade Plutonium and Highly Enriched Uranium with (Laser Inertial Fusion-Fission Energy) LIFE Engine

    SciTech Connect (OSTI)

    Farmer, J C; Diaz de la Rubia, T; Moses, E

    2008-12-23

    The National Ignition Facility (NIF) project, a laser-based Inertial Confinement Fusion (ICF) experiment designed to achieve thermonuclear fusion ignition and burn in the laboratory, is under construction at the Lawrence Livermore National Laboratory (LLNL) and will be completed in April of 2009. Experiments designed to accomplish the NIF's goal will commence in late FY2010 utilizing laser energies of 1 to 1.3 MJ. Fusion yields of the order of 10 to 20 MJ are expected soon thereafter. Laser initiated fusion-fission (LIFE) engines have now been designed to produce nuclear power from natural or depleted uranium without isotopic enrichment, and from spent nuclear fuel from light water reactors without chemical separation into weapons-attractive actinide streams. A point-source of high-energy neutrons produced by laser-generated, thermonuclear fusion within a target is used to achieve ultra-deep burn-up of the fertile or fissile fuel in a sub-critical fission blanket. Fertile fuels including depleted uranium (DU), natural uranium (NatU), spent nuclear fuel (SNF), and thorium (Th) can be used. Fissile fuels such as low-enrichment uranium (LEU), excess weapons plutonium (WG-Pu), and excess highly-enriched uranium (HEU) may be used as well. Based upon preliminary analyses, it is believed that LIFE could help meet worldwide electricity needs in a safe and sustainable manner, while drastically shrinking the nation's and world's stockpile of spent nuclear fuel and excess weapons materials. LIFE takes advantage of the significant advances in laser-based inertial confinement fusion that are taking place at the NIF at LLNL where it is expected that thermonuclear ignition will be achieved in the 2010-2011 timeframe. Starting from as little as 300 to 500 MW of fusion power, a single LIFE engine will be able to generate 2000 to 3000 MWt in steady state for periods of years to decades, depending on the nuclear fuel and engine configuration. Because the fission blanket in a fusion-fission hybrid system is subcritical, a LIFE engine can burn any fertile or fissile nuclear material, including unenriched natural or depleted U and SNF, and can extract a very high percentage of the energy content of its fuel resulting in greatly enhanced energy generation per metric ton of nuclear fuel, as well as nuclear waste forms with vastly reduced concentrations of long-lived actinides. LIFE engines could thus provide the ability to generate vast amounts of electricity while greatly reducing the actinide content of any existing or future nuclear waste and extending the availability of low cost nuclear fuels for several thousand years. LIFE also provides an attractive pathway for burning excess weapons Pu to over 99% FIMA (fission of initial metal atoms) without the need for fabricating or reprocessing mixed oxide fuels (MOX). Because of all of these advantages, LIFE engines offer a pathway toward sustainable and safe nuclear power that significantly mitigates nuclear proliferation concerns and minimizes nuclear waste. An important aspect of a LIFE engine is the fact that there is no need to extract the fission fuel from the fission blanket before it is burned to the desired final level. Except for fuel inspection and maintenance process times, the nuclear fuel is always within the core of the reactor and no weapons-attractive materials are available outside at any point in time. However, an important consideration when discussing proliferation concerns associated with any nuclear fuel cycle is the ease with which reactor fuel can be converted to weapons usable materials, not just when it is extracted as waste, but at any point in the fuel cycle. Although the nuclear fuel remains in the core of the engine until ultra deep actinide burn up is achieved, soon after start up of the engine, once the system breeds up to full power, several tons of fissile material is present in the fission blanket. However, this fissile material is widely dispersed in millions of fuel pebbles, which can be tagged as individual accountable items, and thus made difficult to divert in large quantities. This report discusses the application of the LIFE concept to nonproliferation issues, initially looking at the LIFE (Laser Inertial Fusion-Fission Energy) engine as a means of completely burning WG Pu and HEU. By combining a neutron-rich inertial fusion point source with energy-rich fission, the once-through closed fuel-cycle LIFE concept has the following characteristics: it is capable of efficiently burning excess weapons or separated civilian plutonium and highly enriched uranium; the fission blanket is sub-critical at all times (keff < 0.95); because LIFE can operate well beyond the point at which light water reactors (LWRs) need to be refueled due to burn-up of fissile material and the resulting drop in system reactivity, fuel burn-up of 99% or more appears feasible. The objective of this work is to develop LIFE technology for burning of WG-Pu and HEU.

  12. An Optically Stimulated Luminescence Uranium Enrichment Monitor

    SciTech Connect (OSTI)

    Miller, Steven D.; Tanner, Jennifer E.; Simmons, Kevin L.; Conrady, Matthew M.; Benz, Jacob M.; Greenfield, Bryce A.

    2010-08-11

    The Pacific Northwest National Laboratory (PNNL) has pioneered the use of Optically Stimulated Luminescence (OSL) technology for use in personnel dosimetry and high dose radiation processing dosimetry. PNNL has developed and patented an alumina-based OSL dosimeter that is being used by the majority of medical X-ray and imaging technicians worldwide. PNNL has conceived of using OSL technology to passively measure the level of UF6 enrichment by attaching the prototype OSL monitor to pipes containing UF6 gas within an enrichment facility. The prototype OSL UF6 monitor utilizes a two-element approach with the first element open and unfiltered to measure both the low energy and high energy gammas from the UF6, while the second element utilizes a 3-mm thick tungsten filter to eliminate the low energy gammas and pass only the high energy gammas from the UF6. By placing a control monitor in the room away from the UF6 pipes and other ionizing radiation sources, the control readings can be subtracted from the UF6 pipe monitor measurements. The ratio of the shielded to the unshielded net measurements provides a means to estimate the level of uranium enrichment. PNNL has replaced the commercially available MicroStar alumina-based dosimeter elements with a composite of polyethylene plastic, high-Z glass powder, and BaFBr:Eu OSL phosphor powder at various concentrations. The high-Z glass was added in an attempt to raise the average Z of the composite dosimeter and increase the response. Additionally, since BaFBr:Eu OSL phosphor is optimally excited and emits light at different wavelengths compared to alumina, the commercially available MicroStar reader was modified for reading BaFBr:Eu in a parallel effort to increase reader sensitivity. PNNL will present the design and performance of our novel OSL uranium enrichment monitor based on a combination of laboratory and UF6 test loop measurements. PNNL will also report on the optimization effort to achieve the highest possible performance from both the OSL enrichment monitor and the new custom OSL reader modified for this application. This project has been supported by the US Department of Energys National Nuclear Security Administrations Office of Dismantlement and Transparency (DOE/NNSA/NA-241).

  13. 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...

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

    SciTech Connect (OSTI)

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

    1996-12-31

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

  15. Successful Completion of the Largest Shipment of Russian Research Reactor High-Enriched Uranium Spent Nuclear Fuel from Czech Republic to Russian Federation

    SciTech Connect (OSTI)

    Michael Tyacke; Dr. Igor Bolshinsky; Jeff Chamberlin

    2008-07-01

    On December 8, 2007, the largest shipment of high-enriched uranium spent nuclear fuel was successfully made from a Russian-designed nuclear research reactor in the Czech Republic to the Russian Federation. This accomplishment is the culmination of years of planning, negotiations, and hard work. The United States, Russian Federation, and the International Atomic Energy Agency have been working together on the Russian Research Reactor Fuel Return (RRRFR) Program in support of the Global Threat Reduction Initiative. In February 2003, RRRFR Program representatives met with the Nuclear Research Institute in Re, Czech Republic, and discussed the return of their high-enriched uranium spent nuclear fuel to the Russian Federation for reprocessing. Nearly 5 years later, the shipment was made. This paper discusses the planning, preparations, coordination, and cooperation required to make this important international shipment.

  16. 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.

  17. EA-1977: Acceptance and Disposition of Used Nuclear Fuel Containing U.S.-Origin Highly Enriched Uranium from the Federal Republic of Germany

    Broader source: Energy.gov [DOE]

    This environmental assessment (EA) will evaluate the potential environmental impacts of a DOE proposal to accept used nuclear fuel from the Federal Republic of Germany at DOEs Savannah River Site (SRS) for processing and disposition. This used nuclear fuel is composed of kernels containing thorium and U.S.-origin highly enriched uranium (HEU) embedded in small graphite spheres that were irradiated in nuclear reactors used for research and development purposes.

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

    SciTech Connect (OSTI)

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

    2015-12-15

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

  19. 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

  20. Enrichment Determination of Uranium in Shielded Configurations

    SciTech Connect (OSTI)

    Crye, Jason Michael; Hall, Howard L; McConchie, Seth M; Mihalczo, John T; Pena, Kirsten E

    2011-01-01

    The determination of the enrichment of uranium is required in many safeguards and security applications. Typical methods of determining the enrichment rely on detecting the 186 keV gamma ray emitted by {sup 235}U. In some applications, the uranium is surrounded by external shields, and removal of the shields is undesirable. In these situations, methods relying on the detection of the 186 keV gamma fail because the gamma ray is shielded easily. Oak Ridge National Laboratory (ORNL) has previously measured the enrichment of shielded uranium metal using active neutron interrogation. The method consists of measuring the time distribution of fast neutrons from induced fissions with large plastic scintillator detectors. To determine the enrichment, the measurements are compared to a calibration surface that is created from Monte Carlo simulations where the enrichment in the models is varied. In previous measurements, the geometry was always known. ORNL is extending this method to situations where the geometry and materials present are not known in advance. In the new method, the interrogating neutrons are both time and directionally tagged, and an array of small plastic scintillators measures the uncollided interrogating neutrons. Therefore, the attenuation through the item along many different paths is known. By applying image reconstruction techniques, an image of the item is created which shows the position-dependent attenuation. The image permits estimating the geometry and materials present, and these estimates are used as input for the Monte Carlo simulations. As before, simulations predict the time distribution of induced fission neutrons for different enrichments. Matching the measured time distribution to the closest prediction from the simulations provides an estimate of the enrichment. This presentation discusses the method and provides results from recent simulations that show the importance of knowing the geometry and materials from the imaging system.

  1. 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.

  2. highly enriched uranium

    National Nuclear Security Administration (NNSA)

    NorthStar Medical Radioisotopes to further develop its technology to produce Mo-99 via neutron capture, bringing the total NNSA support to this project to the maximum of 25...

  3. Enterprise Assessments Targeted Review of the Safety System Management of the Secondary Confinement System and Power Distribution Safety System at the Y-12 National Security Complex Highly Enriched Uranium Materials Facility – December 2015

    Broader source: Energy.gov [DOE]

    Targeted Review of the Safety System Management of the Secondary Confinement System and Power Distribution Safety System at the Y-12 National Security Complex Highly Enriched Uranium Materials Facility

  4. Environmental Assessment for the Transportation of Highly Enriched Uranium from the Russian Federation for the Y-12 National Security Complex and Finding of No Significant Impact

    SciTech Connect (OSTI)

    2004-01-01

    The United States (U.S.) Department of Energy (DOE) proposes to transport highly enriched uranium (HEU) from Russia to a secure storage facility in Oak Ridge, TN. This proposed action would allow the United States and Russia to accelerate the disposition of excess nuclear weapons materials in the interest of promoting nuclear disarmament, strengthening nonproliferation, and combating terrorism. The HEU would be used for a non-weapons purpose in the U.S. as fuel in research reactors performing solely peaceful missions.

  5. US developments in technology for uranium enrichment

    SciTech Connect (OSTI)

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

    1982-01-01

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

  6. Toxic Substances Control Act Uranium Enrichment Federal Facilities...

    Office of Environmental Management (EM)

    Toxic Substance Control Act Uranium Enrichment Federal Facilities Compliance Agreement (TSCA-UE- FFCA), February 20, 1992 State Ohio Agreement Type Compliance Agreement Legal ...

  7. Experiments and Simulations of the Use of Time-Correlated Thermal Neutron Counting to Determine the Multiplication of an Assembly of Highly Enriched Uranium

    SciTech Connect (OSTI)

    David L. Chichester; Mathew T. Kinlaw; Scott M. Watson; Jeffrey M. Kalter; Eric C. Miller; William A. Noonan

    2014-11-01

    A series of experiments and numerical simulations using thermal-neutron time-correlated measurements has been performed to determine the neutron multiplication, M, of assemblies of highly enriched uranium available at Idaho National Laboratory. The experiments used up to 14.4 kg of highly-enriched uranium, including bare assemblies and assemblies reflected with high-density polyethylene, carbon steel, and tungsten. A small 252Cf source was used to initiate fission chains within the assembly. Both the experiments and the simulations used 6-channel and 8-channel detector systems, each consisting of 3He proportional counters moderated with polyethylene; data was recorded in list mode for analysis. 'True' multiplication values for each assembly were empirically derived using basic neutron production and loss values determined through simulation. A total of one-hundred and sixteen separate measurements were performed using fifty-seven unique measurement scenarios, the multiplication varied from 1.75 to 10.90. This paper presents the results of these comparisons and discusses differences among the various cases.

  8. 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.

  9. Enterprise Assessments Targeted Review of the Safety System Management of the Secondary Confinement System and Power Distribution Safety System at the Y-12 National Security Complex Highly Enriched Uranium Materials Facility … December 2015

    Energy Savers [EERE]

    Targeted Review of the Safety System Management of the Secondary Confinement System and Safety Significant Power Distribution System at the Y-12 National Security Complex Highly Enriched Uranium Materials Facility December 2015 Office of Nuclear Safety and Environmental Assessments Office of Environment, Safety and Health Assessments Office of Enterprise Assessments U.S. Department of Energy i Table of Contents Acronyms

  10. Toxic Substances Control Act Uranium Enrichment Federal Facilities...

    Office of Environmental Management (EM)

    Thomas L. McCall, Jr. http:www.em.doe.govffaaortsca.html 4252001 Toxic Substances Control Act Uranium Enrichment Federal Facilities Compliance Agree.. Page 12 of 26 Deputy...

  11. Two U.S. University Research Reactors to be Converted From Highly Enriched

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

    Uranium to Low-Enriched Uranium | Department of Energy U.S. University Research Reactors to be Converted From Highly Enriched Uranium to Low-Enriched Uranium Two U.S. University Research Reactors to be Converted From Highly Enriched Uranium to Low-Enriched Uranium April 11, 2005 - 11:34am Addthis WASHINGTON, D.C. - As part of the Bush administration's aggressive effort to reduce the amount of weapons-grade nuclear material worldwide, Secretary of Energy Samuel W. Bodman announced today that

  12. 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.

  13. Perimeter safeguards techniques for uranium-enrichment plants

    SciTech Connect (OSTI)

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

    1981-09-01

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

  14. Uranium enrichment: heading for a cliff

    SciTech Connect (OSTI)

    Norman, C.

    1987-05-22

    Thanks to drastic cost cutting in the past 2 years, US enrichment plants now have the lowest cost production in the world, but US prices are still higher than those of overseas competitors because the business is paying for past mistakes. The most serious difficulty is that the Department of Energy (DOE), which owns and operates the US enrichment enterprise, is paying more than $500 million a year to the Tennessee Valley Authority (TVA) for electricity it once thought it would need but no longer requires. Another is that billions of dollars were spent in the 1970s and early 1980s to build new capacity that is now not needed. As a result, the enrichment enterprise has accumulated a debt to the US Treasury that the General Accounting Office (GAO) estimates at $8.8 billion. This paper presents the background and current debate in Congress about the difficulties facing the enrichment industry. In the midst of this debate over the future of the enterprise, the development of the next generation of enrichment technology is being placed in jeopardy. Known as atomic vapor laser isotope separation, or AVLIS, the process was viewed as the key to the long-term competitiveness of US enrichment. As the federal deficit mounted, however, funding for the AVLIS program was cut back and the timetable was stretched out. The US enrichment program has reached the point at which Congress will be forced to make some politically difficult decisions.

  15. Uranium enrichment: investment options for the long term

    SciTech Connect (OSTI)

    Not Available

    1983-01-01

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

  16. D&D of the French High Enrichment Gaseous Diffusion Plant

    SciTech Connect (OSTI)

    BEHAR, Christophe; GUIBERTEAU, Philippe; DUPERRET, Bernard; TAUZIN, Claude

    2003-02-27

    This paper describes the D&D program that is being implemented at France's High Enrichment Gaseous Diffusion Plant, which was designed to supply France's Military with Highly Enriched Uranium. This plant was definitively shut down in June 1996, following French President Jacques Chirac's decision to end production of Highly Enriched Uranium and dismantle the corresponding facilities.

  17. CRAD, Management- Y-12 Enriched Uranium Operations Oxide Conversion Facility

    Broader source: Energy.gov [DOE]

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a January 2005 assessment of Management program at the Y-12 - Enriched Uranium Operations Oxide Conversion Facility.

  18. CRAD, Training- Y-12 Enriched Uranium Operations Oxide Conversion Facility

    Broader source: Energy.gov [DOE]

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a January 2005 assessment of the Training Program at the Y-12 - Enriched Uranium Operations Oxide Conversion Facility.

  19. Highly Enriched Uranium Materials Facility

    National Nuclear Security Administration (NNSA)

    Chuck Fleischmann (R-Tenn.) as well as NNSA Production Office Deputy Manager Teresa Robbins, Consolidated Nuclear Security President and Chief Executive Officer Jim Haynes, and...

  20. 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.

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

    SciTech Connect (OSTI)

    Sewell, P.G.

    1991-11-01

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

  2. Uranium Enrichment Standards of the Y-12 Nuclear Detection and Sensor Testing Center

    SciTech Connect (OSTI)

    Cantrell, J.

    2012-05-23

    The Y-12 National Security Complex has recently fabricated and characterized a new series of metallic uranium standards for use in the Nuclear Detection and Sensor Testing Center (NDSTC). Ten uranium metal disks with enrichments varying from 0.2 to 93.2% {sup 235}U were designed to provide researchers access to a wide variety of measurement scenarios in a single testing venue. Special care was taken in the selection of the enrichments in order to closely bracket the definitions of reactor fuel at 4% {sup 235}U and that of highly enriched uranium (HEU) at 20% {sup 235}U. Each standard is well characterized using analytical chemistry as well as a series of gamma-ray spectrometry measurements. Gamma-ray spectra of these standards are being archived in a reference library for use by customers of the NDSTC. A software database tool has been created that allows for easier access and comparison of various spectra. Information provided through the database includes: raw count data (including background spectra), regions of interest (ROIs), and full width half maximum calculations. Input is being sought from the user community on future needs including enhancements to the spectral database and additional Uranium standards, shielding configurations and detector types. A related presentation are planned for the INMM 53rd Annual Meeting (Hull, et al.), which describe new uranium chemical compound standards and testing opportunities at Y-12 Nuclear Detection and Sensor Testing Center (NDSTC).

  3. 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...

  4. Natural uranium/conversion services/enrichment services

    SciTech Connect (OSTI)

    1993-12-31

    This article is the 1993 uranium market summary. During this reporting period, there were 50 deals in the concentrates market, 26 deals in the UF6 market, and 14 deals for enrichment services. In the concentrates market, the restricted value closed $0.15 higher at $9.85, and the unrestricted value closed down $0.65 at $7.00. In the UF6 market, restricted prices fluctuated and closed higher at $31.00, and unrestricted prices closed at their initial value of $24.75. The restricted transaction value closed at $10.25 and the unrestricted value closed at $7.15. In the enrichment services market, the restricted value moved steadily higher to close at $84.00 per SWU, and the unrestricted value closed at its initial value of $68.00 per SWU.

  5. Comments on proposed legislation to restructure DOE's uranium enrichment program

    SciTech Connect (OSTI)

    Not Available

    1991-04-01

    This book focuses on H.R.145, H.R.788, and S.210. Each of the proposed bills would restructure DOE's enrichment program as a government corporation with private financing and would encourage the eventual sale of the corporation to the private sector. In doing so, the bills would, among other things, allow the corporation to set prices to maximize long-term returns; establish a fund to meet the costs of decontamination, decommissioning, and other environmental cleanup costs associated with uranium enrichment activities; transfer interest in DOE's new atomic vapor laser isotope separation (AVLIS) process to the new corporation; and, except for H.R. 145, require the government to pay its share of the costs to clean up mill tailings (mining wastes) generated under government contracts.

  6. Laser and gas centrifuge enrichment

    SciTech Connect (OSTI)

    Heinonen, Olli

    2014-05-09

    Principles of uranium isotope enrichment using various laser and gas centrifuge techniques are briefly discussed. Examples on production of high enriched uranium are given. Concerns regarding the possibility of using low end technologies to produce weapons grade uranium are explained. Based on current assessments commercial enrichment services are able to cover the global needs of enriched uranium in the foreseeable future.

  7. 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.

  8. Uranium enrichment. Printed at the request of the Committee on Energy and Natural Resources, United States Senate, May 1982

    SciTech Connect (OSTI)

    Not Available

    1982-01-01

    Two congressional reports outline the need for new uranium-enrichment plants and their costs. Part I, The Need for Additional Uranium Enrichment Capacity to Meet Demand, examines DOE's case for continuing construction of the Portsmouth, Ohio gas centrifuge plant on the basis of projected demand. The report concludes that DOE projections are high and that future demand can be met through preproduction and stockpiling. Part II, Necessity for GCEP (Gas Centrifuge Enrichment Plant) Under Low Nuclear Power Growth Conditions, concludes that continued construction is economically valid because of the uncertainty of demand forecasts. 79 references, 12 tables. (DCK)

  9. Uranium mineralization in fluorine-enriched volcanic rocks

    SciTech Connect (OSTI)

    Burt, D.M.; Sheridan, M.F.; Bikun, J.; Christiansen, E.; Correa, B.; Murphy, B.; Self, S.

    1980-09-01

    Several uranium and other lithophile element deposits are located within or adjacent to small middle to late Cenozoic, fluorine-rich rhyolitic dome complexes. Examples studied include Spor Mountain, Utah (Be-U-F), the Honeycomb Hills, Utah (Be-U), the Wah Wah Mountains, Utah (U-F), and the Black Range-Sierra Cuchillo, New Mexico (Sn-Be-W-F). The formation of these and similar deposits begins with the emplacement of a rhyolitic magma, enriched in lithophile metals and complexing fluorine, that rises to a shallow crustal level, where its roof zone may become further enriched in volatiles and the ore elements. During initial explosive volcanic activity, aprons of lithicrich tuffs are erupted around the vents. These early pyroclastic deposits commonly host the mineralization, due to their initial enrichment in the lithophile elements, their permeability, and the reactivity of their foreign lithic inclusions (particularly carbonate rocks). The pyroclastics are capped and preserved by thick topaz rhyolite domes and flows that can serve as a source of heat and of additional quantities of ore elements. Devitrification, vapor-phase crystallization, or fumarolic alteration may free the ore elements from the glassy matrix and place them in a form readily leached by percolating meteoric waters. Heat from the rhyolitic sheets drives such waters through the system, generally into and up the vents and out through the early tuffs. Secondary alteration zones (K-feldspar, sericite, silica, clays, fluorite, carbonate, and zeolites) and economic mineral concentrations may form in response to this low temperature (less than 200 C) circulation. After cooling, meteoric water continues to migrate through the system, modifying the distribution and concentration of the ore elements (especially uranium).

  10. New Prototype Safeguards Technology Offers Improved Confidence and Automation for Uranium Enrichment Facilities

    SciTech Connect (OSTI)

    Brim, Cornelia P.

    2013-04-01

    An important requirement for the international safeguards community is the ability to determine the enrichment level of uranium in gas centrifuge enrichment plants and nuclear fuel fabrication facilities. This is essential to ensure that countries with nuclear nonproliferation commitments, such as States Party to the Nuclear Nonproliferation Treaty, are adhering to their obligations. However, current technologies to verify the uranium enrichment level in gas centrifuge enrichment plants or nuclear fuel fabrication facilities are technically challenging and resource-intensive. NNSA’s Office of Nonproliferation and International Security (NIS) supports the development, testing, and evaluation of future systems that will strengthen and sustain U.S. safeguards and security capabilities—in this case, by automating the monitoring of uranium enrichment in the entire inventory of a fuel fabrication facility. One such system is HEVA—hybrid enrichment verification array. This prototype was developed to provide an automated, nondestructive assay verification technology for uranium hexafluoride (UF6) cylinders at enrichment plants.

  11. Environmental Survey preliminary report, Portsmouth Uranium Enrichment Complex, Piketon, Ohio

    SciTech Connect (OSTI)

    Not Available

    1987-08-01

    This report presents the preliminary findings from the first phase of the Environmental Survey of the United States Department of Energy (DOE) Portsmouth Uranium Enrichment Complex (PUEC), conducted August 4 through August 15, 1986. The Survey is being conducted by an interdisciplinary team of environmental specialists, led and managed by the Office of Environment, Safety and Health's Office of Environmental Audit. Team specialists are being supplied by a private contractor. The objective of the Survey is to identify environmental problems and areas of environmental regulation. It is being performed in accordance with the DOE Environmental Survey Manual. This phase of the Survey involves the review of existing site environmental data, observations of the operations performed at PUEC, and interviews with site personnel. The Survey team developed a Sampling and Analysis Plan to assist in further assessing certain of the environmental problems identified during its on-site activities. The Sampling and Analysis Plan will be executed by Argonne National Laboratory. When completed, the results will be incorporated into the PUEC Environmental Survey Interim Report. The Interim Report will reflect the final determinations of the PUEC Survey. 55 refs., 22 figs., 21 tabs.

  12. 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.

  13. EA-1255: Project Partnership Transportation of Foreign-Owned Enriched Uranium from the Republic of Georgia

    Broader source: Energy.gov [DOE]

    This EA evaluates the environmental impacts for the proposal to transport 5.26 kilograms of enriched uranium-23 5 in the form of nuclear fuel, from the Republic of Georgia to the United Kingdom.

  14. Uranium Enrichment Decontamination and Decommissioning Fund's Fiscal Year 2011 Financial Statement Audit

    Energy Savers [EERE]

    Uranium Enrichment Decontamination and Decommissioning Fund's Fiscal Year 2011 Financial Statement Audit OAS-FS-13-02 October 2012 September 7, 2012 Mr. Gregory Friedman Inspector General U.S. Department of Energy 1000 Independence Avenue, S.W. Room 5D-039 Washington, DC 20585 Dear Mr. Friedman: We have audited the financial statements of the Department of Energy's (the Department) Uranium Enrichment Decontamination and Decommissioning Fund (D&D Fund) as of and for the year ended September

  15. Toxic Substances Control Act Uranium Enrichment Federal Facilities Compliance Agreement, February 20, 1992 Summary

    Office of Environmental Management (EM)

    Toxic Substance Control Act Uranium Enrichment Federal Facilities Compliance Agreement (TSCA-UE- FFCA), February 20, 1992 State Kentucky Agreement Type Compliance Agreement Legal Driver(s) TSCA Scope Summary Establishes responsibilities and commitments for bringing DOE's former and active Uranium Enrichment Plants in Paducah, Portsmouth, and Oak Ridge into compliance with TSCA and PCB Regulations Parties DOE; U.S. EPA Date 2/20/1992 SCOPE * Establish a plan and the responsibilities and

  16. 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.

  17. highly enriched uranium | National Nuclear Security Administration

    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...

  18. Highly Enriched Uranium Transparency Program | National Nuclear...

    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...

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

    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...

  20. Highly Enriched Uranium Disposition | National Nuclear Security...

    National Nuclear Security Administration (NNSA)

    the economic value of the material by using the resulting LEU as nuclear reactor fuel. ... HEU from Russian nuclear weapons into LEU used as fuel in U.S. commercial power reactors. ...

  1. The Office of Environmental Management Uranium Enrichment D&D | Department

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

    of Energy Uranium Enrichment D&D The Office of Environmental Management Uranium Enrichment D&D PDF icon Microsoft Word - B996F741.doc More Documents & Publications Microsoft Word - PSRP Updates 6-25-10_v2 Above on the left is K-25, at Oak Ridge before and after the 844,000 sq-ft demolition. In addition, on the right: K Cooling Tower at Savannah River Site demolition. Deactivation & Decommissioning (D&D) Program Map 2013 Annual Workforce Analysis and Staffing Plan Report -

  2. Higher Resolution Neutron Velocity Spectrometer Measurements of Enriched Uranium

    DOE R&D Accomplishments [OSTI]

    Rainwater, L. J.; Havens, W. W. Jr.

    1950-08-09

    The slow neutron transmission of a sample of enriched U containing 3.193 gm/cm2 was investigated with a resolution width of 1 microsec/m. Results of transmission measurements are shown graphically. (B.J.H.)

  3. Signatures and Methods for the Automated Nondestructive Assay of UF6 Cylinders at Uranium Enrichment Plants

    SciTech Connect (OSTI)

    Smith, Leon E.; Mace, Emily K.; Misner, Alex C.; Shaver, Mark W.

    2010-08-08

    International Atomic Energy Agency (IAEA) inspectors currently perform periodic inspections at uranium enrichment plants to verify UF6 cylinder enrichment declarations. Measurements are typically performed with handheld high-resolution sensors on a sampling of cylinders taken to be representative of the facility’s entire cylinder inventory. These measurements are time-consuming, expensive, and assay only a small fraction of the total cylinder volume. An automated nondestructive assay system capable of providing enrichment measurements over the full volume of the cylinder could improve upon current verification practices in terms of manpower and assay accuracy. Such a station would use sensors that can be operated in an unattended mode at an industrial facility: medium-resolution scintillators for gamma-ray spectroscopy (e.g., NaI(Tl)) and moderated He-3 neutron detectors. This sensor combination allows the exploitation of additional, more-penetrating signatures beyond the traditional 185-keV emission from U-235: neutrons produced from F-19(α,n) reactions (spawned primarily from U 234 alpha emission) and high-energy gamma rays (extending up to 8 MeV) induced by neutrons interacting in the steel cylinder. This paper describes a study of these non-traditional signatures for the purposes of cylinder enrichment verification. The signatures and the radiation sensors designed to collect them are described, as are proof-of-principle cylinder measurements and analyses. Key sources of systematic uncertainty in the non-traditional signatures are discussed, and the potential benefits of utilizing these non-traditional signatures, in concert with an automated form of the traditional 185-keV-based assay, are discussed.

  4. 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.

  5. CRAD, Conduct of Operations- Y-12 Enriched Uranium Operations Oxide Conversion Facility

    Broader source: Energy.gov [DOE]

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a January, 2005 assessment of Conduct of Operations program at the Y-12 - Enriched Uranium Operations Oxide Conversion Facility.

  6. CRAD, Criticality Safety- Y-12 Enriched Uranium Operations Oxide Conversion Facility

    Broader source: Energy.gov [DOE]

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a January 2005 assessment of the Criticality Safety program at the Y-12 - Enriched Uranium Facility.

  7. CRAD, Emergency Management- Y-12 Enriched Uranium Operations Oxide Conversion Facility

    Broader source: Energy.gov [DOE]

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a January 2005 assessment of Emergency Management program at the Y-12 Enriched Uranium Operations Oxide Conversion Facility.

  8. CRAD, DOE Oversight- Y-12 Enriched Uranium Operations Oxide Conversion Facility

    Broader source: Energy.gov [DOE]

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a DOE independent oversight assessment of the Y-12 Site Office's programs for oversight of its contractors at the Y-12 Enriched Uranium Operations Oxide Conversion Facility.

  9. CRAD, Occupational Safety & Health- Y-12 Enriched Uranium Operations Oxide Conversion Facility

    Broader source: Energy.gov [DOE]

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a January 2005 assessment of Industrial Safety and Industrial Health programs at the Y-12 - Enriched Uranium Operations Oxide Conversion Facility.

  10. CRAD, Radiological Controls- Y-12 Enriched Uranium Operations Oxide Conversion Facility

    Broader source: Energy.gov [DOE]

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a January 2005 assessment of the Radiation Protection Program at the Y-12 - Enriched Uranium Operations Oxide Conversion Facility.

  11. CRAD, Environmental Protection- Y-12 Enriched Uranium Operations Oxide Conversion Facility

    Broader source: Energy.gov [DOE]

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a January 2005 assessment of Environmental Compliance program at the Y-12 - Enriched Uranium Operations Oxide Conversion Facility.

  12. CRAD, Safety Basis- Y-12 Enriched Uranium Operations Oxide Conversion Facility

    Broader source: Energy.gov [DOE]

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a January 2005 assessment of the Safety Basis at the Y-12 - Enriched Uranium Operations Oxide Conversion Facility.

  13. Draft EA for the Acceptance and Disposition of Spent Nuclear Fuel Containing U.S.-Origin Highly Enriched

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

    EA for the Acceptance and Disposition of Spent Nuclear Fuel Containing U.S.-Origin Highly Enriched Uranium from the Federal Republic of Germany DOE/EA-1977 DRAFT ENVIRONMENTAL ASSESSMENT FOR THE ACCEPTANCE AND DISPOSITION OF SPENT NUCLEAR FUEL CONTAINING U.S.-ORIGIN HIGHLY ENRICHED URANIUM FROM THE FEDERAL REPUBLIC OF GERMANY January 2016 U.S. DEPARTMENT OF ENERGY SAVANNAH RIVER OPERATIONS OFFICE AIKEN, SOUTH CAROLINA Draft EA for the Acceptance and Disposition of Spent Nuclear Fuel Containing

  14. 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.

  15. 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.

  16. Operating limit evaluation for disposal of uranium enrichment plant wastes

    SciTech Connect (OSTI)

    Lee, D.W.; Kocher, D.C.; Wang, J.C.

    1996-02-01

    A proposed solid waste landfill at Paducah Gaseous Diffusion Plant (PGDP) will accept wastes generated during normal plant operations that are considered to be non-radioactive. However, nearly all solid waste from any source or facility contains small amounts of radioactive material, due to the presence in most materials of trace quantities of such naturally occurring radionuclides as uranium and thorium. This paper describes an evaluation of operating limits, which are protective of public health and the environment, that would allow waste materials containing small amounts of radioactive material to be sent to a new solid waste landfill at PGDP. The operating limits are expressed as limits on concentrations of radionuclides in waste materials that could be sent to the landfill based on a site-specific analysis of the performance of the facility. These limits are advantageous to PGDP and DOE for several reasons. Most importantly, substantial cost savings in the management of waste is achieved. In addition, certain liabilities that could result from shipment of wastes to a commercial off-site solid waste landfill are avoided. Finally, assurance that disposal operations at the PGDP landfill are protective of public health and the environment is provided by establishing verifiable operating limits for small amounts of radioactive material; rather than relying solely on administrative controls. The operating limit determined in this study has been presented to the Commonwealth of Kentucky and accepted as a condition to be attached to the operating permit for the solid waste landfill.

  17. 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,...

  18. MCNP5 CRITICALITY VALIDATION AND BIAS FOR INTERMEDIATE ENRICHED URANIUM SYSTEMS

    SciTech Connect (OSTI)

    FINFROCK SH

    2009-12-10

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

  19. 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.

  20. 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.

  1. Experimental critical parameters of enriched uranium solution in annular tank geometries

    SciTech Connect (OSTI)

    Rothe, R.E.

    1996-04-01

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

  2. 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.

  3. Critical experiments on an enriched uranium solution system containing periodically distributed strong thermal neutron absorbers

    SciTech Connect (OSTI)

    Rothe, R.E.

    1996-09-30

    A series of 62 critical and critical approach experiments were performed to evaluate a possible novel means of storing large volumes of fissile solution in a critically safe configuration. This study is intended to increase safety and economy through use of such a system in commercial plants which handle fissionable materials in liquid form. The fissile solution`s concentration may equal or slightly exceed the minimum-critical-volume concentration; and experiments were performed for high-enriched uranium solution. Results should be generally applicable in a wide variety of plant situations. The method is called the `Poisoned Tube Tank` because strong neutron absorbers (neutron poisons) are placed inside periodically spaced stainless steel tubes which separate absorber material from solution, keeping the former free of contamination. Eight absorbers are investigated. Both square and triangular pitched lattice patterns are studied. Ancillary topics which closely model typical plant situations are also reported. They include the effect of removing small bundles of absorbers as might occur during inspections in a production plant. Not taking the tank out of service for these inspections would be an economic advantage. Another ancillary topic studies the effect of the presence of a significant volume of unpoisoned solution close to the Poisoned Tube Tank on the critical height. A summary of the experimental findings is that boron compounds were excellent absorbers, as expected. This was true for granular materials such as Gerstley Borate and Borax; but it was also true for the flexible solid composed of boron carbide and rubber, even though only thin sheets were used. Experiments with small bundles of absorbers intentionally removed reveal that quite reasonable tanks could be constructed that would allow a few tubes at a time to be removed from the tank for inspection without removing the tank from production service.

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

    DOE Patents [OSTI]

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

    1995-01-01

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

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

    DOE Patents [OSTI]

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

    1995-05-30

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

  6. 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.

  7. Office of Environmental Management Uranium Enrichment Decontamination and Decommissioning Fund financial statements, September 30, 1995 and 1994

    SciTech Connect (OSTI)

    1996-02-21

    The Energy Policy Act of 1992 (Act) requires the Department of Energy to retain ownership and responsibility for the costs of environmental cleanup resulting from the Government`s operation of the three gaseous diffusion facilities located at the K-25 site in Oak Ridge, Tennessee; Paducah, Kentucky; and Portsmouth, Ohio. The Act transferred the uranium enrichment enterprise to the United States Enrichment Corporation (USEC) as of July 1, 1993, and established the Uranium Enrichment Decontamination and Decommissioning Fund (D&D Fund) to: Pay for the costs of decontamination and decommissioning at the diffusion facilities; pay the annual costs for remedial action at the diffusion facilities to the extent that the amount in the Fund is sufficient; and reimburse uranium/thorium licensees for the costs of decontamination, decommissioning, reclamation, and other remedial actions which are incident to sales to the Government.

  8. Office of Environmental Management uranium enrichment decontamination and decommissioning fund financial statements. September 30, 1994 and 1993

    SciTech Connect (OSTI)

    Marwick, P.

    1994-12-15

    The Energy Policy Act of 1992 (Act) transferred the uranium enrichment enterprise to the United States Enrichment Corporation as of July 1, 1993. However, the Act requires the Department of Energy to retain ownership and responsibility for the costs of environmental cleanup resulting from the Government`s operation of the three gaseous diffusion facilities located in Oak Ridge, Tennessee; Paducah, Kentucky; and Portsmouth, Ohio (diffusion facilities). The Act established the Uranium Enrichment Decontamination and Decommissioning Fund (D&D Fund) to: Pay for the costs of decontamination and decommissioning at the diffusion facilities; Pay the annual costs for remedial action at the diffusion facilities to the extent that the amount in the Fund is sufficient; and Reimburse uranium/thorium licensees for the costs of decontamination, decommissioning, reclamation, and other remedial actions which are incident to sales to the Government.

  9. RADIO FREQUENCY IDENTIFICATION DEVICES: EFFECTIVENESS IN IMPROVING SAFEGUARDS AT GAS-CENTRIFUGE URANIUM-ENRICHMENT PLANTS.

    SciTech Connect (OSTI)

    JOE,J.

    2007-07-08

    Recent advances in radio frequency identification devices (RFIDs) have engendered a growing interest among international safeguards experts. Potentially, RFIDs could reduce inspection work, viz. the number of inspections, number of samples, and duration of the visits, and thus improve the efficiency and effectiveness of international safeguards. This study systematically examined the applications of RFIDs for IAEA safeguards at large gas-centrifuge enrichment plants (GCEPs). These analyses are expected to help identify the requirements and desirable properties for RFIDs, to provide insights into which vulnerabilities matter most, and help formulate the required assurance tests. This work, specifically assesses the application of RFIDs for the ''Option 4'' safeguards approach, proposed by Bruce Moran, U. S. Nuclear Regulatory Commission (NRC), for large gas-centrifuge uranium-enrichment plants. The features of ''Option 4'' safeguards include placing RFIDs on all feed, product and tails (F/P/T) cylinders, along with WID readers in all FP/T stations and accountability scales. Other features of Moran's ''Option 4'' are Mailbox declarations, monitoring of load-cell-based weighing systems at the F/P/T stations and accountability scales, and continuous enrichment monitors. Relevant diversion paths were explored to evaluate how RFIDs improve the efficiency and effectiveness of safeguards. Additionally, the analysis addresses the use of RFIDs in conjunction with video monitoring and neutron detectors in a perimeter-monitoring approach to show that RFIDs can help to detect unidentified cylinders.

  10. 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.

  11. 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

  12. Natural phenomena hazards evaluation of equipment and piping of Gaseous Diffusion Plant Uranium Enrichment Facility

    SciTech Connect (OSTI)

    Singhal, M.K.; Kincaid, J.H.; Hammond, C.R.; Stockdale, B.I.; Walls, J.C.; Brock, W.R.; Denton, D.R.

    1995-12-31

    In support of the Gaseous Diffusion Plant Safety Analysis Report Upgrade program (GDP SARUP), a natural phenomena hazards evaluation was performed for the main process equipment and piping in the uranium enrichment buildings at Paducah and Portsmouth gaseous diffusion plants. In order to reduce the cost of rigorous analyses, the evaluation methodology utilized a graded approach based on an experience data base collected by SQUG/EPRI that contains information on the performance of industrial equipment and piping during past earthquakes. This method consisted of a screening walkthrough of the facility in combination with the use of engineering judgment and simple calculations. By using these screenings combined with evaluations that contain decreasing conservatism, reductions in the time and cost of the analyses were significant. A team of experienced seismic engineers who were trained in the use of the DOE SQUG/EPRI Walkdown Screening Material was essential to the success of this natural phenomena hazards evaluation.

  13. 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.

  14. Enrichment of specific protozoan populations during in situ bioremediation of uranium-contaminated groundwater

    SciTech Connect (OSTI)

    Holmes, Dawn; Giloteaux, L.; Williams, Kenneth H.; Wrighton, Kelly C.; Wilkins, Michael J.; Thompson, Courtney A.; Roper, Thomas J.; Long, Philip E.; Lovley, Derek

    2013-07-28

    The importance of bacteria in the anaerobic bioremediation of groundwater polluted with organic and/or metal contaminants is well-recognized and in some instances so well understood that modeling of the in situ metabolic activity of the relevant subsurface microorganisms in response to changes in subsurface geochemistry is feasible. However, a potentially significant factor influencing bacterial growth and activity in the subsurface that has not been adequately addressed is protozoan predation of the microorganisms responsible for bioremediation. In field experiments at a uranium-contaminated aquifer located in Rifle, CO, acetate amendments initially promoted the growth of metal-reducing Geobacter species followed by the growth of sulfate-reducers, as previously observed. Analysis of 18S rRNA gene sequences revealed a broad diversity of sequences closely related to known bacteriovorous protozoa in the groundwater prior to the addition of acetate. The bloom of Geobacter species was accompanied by a specific enrichment of sequences most closely related to the amoeboid flagellate, Breviata anathema, which at their peak accounted for over 80% of the sequences recovered. The abundance of Geobacter species declined following the rapid emergence of B. anathema. The subsequent growth of sulfate-reducing Peptococcaceae was accompanied by another specific enrichment of protozoa, but with sequences most similar to diplomonadid flagellates from the family Hexamitidae, which accounted for up to 100% of the sequences recovered during this phase of the bioremediation. These results suggest a prey-predator response with specific protozoa responding to increased availability of preferred prey bacteria. Thus, quantifying the influence of protozoan predation on the growth, activity, and composition of the subsurface bacterial community is essential for predictive modeling of in situ uranium bioremediation strategies.

  15. 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.

  16. LABORATORY DEMONSTRATION OF A MULTISENSOR UNATTENDED CYLINDER VERIFICATION STATION FOR URANIUM ENRICHMENT PLANT SAFEGUARDS

    SciTech Connect (OSTI)

    Goodman, David I; Rowland, Kelly L; Smith, Sheriden; Miller, Karen A.; Flynn, Eric B.

    2014-01-10

    The objective of safeguards is the timely detection of the diversion of a significant quantity of nuclear materials, and safeguarding uranium enrichment plants is especially important in preventing the spread of nuclear weapons. The IAEA’s proposed Unattended Cylinder Verification Station (UCVS) for UF6 cylinder verification would combine the operator’s accountancy scale with a nondestructive assay system such as the Passive Neutron Enrichment Meter (PNEM) and cylinder identification and surveillance systems. In this project, we built a laboratory-scale UCVS and demonstrated its capabilities using mock UF6 cylinders. We developed a signal processing algorithm to automate the data collection and processing from four continuous, unattended sensors. The laboratory demonstration of the system showed that the software could successfully identify cylinders, snip sensor data at the appropriate points in time, determine the relevant characteristics of the cylinder contents, check for consistency among sensors, and output the cylinder data to a file. This paper describes the equipment, algorithm and software development, laboratory demonstration, and recommendations for a full-scale UCVS.

  17. 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.

  18. 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)

  19. 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.

  20. Safeguards by design - industry engagement for new uranium enrichment facilities in the United States

    SciTech Connect (OSTI)

    Demuth, Scott F; Grice, Thomas; Lockwood, Dunbar

    2010-01-01

    The United States Department of Energy's (DOE's) Office of Nonproliferation and International Security (NA-24) has initiated a Safeguards by Design (SBD) effort to encourage the incorporation of international (IAEA) safeguards features early in the design phase of a new nuclear facility in order to avoid the need to redesign or retrofit the facility at a later date. The main goals of Safeguards by Design are to (1) make the implementation of international safeguards at new civil nuclear facilities more effective and efficient, (2) avoid costly and time-consuming re-design work or retrofits at such facilities and (3) design such facilities in a way that makes proliferation as technically difficult, as time-consuming, and as detectable as possible. The U.S. Nuclear Regulatory Commission (NRC) has recently hosted efforts to facilitate the use of Safeguards by Design for new uranium enrichment facilities currently being planned for construction in the U.S. While SBD is not a NRC requirement, the NRC is aiding the implementation of SBD by coordinating discussions between DOE's NA-24 and industry's facility design teams. More specifically, during their normal course of licensing discussions the NRC has offered industry the opportunity to engage with NA-24 regarding SBD.

  1. Safeguards Guidance for Designers of Commercial Nuclear Facilities – International Safeguards Requirements for Uranium Enrichment Plants

    SciTech Connect (OSTI)

    Philip Casey Durst; Scott DeMuth; Brent McGinnis; Michael Whitaker; James Morgan

    2010-04-01

    For the past two years, the United States National Nuclear Security Administration, Office of International Regimes and Agreements (NA-243), has sponsored the Safeguards-by-Design Project, through which it is hoped new nuclear facilities will be designed and constructed worldwide more amenable to nuclear safeguards. In the course of this project it was recognized that commercial designer/builders of nuclear facilities are not always aware of, or understand, the relevant domestic and international safeguards requirements, especially the latter as implemented by the International Atomic Energy Agency (IAEA). To help commercial designer/builders better understand these requirements, a report was prepared by the Safeguards-by-Design Project Team that articulated and interpreted the international nuclear safeguards requirements for the initial case of uranium enrichment plants. The following paper summarizes the subject report, the specific requirements, where they originate, and the implications for design and construction. It also briefly summarizes the established best design and operating practices that designer/builder/operators have implemented for currently meeting these requirements. In preparing the subject report, it is recognized that the best practices are continually evolving as the designer/builder/operators and IAEA consider even more effective and efficient means for meeting the safeguards requirements and objectives.

  2. High-temperature gas-cooled reactors: preliminary safety and environmental information document. Volume IV

    SciTech Connect (OSTI)

    Not Available

    1980-01-01

    Information is presented concerning medium-enriched uranium/thorium once-through fuel cycle; medium-enrichment uranium-233/thorium recycle fuel; high-enrichment uranium-235/thorium recycle (spiked) fuel cycle; high-enrichment uranium-233/thorium recycle (spiked) fuel cycle; and gas-turbine high-temperature gas-cooled reactor.

  3. 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.

  4. 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.

  5. 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.

  6. 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

  7. Belgium Highly Enriched Uranium and Plutonium Removals | National...

    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...

  8. NNSA Authorizes Start-Up of Highly Enriched Uranium Materials...

    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...

  9. 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...

  10. Italy Highly Enriched Uranium and Plutonium Removals | National...

    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...

  11. NNSA Highly Enriched Uranium Removal Featured on The Rachel Maddow...

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

    NNSA. Michael Hess Michael Hess Former Digital Communications Specialist, Office of ... the news with Rachel Maddow. The full video can be found on the original NNSA blog post. ...

  12. US, Kazakhstan Cooperate to Eliminate Highly Enriched Uranium...

    National Nuclear Security Administration (NNSA)

    ... application of nuclear science. NNSA maintains and enhances the safety, security, reliability and performance of the U.S. nuclear weapons stockpile without nuclear testing; works ...

  13. 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.

  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. H. R. 788: This Act may be cited as the Uranium Enrichment Reorganization Act, introduced in the House of Representatives, One Hundred Second Congress, First Session, February 4, 1991

    SciTech Connect (OSTI)

    Not Available

    1991-01-01

    This bill would maintain a competitive, financially strong, and secure uranium enrichment capability in the US by reorganizing the uranium enrichment enterprise. The bill amends the Atomic Energy Act of 1954 by establishing the United States Uranium Enrichment Corporation. This bill describes general provisions; the establishment of the corporation; powers and duties of the corporation; organization, finance, and management; licensing, taxation, and miscellaneous provisions; decontamination, decommissioning, and remedial action.

  16. 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)

  17. Transportation of foreign-owned enriched uranium from the Republic of Georgia. Environmental assessment for Project Partnership

    SciTech Connect (OSTI)

    1998-03-31

    The Department of Energy (DOE) Office of Nonproliferation and National Security (NN) has prepared a classified environmental assessment to evaluate the potential environmental impact for the transportation of 5.26 kilograms of enriched uranium-235 in the form of nuclear fuel, from the Republic of Georgia to the United Kingdom. The nuclear fuel consists of primarily fresh fuel, but also consists of a small quantity (less than 1 kilogram) of partially-spent fuel. Transportation of the enriched uranium fuel would occur via US Air Force military aircraft under the control of the Defense Department European Command (EUCOM). Actions taken in a sovereign nation (such as the Republic of Georgia and the United Kingdom) are not subject to analysis in the environmental assessment. However, because the action would involve the global commons of the Black Sea and the North Sea, the potential impact to the global commons has been analyzed. Because of the similarities in the two actions, the Project Sapphire Environmental Assessment was used as a basis for assessing the potential impacts of Project Partnership. However, because Project Partnership involves a small quantity of partially-spent fuel, additional analysis was conducted to assess the potential environmental impacts and to consider reasonable alternatives as required by NEPA. The Project Partnership Environmental Assessment found the potential environmental impacts to be well below those from Project Sapphire.

  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. Evaluation of a RF-Based Approach for Tracking UF6 Cylinders at a Uranium Enrichment Plant

    SciTech Connect (OSTI)

    Pickett, Chris A; Younkin, James R; Kovacic, Donald N; Laughter, Mark D; Hines, Jairus B; Boyer, Brian; Martinez, B.

    2008-01-01

    Approved industry-standard cylinders are used globally to handle and store uranium hexafluoride (UF{sub 6}) feed, product, tails, and samples at uranium enrichment plants. The International Atomic Energy Agency (IAEA) relies on time-consuming physical inspections to verify operator declarations and detect possible diversion of UF{sub 6}. Development of a reliable, automated, and tamper-resistant system for near real-time tracking and monitoring UF{sub 6} cylinders (as they move within an enrichment facility) would greatly improve the inspector function. This type of system can reduce the risk of false or misreported cylinder tare weights, diversion of nuclear material, concealment of excess production, utilization of undeclared cylinders, and misrepresentation of the cylinders contents. This paper will describe a proof-of-concept approach that was designed to evaluate the feasibility of using radio frequency (RF)-based technologies to track individual UF{sub 6} cylinders throughout a portion of their life cycle, and thus demonstrate the potential for improved domestic accountability of materials, and a more effective and efficient method for application of site-level IAEA safeguards. The evaluation system incorporates RF-based identification devices (RFID) which provide a foundation for establishing a reliable, automated, and near real-time tracking system that can be set up to utilize site-specific, rules-based detection algorithms. This paper will report results from a proof-of-concept demonstration at a real enrichment facility that is specifically designed to evaluate both the feasibility of using RF to track cylinders and the durability of the RF equipment to survive the rigors of operational processing and handling. The paper also discusses methods for securely attaching RF devices and describes how the technology can effectively be layered with other safeguard systems and approaches to build a robust system for detecting cylinder diversion. Additionally, concepts for off-site tracking of cylinders are described.

  20. DOE/NNSA Successfully Establishes Uranium Lease and Takeback...

    National Nuclear Security Administration (NNSA)

    the DOENNSA's ongoing support for the establishment of a domestic, commercial Mo-99 production capability that does not use proliferation-sensitive highly enriched uranium (HEU). ...

  1. 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.

  2. 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.

  3. 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

  4. Y-12's uranium canning machine to save money | National Nuclear Security

    National Nuclear Security Administration (NNSA)

    Administration 's uranium canning machine to save money Monday, October 5, 2015 - 10:47am NNSA Blog During a recent visit to the Y-12 National Security Complex, NNSA Uranium Program Manager Tim Driscoll stopped by Building 9204-2E's new direct canning machine, which allows operators to package enriched uranium material removed from dismantled weapons and ship it directly to the Highly Enriched Uranium Materials Facility for long-term storage. An important element of the Uranium

  5. Economic and Non-proliferation Policy Considerations of Uranium Enrichment in Brazil and Argentina

    SciTech Connect (OSTI)

    Short, Steven M.; Phillips, Jon R.; Weimar, Mark R.; Mahy, Heidi A.

    2008-09-01

    The nuclear development programs of both Argentina and Brazil have, since the 1970s, been premised on the desire for self-sufficiency and assurance of nuclear fuel supply. While military rivalry and mutual distrust led to nuclear weapons related development programs in the 1970s and 1980s, both countries have since terminated these programs. Furthermore, the governments of both countries have pledged their commitment to exclusively non-explosive use of nuclear energy and have signed the Non Proliferation Treaty (NPT). Utilizing rights provided for under the NPT, both Argentina and Brazil have nuclear fuel production facilities, with the notable exception of enrichment plants, that provide much of the current indigenous fuel requirements for their nuclear power plants. However, both countries are actively developing enrichment capability to fill this gap. The purpose of this report is to assess the economic basis and non-proliferation policy considerations for indigenous enrichment capability within the context of their desired self-sufficiency and to evaluate possible United States Government policy options.

  6. Assessment of enriched uranium storage safety issues at the Oak Ridge Y-12 Plant

    SciTech Connect (OSTI)

    1996-08-01

    This document is an assessment of the technical safety issues pertaining to the storage of EU at the Oak Ridge Y-12 Plant. The purpose of the assessment is to serve as the basis for defining the technical standards for storage of EU at Y-12. A formal assessment of the Y-12 materials acceptance criteria for EU is currently being conducted by a task force cochaired by B. G. Eddy of DOE Oak Ridge Operations and S. 0. Cox of Y-12 Defense Programs. The mission of this technical assessment for storage is obviously dependent on results of the acceptance assessment. Clearly, the two efforts require coordination to avoid inconsistencies. In addition, both these Assessments must be consistent with the Environmental Assessment for EU storage at Y-12.1 Both the Storage Assessment and the Criteria for Acceptance must take cognizance of the fact that a portion of the EU to be submitted for storage in the future is expected to be derived from foreign sources and to include previously irradiated uranium containing significant levels of transuranics, radioactive daughter products, and unstable uranium isotopes that do not occur in the EU stream of the DOE weapons complex. National security considerations may dictate that these materials be accepted despite the fact that they fail to conform to the Acceptance Criteria. This document will attempt to address the complexities inherent in this situation.

  7. 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...

  8. Safety aspects of gas centrifuge enrichment plants

    SciTech Connect (OSTI)

    Hansen, A.H.

    1987-01-01

    Uranium enrichment by gas centrifuge is a commercially proven, viable technology. Gas centrifuge enrichment plant operations pose hazards that are also found in other industries as well as unique hazards as a result of processing and handling uranium hexafluoride and the handling of enriched uranium. Hazards also found in other industries included those posed by the use of high-speed rotating equipment and equipment handling by use of heavy-duty cranes. Hazards from high-speed rotating equipment are associated with the operation of the gas centrifuges themselves and with the operation of the uranium hexafluoride compressors in the tail withdrawal system. These and related hazards are discussed. It is included that commercial gas centrifuge enrichment plants have been designed to operate safely.

  9. Direct fissile assay of enriched uranium using random self-interrogation and neutron coincidence response

    DOE Patents [OSTI]

    Menlove, Howard O.; Stewart, James E.

    1986-01-01

    Apparatus and method for the direct, nondestructive evaluation of the .sup.235 U nuclide content of samples containing UF.sub.6, UF.sub.4, or UO.sub.2 utilizing the passive neutron self-interrogation of the sample resulting from the intrinsic production of neutrons therein. The ratio of the emitted neutron coincidence count rate to the total emitted neutron count rate is determined and yields a measure of the bulk fissile mass. The accuracy of the method is 6.8% (1.sigma.) for cylinders containing UF.sub.6 with enrichments ranging from 6% to 98% with measurement times varying from 3-6 min. The samples contained from below 1 kg to greater than 16 kg. Since the subject invention relies on fast neutron self-interrogation, complete sampling of the UF.sub.6 takes place, reducing difficulties arising from inhomogeneity of the sample which adversely affects other assay procedures.

  10. Direct fissile assay of enriched uranium using random self-interrogation and neutron coincidence response

    DOE Patents [OSTI]

    Menlove, H.O.; Stewart, J.E.

    1985-02-04

    Apparatus and method for the direct, nondestructive evaluation of the /sup 235/U nuclide content of samples containing UF/sub 6/, UF/sub 4/, or UO/sub 2/ utilizing the passive neutron self-interrogation of the sample resulting from the intrinsic production of neutrons therein. The ratio of the emitted neutron coincidence count rate to the total emitted neutron count rate is determined and yields a measure of the bulk fissile mass. The accuracy of the method is 6.8% (1sigma) for cylinders containing UF/sub 6/ with enrichments ranging from 6% to 98% with measurement times varying from 3-6 min. The samples contained from below 1 kg to greater than 16 kg. Since the subject invention relies on fast neutron self-interrogation, complete sampling of the UF/sub 6/ takes place, reducing difficulties arising from inhomogeneity of the sample which adversely affects other assay procedures. 4 figs., 1 tab.

  11. 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.

  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. Advanced uranium enrichment technologies

    SciTech Connect (OSTI)

    Merriman, R.

    1983-03-10

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

  14. United States Department of Energy, Office of Environmental Management, Uranium Enrichment Decontamination and Decomissioning Fund financial statements, September 30, 1996 and 1995

    SciTech Connect (OSTI)

    1997-05-01

    The Energy Policy Act of 1992 (Act) established the Uranium Enrichment Decontamination and Decommissioning Fund (D and D Fund, or Fund) to pay the costs for decontamination and decommissioning three gaseous diffusion facilities located in Oak Ridge, Tennessee; Paducah, Kentucky; and Portsmouth, Ohio (diffusion facilities). The Act also authorized the Fund to pay remedial action costs associated with the Government`s operation of the facilities and to reimburse uranium and thorium licensees for the costs of decontamination, decommissioning, reclamation, and other remedial actions which are incident to sales to the Government. The report presents the results of the independent certified public accountants` audit of the D and D Fund financial statements as of September 30, 1996. The auditors have expressed an unqualified opinion on the 1996 statement of financial position and the related statements of operations and changes in net position and cash flows.

  15. Conversion and enrichment in the Soviet Union

    SciTech Connect (OSTI)

    1991-04-01

    In the Soviet Union, just as in the West, the civilian nuclear industry emerged from research work undertaken for nuclear weapons development. At first, researchers tried various techniques for physical separation of uranium isotopes: electromagnetic and molecular-kinetic thermo-diffusion methods; gaseous diffusion; and centrifuge methods. All of those methods, which are based primarily on differences in the atomic mass of uranium isotopes, called for extensive research and the development of new, technically unprecedented equipment. Gradually gaseous diffusion and gas centrifuge technology became recognized as most feasible for industrial use, so research on other methods was terminated. Industrial-scale uranium enrichment in the Soviet Union began in 1949 using the gaseous diffusion method; by the early 1960s, centrifuge technology was in use on an industrial scale. All Soviet production of highly-enriched, weapons-grade uranium was halted in 1987. The Soviet Union now has four enrichment plants in operation (at classified locations), solely for civilian nuclear power needs. All four enrichment plants have centrifuge modules, and enrichment provided by gaseous diffusion accounts for less than 5% of their total output. Two of the four enrichment plants also incorporate facilities for conversion to uranium hexafluoride (UF{sub 6}).

  16. 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.

  17. 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

  18. Analysis of organizational options for the uranium enrichment enterprise in relation to asset divesture. [BPA; TVA; SYNFUELS; CONRAIL; British TELECOM; COMSTAT

    SciTech Connect (OSTI)

    Harrer, B.J.; Hattrup, M.P.; Dase, J.E.; Nicholls, A.K.

    1986-08-01

    This report presents a comparison of the characteristics of some prominent examples of independent government corporations and agencies with respect to the Department of Energy's (DOE) uranium enrichment enterprise. The six examples studied were: the Bonneville Power Administration (BPA); the Tennessee Valley Authority (TVA); the Synthetic Fuels Corporation (SYNFUELS); the Consolidated Rail Corporation (CONRAIL); the British Telecommunications Corporation (British TELECOM); and the Communications Satellite Organization (COMSAT), in order of decreasing levels of government ownership and control. They range from BPA, which is organized as an agency within DOE, to COMSAT, which is privately owned and free from almost all regulations common to government agencies. Differences in the degree of government involvement in these corporations and in many other characteristics serve to illustrate that there are no accepted standards for defining the characteristics of government corporations. Thus, historical precedent indicates considerable flexibility would be available in the development of enabling legislation to reorganize the enrichment enterprise as a government corporation or independent government agency.

  19. 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

  20. Structural Behaviour of Uranium Sulfide under High Pressure

    SciTech Connect (OSTI)

    Shareef, F.; Singh, S.; Gour, A.; Bhardwaj, P.; Sarwan, M.; Dubey, R. [High Pressure Research Lab, Department of Physics, Barkatullah University, Bhopal-462026 (India); Singh, R. K. [ITM University, Gurgaon, Haryana-122017 (India)

    2011-07-15

    The study of pressure induced structural phase transition of uranium sulphide, which crystallizes in rock salt (B1) structure, has been performed using the well described three body interaction model (TBIPM). Our present TBIP model consists of long range coulombic interaction, three body interactions, Hafemeister and Flygare type short-range overlap repulsion extended upto the second neighbor ions and the van der Waals interaction. The present results are in good agreement with the available experimental data on the phase transition pressure (Pt = 80.2 GPa). So it can be considered as an adequate and suitable model to perform high pressure studies.

  1. Results from a "Proof-of-Concept" Demonstration of RF-Based Tracking of UF6 Cylinders during a Processing Operation at a Uranium Enrichment Plant

    SciTech Connect (OSTI)

    Pickett, Chris A; Kovacic, Donald N; Whitaker, J Michael; Younkin, James R; Hines, Jairus B; Laughter, Mark D; Morgan, Jim; Carrick, Bernie; Boyer, Brian; Whittle, K.

    2008-01-01

    Approved industry-standard cylinders are used globally for processing, storing, and transporting uranium hexafluoride (UF{sub 6}) at uranium enrichment plants. To ensure that cylinder movements at enrichment facilities occur as declared, the International Atomic Energy Agency (IAEA) must conduct time-consuming periodic physical inspections to validate facility records, cylinder identity, and containment. By using a robust system design that includes the capability for real-time unattended monitoring (of cylinder movements), site-specific rules-based event detection algorithms, and the capability to integrate with other types of monitoring technologies, one can build a system that will improve overall inspector effectiveness. This type of monitoring system can provide timely detection of safeguard events that could be used to ensure more timely and appropriate responses by the IAEA. It also could reduce reliance on facility records and have the additional benefit of enhancing domestic safeguards at the installed facilities. This paper will discuss the installation and evaluation of a radio-frequency- (RF-) based cylinder tracking system that was installed at a United States Enrichment Corporation Centrifuge Facility. This system was installed primarily to evaluate the feasibility of using RF technology at a site and the operational durability of the components under harsh processing conditions. The installation included a basic system that is designed to support layering with other safeguard system technologies and that applies fundamental rules-based event processing methodologies. This paper will discuss the fundamental elements of the system design, the results from this site installation, and future efforts needed to make this technology ready for IAEA consideration.

  2. 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,...

  3. 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.

  4. Highly Enriched Uranyl Nitrate in Annular Tanks with Concrete Reflection: 1 x 3 Line Array of Nested Pairs of Tanks

    SciTech Connect (OSTI)

    James Cleaver; John D. Bess; Nathan Devine; Fitz Trumble

    2009-09-01

    A series of seven experiments were performed at the Rocky Flats Critical Mass Laboratory beginning in August, 1980 (References 1 and 2). Highly enriched uranyl nitrate solution was introduced into a 1-3 linear array of nested stainless steel annular tanks. The tanks were inside a concrete enclosure, with various moderator and absorber materials placed inside and/or between the tanks. These moderators and absorbers included boron-free concrete, borated concrete, borated plaster, and cadmium. Two configurations included placing bottles of highly enriched uranyl nitrate between tanks externally. Another experiment involved nested hemispheres of highly enriched uranium placed between tanks externally. These three configurations are not evaluated in this report. The experiments evaluated here are part of a series of experiments, one set of which is evaluated in HEU-SOL-THERM-033. The experiments in this and HEU-SOL-THERM-033 were performed similarly. They took place in the same room and used the same tanks, some of the same moderators and absorbers, some of the same reflector panels, and uranyl nitrate solution from the same location. There are probably additional similarities that existed that are not identified here. Thus, many of the descriptions in this report are either the same or similar to those in the HEU-SOL-THERM-033 report. Seventeen configurations (sixteen of which were critical) were performed during seven experiments; six of those experiments are evaluated here with thirteen configurations. Two configurations were identical, except for solution height, and were conducted to test repeatability. The solution heights were averaged and the two were evaluated as one configuration, which gives a total of twelve evaluated configurations. One of the seventeen configurations was subcritical. Of the twelve critical configurations evaluated, nine were judged as acceptable as benchmarks.

  5. High-solids enrichment of thermophilic microbial communities and their enzymes on bioenergy feedstocks

    SciTech Connect (OSTI)

    Reddy, A. P.; Allgaier, M.; Singer, S.W.; Hazen, T.C.; Simmons, B.A.; Hugenholtz, P.; VanderGheynst, J.S.

    2011-04-01

    Thermophilic microbial communities that are active in a high-solids environment offer great potential for the discovery of industrially relevant enzymes that efficiently deconstruct bioenergy feedstocks. In this study, finished green waste compost was used as an inoculum source to enrich microbial communities and associated enzymes that hydrolyze cellulose and hemicellulose during thermophilic high-solids fermentation of the bioenergy feedstocks switchgrass and corn stover. Methods involving the disruption of enzyme and plant cell wall polysaccharide interactions were developed to recover xylanase and endoglucanase activity from deconstructed solids. Xylanase and endoglucanase activity increased by more than a factor of 5, upon four successive enrichments on switchgrass. Overall, the changes for switchgrass were more pronounced than for corn stover; solids reduction between the first and second enrichments increased by a factor of four for switchgrass while solids reduction remained relatively constant for corn stover. Amplicon pyrosequencing analysis of small-subunit ribosomal RNA genes recovered from enriched samples indicated rapid changes in the microbial communities between the first and second enrichment with the simplified communities achieved by the third enrichment. The results demonstrate a successful approach for enrichment of unique microbial communities and enzymes active in a thermophilic high-solids environment.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. NEUTRALIZATIONS OF HIGH ALUMINUM LOW URANIUM USED NUCLEAR FUEL SOLUTIONS CONTAINING GADOLINIUM AS A NEUTRON POISON

    SciTech Connect (OSTI)

    Taylor-Pashow, K.

    2011-06-08

    H-Canyon will begin dissolving High Aluminum - Low Uranium (High Al/Low U) Used Nuclear Fuel (UNF) following approval by DOE which is anticipated in CY2011. High Al/Low U is an aluminum/enriched uranium UNF with small quantities of uranium relative to aluminum. The maximum enrichment level expected is 93% {sup 235}U. The High Al/Low U UNF will be dissolved in H-Canyon in a nitric acid/mercury/gadolinium solution. The resulting solution will be neutralized and transferred to Tank 39H in the Tank Farm. To confirm that the solution generated could be poisoned with Gd, neutralized, and discarded to the Savannah River Site (SRS) high level waste (HLW) system without undue nuclear safety concerns the caustic precipitation of simulant solutions was examined. Experiments were performed with three simulant solutions representative of the H-Canyon estimated concentrations in the final solutions after dissolution. The maximum U, Gd, and Al concentration were selected for testing from the range of solution compositions provided. Simulants were prepared in three different nitric acid concentrations, ranging from 0.5 to 1.5 M. The simulant solutions were neutralized to four different endpoints: (1) just before a solid phase was formed (pH 3.5-4), (2) the point where a solid phase was obtained, (3) 0.8 M free hydroxide, and (4) 1.2 M free hydroxide, using 50 wt % sodium hydroxide (NaOH). The settling behavior of the neutralized solutions was found to be slower compared to previous studies, with settling continuing over a one week period. Due to the high concentration of Al in these solutions, precipitation of solids was observed immediately upon addition of NaOH. Precipitation continued as additional NaOH was added, reaching a point where the mixture becomes almost completely solid due to the large amount of precipitate. As additional NaOH was added, some of the precipitate began to redissolve, and the solutions neutralized to the final two endpoints mixed easily and had expected densities of typical neutralized waste. Based on particle size and scanning electron microscopy analyses, the neutralized solids were found to be homogeneous and less than 20 microns in size. The majority of solids were less than 4 microns in size. Compared to previous studies, a larger percentage of the Gd was found to precipitate in the partially neutralized solutions (at pH 3.5-4). In addition the Gd:U mass ratio was found to be at least 1.0 in all of the solids obtained after partial or full neutralization. The hydrogen to U (H:U) molar ratios for two accident scenarios were also determined. The first was for transient neutralization and agitator failure. Experimentally this scenario was determined by measuring the H:U ratio of the settled solids. The minimum H:U molar ratio for solids from fully neutralized solutions was 388:1. The second accident scenario is for the solids drying out in an unagitiated pump box. Experimentally, this scenario was determined by measuring the H:U molar ratio in centrifuged solids. The minimum H:U atom ratios for centrifuged precipitated solids was 250:1. It was determined previously that a 30:1 H:Pu atom ratio was sufficient for a 1:1 Gd:Pu mass ratio. Assuming a 1:1 equivalence with {sup 239}Pu, the results of these experiments show Gd is a viable poison for neutralizing U/Gd solutions with the tested compositions.

  12. A HIGH-RESOLUTION ATLAS OF URANIUM-NEON IN THE H BAND (Journal Article) |

    Office of Scientific and Technical Information (OSTI)

    SciTech Connect A HIGH-RESOLUTION ATLAS OF URANIUM-NEON IN THE H BAND Citation Details In-Document Search Title: A HIGH-RESOLUTION ATLAS OF URANIUM-NEON IN THE H BAND We present a high-resolution (R Almost-Equal-To 50,000) atlas of a uranium-neon (U/Ne) hollow-cathode spectrum in the H band (1454-1638 nm) for the calibration of near-infrared spectrographs. We obtained this U/Ne spectrum simultaneously with a laser-frequency comb spectrum, which we used to provide a first-order calibration to

  13. New generation enrichment monitoring technology for gas centrifuge enrichment plants

    SciTech Connect (OSTI)

    Ianakiev, Kiril D; Alexandrov, Boian S.; Boyer, Brian D.; Hill, Thomas R.; Macarthur, Duncan W.; Marks, Thomas; Moss, Calvin E.; Sheppard, Gregory A.; Swinhoe, Martyn T.

    2008-06-13

    The continuous enrichment monitor, developed and fielded in the 1990s by the International Atomic Energy Agency, provided a go-no-go capability to distinguish between UF{sub 6} containing low enriched (approximately 4% {sup 235}U) and highly enriched (above 20% {sup 235}U) uranium. This instrument used the 22-keV line from a {sup 109}Cd source as a transmission source to achieve a high sensitivity to the UF{sub 6} gas absorption. The 1.27-yr half-life required that the source be periodically replaced and the instrument recalibrated. The instrument's functionality and accuracy were limited by the fact that measured gas density and gas pressure were treated as confidential facility information. The modern safeguarding of a gas centrifuge enrichment plant producing low-enriched UF{sub 6} product aims toward a more quantitative flow and enrichment monitoring concept that sets new standards for accuracy stability, and confidence. An instrument must be accurate enough to detect the diversion of a significant quantity of material, have virtually zero false alarms, and protect the operator's proprietary process information. We discuss a new concept for advanced gas enrichment assay measurement technology. This design concept eliminates the need for the periodic replacement of a radioactive source as well as the need for maintenance by experts. Some initial experimental results will be presented.

  14. 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.

  15. 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.

  16. Safeguards Guidance Document for Designers of Commercial Nuclear Facilities: International Nuclear Safeguards Requirements and Practices For Uranium Enrichment Plants

    SciTech Connect (OSTI)

    Robert Bean; Casey Durst

    2009-10-01

    This report is the second in a series of guidelines on international safeguards requirements and practices, prepared expressly for the designers of nuclear facilities. The first document in this series is the description of generic international nuclear safeguards requirements pertaining to all types of facilities. These requirements should be understood and considered at the earliest stages of facility design as part of a new process called “Safeguards-by-Design.” This will help eliminate the costly retrofit of facilities that has occurred in the past to accommodate nuclear safeguards verification activities. The following summarizes the requirements for international nuclear safeguards implementation at enrichment plants, prepared under the Safeguards by Design project, and funded by the U.S. Department of Energy (DOE) National Nuclear Security Administration (NNSA), Office of NA-243. The purpose of this is to provide designers of nuclear facilities around the world with a simplified set of design requirements and the most common practices for meeting them. The foundation for these requirements is the international safeguards agreement between the country and the International Atomic Energy Agency (IAEA), pursuant to the Treaty on the Non-proliferation of Nuclear Weapons (NPT). Relevant safeguards requirements are also cited from the Safeguards Criteria for inspecting enrichment plants, found in the IAEA Safeguards Manual, Part SMC-8. IAEA definitions and terms are based on the IAEA Safeguards Glossary, published in 2002. The most current specification for safeguards measurement accuracy is found in the IAEA document STR-327, “International Target Values 2000 for Measurement Uncertainties in Safeguarding Nuclear Materials,” published in 2001. For this guide to be easier for the designer to use, the requirements have been restated in plainer language per expert interpretation using the source documents noted. The safeguards agreement is fundamentally a legal document. As such, it is written in a legalese that is understood by specialists in international law and treaties, but not by most outside of this field, including designers of nuclear facilities. For this reason, many of the requirements have been simplified and restated. However, in all cases, the relevant source document and passage is noted so that readers may trace the requirement to the source. This is a helpful living guide, since some of these requirements are subject to revision over time. More importantly, the practices by which the requirements are met are continuously modernized by the IAEA and nuclear facility operators to improve not only the effectiveness of international nuclear safeguards, but also the efficiency. As these improvements are made, the following guidelines should be updated and revised accordingly.

  17. SPERT Destructive Test - I on Aluminum, Highly Enriched Plate Type Core

    SciTech Connect (OSTI)

    2011-04-05

    SPERT - Special Power Excursion Reactor Tests Destructive Test number 1 On Aluminum, Highly Enriched Plate Type Core. A test studying the behavior of the reactor under destructive conditions on a light water moderated pool-type reactor with a plate-type core.

  18. SPERT Destructive Test - I on Aluminum, Highly Enriched Plate Type Core

    ScienceCinema (OSTI)

    None

    2014-05-07

    SPERT - Special Power Excursion Reactor Tests Destructive Test number 1 On Aluminum, Highly Enriched Plate Type Core. A test studying the behavior of the reactor under destructive conditions on a light water moderated pool-type reactor with a plate-type core.

  19. Validation of MCNP with X6.XS cross-section set on the SUN Sparc Station 1+ computer for nominally 5 weight percent {sup 235}U enriched uranium systems

    SciTech Connect (OSTI)

    Lewis, K.D.

    1994-09-01

    The national Atomic Vapor Laser Isotope Separation (AVLIS) project has conducted extensive nuclear criticality safety analyses both in the design of Uranium Demonstration System (UDS) equipment and in AVLIS plant design/plant deployment activities. Currently, the design limit of an AVLIS plant calls for uranium product enriched in {sup 235}U to 5 wt %. Since an objective of an AVLIS plant is to deliver its product in a form readily usable by customers, uranium enriched in {sup 235}U will appear in a variety of forms, including metallic; as oxides, e.g., UO{sub 2}, UO{sub 3}; as fluorides, e.g., UF{sub 6}, UF{sub 4}, UO{sub 2}F{sub 2}; as nitrates or nitrides, e.g., UO{sub 2} (NO{sub 3}){sub 2}; and perhaps as uranium salts mixed with hydrocarbons such as oil. A wide range of neutron moderation levels, ranging from zero to optimal, and beyond can also be anticipated in an AVLIS plant, because of decontamination and cleaning activities and other wet chemistry processes that may be required.

  20. Y-12 uranium storage facility?a dream come true,? part 2

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

    part 2 Last week we introduced Shirley Cox and began a two-part series on her career at Y-12 leading up to the recommendation that the Highly Enriched Uranium Materials Facility...

  1. 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.

  2. R and D of crystal scintillators from enriched isotopes for high sensitivity double ? decay experiments

    SciTech Connect (OSTI)

    Danevich, F. A. [Institute for Nuclear Research, Kyiv (Ukraine)

    2013-12-30

    Experiments to search for neutrinoless double beta decay enters to a new phase when a sensitivity on the level of T{sub 1/2}?10{sup 26}?10{sup 28} yr is required. Scintillating low temperature detectors possess important properties required for high-sensitivity double beta decay experiments: presence of elements of interest, high energy resolution and detection efficiency, low level of background thanks to excellent particle discrimination ability. High concentration of isotope of interest and as low as possible radioactive contamination are important requirements to crystal scintillators. Other crucial issues are maximal output of detectors and minimal loss of enriched materials. Prospects of several scintillation materials, enriched in isotopes promising for double beta decay experiments, are discussed.

  3. DOE to Remove 200 Metric Tons of Highly Enriched Uranium from...

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

    Will Be Redirected to Naval Reactors, Down-blended or Used for Space Programs WASHINGTON, ... Approximately 20 MT will be reserved for space and research reactors that currently use ...

  4. 2 x 2 Polyethylene Reflected and Moderated Highly Enriched Uranium System with Rhenium

    SciTech Connect (OSTI)

    A. Nichole Ellis; Jesson Hutchinson; John D. Bess; Dmitry N. Polyakov; Evgeny S. Glushkov; Alexey E. Glushkov

    2010-09-01

    The 2 2 array HEU-Re experiment was performed on the Planet universal critical assembly machine on November 4th, 2003 at the Los Alamos Critical Experiments Facility (LACEF) at Los Alamos National Laboratory (LANL). For this experiment, there were 10 units, each full unit containing four HEU foils and two rhenium foils. The top unit contained only two HEU foils and two rhenium foils. A total of 42 HEU foils were used for this experiment. Rhenium is a desirable cladding material for space nuclear power applications. This experiment consisted of HEU foils interleaved with rhenium foils and is moderated and reflected by polyethylene plates. A unit consisted of a polyethylene plate, which has a recess for rhenium foils, and four HEU foils in a single layer in the top recess of each polyethylene plate. The Planet universal criticality assembly machine has been previously used in experiments containing HEU foils interspersed with SiO2 (HEU-MET-THERM-001), Al (HEU-MET-THERM-008), MgO (HEU-MET-THERM-009), Gd foils (HEU-MET-THERM-010), 2 2 26 Al (HEU-MET-THERM-012), Fe (HEU-MET-THERM-013 and HEU-MET-THERM-015), 2 2 23 SiO2 (HEU-MET-THERM-014), 2 2 11 hastalloy plates (HEU-MET-THERM-016), and concrete (HEU-MET-THERM-018). The 2 2 array of HEU-Re is considered acceptable for use as a benchmark critical experiment.

  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. Workshop on Preserving High Purity Uranium-233 (Conference) | SciTech

    Office of Scientific and Technical Information (OSTI)

    Connect Conference: Workshop on Preserving High Purity Uranium-233 Citation Details In-Document Search Title: Workshop on Preserving High Purity Uranium-233 A workshop was held on at the MARC X conference to provide a forum for the scientific community to communicate needs for high-purity 233U and its by-products in order to preserve critical items otherwise slated for downblending and disposal. Currently, only a small portion of the U.S. holdings of separated 233U is being preserved.

  7. Peak fitting applied to low-resolution enrichment measurements

    SciTech Connect (OSTI)

    Bracken, D.; McKown, T.; Sprinkle, J.K. Jr.; Gunnink, R.; Kartoshov, M.; Kuropatwinski, J.; Raphina, G.; Sokolov, G.

    1998-12-01

    Materials accounting at bulk processing facilities that handle low enriched uranium consists primarily of weight and uranium enrichment measurements. Most low enriched uranium processing facilities draw separate materials balances for each enrichment handled at the facility. The enrichment measurement determines the isotopic abundance of the {sup 235}U, thereby determining the proper strata for the item, while the weight measurement generates the primary accounting value for the item. Enrichment measurements using the passive gamma radiation from uranium were developed for use in US facilities a few decades ago. In the US, the use of low-resolution detectors was favored because they cost less, are lighter and more robust, and don`t require the use of liquid nitrogen. When these techniques were exported to Europe, however, difficulties were encountered. Two of the possible root causes were discovered to be inaccurate knowledge of the container wall thickness and higher levels of minor isotopes of uranium introduced by the use of reactor returns in the enrichment plants. the minor isotopes cause an increase in the Compton continuum under the 185.7 keV assay peak and the observance of interfering 238.6 keV gamma rays. The solution selected to address these problems was to rely on the slower, more costly, high-resolution gamma ray detectors when the low-resolution method failed. Recently, these gamma ray based enrichment measurement techniques have been applied to Russian origin material. The presence of interfering gamma radiation from minor isotopes was confirmed. However, with the advent of fast portable computers, it is now possible to apply more sophisticated analysis techniques to the low-resolution data in the field. Explicit corrections for Compton background, gamma rays from {sup 236}U daughters, and the attenuation caused by thick containers can be part of the least squares fitting routine. Preliminary results from field measurements in Kazakhstan will be discussed.

  8. 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

  9. Conceptual design of a high throughput electrorefining of a uranium by using graphite cathode

    SciTech Connect (OSTI)

    Lee, J.H.; Kang, Y.H.; Hwang, S.C.; Park, S.B.; Shim, J.B.; Lee, H.S.; Kim, E.H.; Park, S.W.

    2007-07-01

    Conceptual designing of a high throughput electro-refiner was performed by using basic experimental data and a commercial computational fluid dynamic code, CFX. An electro-refiner concept equipped with a graphite cathode bundle was designed to recover a high purity uranium product continuously without a noble metal contamination. The performance of the process for a decontamination of a noble metal in a uranium product was evaluated as a function of the process parameters such as the rotation speeds of the stirrer and the anode basket. (authors)

  10. Need for High-Purity Uranium-233 (Conference) | SciTech Connect

    Office of Scientific and Technical Information (OSTI)

    SciTech Connect Search Results Conference: Need for High-Purity Uranium-233 Citation Details In-Document Search Title: Need for High-Purity Uranium-233 Abstract is at the beginning of the paper. Authors: Krichinsky, Alan M [1] ; Canaan, R Douglas [2] ; Giaquinto, Joseph [1] ; Hexel, Cole R [1] + Show Author Affiliations ORNL {Doug} [ORNL Publication Date: 2014-01-01 OSTI Identifier: 1185625 DOE Contract Number: DE-AC05-00OR22725 Resource Type: Conference Resource Relation: Conference: 8th

  11. 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.

  12. Unattended Environmental Sampling and Laser-based Enrichment Assay for Detection of Undeclared HEU Production in Enrichment Plants

    SciTech Connect (OSTI)

    Anheier, Norman C.; Bushaw, Bruce A.

    2010-04-15

    Nuclear power is enjoying rapid growth as government energy policies and public demand shift toward carbon neutral energy production. Accompanying the growth in nuclear power is the requirement for increased nuclear fuel production, including a significant expansion in uranium enrichment capacity. Essential to the success of the nuclear energy renaissance is the development and implementation of sustainable, proliferation-resistant nuclear power generation. Unauthorized production of highly enriched uranium (HEU) remains the primary proliferation concern for modern gaseous centrifuge enrichment plants (GCEPs). While to date there has been no indication of declared, safeguarded GCEPs producing HEU, the massive separative work unit (SWU) processing power of modern GCEPs presents a significant latent risk of nuclear breakout and suggests the need for more timely detection of potential facility misuse. The Pacific Northwest National Laboratory is developing an unattended safeguards instrument, combining continuous aerosol particulate collection with uranium isotope assay, to provide timely HEU detection within a GCEP. This approach is based on laser vaporization of aerosol particulates, followed by laser spectroscopy to characterize the uranium enrichment level. We demonstrate enrichment assay, with relative isotope abundance uncertainty <5%, on individual micron-sized particles that are trace components within a mixture ‘background’ particles

  13. Unattended Monitoring of HEU Production in Gaseous Centrifuge Enrichment Plants using Automated Aerosol Collection and Laser-based Enrichment Assay

    SciTech Connect (OSTI)

    Anheier, Norman C.; Bushaw, Bruce A.

    2010-08-11

    Nuclear power is enjoying rapid growth as government energy policies and public demand shift toward low carbon energy production. Pivotal to the global nuclear power renaissance is the development and deployment of robust safeguards instrumentation that allows the limited resources of the IAEA to keep pace with the expansion of the nuclear fuel cycle. Undeclared production of highly enriched uranium (HEU) remains a primary proliferation concern for modern gaseous centrifuge enrichment plants (GCEPs), due to their massive separative work unit (SWU) processing power and comparably short cascade equilibrium timescale. The Pacific Northwest National Laboratory is developing an unattended safeguards instrument, combining continuous aerosol particulate collection with uranium isotope assay, to provide timely detection of HEU production within a GCEP. This approach is based on laser vaporization of aerosol particulates, followed by laser spectroscopy to characterize the uranium enrichment level. Our prior investigation demonstrated single-shot detection sensitivity approaching the femtogram range and relative isotope ratio uncertainty better than 10% using gadolinium as a surrogate for uranium. In this paper we present measurement results on standard samples containing traces of depleted, natural, and low enriched uranium, as well as measurements on aerodynamic size uranium particles mixed in background materials (e.g., dust, minerals, soils). Improvements and optimizations in the detection electronics, signal timing, calibration, and laser alignment have lead to significant improvements in detection sensitivity and enrichment accuracy, contributing to an overall reduction in the false alarm probability. The sample substrate media was also found to play a significant role in facilitating laser-induced vaporization and the production of energetic plasma conditions, resulting in ablation optimization and further improvements in the isotope abundance sensitivity.

  14. Incidence of High Nitrogen in Sintered Uranium Dioxide: A Case Study

    SciTech Connect (OSTI)

    Balakrishna, Palanki; Murty, B. Narasimha; Anuradha, M.; Yadav, R.B.; Jayaraj, R.N

    2005-05-15

    Nitrogen content, above the specified limit of 75 {mu}g(gU){sup -1}, was encountered in sintered uranium dioxide in the course of its manufacture. The cause was traced to the sintering process, wherein carbon, a degradation product of the die wall or admixed lubricant, was retained in the compact as a result of inadvertent reversal of gas flow in the sintering furnace. In the presence of carbon, the uranium dioxide reacted with nitrogen from the furnace atmosphere to form nitride. The compacts with high nitrogen were also those with low sintered density, arising from low green density. The low green density was due to filling problems of an inhomogeneous powder. The experiments carried out establish the causes of high nitrogen to be the carbon residue from lubricant when the UO{sub 2} is sintered in a cracked ammonia atmosphere.

  15. PHASE ANALYSES OF URANIUM-BEARING MINERALS FROM THE HIGH GRADE ORE, NOPAL I, PENA BLANCA, MEXICO

    SciTech Connect (OSTI)

    M. Ren; P. Goodell; A. Kelts; E.Y. Anthony; M. Fayek; C. Fan; C. Beshears

    2005-07-11

    The Nopal I uranium deposit is located in the Pena Blanca district, approximately 40 miles north of Chihuahua City, Mexico. The deposit was formed by hydrothermal processes within the fracture zone of welded silicic volcanic tuff. The ages of volcanic formations are between 35 to 44 m.y. and there was secondary silicification of most of the formations. After the formation of at least part of the uranium deposit, the ore body was uplifted above the water table and is presently exposed at the surface. Detailed petrographic characterization, electron microprobe backscatter electron (BSE) imagery, and selected x-ray maps for the samples from Nopal I high-grade ore document different uranium phases in the ore. There are at least two stages of uranium precipitation. A small amount of uraninite is encapsulated in silica. Hexavalent uranium may also have been a primary precipitant. The uranium phases were precipitated along cleavages of feldspars, and along fractures in the tuff. Energy dispersive spectrometer data and x-ray maps suggest that the major uranium phases are uranophane and weeksite. Substitutions of Ca and K occur in both phases, implying that conditions were variable during the mineralization/alteration process, and that compositions of the original minerals have a major influence on later stage alteration. Continued study is needed to fully characterize uranium behavior in these semi-arid to arid conditions.

  16. Uranium atomic vapor laser isotope separation (AVL1S)

    SciTech Connect (OSTI)

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

    1992-12-01

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

  17. 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

  18. Uranium Transport in a High-Throughput Electrorefiner for EBR-II Blanket Fuel

    SciTech Connect (OSTI)

    Ahluwalia, Rajesh K.; Hua, Thanh Q.; Vaden, DeeEarl

    2004-01-15

    A unique high-throughput Mk-V electrorefiner is being used in the electrometallurgical treatment of the metallic sodium-bonded blanket fuel from the Experimental Breeder Reactor II. Over many cycles, it transports uranium back and forth between the anodic fuel dissolution baskets and the cathode tubes until, because of imperfect adherence of the dendrites, it all ends up in the product collector at the bottom. The transport behavior of uranium in the high-throughput electrorefiner can be understood in terms of the sticking coefficients for uranium adherence to the cathode tubes in the forward direction and to the dissolution baskets in the reverse direction. The sticking coefficients are inferred from the experimental voltage and current traces and are correlated in terms of a single parameter representing the ratio of the cell current to the limiting current at the surface acting as the cathode. The correlations are incorporated into an engineering model that calculates the transport of uranium in the different modes of operation. The model also uses the experimentally derived electrorefiner operating maps that describe the relationship between the cell voltage and the cell current for the three principal transport modes. It is shown that the model correctly simulates the cycle-to-cycle variation of the voltage and current profiles. The model is used to conduct a parametric study of electrorefiner throughput rate as a function of the principal operating parameters. The throughput rate is found to improve with lowering of the basket rotation speed, reduction of UCl{sub 3} concentration in salt, and increasing the maximum cell current or cut-off voltage. Operating conditions are identified that can improve the throughput rate by 60 to 70% over that achieved at present.

  19. Uranium Oxide as a Highly Reflective Coating from 100-400 eV

    SciTech Connect (OSTI)

    Sandberg, Richard L.; Allred, David D.; Bissell, Luke J.; Johnson, Jed E.; Turley, R. Steven

    2004-05-12

    We present the measured reflectances (Beamline 6.3.2, ALS at LBNL) of naturally oxidized uranium and naturally oxidized nickel thin films from 100-460 eV (2.7 to 11.6 nm) at 5 and 15 degrees grazing incidence. These show that uranium, as UO2, can fulfill its promise as the highest known single surface reflector for this portion of the soft x-ray region, being nearly twice as reflective as nickel in the 124-250 eV (5-10 nm) region. This is due to its large index of refraction coupled with low absorption. Nickel is commonly used in soft x-ray applications in astronomy and synchrotrons. (Its reflectance at 10 deg. exceeds that of Au and Ir for most of this range.) We prepared uranium and nickel thin films via DC-magnetron sputtering of a depleted U target and resistive heating evaporation respectively. Ambient oxidation quickly brought the U sample to UO2 (total thickness about 30 nm). The nickel sample (50 nm) also acquired a thin native oxide coating (<2nm). Though the density of U in UO2 is only half of the metal, its reflectance is high and it is relatively stable against further changes.

  20. Advanced uranium enrichment technologies. Hearing before the Subcommittee on Energy Research and Production of the Committee on Science and Technology, US House of Representatives, Ninety-Sixth Congress, first session, September 22, 1979

    SciTech Connect (OSTI)

    Not Available

    1980-01-01

    This hearing was to learn about projected requirements for enriched uranium. The gas centrifuge work at Oak Ridge, Tennessee, and Portsmouth, Ohio, needed assessing. Laser isotope separation technique needed to be reviewed. Three technologies currently being emphasized in the Department of Energy's Advanced Isotope Separation (AIS) program were discussed; these included the Molecular Laser Isotope Separation (MLIS), Livermore's process called Atomic Vapor Laser Isotope Separation (AVLIS), and Plasma Separation Process (PSP). The status of each process was given. The present DOE AIS program calls for a process selection at the end of FY 1981, development module operation starting in the mid-1980's, pilot plant operations through the late 1980's and early 1990's, and a first production plant in the mid-1990's. (DP)

  1. Preliminary study on weapon grade uranium utilization in molten salt reactor miniFUJI

    SciTech Connect (OSTI)

    Aji, Indarta Kuncoro; Waris, A.

    2014-09-30

    Preliminary study on weapon grade uranium utilization in 25MWth and 50MWth of miniFUJI MSR (molten salt reactor) has been carried out. In this study, a very high enriched uranium that we called weapon grade uranium has been employed in UF{sub 4} composition. The {sup 235}U enrichment is 90 - 95 %. The results show that the 25MWth miniFUJI MSR can get its criticality condition for 1.56 %, 1.76%, and 1.96% of UF{sub 4} with {sup 235}U enrichment of at least 93%, 90%, and 90%, respectively. In contrast, the 50 MWth miniFUJI reactor can be critical for 1.96% of UF{sub 4} with {sup 235}U enrichment of at smallest amount 95%. The neutron spectra are almost similar for each power output.

  2. Consideration of critically when directly disposing highly enriched spent nuclear fuel in unsaturated tuff: Bounding estimates

    SciTech Connect (OSTI)

    Rechard, R.P.; Tierney, M.S.; Sanchez, L.C.; Martell, M.-A.

    1996-05-01

    This report presents one of 2 approaches (bounding calculations) which were used in a 1994 study to examine the possibility of a criticality in a repository. Bounding probabilities, although rough, point to the difficulty of creating conditions under which a critical mass could be assembled (container corrosion, separation of neutron absorbers from fissile material, collapse or precipitation of fissile material) and how significant the geochemical and hydrologic phenomena are. The study could not conceive of a mechanism consistent with conditions under which an atomic explosion could occur. Should a criticality occur in or near a container in the future, boundary consequence calculations showed that fissions from one critical event (<10{sup 20} fissions, if similar to aqueous and metal accidents and experiments) are quite small compared to the amount of fissions represented by the spent fuel itself. If it is assumed that the containers necessary to hold the highly enriched spent fuel went critical once per day for 1 million years, creating an energy release of about 10{sup 20} fissions, the number of fissions equals about 10{sup 28}, which corresponds to only 1% of the fission inventory in a repository containing 70,000 metric tons of heavy metal, the expected size for the proposed repository at Yucca Mountain, Nevada.

  3. Detection of illicit HEU production in gaseous centrifuge enrichment plants using neutron counting techniques on product cylinders

    SciTech Connect (OSTI)

    Freeman, Corey R; Geist, William H

    2010-01-01

    Innovative and novel safeguards approaches are needed for nuclear energy to meet global energy needs without the threat of nuclear weapons proliferation. Part of these efforts will include creating verification techniques that can monitor uranium enrichment facilities for illicit production of highly-enriched uranium (HEU). Passive nondestructive assay (NDA) techniques will be critical in preventing illicit HEU production because NDA offers the possibility of continuous and unattended monitoring capabilities with limited impact on facility operations. Gaseous centrifuge enrichment plants (GCEP) are commonly used to produce low-enriched uranium (LEU) for reactor fuel. In a GCEP, gaseous UF{sub 6} spins at high velocities in centrifuges to separate the molecules containing {sup 238}U from those containing the lighter {sup 235}U. Unfortunately, the process for creating LEU is inherently the same as HEU, creating a proliferation concern. Insuring that GCEPs are producing declared enrichments poses many difficult challenges. In a GCEP, large cascade halls operating thousands of centrifuges work together to enrich the uranium which makes effective monitoring of the cascade hall economically prohibitive and invasive to plant operations. However, the enriched uranium exiting the cascade hall fills product cylinders where the UF{sub 6} gas sublimes and condenses for easier storage and transportation. These product cylinders hold large quantities of enriched uranium, offering a strong signal for NDA measurement. Neutrons have a large penetrability through materials making their use advantageous compared to gamma techniques where the signal is easily attenuated. One proposed technique for detecting HEU production in a GCEP is using neutron coincidence counting at the product cylinder take off stations. This paper discusses findings from Monte Carlo N-Particle eXtended (MCNPX) code simulations that examine the feasibility of such a detector.

  4. Development of the laser isotope separation method (AVLIS) for obtaining weight amounts of highly enriched {sup 150}Nd isotope

    SciTech Connect (OSTI)

    Babichev, A P; Grigoriev, Igor' S; Grigoriev, A I; Dorovskii, A P; D'yachkov, Aleksei B; Kovalevich, S K; Kochetov, V A; Kuznetsov, V A; Labozin, Valerii P; Matrakhov, A V; Mironov, Sergei M; Nikulin, Sergei A; Pesnya, A V; Timofeev, N I; Firsov, Valerii A; Tsvetkov, G O; Shatalova, G G

    2005-10-31

    Results obtained at the first stage of development of the experimental technique for obtaining weight amounts of the highly enriched {sup 150}Nd isotope by laser photoionisation are presented. The vaporiser and the laser are designed, and various methods of irradiation of neodymium vapour and extraction of photoions are tested. The product yield {approx}40 mg h{sup -1} for the {approx}60% enrichment and 25 mg h{sup -1} for the {approx}65% enrichment is achieved for a vaporiser of length 27 cm. The cost of constructing the facility for preparing 50 kg of the {sup 150}Nd isotope, intended for determining the neutrino mass, is estimated. This estimate shows that the cost of production can be lowered by a factor of 5-7 compared to the electromagnetic method. (invited paper)

  5. A HIGH-RESOLUTION ATLAS OF URANIUM-NEON IN THE H BAND

    SciTech Connect (OSTI)

    Redman, Stephen L.; Terrien, Ryan; Mahadevan, Suvrath; Ramsey, Lawrence W.; Bender, Chad F. [Department of Astronomy and Astrophysics, Pennsylvania State University, University Park, PA 16802 (United States); Ycas, Gabriel G. [Department of Physics, University of Colorado, Boulder, CO 80309 (United States); Osterman, Steven N. [Center for Astrophysics and Space Astronomy, University of Colorado, Boulder, CO 80309 (United States); Diddams, Scott A.; Quinlan, Franklyn [Time and Frequency Division, National Institute of Standards and Technology, Boulder, CO 80305 (United States); Lawler, James E. [Department of Physics, University of Wisconsin, 1150 University Avenue, Madison, WI 53706 (United States); Nave, Gillian [Atomic Physics Division, National Institute of Standards and Technology, Gaithersburg, MD 20899 (United States)

    2012-03-01

    We present a high-resolution (R Almost-Equal-To 50,000) atlas of a uranium-neon (U/Ne) hollow-cathode spectrum in the H band (1454-1638 nm) for the calibration of near-infrared spectrographs. We obtained this U/Ne spectrum simultaneously with a laser-frequency comb spectrum, which we used to provide a first-order calibration to the U/Ne spectrum. We then calibrated the U/Ne spectrum using the recently published uranium line list of Redman et al., which is derived from high-resolution Fourier transform spectrometer measurements. These two independent calibrations allowed us to easily identify emission lines in the hollow-cathode lamp that do not correspond to known (classified) lines of either uranium or neon, and to compare the achievable precision of each source. Our frequency comb precision was limited by modal noise and detector effects, while the U/Ne precision was limited primarily by the signal-to-noise ratio (S/N) of the observed emission lines and our ability to model blended lines. The standard deviation in the dispersion solution residuals from the S/N-limited U/Ne hollow-cathode lamp was 50% larger than the standard deviation of the dispersion solution residuals from the modal-noise-limited laser-frequency comb. We advocate the use of U/Ne lamps for precision calibration of near-infrared spectrographs, and this H-band atlas makes these lamps significantly easier to use for wavelength calibration.

  6. Fiscal year 1986 Department of Energy Authorization (uranium enrichment and electric energy systems, energy storage and small-scale hydropower programs). Volume VI. Hearings before the Subcommittee on Energy Research and Production of the Committee on Science and Technology, US House of Representatives, Ninety-Ninth Congress, First Session, February 28; March 5, 7, 1985

    SciTech Connect (OSTI)

    Not Available

    1985-01-01

    Volume VI of the hearing record covers three days of testimony on the future of US uranium enrichment and on programs involving electric power and energy storage. There were four areas of concern about uranium enrichment: the choice between atomic vapor laser isotope separation (AVLIS) and the advanced gas centrifuge (AGC) technologies, cost-effective operation of gaseous diffusion plants, plans for a gas centrifuge enrichment plant, and how the DOE will make its decision. The witnesses represented major government contractors, research laboratories, and energy suppliers. The discussion on the third day focused on the impact of reductions in funding for electric energy systems and energy storage and a small budget increase to encourage small hydropower technology transfer to the private sector. Two appendices with additional statements and correspondence follow the testimony of 17 witnesses.

  7. Assessment of Controlling Processes for Field-Scale Uranium Reactive Transport under Highly Transient Flow Conditions

    SciTech Connect (OSTI)

    Ma, Rui; Zheng, Chunmiao; Liu, Chongxuan; Greskowiak, Janek; Prommer, Henning; Zachara, John M.

    2014-02-13

    This paper presents the results of a comprehensive model-based analysis of a uranium tracer test conducted at the U.S Department of Energy Hanford 300 Area (300A) IFRC site. A three-dimensional multi-component reactive transport model was employed to assess the key factors and processes that control the field-scale uranium reactive transport. Taking into consideration of relevant physical and chemical processes, the selected conceptual/numerical model replicates the spatial and temporal variations of the observed U(VI) concentrations reasonably well in spite of the highly complex field conditions. A sensitivity analysis was performed to interrogate the relative importance of various processes and factors for reactive transport of U(VI) at the field-scale. The results indicate that multi-rate U(VI) sorption/desorption, U(VI) surface complexation reactions, and initial U(VI) concentrations were the most important processes and factors controlling U(VI) migration. On the other hand, cation exchange reactions, the choice of the surface complexation model, and dual-domain mass transfer processes, which were previously identified to be important in laboratory experiments, played less important roles under the field-scale experimental condition at the 300A site. However, the model simulations also revealed that the groundwater chemistry was relatively stable during the uranium tracer experiment and therefore presumably not dynamic enough to appropriately assess the effects of ion exchange reaction and the choice of surface complexation models on U(VI) sorption and desorption. Furthermore, it also showed that the field experimental duration (16 days) was not sufficiently long to precisely assess the role of a majority of the sorption sites that were accessed by slow kinetic processes within the dual domain model. The sensitivity analysis revealed the crucial role of the intraborehole flow that occurred within the long-screened monitoring wells and thus significantly affected both field-scale measurements and simulated U(VI) concentrations as a combined effect of aquifer heterogeneity and highly dynamic flow conditions. Overall, this study, which provides one of the few detailed and highly data-constrained uranium transport simulations, highlights the difference in controlling processes between laboratory and field scale that prevent a simple direct upscaling of laboratory-scale models.

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

    SciTech Connect (OSTI)

    Anheier, Norman C.; Bushaw, Bruce A.

    2009-07-01

    The rapid global expansion of nuclear energy is motivating the expedited development of new safeguards technology to mitigate potential proliferation threats arising from monitoring gaps within the uranium enrichment process. Current onsite enrichment level monitoring methods are limited by poor sensitivity and accuracy performance. Offsite analysis has better performance, but this approach requires onsite hand sampling followed by time-consuming and costly post analysis. These limitations make it extremely difficult to implement comprehensive safeguards accounting measures that can effectively counter enrichment facility misuse. In addition, uranium enrichment by modern centrifugation leads to a significant proliferation threat, since the centrifuge cascades can quickly produce a significant quantity of highly enriched uranium (HEU). The Pacific Northwest National Laboratory is developing an engineered safeguards approach having continuous aerosol particulate collection and uranium isotope analysis to provide timely detection of HEU production in a low enriched uranium facility. This approach is based on laser vaporization of aerosol particulate samples, followed by wavelength tuned laser diode spectroscopy, to characterize the 235U/238U isotopic ratio by subtle differences in atomic absorption wavelengths arising from differences in each isotopes nuclear mass, volume, and spin (hyperfine structure for 235U). Environmental sampling media is introduced into a small, reduced pressure chamber, where a focused pulsed laser vaporizes a 10 to 20-m sample diameter. The ejected plasma forms a plume of atomic vapor. A plume for a sample containing uranium has atoms of the 235U and 238U isotopes present. Tunable diode lasers are directed through the plume to selectively excite each isotope and their presence is detected by monitoring absorbance signals on a shot-to-shot basis. Single-shot detection sensitivity approaching the femtogram range and abundance uncertainty less than 10% have been demonstrated with measurements on surrogate materials. In this paper we present measurement results on samples containing background materials (e.g., dust, minerals, soils) laced with micron-sized target particles having isotopic ratios ranging from 1 to 50%.

  9. Automated UF6 Cylinder Enrichment Assay: Status of the Hybrid Enrichment Verification Array (HEVA) Project: POTAS Phase II

    SciTech Connect (OSTI)

    Jordan, David V.; Orton, Christopher R.; Mace, Emily K.; McDonald, Benjamin S.; Kulisek, Jonathan A.; Smith, Leon E.

    2012-06-01

    Pacific Northwest National Laboratory (PNNL) intends to automate the UF6 cylinder nondestructive assay (NDA) verification currently performed by the International Atomic Energy Agency (IAEA) at enrichment plants. PNNL is proposing the installation of a portal monitor at a key measurement point to positively identify each cylinder, measure its mass and enrichment, store the data along with operator inputs in a secure database, and maintain continuity of knowledge on measured cylinders until inspector arrival. This report summarizes the status of the research and development of an enrichment assay methodology supporting the cylinder verification concept. The enrichment assay approach exploits a hybrid of two passively-detected ionizing-radiation signatures: the traditional enrichment meter signature (186-keV photon peak area) and a non-traditional signature, manifested in the high-energy (3 to 8 MeV) gamma-ray continuum, generated by neutron emission from UF6. PNNL has designed, fabricated, and field-tested several prototype assay sensor packages in an effort to demonstrate proof-of-principle for the hybrid assay approach, quantify the expected assay precision for various categories of cylinder contents, and assess the potential for unsupervised deployment of the technology in a portal-monitor form factor. We refer to recent sensor-package prototypes as the Hybrid Enrichment Verification Array (HEVA). The report provides an overview of the assay signatures and summarizes the results of several HEVA field measurement campaigns on populations of Type 30B UF6 cylinders containing low-enriched uranium (LEU), natural uranium (NU), and depleted uranium (DU). Approaches to performance optimization of the assay technique via radiation transport modeling are briefly described, as are spectroscopic and data-analysis algorithms.

  10. Active Interrogation Observables for Enrichment Determination of DU Shielded HEU Metal Assemblies with Limited Geometrical Information

    SciTech Connect (OSTI)

    Pena, Kirsten E; McConchie, Seth M; Crye, Jason Michael; Mihalczo, John T

    2011-01-01

    Determining the enrichment of highly enriched uranium (HEU) metal assemblies shielded by depleted uranium (DU) proves a unique challenge to currently employed measurement techniques. Efforts to match time-correlated neutron distributions obtained through active interrogation to Monte Carlo simulations of the assemblies have shown promising results, given that the exact geometries of both the HEU metal assemblies and DU shields are known from imaging and fission site mapping. In certain situations, however, it is desirable to obtain enrichment with limited or no geometrical information of the assemblies being measured. This paper explores the possibility that the utilization of observables in the interrogation of assemblies by time-tagged D-T neutrons, including time-correlated distribution of neutrons and gammas using liquid scintillators operating on the fission chain time scale, can lead to enrichment determination without a complete set of geometrical information.

  11. 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.

  12. Direct fissile assay of highly enriched UF/sub 6/ using random self-interrogation and neutron coincidence response

    SciTech Connect (OSTI)

    Stewart, J.E.; Menlove, H.O.

    1983-01-01

    A new nondestructive method for direct assay of /sup 235/U mass contained in Model 5A uranium hexafluoride (UF/sub 6/) product storage cylinders has been successfully tested in the laboratory and under field conditions. The technique employs passive neutron self-interrogation and uses the ratio of coincidences-to-totals counts as a measure of bulk fissile mass. The accuracy of the method is 6.8% (1 sigma) based on field measurements of 44 Model 5A cylinders, 11 of which were either only partially filled or contained reactor return material. The cylinders contained UF/sub 6/ with enrichments from 5.96% to 97.6%. Count times were 3 to 6 min depending on /sup 235/U mass. Samples ranged from below 1 kg to over 16 kg of /sup 235/U. Because the method relies primarily on fast neutron self-interrogation, complete sampling of the UF/sub 6/ takes place. This feature alleviates inhomogeneity problems and offers increased assurance of the presence of stated amounts of bulk fissile material as compared with current verification methods.

  13. Tomographic gamma scanning of uranium-contaminated waste at Rocky Flats

    SciTech Connect (OSTI)

    Mercer, D.J.; Betts, S.E.; Prettyman, T.H.; Rael, C.D.

    1998-12-31

    A tomographic gamma-ray scanning (TGS) instrument was deployed at Rocky Flats Environmental Technology Site (RFETS) to assist with the deactivation of Building 886. Many 208-L drums containing waste contaminated with highly enriched uranium were measured in order to certify these sites for shipment and disposal. This project marks a successful cooperation between RFETS and Los Alamos National Laboratory and is the first major field experience using TGS technology to assay uranium.

  14. High-resolution mineralogical characterization and biogeochemical modeling of uranium reaction pathways at the FRC

    SciTech Connect (OSTI)

    Chen Zhu

    2006-06-15

    High-Resolution Mineralogical Characterization and Biogeochemical Modeling of Uranium Reduction Pathways at the Oak Ridge Field-Research Center (FRC) Chen Zhu, Indiana University, David R. Veblen, Johns Hopkins University We have successfully completed a proof-of-concept, one-year grant on a three-year proposal from the former NABIR program, and here we seek additional two-year funding to complete and publish the research. Using a state-of-the-art 300-kV, atomic resolution, Field Emission Gun Transmission Electron Microscope (TEM), we have successfully identified three categories of mineral hosts for uranium in contaminated soils: (1) iron oxides; (2) mixed manganese-iron oxides; and (3) uranium phosphates. Method development using parallel electron energy loss spectroscopy (EELS) associated with the TEM shows great promise for characterizing the valence states of immobilized U during bioremediation. We have also collected 27 groundwater samples from two push-pull field biostimulation tests, which form two time series from zero to approximately 600 hours. The temporal evolution in major cations, anions, trace elements, and the stable isotopes 34S, 18O in sulfate, 15N in nitrate, and 13C in dissolved inorganic carbon (DIC) clearly show that biostimulation resulted in reduction of nitrate, Mn(IV), Fe(III), U(VI), sulfate, and Tc(VII), and these reduction reactions were intimately coupled with a complex network of inorganic reactions evident from alkalinity, pH, Na, K, Mg, and Ca concentrations. From these temporal trends, apparent zero order rates were regressed. However, our extensive suite of chemical and isotopic data sets, perhaps the first and only comprehensive data set available at the FRC, show that the derived rates from these field biostimulation experiments are composite and lump-sum rates. There were several reactions that were occurring at the same time but were masked by these pseudo-zero order rates. A reaction-path model comprising a total of nine redox couples (NO3–/NH4+, MnO2(s)/Mn2+, Fe(OH)3(s) /Fe2+, TcO4–/TcO2(s), UO22+/UO2(s), SO42–/HS–, CO2/CH4, ethanol/acetate, and H+/H2.) is used to simulate the temporal biogeochemical evolution observed in the field tests. Preliminary results show that the models based on thermodynamics and more complex rate laws can generate the apparent zero order rates when several concurrent or competing reactions occur. Professor Alex Halliday of Oxford University, UK, and his postdoctoral associates are measuring the uranium isotopes in our groundwater samples. Newly developed state-of-the-art analytical techniques in measuring variability in 235U/238U offer the potential to distinguish biotic and abiotic uranium reductive mechanisms.

  15. Determination of uranium and thorium in semiconductor memory materials by high fluence neutron activation analysis

    SciTech Connect (OSTI)

    Dyer, F.F.; Emery, J.F.; Northcutt, K.J.; Scott, R.M.

    1981-01-01

    Uranium and thorium were measured by absolute neutron activation analysis in high-purity materials used to manufacture semiconductor memories. The main thrust of the study concerned aluminum and aluminum alloys used as sources for thin film preparation, evaporated metal films, and samples from the Czochralski silicon crystal process. Average levels of U and Th were found for the source alloys to be approx. 65 and approx. 45 ppB, respectively. Levels of U and Th in silicon samples fell in the range of a few parts per trillion. Evaporated metal films contained about 1 ppB U and Th, but there is some question about these results due to the possibility of contamination.

  16. 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

  17. Development of Novel Sorbents for Uranium Extraction from Seawater

    SciTech Connect (OSTI)

    Lin, Wenbin; Taylor-Pashow, Kathryn

    2014-01-08

    As the uranium resource in terrestrial ores is limited, it is difficult to ensure a long-term sustainable nuclear energy technology. The oceans contain approximately 4.5 billion tons of uranium, which is one thousand times the amount of uranium in terrestrial ores. Development of technologies to recover the uranium from seawater would greatly improve the uranium resource availability, sustaining the fuel supply for nuclear energy. Several methods have been previously evaluated including solvent extraction, ion exchange, flotation, biomass collection, and adsorption; however, none have been found to be suitable for reasons such as cost effectiveness, long term stability, and selectivity. Recent research has focused on the amidoxime functional group as a promising candidate for uranium sorption. Polymer beads and fibers have been functionalized with amidoxime functional groups, and uranium adsorption capacities as high as 1.5 g U/kg adsorbent have recently been reported with these types of materials. As uranium concentration in seawater is only ~3 ppb, great improvements to uranium collection systems must be made in order to make uranium extraction from seawater economically feasible. This proposed research intends to develop transformative technologies for economic uranium extraction from seawater. The Lin group will design advanced porous supports by taking advantage of recent breakthroughs in nanoscience and nanotechnology and incorporate high densities of well-designed chelators into such nanoporous supports to allow selective and efficient binding of uranyl ions from seawater. Several classes of nanoporous materials, including mesoporous silica nanoparticles (MSNs), mesoporous carbon nanoparticles (MCNs), meta-organic frameworks (MOFs), and covalent-organic frameworks (COFs), will be synthesized. Selective uranium-binding liagnds such as amidoxime will be incorporated into the nanoporous materials to afford a new generation of sorbent materials that will be evaluated for their uranium extraction efficiency. The initial testing of these materials for uranium binding will be carried out in the Lin group, but more detailed sorption studies will be carried out by Dr. Taylor-Pashow of Savannah River National Laboratory in order to obtain quantitative uranyl sorption selectivity and kinetics data for the proposed materials. The proposed nanostructured sorbent materials are expected to have higher binding capacities, enhanced extraction kinetics, optimal stripping efficiency for uranyl ions, and enhanced mechanical and chemical stabilities. This transformative research will significantly impact uranium extraction from seawater as well as benefit DOE’s efforts on environmental remediation by developing new materials and providing knowledge for enriching and sequestering ultralow concentrations of other metals.

  18. Laboratory Enrichment of Radioactive Assemblages and Estimation of Thorium

    Office of Scientific and Technical Information (OSTI)

    and Uranium Radioactivity in Fractions Separated from Placer Sands in Southeast Bangladesh (Journal Article) | SciTech Connect SciTech Connect Search Results Journal Article: Laboratory Enrichment of Radioactive Assemblages and Estimation of Thorium and Uranium Radioactivity in Fractions Separated from Placer Sands in Southeast Bangladesh Citation Details In-Document Search Title: Laboratory Enrichment of Radioactive Assemblages and Estimation of Thorium and Uranium Radioactivity in

  19. Environmental acceptability of high-performance alternatives for depleted uranium penetrators

    SciTech Connect (OSTI)

    Kerley, C.R.; Easterly, C.E.; Eckerman, K.F.

    1996-08-01

    The Army`s environmental strategy for investigating material substitution and management is to measure system environmental gains/losses in all phases of the material management life cycle from cradle to grave. This study is the first in a series of new investigations, applying material life cycle concepts, to evaluate whether there are environmental benefits from increasing the use of tungsten as an alternative to depleted uranium (DU) in Kinetic Energy Penetrators (KEPs). Current military armor penetrators use DU and tungsten as base materials. Although DU alloys have provided the highest performance of any high-density alloy deployed against enemy heavy armor, its low-level radioactivity poses a number of environmental risks. These risks include exposures to the military and civilian population from inhalation, ingestion, and injection of particles. Depleted uranium is well known to be chemically toxic (kidney toxicity), and workplace exposure levels are based on its renal toxicity. Waste materials containing DU fragments are classified as low-level radioactive waste and are regulated by the Nuclear Regulatory Commission. These characteristics of DU do not preclude its use in KEPs. However, long-term management challenges associated with KEP deployment and improved public perceptions about environmental risks from military activities might be well served by a serious effort to identify, develop, and substitute alternative materials that meet performance objectives and involve fewer environmental risks. Tungsten, a leading candidate base material for KEPS, is potentially such a material because it is not radioactive. Tungsten is less well studied, however, with respect to health impacts and other environmental risks. The present study is designed to contribute to the understanding of the environmental behavior of tungsten by synthesizing available information that is relevant to its potential use as a penetrator.

  20. Preparation of High Purity, High Molecular-Weight Chitin from Ionic Liquids for Use as an Adsorbate for the Extraction of Uranium from Seawater

    SciTech Connect (OSTI)

    Rogers, Robin

    2013-12-21

    Ensuring a domestic supply of uranium is a key issue facing the wider implementation of nuclear power. Uranium is mostly mined in Kazakhstan, Australia, and Canada, and there are few high-grade uranium reserves left worldwide. Therefore, one of the most appealing potential sources of uranium is the vast quantity dissolved in the oceans (estimated to be 4.4 billion tons worldwide). There have been research efforts centered on finding a means to extract uranium from seawater for decades, but so far none have resulted in an economically viable product, due in part to the fact that the materials that have been successfully demonstrated to date are too costly (in terms of money and energy) to produce on the necessary scale. Ionic Liquids (salts which melt below 100{degrees}C) can completely dissolve raw crustacean shells, leading to recovery of a high purity, high molecular weight chitin powder and to fibers and films which can be spun directly from the extract solution suggesting that continuous processing might be feasible. The work proposed here will utilize the unprecedented control this makes possible over the chitin fiber a) to prepare electrospun nanofibers of very high surface area and in specific architectures, b) to modify the fiber surfaces chemically with selective extractant capacity, and c) to demonstrate their utility in the direct extraction and recovery of uranium from seawater. This approach will 1) provide direct extraction of chitin from shellfish waste thus saving energy over the current industrial process for obtaining chitin; 2) allow continuous processing of nanofibers for very high surface area fibers in an economical operation; 3) provide a unique high molecular weight chitin not available from the current industrial process, leading to stronger, more durable fibers; and 4) allow easy chemical modification of the large surface areas of the fibers for appending uranyl selective functionality providing selectivity and ease of stripping. The resulting sorbent should prove economically feasible, as well as providing an overall net energy gain.

  1. 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.

  2. 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.

  3. 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.

  4. Gas centrifuge enrichment plants inspection frequency and remote monitoring issues for advanced safeguards implementation

    SciTech Connect (OSTI)

    Boyer, Brian David; Erpenbeck, Heather H; Miller, Karen A; Ianakiev, Kiril D; Reimold, Benjamin A; Ward, Steven L; Howell, John

    2010-09-13

    Current safeguards approaches used by the IAEA at gas centrifuge enrichment plants (GCEPs) need enhancement in order to verify declared low enriched uranium (LEU) production, detect undeclared LEU production and detect high enriched uranium (BEU) production with adequate probability using non destructive assay (NDA) techniques. At present inspectors use attended systems, systems needing the presence of an inspector for operation, during inspections to verify the mass and {sup 235}U enrichment of declared cylinders of uranium hexafluoride that are used in the process of enrichment at GCEPs. This paper contains an analysis of how possible improvements in unattended and attended NDA systems including process monitoring and possible on-site destructive analysis (DA) of samples could reduce the uncertainty of the inspector's measurements providing more effective and efficient IAEA GCEPs safeguards. We have also studied a few advanced safeguards systems that could be assembled for unattended operation and the level of performance needed from these systems to provide more effective safeguards. The analysis also considers how short notice random inspections, unannounced inspections (UIs), and the concept of information-driven inspections can affect probability of detection of the diversion of nuclear material when coupled to new GCEPs safeguards regimes augmented with unattended systems. We also explore the effects of system failures and operator tampering on meeting safeguards goals for quantity and timeliness and the measures needed to recover from such failures and anomalies.

  5. Two U.S. University Research Reactors to be Converted From Highly...

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

    ... the NNSA. NNSA Highly Enriched Uranium Removal Featured on The Rachel Maddow Show US, Russian Federation Sign Joint Statement on Reactor Conversion DOE to Remove 200 Metric Tons of ...

  6. 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.

  7. Production of highly-enriched 134Ba for a reference material for isotope dilution mass spectrometry measurements

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

    Horkley, J. J.; Carney, K. P.; Gantz, E. M.; Davies, J. E.; Lewis, R. R.; Crow, J. P.; Poole, C. A.; Grimes, T. S.; Giglio, J. J.

    2015-03-17

    Isotope dilution mass spectrometry (IDMS) is an analytical technique capable of providing accurate and precise quantitation of trace isotope abundance and assay providing measurement uncertainties below 1 %. To achieve these low uncertainties, the IDMS method ideally utilizes chemically pure “spike” solutions that consist of a single highly enriched isotope that is well-characterized relating to the abundance of companion isotopes and concentration in solution. To address a current demand for accurate 137Cs/137Ba ratio measurements for “age” determination of radioactive 137Cs sources, Idaho National Laboratory (INL) is producing enriched 134Ba isotopes that are tobe used for IDMS spikes to accurately determinemore » 137Ba accumulation from the decay of 137Cs. The final objective of this work it to provide a homogenous set of reference materials that the National Institute of Standards and Technology can certify as standard reference materials used for IDMS. The process that was developed at INL for the separation and isolation of Ba isotopes, chemical purification of the isotopes in solution, and the encapsulation of the materials will be described.« less

  8. Production of highly-enriched 134Ba for a reference material for isotope dilution mass spectrometry measurements

    SciTech Connect (OSTI)

    Horkley, J. J.; Carney, K. P.; Gantz, E. M.; Davies, J. E.; Lewis, R. R.; Crow, J. P.; Poole, C. A.; Grimes, T. S.; Giglio, J. J.

    2015-03-17

    Isotope dilution mass spectrometry (IDMS) is an analytical technique capable of providing accurate and precise quantitation of trace isotope abundance and assay providing measurement uncertainties below 1 %. To achieve these low uncertainties, the IDMS method ideally utilizes chemically pure “spike” solutions that consist of a single highly enriched isotope that is well-characterized relating to the abundance of companion isotopes and concentration in solution. To address a current demand for accurate 137Cs/137Ba ratio measurements for “age” determination of radioactive 137Cs sources, Idaho National Laboratory (INL) is producing enriched 134Ba isotopes that are tobe used for IDMS spikes to accurately determine 137Ba accumulation from the decay of 137Cs. The final objective of this work it to provide a homogenous set of reference materials that the National Institute of Standards and Technology can certify as standard reference materials used for IDMS. The process that was developed at INL for the separation and isolation of Ba isotopes, chemical purification of the isotopes in solution, and the encapsulation of the materials will be described.

  9. Production of highly-enriched 134Ba for a reference material for isotope dilution mass spectrometry measurements

    SciTech Connect (OSTI)

    J.J. Horkley; K.P E.M. Gantz; J.E. Davis; R.R. Lewis; J.P. Crow; C.A. Poole; T.S. Grimes; J.J. Giglio

    2015-03-01

    t Isotope dilution mass spectrometry (IDMS) is an analytical technique capable of providing accurate and precise quantitation of trace isotope abundance and assay providing measurement uncertainties below 1 %. To achieve these low uncertainties, the IDMS method ideally utilizes chemically pure spike solutions that consist of a single highly enriched isotope that is well-characterized relating to the abundance of companion isotopes and concentration in solution. To address a current demand for accurate 137Cs/137Ba ratio measurements for age determination of radioactive 137Cs sources, Idaho National Laboratory (INL) is producing enriched 134Ba isotopes that are tobe used for IDMS spikes to accurately determine 137Ba accumulation from the decay of 137Cs. The final objective of this work it to provide a homogenous set of reference materials that the National Institute of Standards and Technology can certify as standard reference materials used for IDMS. The process that was developed at INL for the separation and isolation of Ba isotopes, chemical purification of the isotopes in solution,

  10. Production of highly-enriched 134Ba for a reference material for isotope dilution mass spectrometry measurements

    SciTech Connect (OSTI)

    Horkley, J. J.; Carney, K. P.; Gantz, E. M.; Davies, J. E.; Lewis, R. R.; Crow, J. P.; Poole, C. A.; Grimes, T. S.; Giglio, J. J.

    2015-03-17

    Isotope dilution mass spectrometry (IDMS) is an analytical technique capable of providing accurate and precise quantitation of trace isotope abundance and assay providing measurement uncertainties below 1 %. To achieve these low uncertainties, the IDMS method ideally utilizes chemically pure spike solutions that consist of a single highly enriched isotope that is well-characterized relating to the abundance of companion isotopes and concentration in solution. To address a current demand for accurate 137Cs/137Ba ratio measurements for age determination of radioactive 137Cs sources, Idaho National Laboratory (INL) is producing enriched 134Ba isotopes that are tobe used for IDMS spikes to accurately determine 137Ba accumulation from the decay of 137Cs. The final objective of this work it to provide a homogenous set of reference materials that the National Institute of Standards and Technology can certify as standard reference materials used for IDMS. The process that was developed at INL for the separation and isolation of Ba isotopes, chemical purification of the isotopes in solution, and the encapsulation of the materials will be described.

  11. 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.

  12. Qualification and initial characterization of a high-purity 233U spike for use in uranium analyses

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

    Mathew, K. J.; Canaan, R. D.; Hexel, C.; Giaquinto, J.; Krichinsky, A. M.

    2015-08-20

    Several high-purity 233U items potentially useful as isotope dilution mass spectrometry standards for safeguards, non-proliferation, and nuclear forensics measurements are identified and rescued from downblending. By preserving the supply of 233U materials of different pedigree for use as source materials for certified reference materials (CRMs), it is ensured that the safeguards community has high quality uranium isotopic standards required for calibration of the analytical instruments. One of the items identified as a source material for a high-purity CRM is characterized for the uranium isotope-amount ratios using thermal ionization mass spectrometry (TIMS). Additional verification measurements on this material using quadrupole inductivelymore » coupled plasma mass spectrometry (ICPMS) are also performed. As a result, the comparison of the ICPMS uranium isotope-amount ratios with the TIMS data, with much smaller uncertainties, validated the ICPMS measurement practices. ICPMS is proposed for the initial screening of the purity of items in the rescue campaign.« less

  13. Qualification and initial characterization of a high-purity 233U spike for use in uranium analyses

    SciTech Connect (OSTI)

    Mathew, K. J.; Canaan, R. D.; Hexel, C.; Giaquinto, J.; Krichinsky, A. M.

    2015-08-20

    Several high-purity 233U items potentially useful as isotope dilution mass spectrometry standards for safeguards, non-proliferation, and nuclear forensics measurements are identified and rescued from downblending. By preserving the supply of 233U materials of different pedigree for use as source materials for certified reference materials (CRMs), it is ensured that the safeguards community has high quality uranium isotopic standards required for calibration of the analytical instruments. One of the items identified as a source material for a high-purity CRM is characterized for the uranium isotope-amount ratios using thermal ionization mass spectrometry (TIMS). Additional verification measurements on this material using quadrupole inductively coupled plasma mass spectrometry (ICPMS) are also performed. As a result, the comparison of the ICPMS uranium isotope-amount ratios with the TIMS data, with much smaller uncertainties, validated the ICPMS measurement practices. ICPMS is proposed for the initial screening of the purity of items in the rescue campaign.

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

    SciTech Connect (OSTI)

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

    2002-01-01

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

  15. Uranium Mining, Conversion, and Enrichment Industries

    Office of Environmental Management (EM)

    ... Down-blending of HEU NNSA contracts with WesDyne International for down-blending of HEU to ... Under the terms of the contract with WesDyne, DOE can use a mix of money and ...

  16. Safeguards Verification Measurements using Laser Ablation, Absorbance Ratio Spectrometry in Gaseous Centrifuge Enrichment Plants

    SciTech Connect (OSTI)

    Anheier, Norman C.; Cannon, Bret D.; Kulkarni, Gourihar R.; Munley, John T.; Nelson, Danny A.; Qiao, Hong; Phillips, Jon R.

    2012-07-17

    Laser Ablation Absorbance Ratio Spectrometry (LAARS) is a new verification measurement technology under development at the US Department of Energy (DOE) Pacific Northwest National Laboratory (PNNL). LAARS uses three lasers to ablate and then measure the relative isotopic abundance of uranium compounds. An ablation laser is tightly focused on uranium-bearing solids, producing a small atomic uranium vapor plume. Two collinear wavelength-tuned spectrometry lasers transit through the plume and the absorbance of U-235 and U-238 isotopes are measured to determine U-235 enrichment. The measurement is independent of chemical form and degree of dilution with nuisance dust and other materials. LAARS has high relative precision and detection limits approaching the femtogram range for U-235. The sample is scanned and assayed point-by-point at rates reaching 1 million measurements/hour, enabling LAARS to detect and analyze uranium in trace samples. The spectrometer is assembled using primarily commercially available components and features a compact design and automated analysis.Two specific gaseous centrifuge enrichment plant (GCEP) applications of the spectrometer are currently under development: 1) LAARS-Environmental Sampling (ES), which collects and analyzes aerosol particles for GCEP misuse detection and 2) LAARS-Destructive Assay (DA), which enables onsite enrichment DA sample collection and analysis for protracted diversion detection. The two applications propose game-changing technological advances in GCEP safeguards verification.

  17. Demonstration of High Performance in Layered Deuterium-Tritium Capsule Implosions in Uranium Hohlraums at the National Ignition Facility

    SciTech Connect (OSTI)

    Döppner, T.; Callahan, D. A.; Hurricane, O. A.; Hinkel, D. E.; Ma, T.; Park, H. -S.; Berzak Hopkins, L. F.; Casey, D. T.; Celliers, P. P.; Dewald, E. L.; Dittrich, T. R.; Haan, S.; Kritcher, A. L.; MacPhee, A.; Le Pape, S.; Pak, A.; Patel, P. K.; Springer, P. T.; Salmonson, J. D.; Tommasini, R.; Benedetti, L. R.; Bond, E.; Bradley, D. K.; Caggiano, J.; Church, J.; Dixit, S.; Edgell, D.; Edwards, M. J.; Fittinghoff, D. N.; Frenje, J.; Gatu Johnson, M.; Grim, G.; Hatarik, R.; Havre, M.; Herrmann, H.; Izumi, N.; Khan, S. F.; Kline, J. L.; Knauer, J.; Kyrala, G. A.; Landen, O. L.; Merrill, F. E.; Moody, J.; Moore, A. S.; Nikroo, A.; Ralph, J. E.; Remington, B. A.; Robey, H.; Sayre, D.; Schneider, M.; Streckert, H.; Town, R.; Turnbull, D.; Volegov, P. L.; Wan, A.; Widmann, K.; Wilde, C. H.; Yeamans, C.

    2015-07-28

    We report on the first layered deuterium-tritium (DT) capsule implosions indirectly driven by a “highfoot” laser pulse that were fielded in depleted uranium hohlraums at the National Ignition Facility. Recently, high-foot implosions have demonstrated improved resistance to ablation-front Rayleigh-Taylor instability induced mixing of ablator material into the DT hot spot [Hurricane et al., Nature (London) 506, 343 (2014)]. Uranium hohlraums provide a higher albedo and thus an increased drive equivalent to an additional 25 TW laser power at the peak of the drive compared to standard gold hohlraums leading to higher implosion velocity. Additionally, we observe an improved hot-spot shape closer to round which indicates enhanced drive from the waist. In contrast to findings in the National Ignition Campaign, now all of our highest performing experiments have been done in uranium hohlraums and achieved total yields approaching 1016 neutrons where more than 50% of the yield was due to additional heating of alpha particles stopping in the DT fuel.

  18. Demonstration of High Performance in Layered Deuterium-Tritium Capsule Implosions in Uranium Hohlraums at the National Ignition Facility

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

    Döppner, T.; Callahan, D. A.; Hurricane, O. A.; Hinkel, D. E.; Ma, T.; Park, H. -S.; Berzak Hopkins, L. F.; Casey, D. T.; Celliers, P. P.; Dewald, E. L.; et al

    2015-07-28

    We report on the first layered deuterium-tritium (DT) capsule implosions indirectly driven by a “highfoot” laser pulse that were fielded in depleted uranium hohlraums at the National Ignition Facility. Recently, high-foot implosions have demonstrated improved resistance to ablation-front Rayleigh-Taylor instability induced mixing of ablator material into the DT hot spot [Hurricane et al., Nature (London) 506, 343 (2014)]. Uranium hohlraums provide a higher albedo and thus an increased drive equivalent to an additional 25 TW laser power at the peak of the drive compared to standard gold hohlraums leading to higher implosion velocity. Additionally, we observe an improved hot-spot shapemore » closer to round which indicates enhanced drive from the waist. In contrast to findings in the National Ignition Campaign, now all of our highest performing experiments have been done in uranium hohlraums and achieved total yields approaching 1016 neutrons where more than 50% of the yield was due to additional heating of alpha particles stopping in the DT fuel.« less

  19. DOE Signs Advanced Enrichment Technology License and Facility Lease |

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

    Department of Energy Advanced Enrichment Technology License and Facility Lease DOE Signs Advanced Enrichment Technology License and Facility Lease December 8, 2006 - 9:34am Addthis Announces Agreements with USEC Enabling Deployment of Advanced Domestic Technology for Uranium Enrichment WASHINGTON, DC - U.S. Secretary of Energy Samuel W. Bodman today announced the signing of a lease agreement with the United States Enrichment Corporation, Inc. (USEC) for their use of the Department's gas

  20. Selective Extraction of Uranium from Liquid or Supercritical Carbon Dioxide

    SciTech Connect (OSTI)

    Farawila, Anne F.; O'Hara, Matthew J.; Wai, Chien M.; Taylor, Harry Z.; Liao, Yu-Jung

    2012-07-31

    Current liquid-liquid extraction processes used in recycling irradiated nuclear fuel rely on (1) strong nitric acid to dissolve uranium oxide fuel, and (2) the use of aliphatic hydrocarbons as a diluent in formulating the solvent used to extract uranium. The nitric acid dissolution process is not selective. It dissolves virtually the entire fuel meat which complicates the uranium extraction process. In addition, a solvent washing process is used to remove TBP degradation products, which adds complexity to the recycling plant and increases the overall plant footprint and cost. A liquid or supercritical carbon dioxide (l/sc -CO2) system was designed to mitigate these problems. Indeed, TBP nitric acid complexes are highly soluble in l/sc -CO2 and are capable of extracting uranium directly from UO2, UO3 and U3O8 powders. This eliminates the need for total acid dissolution of the irradiated fuel. Furthermore, since CO2 is easily recycled by evaporation at room temperature and pressure, it eliminates the complex solvent washing process. In this report, we demonstrate: (1) A reprocessing scheme starting with the selective extraction of uranium from solid uranium oxides into a TBP-HNO3 loaded Sc-CO2 phase, (2) Back extraction of uranium into an aqueous phase, and (3) Conversion of recovered purified uranium into uranium oxide. The purified uranium product from step 3 can be disposed of as low level waste, or mixed with enriched uranium for use in a reactor for another fuel cycle. After an introduction on the concept and properties of supercritical fluids, we first report the characterization of the different oxides used for this project. Our extraction system and our online monitoring capability using UV-Vis absorbance spectroscopy directly in sc-CO2 is then presented. Next, the uranium extraction efficiencies and kinetics is demonstrated for different oxides and under different physical and chemical conditions: l/sc -CO2 pressure and temperature, TBP/HNO3 complex used, reductant or complexant used for selectivity, and ionic liquids used as supportive media. To complete the extraction and recovery cycle, we then demonstrate uranium back extraction from the TBP loaded sc-CO2 phase into an aqueous phase and the characterization of the uranium complex formed at the end of this process. Another aspect of this project was to limit proliferation risks by either co-extracting uranium and plutonium, or by leaving plutonium behind by selectively extracting uranium. We report that the former is easily achieved, since plutonium is in the tetravalent or hexavalent oxidation state in the oxidizing environment created by the TBP-nitric acid complex, and is therefore co-extracted. The latter is more challenging, as a reductant or complexant to plutonium has to be used to selectively extract uranium. After undertaking experiments on different reducing or complexing systems (e.g., AcetoHydroxamic Acid (AHA), Fe(II), ascorbic acid), oxalic acid was chosen as it can complex tetravalent actinides (Pu, Np, Th) in the aqueous phase while allowing the extraction of hexavalent uranium in the sc-CO2 phase. Finally, we show results using an alternative media to commonly used aqueous phases: ionic liquids. We show the dissolution of uranium in ionic liquids and its extraction using sc-CO2 with and without the presence of AHA. The possible separation of trivalent actinides from uranium is also demonstrated in ionic liquids using neodymium as a surrogate and diglycolamides as the extractant.

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

    DOE Patents [OSTI]

    Travelli, A.

    1985-10-25

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

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

    DOE Patents [OSTI]

    Travelli, Armando

    1988-01-01

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

  3. Reaction of uranium oxides with chlorine and carbon or carbon monoxide to prepare uranium chlorides

    SciTech Connect (OSTI)

    Haas, P.A.; Lee, D.D.; Mailen, J.C.

    1991-11-01

    The preferred preparation concept of uranium metal for feed to an AVLIS uranium enrichment process requires preparation of uranium tetrachloride (UCI{sub 4}) by reacting uranium oxides (UO{sub 2}/UO{sub 3}) and chlorine (Cl{sub 2}) in a molten chloride salt medium. UO{sub 2} is a very stable metal oxide; thus, the chemical conversion requires both a chlorinating agent and a reducing agent that gives an oxide product which is much more stable than the corresponding chloride. Experimental studies in a quartz reactor of 4-cm ID have demonstrated the practically of some chemical flow sheets. Experimentation has illustrated a sequence of results concerning the chemical flow sheets. Tests with a graphite block at 850{degrees}C demonstrated rapid reactions of Cl{sub 2} and evolution of carbon dioxide (CO{sub 2}) as a product. Use of carbon monoxide (CO) as the reducing agent also gave rapid reactions of Cl{sub 2} and formation of CO{sub 2} at lower temperatures, but the reduction reactions were slower than the chlorinations. Carbon powder in the molten salt melt gave higher rates of reduction and better steady state utilization of Cl{sub 2}. Addition of UO{sub 2} feed while chlorination was in progress greatly improved the operation by avoiding the plugging effects from high UO{sub 2} concentrations and the poor Cl{sub 2} utilizations from low UO{sub 2} concentrations. An UO{sub 3} feed gave undesirable effects while a feed of UO{sub 2}-C spheres was excellent. The UO{sub 2}-C spheres also gave good rates of reaction as a fixed bed without any molten chloride salt. Results with a larger reactor and a bottom condenser for volatilized uranium show collection of condensed uranium chlorides as a loose powder and chlorine utilizations of 95--98% at high feed rates. 14 refs., 7 figs., 14 tabs.

  4. Thermal breeder fuel enrichment zoning

    DOE Patents [OSTI]

    Capossela, Harry J.; Dwyer, Joseph R.; Luce, Robert G.; McCoy, Daniel F.; Merriman, Floyd C.

    1992-01-01

    A method and apparatus for improving the performance of a thermal breeder reactor having regions of higher than average moderator concentration are disclosed. The fuel modules of the reactor core contain at least two different types of fuel elements, a high enrichment fuel element and a low enrichment fuel element. The two types of fuel elements are arranged in the fuel module with the low enrichment fuel elements located between the high moderator regions and the high enrichment fuel elements. Preferably, shim rods made of a fertile material are provided in selective regions for controlling the reactivity of the reactor by movement of the shim rods into and out of the reactor core. The moderation of neutrons adjacent the high enrichment fuel elements is preferably minimized as by reducing the spacing of the high enrichment fuel elements and/or using a moderator having a reduced moderating effect.

  5. Defect structures induced by high-energy displacement cascades in ? uranium

    SciTech Connect (OSTI)

    Miao, Yinbin; Beeler, Benjamin; Deo, Chaitanya; Baskes, Michael I.; Okuniewski, Maria A.; Stubbins, James F.

    2015-01-01

    Displacement cascade simulations were conducted for the c uranium system based on molecular dynamics. A recently developed modified embedded atom method (MEAM) potential was employed to replicate the atomic interactions while an embedded atom method (EAM) potential was adopted to help characterize the defect structures induced by the displacement cascades. The atomic displacement process was studied by providing primary knock-on atoms (PKAs) with kinetic energies from 1 keV to 50 keV. The influence of the PKA incident direction was examined. The defect structures were analyzed after the systems were fully relaxed. The states of the self-interstitial atoms (SIAs) were categorized into various types of dumbbells, the crowdion, and the octahedral interstitial. The voids were determined to have a polyhedral shape with {110} facets. The size distribution of the voids was also obtained. The results of this study not only expand the knowledge of the microstructural evolution in irradiated c uranium, but also provide valuable references for the radiation-induced defects in uranium alloy fuels.

  6. Enrichment Assay Methods for a UF6 Cylinder Verification Station

    SciTech Connect (OSTI)

    Smith, Leon E.; Jordan, David V.; Misner, Alex C.; Mace, Emily K.; Orton, Christopher R.

    2010-11-30

    International Atomic Energy Agency (IAEA) inspectors currently perform periodic inspections at uranium enrichment plants to verify UF6 cylinder enrichment declarations. Measurements are typically performed with handheld high-resolution sensors on a sampling of cylinders taken to be representative of the facility’s entire cylinder inventory. These enrichment assay methods interrogate only a small fraction of the total cylinder volume, and are time-consuming and expensive to execute for inspectors. Pacific Northwest National Laboratory (PNNL) is developing an unattended measurement system capable of automated enrichment measurements over the full volume of Type 30B and Type 48 cylinders. This Integrated Cylinder Verification System (ICVS) could be located at key measurement points to positively identify each cylinder, measure its mass and enrichment, store the collected data in a secure database, and maintain continuity of knowledge on measured cylinders until IAEA inspector arrival. The focus of this paper is the development of nondestructive assay (NDA) methods that combine “traditional” enrichment signatures (e.g. 185-keV emission from U-235) and more-penetrating “non-traditional” signatures (e.g. high-energy neutron-induced gamma rays spawned primarily from U-234 alpha emission) collected by medium-resolution gamma-ray spectrometers (i.e. sodium iodide or lanthanum bromide). The potential of these NDA methods for the automated assay of feed, tail and product cylinders is explored through MCNP modeling and with field measurements on a cylinder population ranging from 0.2% to 5% in U-235 enrichment.

  7. Energy Department Selects Global Laser Enrichment for Future Operations at

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

    Paducah Site | Department of Energy Global Laser Enrichment for Future Operations at Paducah Site Energy Department Selects Global Laser Enrichment for Future Operations at Paducah Site November 27, 2013 - 12:00pm Addthis Workers inspect cylinders containing depleted uranium hexafluoride. Workers inspect cylinders containing depleted uranium hexafluoride. Media Contact (202) 586-4940 Washington, D.C. - The U.S. Department of Energy announced today that it will open negotiations with Global

  8. 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.

  9. 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)

  10. 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.

  11. Fabrication of Cerium Oxide and Uranium Oxide Microspheres for Space Nuclear Power Applications

    SciTech Connect (OSTI)

    Jeffrey A. Katalenich; Michael R. Hartman; Robert C. O'Brien

    2013-02-01

    Cerium oxide and uranium oxide microspheres are being produced via an internal gelation sol-gel method to investigate alternative fabrication routes for space nuclear fuels. Depleted uranium and non-radioactive cerium are being utilized as surrogates for plutonium-238 (Pu-238) used in radioisotope thermoelectric generators and for enriched uranium required by nuclear thermal rockets. While current methods used to produce Pu-238 fuels at Los Alamos National Laboratory (LANL) involve the generation of fine powders that pose a respiratory hazard and have a propensity to contaminate glove boxes, the sol-gel route allows for the generation of oxide microsphere fuels through an aqueous route. The sol-gel method does not generate fine powders and may require fewer processing steps than the LANL method with less operator handling. High-quality cerium dioxide microspheres have been fabricated in the desired size range and equipment is being prepared to establish a uranium dioxide microsphere production capability.

  12. A Robust Infrastructure Design for Gas Centrifuge Enrichment Plant Unattended Online Enrichment Monitoring

    SciTech Connect (OSTI)

    Younkin, James R; Rowe, Nathan C; Garner, James R

    2012-01-01

    An online enrichment monitor (OLEM) is being developed to continuously measure the relative isotopic composition of UF6 in the unit header pipes of a gas centrifuge enrichment plant (GCEP). From a safeguards perspective, OLEM will provide early detection of a facility being misused for production of highly enriched uranium. OLEM may also reduce the number of samples collected for destructive assay and if coupled with load cell monitoring can provide isotope mass balance verification. The OLEM design includes power and network connections for continuous monitoring of the UF6 enrichment and state of health of the instrument. Monitoring the enrichment on all header pipes at a typical GCEP could require OLEM detectors on each of the product, tails, and feed header pipes. If there are eight process units, up to 24 detectors may be required at a modern GCEP. Distant locations, harsh industrial environments, and safeguards continuity of knowledge requirements all place certain demands on the network robustness and power reliability. This paper describes the infrastructure and architecture of an OLEM system based on OLEM collection nodes on the unit header pipes and power and network support nodes for groupings of the collection nodes. A redundant, self-healing communications network, distributed backup power, and a secure communications methodology. Two candidate technologies being considered for secure communications are the Object Linking and Embedding for Process Control Unified Architecture cross-platform, service-oriented architecture model for process control communications and the emerging IAEA Real-time And INtegrated STream-Oriented Remote Monitoring (RAINSTORM) framework to provide the common secure communication infrastructure for remote, unattended monitoring systems. The proposed infrastructure design offers modular, commercial components, plug-and-play extensibility for GCEP deployments, and is intended to meet the guidelines and requirements for unattended and remotely monitored safeguards systems.

  13. In Situ Synthesis of Uranium Carbide and its High Temperature Cubic Phase

    SciTech Connect (OSTI)

    Reiche, Helmut Matthias; Vogel, Sven C.

    2015-03-25

    New in situ data for the U-C system are presented, with the goal of improving knowledge of the phase diagram to enable production of new ceramic fuels. The none quenchable, cubic, ?-phase, which in turn is fundamental to computational methods, was identified. Rich datasets of the formation synthesis of uranium carbide yield kinetics data which allow the benchmarking of modeling, thermodynamic parameters etc. The order-disorder transition (carbon sublattice melting) was observed due to equal sensitivity of neutrons to both elements. This dynamic has not been accurately described in some recent simulation-based publications.

  14. 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.

  15. Analysis of the effectiveness of gas centrifuge enrichment plants advanced safeguards

    SciTech Connect (OSTI)

    Boyer, Brian David; Erpenbeck, Heather H; Miller, Karen A; Swinjoe, Martyn T; Ianakiev, Kiril D; Marlow, Johnna B

    2010-01-01

    Current safeguards approaches used by the International Atomic Energy Agency (IAEA) at gas centrifuge enrichment plants (GCEPs) need enhancement in order to verify declared low-enriched uranium (LEU) production, detect undeclared LEU production and detect highly enriched uranium (HEU) production with adequate detection probability using non destructive assay (NDA) techniques. At present inspectors use attended systems, systems needing the presence of an inspector for operation, during inspections to verify the mass and 235U enrichment of declared UF6 containers used in the process of enrichment at GCEPs. This paper contains an analysis of possible improvements in unattended and attended NDA systems including process monitoring and possible on-site destructive assay (DA) of samples that could reduce the uncertainty of the inspector's measurements. These improvements could reduce the difference between the operator's and inspector's measurements providing more effective and efficient IAEA GCEPs safeguards. We also explore how a few advanced safeguards systems could be assembled for unattended operation. The analysis will focus on how unannounced inspections (UIs), and the concept of information-driven inspections (IDS) can affect probability of detection of the diversion of nuclear materials when coupled to new GCEPs safeguards regimes augmented with unattended systems.

  16. 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

  17. IMPACT OF URANIUM AND THORIUM ON HIGH TIO2 CONCENTRATION NUCLEAR WASTE GLASSES

    SciTech Connect (OSTI)

    Fox, K.; Edwards, T.

    2012-01-11

    This study focused on the potential impacts of the addition of Crystalline Silicotitanate (CST) and Monosodium Titanate (MST) from the Small Column Ion Exchange (SCIX) process on the Defense Waste Processing Facility (DWPF) glass waste form and the applicability of the DWPF process control models. MST from the Salt Waste Processing Facility (SWPF) is also considered in the study. The KT08-series of glasses was designed to evaluate any impacts of the inclusion of uranium and thorium in glasses containing the SCIX components. All but one of the study glasses were found to be amorphous by X-ray diffraction (XRD). One of the slowly cooled glasses contained a small amount of trevorite, which is typically found in DWPF-type glasses and had no practical impact on the durability of the glass. The measured Product Consistency Test (PCT) responses for the study glasses and the viscosities of the glasses were well predicted by the current DWPF models. No unexpected issues were encountered when uranium and thorium were added to the glasses with SCIX components.

  18. Safeguards Verification Measurements using Laser Ablation, Absorbance Ratio Spectrometry in Gaseous Centrifuge Enrichment Plants

    SciTech Connect (OSTI)

    Anheier, Norman C.; Cannon, Bret D.; Qiao, Hong; Phillips, Jon R.

    2012-07-01

    Laser Ablation Absorbance Ratio Spectrometry (LAARS) is a new verification measurement technology under development at the US Department of Energy’s (DOE) Pacific Northwest National Laboratory (PNNL). LAARS uses three lasers to ablate and then measure the relative isotopic abundance of uranium compounds. An ablation laser is tightly focused on uranium-bearing solids producing a small plume containing uranium atoms. Two collinear wavelength-tuned spectrometry lasers transit through the plume and the absorbance of U-235 and U-238 isotopes are measured to determine U-235 enrichment. The measurement has high relative precision and detection limits approaching the femtogram range for uranium. It is independent of chemical form and degree of dilution with nuisance dust and other materials. High speed sample scanning and pinpoint characterization allow measurements on millions of particles/hour to detect and analyze the enrichment of trace uranium in samples. The spectrometer is assembled using commercially available components at comparatively low cost, and features a compact and low power design. Future designs can be engineered for reliable, autonomous deployment within an industrial plant environment. Two specific applications of the spectrometer are under development: 1) automated unattended aerosol sampling and analysis and 2) on-site small sample destructive assay measurement. The two applications propose game-changing technological advances in gaseous centrifuge enrichment plant (GCEP) safeguards verification. The aerosol measurement instrument, LAARS-environmental sampling (ES), collects aerosol particles from the plant environment in a purpose-built rotating drum impactor and then uses LAARS-ES to quickly scan the surface of the impactor to measure the enrichments of the captured particles. The current approach to plant misuse detection involves swipe sampling and offsite analysis. Though this approach is very robust it generally requires several months to obtain results from a given sample collection. The destructive assay instrument, LAARS-destructive assay (DA), uses a simple purpose-built fixture with a sampling planchet to collect adsorbed UF6 gas from a cylinder valve or from a process line tap or pigtail. A portable LAARS-DA instrument scans the microgram quantity of uranium collected on the planchet and the assay of the uranium is measured to ~0.15% relative precision. Currently, destructive assay samples for bias defect measurements are collected in small sample cylinders for offsite mass spectrometry measurement.

  19. 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...

  20. Portable NDA Equipment for Enrichment Measurements in the HEU Transparency Program

    SciTech Connect (OSTI)

    Decman, D J; Bandong, B B; Wong, J L; Valentine, J D; Luke, S J

    2008-06-02

    The Highly Enriched Uranium (HEU) Transparency Program has used portable nondestructive assay (NDA) equipment to measure the {sup 235}U enrichment of material subject to the transparency agreement since 1997. The equipment is based on the 'enrichment meter' method and uses low-resolution sodium iodide (NaI(Tl)) detectors. Although systems using high-purity germanium (HPGe) detectors can produce more accurate results we have found that the results with NaI(Tl) detectors are quite adequate for the requirements of the transparency agreement. This paper will describe the details of the equipment's operation, calibration, testing, and deployment in Russia. We will also provide a comparison of the units originally deployed in 1997 with the upgraded systems that were deployed in 2003.

  1. 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)

  2. Enrichment Assay Methods Development for the Integrated Cylinder Verification System

    SciTech Connect (OSTI)

    Smith, Leon E.; Misner, Alex C.; Hatchell, Brian K.; Curtis, Michael M.

    2009-10-22

    International Atomic Energy Agency (IAEA) inspectors currently perform periodic inspections at uranium enrichment plants to verify UF6 cylinder enrichment declarations. Measurements are typically performed with handheld high-resolution sensors on a sampling of cylinders taken to be representative of the facility's entire product-cylinder inventory. Pacific Northwest National Laboratory (PNNL) is developing a concept to automate the verification of enrichment plant cylinders to enable 100 percent product-cylinder verification and potentially, mass-balance calculations on the facility as a whole (by also measuring feed and tails cylinders). The Integrated Cylinder Verification System (ICVS) could be located at key measurement points to positively identify each cylinder, measure its mass and enrichment, store the collected data in a secure database, and maintain continuity of knowledge on measured cylinders until IAEA inspector arrival. The three main objectives of this FY09 project are summarized here and described in more detail in the report: (1) Develop a preliminary design for a prototype NDA system, (2) Refine PNNL's MCNP models of the NDA system, and (3) Procure and test key pulse-processing components. Progress against these tasks to date, and next steps, are discussed.

  3. 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 =

  4. Defining the needs for gas centrifuge enrichment plants advanced safeguards

    SciTech Connect (OSTI)

    Boyer, Brian David; Erpenbeck, Heather H; Miller, Karen A; Swinhoe, Martyn T; Ianakiev, Kiril; Marlow, Johnna B

    2010-04-05

    Current safeguards approaches used by the International Atomic Energy Agency (IAEA) at gas centrifuge enrichment plants (GCEPs) need enhancement in order to verify declared low-enriched (LEU) production, detect undeclared LEU production and detect highly enriched uranium (HEU) production with adequate detection probability using nondestructive assay (NDA) techniques. At present inspectors use attended systems, systems needing the presence of an inspector for operation, during inspections to verify the mass and {sup 235}U enrichment of declared UF{sub 6} containers used in the process of enrichment at GCEPs. In verifying declared LEU production, the inspectors also take samples for off-site destructive assay (DA) which provide accurate data, with 0.1% to 0.5% measurement uncertainty, on the enrichment of the UF{sub 6} feed, tails, and product. However, taking samples of UF{sub 6} for off-site analysis is a much more labor and resource intensive exercise for the operator and inspector. Furthermore, the operator must ship the samples off-site to the IAEA laboratory which delays the timeliness of results and interruptions to the continuity of knowledge (CofK) of the samples during their storage and transit. This paper contains an analysis of possible improvements in unattended and attended NDA systems such as process monitoring and possible on-site analysis of DA samples that could reduce the uncertainty of the inspector's measurements and provide more effective and efficient IAEA GCEPs safeguards. We also introduce examples advanced safeguards systems that could be assembled for unattended operation.

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

    SciTech Connect (OSTI)

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

    2012-12-15

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

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

    SciTech Connect (OSTI)

    1995-12-01

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

  7. Determination of the 235U Mass and Enrichment within Small UF6 Cylinders via a Neutron Coincidence Well Counting System

    SciTech Connect (OSTI)

    McElroy, Robert Dennis; Croft, Dr. Stephen; Young, Brian M; Venkataraman, Ram

    2011-01-01

    The construction of three new uranium enrichment facilities in the United States has sparked renewed interest in the development and enhancement of methods to determine the enrichment and fissile mass content of UF6 cylinders. We describe the design and examine the expected performance of a UF6 bottle counter developed for the assay of Type 5A cylinders. The counter, as designed and subsequently constructed, is a tall passive neutron well counter with a clam-shell configuration and graphite end plugs operated in fast neutron mode. Factory performance against expectation is described. The relatively high detection efficiency and effectively 4 detection geometry provide a near-ideal measurement configuration, making the UF6 bottle counter a valuable tool for the evaluation of the neutron coincidence approach to UF6 cylinder assay. The impacts of non-uniform filling, voids, enrichment, and mixed enrichments are examined

  8. Proceedings of the 1990 International Meeting on Reduced Enrichment for Research and Test Reactors

    SciTech Connect (OSTI)

    Not Available

    1993-07-01

    The global effort to reduce, and possibly, eliminate the international traffic in highly-enriched uranium caused by its use in research reactors requires extensive cooperation and free exchange of information among all participants. To foster this free exchange of information, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the thirteenth of a series which began in 1978. The common effort brought together, past, a large number of specialists from many countries. On hundred twenty-three participants from 26 countries, including scientists, reactor operators, and personnel from commercial fuel suppliers, research centers, and government organizations, convened in Newport, Rhode Island to discuss their results, their activities, and their plans relative to converting research reactors to low-enriched fuels. As more and more reactors convert to the use of low-enriched uranium, the emphasis of our effort has begun to shift from research and development to tasks more directly related to implementation of the new fuels and technologies that have been developed, and to refinements of those fuels and technologies. It is appropriate, for this reason, that the emphasis of this meeting was placed on safety and on conversion experiences. This individual papers in this report have been cataloged separately.

  9. 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.

  10. 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.

  11. 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. ...

  12. 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.

  13. 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

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

    SciTech Connect (OSTI)

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

    1980-10-01

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

  15. Modeling and Simulations for the High Flux Isotope Reactor Cycle 400

    Office of Scientific and Technical Information (OSTI)

    (Technical Report) | SciTech Connect Modeling and Simulations for the High Flux Isotope Reactor Cycle 400 Citation Details In-Document Search Title: Modeling and Simulations for the High Flux Isotope Reactor Cycle 400 A concerted effort over the past few years has been focused on enhancing the core model for the High Flux Isotope Reactor (HFIR), as part of a comprehensive study for HFIR conversion from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel. At this time, the core

  16. Microsoft Word - L15 01-22 Uranium Tranfers

    Energy Savers [EERE]

    To: Office of Nuclear Energy Department of Energy 1000 Independence Ave., SW Washington, DC 20585 From: Nan Swift Federal Affairs Manager National Taxpayers Union 108 N. Alfred Street Alexandria, VA 22314 Subject: Request for Information: Excess Uranium Management: Effects of DOE Transfers of Excess Uranium on Domestic Uranium Mining, Conversion, and Enrichment Industries To whom it may concern: On behalf of the members of the National Taxpayers Union (NTU), I write to express our concerns

  17. Uranium Track Team | Y-12 National Security Complex

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

    Track Team Uranium Track Team Posted: July 22, 2013 - 3:31pm | Y-12 Report | Volume 10, Issue 1 | 2013 Y-12's inventory and accounting processes for enriched uranium are more meticulous than for any other metal. "We have to track every gram of uranium-235," said Amy Wilson of Nuclear Materials Control and Accountability. Accordingly, even the smallest byproducts from the manufacturing processes, as well as piece parts from disassembly and dismantlement activities, are collected and

  18. Isotope Enrichment Detection by Laser Ablation - Laser Absorption Spectrometry: Automated Environmental Sampling and Laser-Based Analysis for HEU Detection

    SciTech Connect (OSTI)

    Anheier, Norman C.; Bushaw, Bruce A.

    2010-01-01

    The global expansion of nuclear power, and consequently the uranium enrichment industry, requires the development of new safeguards technology to mitigate proliferation risks. Current enrichment monitoring instruments exist that provide only yes/no detection of highly enriched uranium (HEU) production. More accurate accountancy measurements are typically restricted to gamma-ray and weight measurements taken in cylinder storage yards. Analysis of environmental and cylinder content samples have much higher effectiveness, but this approach requires onsite sampling, shipping, and time-consuming laboratory analysis and reporting. Given that large modern gaseous centrifuge enrichment plants (GCEPs) can quickly produce a significant quantity (SQ ) of HEU, these limitations in verification suggest the need for more timely detection of potential facility misuse. The Pacific Northwest National Laboratory (PNNL) is developing an unattended safeguards instrument concept, combining continuous aerosol particulate collection with uranium isotope assay, to provide timely analysis of enrichment levels within low enriched uranium facilities. This approach is based on laser vaporization of aerosol particulate samples, followed by wavelength tuned laser diode spectroscopy to characterize the uranium isotopic ratio through subtle differences in atomic absorption wavelengths. Environmental sampling (ES) media from an integrated aerosol collector is introduced into a small, reduced pressure chamber, where a focused pulsed laser vaporizes material from a 10 to 20-m diameter spot of the surface of the sampling media. The plume of ejected material begins as high-temperature plasma that yields ions and atoms, as well as molecules and molecular ions. We concentrate on the plume of atomic vapor that remains after the plasma has expanded and then cooled by the surrounding cover gas. Tunable diode lasers are directed through this plume and each isotope is detected by monitoring absorbance signals on a shot-to-shot basis. The media is translated by a micron resolution scanning system, allowing the isotope analysis to cover the entire sample surface. We also report, to the best of our knowledge, the first demonstration of laser-based isotopic measurements on individual micron-sized particles that are minor target components in a much larger heterogeneous mix of background particles. This composition is consistent with swipe and environmental aerosol samples typically collected for safeguards ES purposes. Single-shot detection sensitivity approaching the femtogram range and relative isotope abundance uncertainty better than 10% has been demonstrated using gadolinium isotopes as surrogate materials.

  19. High Field Magnetization measurements of uranium dioxide single crystals (P08358- E003-PF)

    SciTech Connect (OSTI)

    Gofryk, K.; Harrison, N.; Jaime, M.

    2014-12-01

    Our preliminary high field magnetic measurements of UO2 are consistent with a complex nature of the magnetic ordering in this material, compatible with the previously proposed non-collinear 3-k magnetic structure. Further extensive magnetic studies on well-oriented (<100 > and <111>) UO2 crystals are planned to address the puzzling behavior of UO2 in both antiferromagnetic and paramagnetic states at high fields.

  20. Uranium in geologic fluids: Estimates of standard partial molal properties, oxidation potentials, and hydrolysis constants at high temperatures and pressures

    SciTech Connect (OSTI)

    Shock, E.L.; Sassani, D.C.; Betz, H.

    1997-10-01

    Theoretical methods are used with the available experimental data to provide estimates of parameters for the revised-HKF equations of state for aqueous uranium species. These parameters are used with standard state thermodynamic data at 25{degrees}C and 1 bar to calculate equilibrium constants for redox reactions among the four most common oxidation states of uranium (U(III), U(IV), U(V), and U(VI)), and their hydrolysis reactions at temperatures to 1000{degrees}C and pressures to 5 kb. A total of nineteen aqueous uranium species are included. The predicted equilibrium constants are used to construct oxidation potential-pH diagrams at elevated temperatures and pressures and to calculate the solubilities of uraninite as functions of temperature and pH, which are compared to experimental data. Oxidation potential-pH diagrams illustrate the relative stabilities of aqueous uranium species and indicate that U(IV) and U(VI) species predominate in aqueous solution in the U-O-H system. Increasing temperature stabilizes U(VI) and U(III) species relative to U(IV) species, but U(IV) species dominate at oxidation states consistent and mineral-buffer assemblages and near-neutral pH. At low pH, U(VI) is stabilized relative to U(IV) suggesting that uranium transport in hydrothermal systems requires either acidic solutions or potent complexes of U(IV). 40 refs., 15 figs., 3 tabs.

  1. The Effect of U-234 Content on the Neutronic Behavior of Uranium Systems

    SciTech Connect (OSTI)

    Busch, Robert D.; Bledsoe, Keith C

    2011-01-01

    When analyzing uranium systems, the usual rule of thumb is to ignore the U-234 by assuming that it behaves neutronically like U-238. Thus for uranium systems, the uranium is evaluated as U-235 with everything else being U-238. The absorption cross section of U-234 is indeed qualitatively very similar to that of U-238. However, thermal absorption cross section of U-234 is about 100 times that of U-238. At low U-235 enrichments, the amount of U-234 is quite small so the impact of assuming it is U-238 is minimal. However, at high enrichments, the relative ratio of U-234 to U-238 is quite large (maybe as much as 1 to 5). Thus, one would expect that some effect of using the rule of thumb might be seen in higher enriched systems. Analyses were performed on three uranium systems from the set of Benchmarks [1]. Although the benchmarks are adequately characterized as to the U-234 content, often, materials used in processing are not as well characterized. This issue may become more important with the advent of laser enrichment processes, which have little or no effect on the U-234 content. Analytical results based on the relationship of U-234 activity to that of U-235 have shown good predictive capability but with large variability in the uncertainties [2]. Rucker and Johnson noted that the actual isotopics vary with enrichment, design of the enrichment cascade, composition of the feed material, and on blending of enrichments so there is considerable uncertainty in the use of models to determine isotopics. Thus, it is important for criticality personnel to understand the effects of variation of U-234 content in fissile systems and the impact of different modeling assumptions in handling the U-234. Analyses were done on LEU, IEU and HEU benchmarks from the International Handbook. These indicate that the effect of ignoring U-234 in HEU metal systems is non-conservative while it seems to be conservative for HEU solution systems. The magnitude of change in k-effective was as high as 0.4%, which has implications on selection of administrative margins and the determination of the upper subcriticality limit.

  2. Description of the Portsmouth Gas Centrifuge Enrichment Plant

    SciTech Connect (OSTI)

    Arthur, W.B.

    1980-12-16

    The Portsmouth Gas Centrifuge Enrichment Plant (GCEP) will be located at the site of the Portsmouth Gaseous Diffusion Plant in Piketon, Ohio. The purpose of the facility is to provide enriching services for the production of low assay enriched uranium for civilian nuclear power reactors. The construction and operation of the GCEP is administered by the US Department of Energy. The facility will be operated under contract from the US Government. Control of the GCEP rests solely with the US Government, which holds and controls access to the technology. Construction of GCEP is expected to be completed in the mid-1990's. Many facility design and operating procedures are subject to change. Nonetheless, the design described in this report does reflect current thinking. Descriptions of the general facility and major buildings such as the process buildings, feed and withdrawal building, cylinder storage and transfer, recycle/assembly building, and a summary of the centrifuge uranium enriching process are provided in this report.

  3. Method and apparatus for measuring enrichment of UF6

    DOE Patents [OSTI]

    Hill, Thomas Roy; Ianakiev, Kiril Dimitrov

    2011-06-07

    A system and method are disclosed for determining the enrichment of .sup.235U in Uranium Hexafluoride (UF6) utilizing synthesized X-rays which are directed at a container test zone containing a sample of UF6. A detector placed behind the container test zone then detects and counts the X-rays which pass through the container and the UF6. In order to determine the portion of the attenuation due to the UF6 gas alone, this count rate may then be compared to a calibration count rate of X-rays passing through a calibration test zone which contains a vacuum, the test zone having experienced substantially similar environmental conditions as the actual test zone. Alternatively, X-rays of two differing energy levels may be alternately directed at the container, where either the container or the UF6 has a high sensitivity to the difference in the energy levels, and the other having a low sensitivity.

  4. Spin-lattice coupling in uranium dioxide probed by magnetostriction measurements at high magnetic fields (P08358-E001-PF)

    SciTech Connect (OSTI)

    Gofryk, K.; Jaime, M.

    2014-12-01

    Our preliminary magnetostriction measurements have already shown a strong interplay of lattice dynamic and magnetism in both antiferromagnetic and paramagnetic states, and give unambiguous evidence of strong spin- phonon coupling in uranium dioxide. Further studies are planned to address the puzzling behavior of UO2 in magnetic and paramagnetic states and details of the spin-phonon coupling.

  5. 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

  6. 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

  7. 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.

  8. Hybrid interferometric/dispersive atomic spectroscopy of laser-induced uranium plasma

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

    Morgan, Phyllis K.; Scott, Jill R.; Jovanovic, Igor

    2015-12-19

    An established optical emission spectroscopy technique, laser-induced breakdown spectroscopy (LIBS), holds promise for detection and rapid analysis of elements relevant for nuclear safeguards, nonproliferation, and nuclear power, including the measurement of isotope ratios. One such important application of LIBS is the measurement of uranium enrichment (235U/238U), which requires high spectral resolution (e.g., 25 pm for the 424.4 nm U II line). High-resolution dispersive spectrometers necessary for such measurements are typically bulky and expensive. We demonstrate the use of an alternative measurement approach, which is based on an inexpensive and compact Fabry–Perot etalon integrated with a low to moderate resolution Czerny–Turnermore » spectrometer, to achieve the resolution needed for isotope selectivity of LIBS of uranium in ambient air. Furthermore, spectral line widths of ~ 10 pm have been measured at a center wavelength 424.437 nm, clearly discriminating the natural from the highly enriched uranium.« less

  9. Paducah Plant Begins Enrichment Operations after Five Parties Strike Agreement

    Broader source: Energy.gov [DOE]

    PADUCAH, Ky. – On this Friday, June 1, the first of about 1,100 14-ton cylinders filled with depleted uranium tails will be re-fed into the Paducah Gaseous Diffusion Plant, officially marking the start of a project that will extend enrichment operations there for a year.

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

    SciTech Connect (OSTI)

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

    2007-07-01

    Lowry Air Force Base (Lowry) was closed in September 1994 as part of the Base Realignment and Closure (BRAC) program and the base was transferred to the Lowry Redevelopment Authority in 1995. As part of the due diligence activities conducted by the Air Force, a series of remedial investigations were conducted across the base. A closed waste landfill, designated Operable Unit 2 (OU 2), was initially assessed in a 1990 Remedial Investigation (RI; [1]). A Supplemental Remedial Investigation was conducted in 1995 [2] and additional studies were conducted in a 1998 Focused Feasibility Study. [3] The three studies indicated that gross alpha, gross beta, and uranium concentrations were consistently above regulatory standards and that there were detections of low concentrations other radionuclides. Results from previous investigations at OU 2 have shown elevated gross alpha, gross beta, and uranium concentrations in groundwater, surface water, and sediments. The US Air Force has sought to understand the provenance of these radionuclides in order to determine if they could be due to leachates from buried radioactive materials within the landfill or whether they are naturally-occurring. The Air Force and regulators agreed to use a one-year monitoring and sampling program to seek to explain the origins of the radionuclides. Over the course of the one-year program, dissolved uranium levels greater than the 30 {mu}g/L Maximum Contaminant Level (MCL) were consistently found in both up-gradient and down-gradient wells at OU 2. Elevated Gross Alpha and Gross Beta measurements that were observed during prior investigations and confirmed during the LTM were found to correlate with high dissolved uranium content in groundwater. If Gross Alpha values are corrected to exclude uranium and radon contributions in accordance with US EPA guidance, then the 15 pCi/L gross alpha level is not exceeded. The large dataset also allowed development of gross alpha to total uranium correlation factors so that gross alpha action levels can be applied to future long-term landfill monitoring to track radiological conditions at lower cost. Ratios of isotopic uranium results were calculated to test whether the elevated uranium displayed signatures indicative of military use. Results of all ratio testing strongly supports the conclusion that the uranium found in groundwater, surface water, and sediment at OU 2 is naturally-occurring and has not undergone anthropogenic enrichment or processing. U-234:U-238 ratios also show that a disequilibrium state, i.e., ratio greater than 1, exists throughout OU 2 which is indicative of long-term aqueous transport in aged aquifers. These results all support the conclusion that the elevated uranium observed at OU 2 is due to the high concentrations in the regional watershed. Based on the results of this monitoring program, we concluded that the elevated uranium concentrations measured in OU 2 groundwater, surface water, and sediment are due to the naturally-occurring uranium content of the regional watershed and are not the result of waste burials in the former landfill. Several lines of evidence indicate that natural uranium has been naturally concentrated beneath OU 2 in the geologic past and the higher of uranium concentrations in down-gradient wells is the result of geochemical processes and not the result of a uranium ore disposal. These results therefore provide the data necessary to support radiological closure of OU 2. (authors)

  11. Enriching stable isotopes: Alternative use for Urenco technology

    SciTech Connect (OSTI)

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

    1996-12-31

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

  12. 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.

  13. Status of the RERTR (Reduced Enrichment Research and Test Reactor) Program

    SciTech Connect (OSTI)

    Travelli, A.

    1988-01-01

    The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. After a brief summary of the results which the RERTR Program, in collaboration with its many international partners, had achieved by the end of 1987, the major events, findings and activities of 1988 are reviewed. The US Nuclear Regulatory Commission issued a formal and generic approval of the use of U3Si2-Al dispersion fuel in research and test reactors, with densities up to 4.8 g U/cmT. New significant findings from postirradiation examinations, from ion-beam irradiations, and from analytical modeling, have raised serious doubts about the potential of LEU U3Si-Al dispersion fuel for applications requiring very high uranium densities and high burnups (>6 g U/cmT, >50% burnup). As a result of these findings, the fuel development efforts have been redirected towards three new initiatives: (1) a systematic application of ion-beam irradiations to screen new materials; (2) application of Hot Isostatic Pressing (HIP) procedures to produce U3Si2-Al plates with high uranium densities and thin uniform cladding; and (3) application of HIP procedures to produce plates with U3Si wires imbedded in an aluminum matrix, achieving stability, high uranium density, and thin uniform cladding. The new fuel concepts hold the promise of extraordinary performance potential and require approximately five years to develop.

  14. 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.

  15. Multiple recycle of REMIX fuel based on reprocessed uranium and plutonium mixture in thermal reactors

    SciTech Connect (OSTI)

    Fedorov, Y.S.; Bibichev, B.A.; Zilberman, B.Y.; Baryshnikov, M.V.; Kryukov, O.V.; Khaperskaya, A.V.

    2013-07-01

    REMIX fuel consumption in WWER-1000 is considered. REMIX fuel is fabricated from non-separated mixture of uranium and plutonium obtained during NPP spent fuel reprocessing with further makeup by enriched natural uranium. It makes possible to recycle several times the total amount of uranium and plutonium obtained from spent fuel with 100% loading of the WWER-1000 core. The stored SNF could be also involved in REMIX fuel cycle by enrichment of regenerated uranium. The same approach could be applied to closing the fuel cycle of CANDU reactors. (authors)

  16. 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.

  17. 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.

  18. 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 ...

  19. Safeguards training course: Nuclear material safeguards for enrichment plants

    SciTech Connect (OSTI)

    Not Available

    1990-06-01

    The main objective of this course is to provide the course participants with the necessary skills to perform their inspection activities at enrichment plants. As background information, a variety of enrichment technologies will first be characterized and compared followed by a review of basic cascade, gas centrifuge, and gaseous diffusion theory. To focus on gas centrifuge and gaseous diffusion technology, the major components and system of gas centrifuge and gaseous diffusion enrichment plants including their function in routine LEU production will be identified. The objectives of safeguards at an enrichment plant, including those agreed to in the Hexapartite Safeguards Project, will then be described. Discussions will then focus on potential diversion scenarios at both a centrifuge and diffusion enrichment facility and applicable safeguards inspection activities for detecting these scenarios. This report presents a discussion on basic separation and cascade theory, uranium hexafluoride, and detailed separation theory, including gas centrifuge and gaseous diffusion.

  20. Safeguards design strategies: designing and constructing new uranium and plutonium processing facilities in the United States

    SciTech Connect (OSTI)

    Scherer, Carolynn P; Long, Jon D

    2010-09-28

    In the United States, the Department of Energy (DOE) is transforming its outdated and oversized complex of aging nuclear material facilities into a smaller, safer, and more secure National Security Enterprise (NSE). Environmental concerns, worker health and safety risks, material security, reducing the role of nuclear weapons in our national security strategy while maintaining the capability for an effective nuclear deterrence by the United States, are influencing this transformation. As part of the nation's Uranium Center of Excellence (UCE), the Uranium Processing Facility (UPF) at the Y-12 National Security Complex in Oak Ridge, Tennessee, will advance the U.S.'s capability to meet all concerns when processing uranium and is located adjacent to the Highly Enriched Uranium Materials Facility (HEUMF), designed for consolidated storage of enriched uranium. The HEUMF became operational in March 2010, and the UPF is currently entering its final design phase. The designs of both facilities are for meeting anticipated security challenges for the 21st century. For plutonium research, development, and manufacturing, the Chemistry and Metallurgy Research Replacement (CMRR) building at the Los Alamos National Laboratory (LANL) in Los Alamos, New Mexico is now under construction. The first phase of the CMRR Project is the design and construction of a Radiological Laboratory/Utility/Office Building. The second phase consists of the design and construction of the Nuclear Facility (NF). The National Nuclear Security Administration (NNSA) selected these two sites as part of the national plan to consolidate nuclear materials, provide for nuclear deterrence, and nonproliferation mission requirements. This work examines these two projects independent approaches to design requirements, and objectives for safeguards, security, and safety (3S) systems as well as the subsequent construction of these modern processing facilities. Emphasis is on the use of Safeguards-by-Design (SBD), incorporating Systems Engineering (SE) principles for these two projects.

  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. 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.

  3. 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.

  4. 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.

  5. Manhattan Project: The Uranium Path to the Bomb, 1942-1944

    Office of Scientific and Technical Information (OSTI)

    Alpha Racetrack, Y-12 Electromagnetic Plant, Oak Ridge THE URANIUM PATH TO THE BOMB (1942-1944) Events > The Uranium Path to the Bomb, 1942-1944 Y-12: Design, 1942-1943 Y-12: Construction, 1943 Y-12: Operation, 1943-1944 Working K-25 into the Mix, 1943-1944 The Navy and Thermal Diffusion, 1944 The uranium path to the atomic bomb ran through Oak Ridge, Tennessee. Only if the new plants built at Oak Ridge produced enough enriched uranium-235 would a uranium bomb be possible. General Groves

  6. Gas Centrifuge Enrichment Plant Safeguards System Modeling

    SciTech Connect (OSTI)

    Elayat, H A; O'Connell, W J; Boyer, B D

    2006-06-05

    The U.S. Department of Energy (DOE) is interested in developing tools and methods for potential U.S. use in designing and evaluating safeguards systems used in enrichment facilities. This research focuses on analyzing the effectiveness of the safeguards in protecting against the range of safeguards concerns for enrichment plants, including diversion of attractive material and unauthorized modes of use. We developed an Extend simulation model for a generic medium-sized centrifuge enrichment plant. We modeled the material flow in normal operation, plant operational upset modes, and selected diversion scenarios, for selected safeguards systems. Simulation modeling is used to analyze both authorized and unauthorized use of a plant and the flow of safeguards information. Simulation tracks the movement of materials and isotopes, identifies the signatures of unauthorized use, tracks the flow and compilation of safeguards data, and evaluates the effectiveness of the safeguards system in detecting misuse signatures. The simulation model developed could be of use to the International Atomic Energy Agency IAEA, enabling the IAEA to observe and draw conclusions that uranium enrichment facilities are being used only within authorized limits for peaceful uses of nuclear energy. It will evaluate improved approaches to nonproliferation concerns, facilitating deployment of enhanced and cost-effective safeguards systems for an important part of the nuclear power fuel cycle.

  7. 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 -...

  8. Predicting 232U Content in Uranium (Technical Report) | SciTech...

    Office of Scientific and Technical Information (OSTI)

    The primary advantage of 232 U as an indicator of the presence of enriched uranium is the plentiful and penetrating nature of the radiation emitted by its daughter radionuclide ...

  9. 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.

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

    SciTech Connect (OSTI)

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

    1997-07-01

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

  11. 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.

  12. 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)

  13. Determination of the Relative Amount of Fluorine in Uranium Oxyfluoride Particles using Secondary Ion Mass Spectrometry and Optical Spectroscopy

    SciTech Connect (OSTI)

    Kips, R; Kristo, M J; Hutcheon, I D; Amonette, J; Wang, Z; Johnson, T; Gerlach, D; Olsen, K B

    2009-05-29

    Both nuclear forensics and environmental sampling depend upon laboratory analysis of nuclear material that has often been exposed to the environment after it has been produced. It is therefore important to understand how those environmental conditions might have changed the chemical composition of the material over time, particularly for chemically sensitive compounds. In the specific case of uranium enrichment facilities, uranium-bearing particles stem from small releases of uranium hexafluoride, a highly reactive gas that hydrolyzes upon contact with moisture from the air to form uranium oxyfluoride (UO{sub 2}F{sub 2}) particles. The uranium isotopic composition of those particles is used by the International Atomic Energy Agency (IAEA) to verify whether a facility is compliant with its declarations. The present study, however, aims to demonstrate how knowledge of time-dependent changes in chemical composition, particle morphology and molecular structure can contribute to an even more reliable interpretation of the analytical results. We prepared a set of uranium oxyfluoride particles at the Institute for Reference Materials and Measurements (IRMM, European Commission, Belgium) and followed changes in their composition, morphology and structure with time to see if we could use these properties to place boundaries on the particle exposure time in the environment. Because the rate of change is affected by exposure to UV-light, humidity levels and elevated temperatures, the samples were subjected to varying conditions of those three parameters. The NanoSIMS at LLNL was found to be the optimal tool to measure the relative amount of fluorine in individual uranium oxyfluoride particles. At PNNL, cryogenic laser-induced time-resolved U(VI) fluorescence microspectroscopy (CLIFS) was used to monitor changes in the molecular structure.

  14. 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.

  15. 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.

  16. Oak Ridge, Tenn. Selected as Uranium Enrichment Site | National...

    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...

  17. Uranium enrichment decontamination and decommissioning fund, 1995 report

    SciTech Connect (OSTI)

    1996-11-01

    This report describes strategies for the decontamination and decommissioning of gaseous diffusion plants. Progress in remedial action activities are discussed.

  18. Y-12 and the super enriched Uranium 235?

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

    of them used Bill's walking stick to get in the house. It was fun listening to them coach each other about how to walk without falling. I tell you this to help you know that...

  19. Oak Ridge, Tenn. Selected as Uranium Enrichment Site | National...

    National Nuclear Security Administration (NNSA)

    K-25 area of the site, in the electromagnetic plant in the Y-12 area, and in the liquid thermal diffusion plant. A pilot pile (reactor) and plutonium separation facility are...

  20. Current status and future plan of uranium enrichment technology

    SciTech Connect (OSTI)

    Yonekawa, S.; Yamamoto, F.; Yato, Y.; Kishimoto, Y.

    1994-12-31

    The Power Reactor and Nuclear Fuel Development Corporation (PNC) has been conducting extensive research and development (R&D) on the centrifuge process for more than a quarter of a century. This development program, designated as a national project in 1972, has resulted in the construction and operation of a pilot plant with a capacity of 50 t separative work unit (SWU) per year as well as a demonstration plant with a capacity of 200 t SWU/yr. Under the basic agreement of cooperation concluded in 1985, the technology developed in this program has been transferred to Japan Nuclear Fuel Limited (JNFL), which is now constructing and operating the commercial plant with a capacity of 1500 t SWU/yr at Rokkasho, Aomori. This paper describes the operational experiences of the demonstration plant, the status of a new material centrifuge, which will be introduced at a later stage of construction of the commercial plant, the development of an advanced centrifuge as a next-generation machine, and the research of a superadvanced centrifuge.

  1. Advanced Neutron Source enrichment study -- Volume 1: Main report. Final report, Revision 12/94

    SciTech Connect (OSTI)

    Bari, R.A.; Ludewig, H.; Weeks, J.

    1994-12-31

    A study has been performed of the impact on performance of using low enriched uranium (20% {sup 235}U) or medium enriched uranium (35% {sup 235}U) as an alternative fuel for the Advanced Neutron Source, which is currently designed to use uranium enriched to 93% {sup 235}U. Higher fuel densities and larger volume cores were evaluated at the lower enrichments in terms of impact on neutron flux, safety, safeguards, technical feasibility, and cost. The feasibility of fabricating uranium silicide fuel at increasing material density was specifically addressed by a panel of international experts on research reactor fuels. The most viable alternative designs for the reactor at lower enrichments were identified and discussed. Several sensitivity analyses were performed to gain an understanding of the performance of the reactor at parametric values of power, fuel density, core volume, and enrichment that were interpolations between the boundary values imposed on the study or extrapolations from known technology. Volume 2 of this report contains 26 appendices containing results, meeting minutes, and fuel panel presentations.

  2. final ERI-2142 18-1501 Analysis of Potential Effects on Domestic Industries of DOE Excess Uranium Inventory 2015-2024.docx

    Energy Savers [EERE]

    ERI-2142.18-1501 Analysis of the Potential Effects on the Domestic Uranium Mining, Conversion and Enrichment Industries of the Introduction of DOE Excess Uranium Inventory During CY 2015 Through 2024 ENERGY RESOURCES INTERNATIONAL, INC. 1015 18 th Street, NW, Suite 650 Washington, DC 20036 USA Telephone: (202) 785-8833 Facsimile: (202) 785-8834 ERI-2142.18-1501 Analysis of the Potential Effects on the Domestic Uranium Mining, Conversion and Enrichment Industries of the Introduction of DOE

  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. A Robust and Flexible Design for GCEP Unattended Online Enrichment Monitoring: An OLEM Collection Node Network

    SciTech Connect (OSTI)

    Younkin, James R; March-Leuba, Jose A; Garner, James R

    2013-01-01

    Oak Ridge National Laoratory (ORNL) has engineered an on-line enrichment monitor (OLEM) to continuously measure U-235 emissions from the UF6 gas flowing through a unit header pipe of a gas centrifuge enrichment plant (GCEP) as a component of the International Atomic Energy Agency s (IAEA) new generation of technology to support enrichment plant safeguards1. In contrast to other enrichment monitoring approaches, OLEM calibrates and corrects for the pressure and temperature dependent UF6 gas-density without external radiation sources by using the inherent unit header pipe pressure dynamics and combining U-235 gamma-ray spectrometery using a shielded NaI detector with gas pressure and temperature data near the spectrum measurement point to obtain the enrichment of the gas as a function of time. From a safeguards perspective, OLEM can provide early detection of a GCEP being misused for production of highly enriched uranium, but would not detect directly the isolation and use of a cascade within the production unit to produce HEU. OLEM may also reduce the number of samples collected for destructive assay and, if coupled with load cell monitoring, could support isotope mass balance verification and unattended cylinder verification. The earlier paper presented OLEM as one component along with shared load cells and unattended cylinder verification, in the IAEA emering toolbox for unattended instruments at GCEPs1 and described the OLEM concept and how previous modeling studies and field measurements helped confirm the viability of a passive on-line enrichment monitor for meeting IAEA objectives and to support the development of performance targets. Phase I of the United States Support Program (USSP) OLEM project completed a preliminary hardware, software and communications design; phase II will build and test field prototypes in controlled laboratory settings and then at an operational facility. That paper also discussed many of the OLEM collection node commercial off the shelf (COTS) components and summarized the OLEM collection node data security provisions. This paper will discuss a secure and redundant network of OLEM collection nodes, auxiliary detection units and supporting junction boxes distributed throughout a facility for monitoring enrichment on product, feed and tails unit header pipes; the purpose and capability of the built-in Electronic Optical Sealing System (EOSS) network gateway; and a network approach for obtaining reliable and authenticated pressure measurements.

  7. EIS-0471: Department of Energy Loan Guarantee to Support Proposed Eagle Rock Enrichment Facility in Bonneville County, Idaho

    Broader source: Energy.gov [DOE]

    This EIS evaluates the environmental impacts of construction, operation, and decommissioning of the proposed Eagle Rock Enrichment Facility (EREF), a gas centrifuge uranium enrichment facility to be located in a rural area in western Bonneville County, Idaho. (DOE adopted this EIS issued by NRC on 04/13/2007.)

  8. Benchmark Evaluation of Uranium Metal Annuli and Cylinders with Beryllium Reflectors

    SciTech Connect (OSTI)

    John D. Bess

    2010-06-01

    An extensive series of delayed critical experiments were performed at the Oak Ridge Critical Experiments Facility using enriched uranium metal during the 1960s and 1970s in support of criticality safety operations at the Y-12 Plant. These experiments were designed to evaluate the storage, casting, and handling limits of the Y-12 Plant and to provide data for the verification of cross sections and calculation methods utilized in nuclear criticality safety applications. Many of these experiments have already been evaluated and included in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook: unreflected (HEU-MET-FAST-051), graphite-reflected (HEU-MET-FAST-071), and polyethylene-reflected (HEU-MET-FAST-076). Three of the experiments consisted of highly-enriched uranium (HEU, ~93.2% 235U) metal parts reflected by beryllium metal discs. The first evaluated experiment was constructed from a stack of 7-in.-diameter, 4-1/8-in.-high stack of HEU discs top-reflected by a 7-in.-diameter, 5-9/16-in.-high stack of beryllium discs. The other two experiments were formed from stacks of concentric HEU metal annular rings surrounding a 7-in.diameter beryllium core. The nominal outer diameters were 13 and 15 in. with a nominal stack height of 5 and 4 in., respectively. These experiments have been evaluated for inclusion in the ICSBEP Handbook.

  9. Energy Department Selects Global Laser Enrichment for Future Operations at Paducah Site

    Broader source: Energy.gov [DOE]

    Washington, D.C. – The U.S. Department of Energy announced today that it will open negotiations with Global Laser Enrichment (GLE) for the sale of the depleted uranium hexafluoride inventory. The Department determined that GLE offered the greatest benefit to the government among those who responded to a Request for Offers (RFO) released earlier this year. Through the RFO review process, the Department also decided to enter into negotiations with AREVA for the off-specification uranium hexafluoride inventory.

  10. 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, ...

  11. Stable Isotope Enrichment Capabilities at ORNL

    SciTech Connect (OSTI)

    Egle, Brian; Aaron, W Scott; Hart, Kevin J

    2013-01-01

    The Oak Ridge National Laboratory (ORNL) and the US Department of Energy Nuclear Physics Program have built a high-resolution Electromagnetic Isotope Separator (EMIS) as a prototype for reestablishing a US based enrichment capability for stable isotopes. ORNL has over 60 years of experience providing enriched stable isotopes and related technical services to the international accelerator target community, as well as medical, research, industrial, national security, and other communities. ORNL is investigating the combined use of electromagnetic and gas centrifuge isotope separation technologies to provide research quantities (milligram to several kilograms) of enriched stable isotopes. In preparation for implementing a larger scale production facility, a 10 mA high-resolution EMIS prototype has been built and tested. Initial testing of the device has simultaneously collected greater than 98% enriched samples of all the molybdenum isotopes from natural abundance feedstock.

  12. U.S. Uranium Reserves Estimates

    Gasoline and Diesel Fuel Update (EIA)

    Major U.S. Uranium Reserves

  13. In-line assay monitor for uranium hexafluoride

    DOE Patents [OSTI]

    Wallace, Steven A.

    1981-01-01

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

  14. 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.

  15. 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.

  16. 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.

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

    SciTech Connect (OSTI)

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

    2007-07-01

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

  18. 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.

  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 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

  20. Separation of uranium from (Th,U)O.sub.2 solid solutions

    DOE Patents [OSTI]

    Chiotti, Premo; Jha, Mahesh Chandra

    1976-09-28

    Uranium is separated from mixed oxides of thorium and uranium by a pyrometallurgical process in which the oxides are mixed with a molten chloride salt containing thorium tetrachloride and thorium metal which reduces the uranium oxide to uranium metal which can then be recovered from the molten salt. The process is particularly useful for the recovery of uranium from generally insoluble high-density sol-gel thoria-urania nuclear reactor fuel pellets.

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

    SciTech Connect (OSTI)

    Schneider, Erich; Scopatza, Anthony; Deinert, Mark

    2007-07-01

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

  2. Rescuing a Treasure Uranium-233 (Conference) | SciTech Connect

    Office of Scientific and Technical Information (OSTI)

    and characterizing natural uranium isotopes for domestic and international safeguards. ... and the high cost of returning to operation this currently shut down capability. ...

  3. 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

  4. Uranium Marketing Annual Report - Release Date: May 31, 2011

    Gasoline and Diesel Fuel Update (EIA)

    4. Uranium purchased by owners and operators of U.S. civilian nuclear power reactors by origin and material type, 2014 deliveries thousand pounds U3O8 equivalent; dollars per pound U3O8 equivalent Deliveries Uranium concentrate Natural UF6 Enriched UF6 Total U.S.-origin uranium Purchases 2,004 1,312 0 3,316 Weighted-average price 49.87 45.41 -- 48.11 Foreign-origin uranium Purchases 27,257 16,491 6,285 50,033 Weighted-average price 44.56 48.98 44.69 46.03 Total Purchases 29,260 17,803 6,285

  5. The Enriched Xenon Observatory

    SciTech Connect (OSTI)

    Dolinski, M. J. [Stanford University Physics Department, 382 Via Pueblo Mall, Stanford, CA 94305-4060 (United States)

    2009-12-17

    The Enriched Xenon Observatory (EXO) experiment will search for neutrinoless double beta decay of {sup 136}Xe. The EXO Collaboration is actively pursuing both liquid-phase and gas-phase Xe detector technologies with scalability to the ton-scale. The search for neutrinoless double beta decay of {sup 136}Xe is especially attractive because of the possibility of tagging the resulting Ba daughter ion, eliminating all sources of background other than the two neutrino decay mode. EXO-200, the first phase of the project, is a liquid Xe time projection chamber with 200 kg of Xe enriched to 80% in {sup 136}Xe. EXO-200, which does not include Ba-tagging, will begin taking data in 2009, with two-year sensitivity to the half-life for neutrinoless double beta decay of 6.4x10{sup 25} years. This corresponds to an effective Majorana neutrino mass of 0.13 to 0.19 eV.

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

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

    Dewji, Shaheen A.; Lee, Denise L.; Croft, Stephen; Hertel, Nolan E.; Chapman, Jeffrey Allen; McElroy, Jr., Robert Dennis; Cleveland, S.

    2016-03-28

    Recent IAEA circulars and policy papers have sought to implement safeguards when any purified aqueous uranium solution or uranium oxides suitable for isotopic enrichment or fuel fabrication exists. Under the revised policy, IAEA Policy Paper 18, the starting point for nuclear material under safeguards was reinterpreted, suggesting that purified uranium compounds should be subject to safeguards procedures no later than the first point in the conversion process. In response to this technical need, a combination of simulation models and experimental measurements were employed to develop and validate concepts of nondestructive assay monitoring systems in a natural uranium conversion plant (NUCP).more » In particular, uranyl nitrate (UO2(NO3)2) solution exiting solvent extraction was identified as a key measurement point (KMP), where gamma-ray spectroscopy was selected as the process monitoring tool. The Uranyl Nitrate Calibration Loop Equipment (UNCLE) facility at Oak Ridge National Laboratory was employed to simulate the full-scale operating conditions of a purified uranium-bearing aqueous stream exiting the solvent extraction process in an NUCP. Nondestructive assay techniques using gamma-ray spectroscopy were evaluated to determine their viability as a technical means for drawing safeguards conclusions at NUCPs, and if the IAEA detection requirements of 1 significant quantity (SQ) can be met in a timely way. This work investigated gamma-ray signatures of uranyl nitrate circulating in the UNCLE facility and evaluated various gamma-ray detector sensitivities to uranyl nitrate. These detector validation activities include assessing detector responses to the uranyl nitrate gamma-ray signatures for spectrometers based on sodium iodide, lanthanum bromide, and high-purity germanium detectors. The results of measurements under static and dynamic operating conditions at concentrations ranging from 10–90 g U/L of natural uranyl nitrate are presented. A range of gamma-ray lines is examined, including attenuation for transmission measurement of density and concentration. It was determined that transmission-corrected gamma-ray spectra provide a reliable way to monitor the 235U concentration of uranyl nitrate solution in transfer pipes in NUCPs. Furthermore, existing predictive and analysis methods are adequate to design and realize practical designs. The 137Cs transmission source employed in this work is viable but not optimal for 235U densitometry determination. Validated simulations assessed the viability of 133Ba and 57Co as alternative densitometry sources. All three gamma-ray detectors are viable for monitoring natural uranium feed; although high-purity germanium is easiest to interpret, it is, however, the least attractive as an installation instrument. Overall, for monitoring throughput in a facility such as UNCLE, emulating the uranium concentration and pump speeds of the Springfields conversion facility in the United Kingdom, an uncertainty of less than 0.17% is required in order to detect the diversion of 1 SQ of uranyl nitrate through changes in uranium concentration over an accountancy period of one year with a detection probability of 50%. As a result, calibrated gamma-ray detection systems are capable of determining the concentration of uranium content in NUCPs, it is only in combination with verifiable operator declarations and supporting data, such as flow rate and enrichment, that safeguards conclusions can be drawn.« less

  7. Protein scaffolds for selective enrichment of metal ions

    DOE Patents [OSTI]

    He, Chuan; Zhou, Lu; Bosscher, Michael

    2016-02-09

    Polypeptides comprising high affinity for the uranyl ion are provided. Methods for binding uranyl using such proteins are likewise provided and can be used, for example, in methods for uranium purification or removal.

  8. 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.

  9. 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.

  10. 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 ...

  11. 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 ...

  12. 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 ...

  13. 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.

  14. 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. ...

  15. 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

  16. 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.

  17. 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.

  18. 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 )

  19. Performance and safety parameters for the high flux isotope reactor

    SciTech Connect (OSTI)

    Ilas, G. [Oak Ridge National Laboratory, 1 Bethel Valley Road, Oak Ridge, TN 37831-6172 (United States); Primm III, T. [Oak Ridge National Laboratory, 1 Bethel Valley Road, Oak Ridge, TN 37831-6172 (United States); Primm Consulting, LLC, 945 Laurel Hill Road, Knoxville, TN 37923 (United States)

    2012-07-01

    A Monte Carlo depletion model for the High Flux Isotope Reactor (HFIR) Cycle 400 and its use in calculating parameters of relevance to the reactor performance and safety during the reactor cycle are presented in this paper. This depletion model was developed to serve as a reference for the design of a low-enriched uranium (LEU) fuel for an ongoing study to convert HFIR from high-enriched uranium (HEU) to LEU fuel; both HEU and LEU depletion models use the same methodology and ENDF/B-VII nuclear data as discussed in this paper. The calculated HFIR Cycle 400 parameters, which are compared with measurement data from critical experiments performed at HFIR, data included in the HFIR Safety Analysis Report (SAR), or data reported by previous calculations, provide a basis for verification or updating of the corresponding SAR data. (authors)

  20. Performance and Safety Parameters for the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Ilas, Germina [ORNL; Primm, Trent [Primm Consulting, LLC

    2012-01-01

    A Monte Carlo depletion model for the High Flux Isotope Reactor (HFIR) Cycle 400 and its use in calculating parameters of relevance to the reactor performance and safety during the reactor cycle are presented in this paper. This depletion model was developed to serve as a reference for the design of a low-enriched uranium (LEU) fuel for an ongoing study to convert HFIR from high-enriched uranium (HEU) to LEU fuel; both HEU and LEU depletion models use the same methodology and ENDV/B-VII nuclear data as discussed in this paper. The calculated HFIR Cycle 400 parameters, which are compared when available with measurement data from critical experiments performed at HFIR, data included in the HFIR Safety Analysis Report (SAR), or data reported by previous calculations, provide a basis for verification or updating of the corresponding SAR data.

  1. 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...

  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 thousand ...

  3. 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 ...

  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 Deliveries 2011 2012 2013 2014 2015 Purchases of ...

  5. 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) ...

  6. 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- ...

  7. 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 ...

  8. 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

  9. 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

  10. 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 ...

  11. 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 ...

  12. 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 ...

  13. 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 ...

  14. 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 ...

  15. 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 ...

  16. 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 ...

  17. 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

  18. 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 ...

  19. 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 ...

  20. 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 ...

  1. 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

  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 Minimum ...

  3. 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 ...

  4. 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 ...

  5. 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 ...

  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 1

  7. 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

  8. 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.

  9. Using Process Load Cell Information for IAEA Safeguards at Enrichment Plants

    SciTech Connect (OSTI)

    Laughter, Mark D; Whitaker, J Michael; Howell, John

    2010-01-01

    Uranium enrichment service providers are expanding existing enrichment plants and constructing new facilities to meet demands resulting from the shutdown of gaseous diffusion plants, the completion of the U.S.-Russia highly enriched uranium downblending program, and the projected global renaissance in nuclear power. The International Atomic Energy Agency (IAEA) conducts verification inspections at safeguarded facilities to provide assurance that signatory States comply with their treaty obligations to use nuclear materials only for peaceful purposes. Continuous, unattended monitoring of load cells in UF{sub 6} feed/withdrawal stations can provide safeguards-relevant process information to make existing safeguards approaches more efficient and effective and enable novel safeguards concepts such as information-driven inspections. The IAEA has indicated that process load cell monitoring will play a central role in future safeguards approaches for large-scale gas centrifuge enrichment plants. This presentation will discuss previous work and future plans related to continuous load cell monitoring, including: (1) algorithms for automated analysis of load cell data, including filtering methods to determine significant weights and eliminate irrelevant impulses; (2) development of metrics for declaration verification and off-normal operation detection ('cylinder counting,' near-real-time mass balancing, F/P/T ratios, etc.); (3) requirements to specify what potentially sensitive data is safeguards relevant, at what point the IAEA gains on-site custody of the data, and what portion of that data can be transmitted off-site; (4) authentication, secure on-site storage, and secure transmission of load cell data; (5) data processing and remote monitoring schemes to control access to sensitive and proprietary information; (6) integration of process load cell data in a layered safeguards approach with cross-check verification; (7) process mock-ups constructed to provide simulated load cell data; (8) hardware and software implementation for process load cell data collection; (9) costs associated with unattended monitoring of load cells (for both operator and inspector) weighed against the potential benefits of having access to such data; (10) results from field tests of load cell data collection systems in operating facilities; and (11) use of unattended load cell data to increase efficiency of on-site inspection schedules and activities.

  10. Application of cadmium-zinc-telluride detectors in U-235 enrichment measurements

    SciTech Connect (OSTI)

    Ruhter, W.D.; Gunnink, R.

    1994-04-01

    High-resolution, gamma- and x-ray spectrometry are used routinely in nuclear safeguards verification measurements of plutonium and uranium in the field. These measurements are mostly performed with high-purity germanium (HPGe) detectors, that require cooling to liquid-nitrogen temperatures, thus limiting their utility in field and unattended safeguards measurement applications. NaI scintillation detectors do not require cooling, but their moderate energy resolution (10% at 122 keV) is insufficient in most cases for reliable verification measurements. Semiconductor detectors that operate at room temperature, such as cadmium-zinc-telluride (CdZnTe) detectors, with energy resolution performance reaching 2.0% at 122 keV may complement HPGe detectors for certain safeguards verification applications. The authors used a 5x5x5 mm CdZnTe detector to measure U-235 enrichments ranging from 3% to 75%. They use a spectrum analysis technique that fits U-235, U-238, and U K x-ray response profiles to data in the 89- to 100-keV region of gamma-ray spectrum. From the relative magnitudes of the U-235 and U-238 profiles they determine the U-235 enrichment with an accuracy of about 10% with CdZnTe detectors.

  11. 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.

  12. Ab-initio study of high temperature lattice dynamics of BCC zirconium (β-Zr) and uranium (γ-U)

    SciTech Connect (OSTI)

    Ghosh, Partha S. Arya, A. Dey, G. K.

    2014-04-24

    Using self consistent ab-initio lattice dynamics calculations, we show that bcc structures of Zr and U phases become stable at high temperature by phonon-phonon interactions. The calculated temperature dependent phonon dispersion curve (PDC) of β-Zr match excellently with experimental PDC. But the calculated PDC for γ-U shows negative phonon frequencies even at solid to liquid transition temperature. We show that this discrepancy is due to an overestimation of instability depth of bcc U phase which is removed by incorporation of spin-orbit coupling in the electronic structure calculations.

  13. 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/.

  14. 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)

  15. A Non-Proliferating Fuel Cycle: No Enrichment, Reprocessing or Accessible Spent Fuel - 12375

    SciTech Connect (OSTI)

    Parker, Frank L.

    2012-07-01

    Current fuel cycles offer a number of opportunities for access to plutonium, opportunities to create highly enriched uranium and access highly radioactive wastes to create nuclear weapons and 'dirty' bombs. The non-proliferating fuel cycle however eliminates or reduces such opportunities and access by eliminating the mining, milling and enrichment of uranium. The non-proliferating fuel cycle also reduces the production of plutonium per unit of energy created, eliminates reprocessing and the separation of plutonium from the spent fuel and the creation of a stream of high-level waste. It further simplifies the search for land based deep geologic repositories and interim storage sites for spent fuel in the USA by disposing of the spent fuel in deep sub-seabed sediments after storing the spent fuel at U.S. Navy Nuclear Shipyards that have the space and all of the necessary equipment and security already in place. The non-proliferating fuel cycle also reduces transportation risks by utilizing barges for the collection of spent fuel and transport to the Navy shipyards and specially designed ships to take the spent fuel to designated disposal sites at sea and to dispose of them there in deep sub-seabed sediments. Disposal in the sub-seabed sediments practically eliminates human intrusion. Potential disposal sites include Great Meteor East and Southern Nares Abyssal Plain. Such sites then could easily become international disposal sites since they occur in the open ocean. It also reduces the level of human exposure in case of failure because of the large physical and chemical dilution and the elimination of a major pathway to man-seawater is not potable. Of course, the recovery of uranium from sea water and the disposal of spent fuel in sub-seabed sediments must be proven on an industrial scale. All other technologies are already operating on an industrial scale. If externalities, such as reduced terrorist threats, environmental damage (including embedded emissions), long term care, reduced access to 'dirty' bomb materials, the social and political costs of siting new facilities and the psychological impact of no solution to the nuclear waste problem, were taken into account, the costs would be far lower than those of the present fuel cycle. (authors)

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

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

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

    2015-09-04

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

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

    SciTech Connect (OSTI)

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

    2015-09-04

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

  18. 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.

  19. 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

  20. 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.

  1. 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.

  2. 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.

  3. Overview of enrichment plant safeguards

    SciTech Connect (OSTI)

    Swindle, D.W. Jr.; Wheeler, L.E.

    1982-01-01

    The relationship of enrichment plant safeguards to US nonproliferation objectives and to the operation and management of enrichment facilities is reviewed. During the review, the major components of both domestic and international safeguards systems for enrichment plants are discussed. In discussing domestic safeguards systems, examples of the technology currently in use to support nuclear materials accountability are described including the measurement methods, procedures and equipment used for weighing, sampling, chemical and isotopic analyses and nondestructive assay techniques. Also discussed is how the information obtained as part of the nuclear material accountancy task is useful to enrichment plant operations. International material accountancy verification and containment/surveillance concepts for enrichment plants are discussed, and the technologies presently being developed for international safeguards in enrichment plants are identified and the current development status is reported.

  4. Uranium isotopes in ground water as a prospecting technique

    SciTech Connect (OSTI)

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

    1980-02-01

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

  5. 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.

  6. 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...

  7. 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...

  8. 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...

  9. 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...

  10. 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, ...

  11. 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.

  12. 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.

  13. Laser Isotope Enrichment for Medical and Industrial Applications

    SciTech Connect (OSTI)

    Leonard Bond

    2006-07-01

    Laser Isotope Enrichment for Medical and Industrial Applications by Jeff Eerkens (University of Missouri), Jay Kunze (Idaho State University), and Leonard Bond (Idaho National Laboratory) The principal isotope enrichment business in the world is the enrichment of uranium for commercial power reactor fuels. However, there are a number of other needs for separated isotopes. Some examples are: 1) Pure isotopic targets for irradiation to produce medical radioisotopes. 2) Pure isotopes for semiconductors. 3) Low neutron capture isotopes for various uses in nuclear reactors. 4) Isotopes for industrial tracer/identification applications. Examples of interest to medicine are targets to produce radio-isotopes such as S-33, Mo-98, Mo-100, W-186, Sn-112; while for MRI diagnostics, the non-radioactive Xe-129 isotope is wanted. For super-semiconductor applications some desired industrial isotopes are Si-28, Ga-69, Ge-74, Se-80, Te-128, etc. An example of a low cross section isotope for use in reactors is Zn-68 as a corrosion inhibitor material in nuclear reactor primary systems. Neutron activation of Ar isotopes is of interest in industrial tracer and diagnostic applications (e.g. oil-logging). . In the past few years there has been a sufficient supply of isotopes in common demand, because of huge Russian stockpiles produced with old electromagnetic and centrifuge separators previously used for uranium enrichment. Production of specialized isotopes in the USA has been largely accomplished using old ”calutrons” (electromagnetic separators) at Oak Ridge National Laboratory. These methods of separating isotopes are rather energy inefficient. Use of lasers for isotope separation has been considered for many decades. None of the proposed methods have attained sufficient proof of principal status to be economically attractive to pursue commercially. Some of the authors have succeeded in separating sulfur isotopes using a rather new and different method, known as condensation repression. In this scheme a gas, of the selected isotopes for enrichment, is irradiated with a laser at a particular wavelength that would excite only one of the isotopes. The entire gas is subject to low temperatures sufficient to cause condensation on a cold surface. Those molecules in the gas that the laser excited are not as likely to condense as are the unexcited molecules. Hence the gas drawn out of the system will be enriched in the isotope that was excited by the laser. We have evaluated the relative energy required in this process if applied on a commercial scale. We estimate the energy required for laser isotope enrichment is about 20% of that required in centrifuge separations, and 2% of that required by use of "calutrons".

  14. 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.

  15. 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.

  16. Uranium Marketing Annual Report -

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

    AREVA NC, Inc AREVA Enrichment Services, LLC AREVA NC, Inc. AREVA NC, Inc. .CNEIC (China Nuclear Energy Industry Corporation) CNEIC (China Nuclear Energy Industry Corporation) ...

  17. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    7. Purchases of enrichment services by owners and operators of U.S. civilian nuclear power reactors by contract type in delivery year, 2014 thousand separative work units (SWU) Enrichment service contract type U.S. enrichment Foreign enrichment Total Spot W W 628 Long-term W W 12,310 Total 3,773 9,165 12,939 W = Data withheld to avoid disclosure of individual company data. Note: Totals may not equal sum of components because of independent rounding. Source: U.S. Energy Information

  18. 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.

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

    SciTech Connect (OSTI)

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

    1997-08-01

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

  20. 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

  1. Enterprise Assessments Targeted Review of the Safety System Management...

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

    System at the Y-12 National Security Complex Highly Enriched Uranium Materials ... System at the Y-12 National Security Complex Highly Enriched Uranium Materials ...

  2. Influence of Acidic and Alkaline Waste Solution Properties on Uranium Migration in Subsurface Sediments

    SciTech Connect (OSTI)

    Szecsody, James E.; Truex, Michael J.; Qafoku, Nikolla; Wellman, Dawn M.; Resch, Charles T.; Zhong, Lirong

    2013-08-01

    This study shows that acidic and alkaline wastes co-disposed with uranium into subsurface sediments has significant impact on changes in uranium retardation, concentration, and mass during downward migration. For uranium co-disposal with acidic wastes, significant rapid (i.e., hours) carbonate and slow (i.e., 100s of hours) clay dissolution resulted, releasing significant sediment-associated uranium, but the extent of uranium release and mobility change was controlled by the acid mass added relative to the sediment proton adsorption capacity. Mineral dissolution in acidic solutions (pH 2) resulted in a rapid (< 10 h) increase in aqueous carbonate (with Ca2+, Mg2+) and phosphate and a slow (100s of hours) increase in silica, Al3+, and K+, likely from 2:1 clay dissolution. Infiltration of uranium with a strong acid resulted in significant shallow uranium mineral dissolution and deeper uranium precipitation (likely as phosphates and carbonates) with downward uranium migration of three times greater mass at a faster velocity relative to uranium infiltration in pH neutral groundwater. In contrast, mineral dissolution in an alkaline environment (pH 13) resulted in a rapid (< 10 h) increase in carbonate, followed by a slow (10s to 100s of hours) increase in silica concentration, likely from montmorillonite, muscovite, and kaolinite dissolution. Infiltration of uranium with a strong base resulted in uranium-silicate precipitation (presumed Na-boltwoodite) but also desorption of natural uranium on the sediment due to the high ionic strength solution, or 60% greater mass with greater retardation compared with groundwater. Overall, these results show that acidic or alkaline co-contaminant disposal with uranium can result in complex depth- and time-dependent changes in uranium dissolution/precipitation reactions and uranium sorption, which alter the uranium migration mass, concentration, and velocity.

  3. Phosphorus Enrichment as a New Composition in the Solid Electrolyte Interphase of High-Voltage Cathodes and Its Effects on Battery Cycling

    SciTech Connect (OSTI)

    Yan, Pengfei; Zheng, Jianming; Kuppan, Saravanan; Li, Qiuyan; Lv, Dongping; Xiao, Jie; Chen, Guoying; Zhang, Jiguang; Wang, Chong M.

    2015-11-10

    Immersion of a solid into liquid often leads to the modification of both the structure and chemistry of surface of the solid, which subsequently affects the chemical and physical properties of the system. For the case of the rechargeable lithium ion battery, such a surface modification is termed as solid electrolyte interphase (SEI) layer, which has been perceived to play critical role for the stable operation of the batteries. However, the structure and chemical composition of SEI layer and its spatial distribution and dependence on the battery operating condition remain unclear. By using aberration corrected scanning transmission electron microscopy coupled with ultra-high sensitive energy dispersive x-ray spectroscopy, we probed the structure and chemistry of SEI layer on several high voltage cathodes. We show that layer-structured cathodes, when cycled at a high cut off voltage, can form a P-rich SEI layer on their surface, which is a direct evidence of Li-salt (LiPF6) decomposition. Our systematical investigations indicate such cathode/Li-salt side reaction shows strong dependence on structure of the cathode materials, operating voltage and temperature, indicating the feasibility of SEI engineering. These findings provide us valuable insights into the complex interface between the high-voltage cathode and the electrolyte.

  4. High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

    SciTech Connect (OSTI)

    Sterbentz, James William; Bayless, Paul David; Nelson, Lee Orville; Gougar, Hans David; Strydom, Gerhard

    2016-01-01

    A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

  5. S. 210: This Act may be referred to as the Comprehensive Uranium Act of 1991, introduced in the Senate of the United States, One Hundred Second Congress, First Session, January 15, 1991

    SciTech Connect (OSTI)

    Not Available

    1991-01-01

    This bill would establish the United States Enrichment Corporation to operate the Federal uranium enrichment program on a profitable and efficient basis in order to maximize the long term economic value to the US, provide assistance to the domestic uranium industry, and provide a Federal contribution for the reclamation of mill tailings generated pursuant to Federal defense contracts at active uranium and thorium processing sites. The bill describes congressional findings; definitions, establishment of corporation, and purposes; corporate offices; powers and duties of the corporation; organization, finance, and management; licensing, taxation, and miscellaneous provisions; decontamination and decommissioning; and uranium security and tailing reclamation.

  6. 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.

  7. 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.

  8. 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.

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

  10. 2015 Uranium Marketing Annual Report

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

    7 2015 Uranium Marketing Annual Report Release Date: May 24, 2016 Next Release Date: May ... Annual, Tables 28, 29, 30 and 31. 2003-15-Form EIA-858, "Uranium Marketing Annual Survey". ...

  11. 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 ...

  12. 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 ...

  13. 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 ...

  14. 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.

  15. 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.

  16. 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)

  17. 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

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. 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

  4. 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

  5. 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

  6. 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.

  7. 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)

  8. Status Report on the Passive Neutron Enrichment Meter (PNEM) for UF6 Cylinder Assay

    SciTech Connect (OSTI)

    Miller, Karen A.; Swinhoe, Martyn T.; Menlove, Howard O.; Marlow, Johnna B.

    2012-05-02

    The Passive Neutron Enrichment Meter (PNEM) is a nondestructive assay (NDA) system being developed at Los Alamos National Laboratory (LANL). It was designed to determine {sup 235}U mass and enrichment of uranium hexafluoride (UF{sub 6}) in product, feed, and tails cylinders (i.e., 30B and 48Y cylinders). These cylinders are found in the nuclear fuel cycle at uranium conversion, enrichment, and fuel fabrication facilities. The PNEM is a {sup 3}He-based neutron detection system that consists of two briefcase-sized detector pods. A photograph of the system during characterization at LANL is shown in Fig. 1. Several signatures are currently being studied to determine the most effective measurement and data reduction technique for unfolding {sup 235}U mass and enrichment. The system collects total neutron and coincidence data for both bare and cadmium-covered detector pods. The measurement concept grew out of the success of the Uranium Cylinder Assay System (UCAS), which is an operator system at Rokkasho Enrichment Plant (REP) that uses total neutron counting to determine {sup 235}U mass in UF{sub 6} cylinders. The PNEM system was designed with higher efficiency than the UCAS in order to add coincidence counting functionality for the enrichment determination. A photograph of the UCAS with a 48Y cylinder at REP is shown in Fig. 2, and the calibration measurement data for 30B product and 48Y feed and tails cylinders is shown in Fig. 3. The data was collected in a low-background environment, meaning there is very little scatter in the data. The PNEM measurement concept was first presented at the 2010 Institute of Nuclear Materials Management (INMM) Annual Meeting. The physics design and uncertainty analysis were presented at the 2010 International Atomic Energy Agency (IAEA) Safeguards Symposium, and the mechanical and electrical designs and characterization measurements were published in the ESARDA Bulletin in 2011.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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

  15. 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.

  16. 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

  17. 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.

  18. A review of the Model 5A uranium hexafluoride cylinder

    SciTech Connect (OSTI)

    Dorning, R.E. II

    1989-05-23

    Both the Model 5A (Monel 400) and 5A (Monel 400) Modified five-inch cylinders have been used at the Portsmouth GDP to withdraw, store, and ship highly enriched uranium hexafluoride. As a result of a generic cracking problem with Monel 400 valve-boss material, a cylinder modification was implemented in the mid 1970s. This modification resulted in the violation of the ASME ''Code'' stamp status of the Model 5A Modified cylinder. Hydrostatic testing-to- rupture data indicated that the Model 5A Modified cylinders had ruptured strengths equivalent to that of the original Model 5A cylinders. An independent consultant reviewed the available information and confirmed that the Model 5A Modified cylinders ''will with proper maintenance continue to perform satisfactorily for many additional years of service.'' Based on the test data and consultant's review, DOE approved continued use of the 5A Modified cylinder and also requested procurement of replacement 5B cylinders be expedited. Currently, the 5A modified cylinders are in the production, storage, shipment cycle, and a sufficient number of 5B cylinders has been ordered to accommodate the projected product shipping requirements for the Navy flow. 3 tabs.

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

    SciTech Connect (OSTI)

    Forsberg, C.W.

    1996-01-20

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

  20. 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 ...