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Sample records for uranium enrichment decontamination

  1. 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...

  2. 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

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

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

    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

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. Results of chemical decontamination of DOE`s uranium-enrichment scrap metal

    SciTech Connect (OSTI)

    Levesque, R.G.

    1997-02-01

    The CORPEX{reg_sign} Nuclear Decontamination Processes were used to decontaminate representative scrap metal specimens obtained from the existing scrap metal piles located at the Department of Energy (DOE) Portsmouth Gaseous Diffusion Plant (PORTS), Piketon, Ohio. In September 1995, under contract to Lockheed Martin Energy Systems, MELE Associates, Inc. performed the on-site decontamination demonstration. The decontamination demonstration proved that significant amounts of the existing DOE scrap metal can be decontaminated to levels where the scrap metal could be economically released by DOE for beneficial reuse. This simple and environmentally friendly process can be used as an alternative, or in addition to, smelting radiologically contaminated scrap metal.

  11. 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.

  12. DECONTAMINATION OF URANIUM

    DOE Patents [OSTI]

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

    1958-02-01

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

  13. highly enriched uranium

    National Nuclear Security Administration (NNSA)

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

  14. 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. Highly Enriched Uranium Materials Facility

    National Nuclear Security Administration (NNSA)

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

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

    Office of Environmental Management (EM)

    Toxic Substances Control Act (TSCA) Uranium Enrichment Federal Facility Compliance Agreement establishes a plan to bring DOE's Uranium Enrichment Plants (and support facilities) ...

  17. 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...

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

    Office of Environmental Management (EM)

    Toxic Substance Control Act Uranium Enrichment Federal Facilities Compliance Agreement ... for bringing DOE's former and active Uranium Enrichment Plants in Paducah, Portsmouth, ...

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

    Broader source: Energy.gov [DOE]

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

  20. highly enriched uranium | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    highly enriched uranium NNSA Announces Elimination of Highly Enriched Uranium (HEU) from Indonesia All of Southeast Asia Now HEU-Free (WASHINGTON, D.C.) - The U.S. Department of Energy's National Nuclear Security Administration (DOE/NNSA), Indonesian Nuclear Industry, LLC (PT INUKI), the National Nuclear Energy Agency (BATAN), and the Nuclear Energy Regulatory Agency (BAPETEN) of the... NNSA deputy administrator travels to Ukraine Earlier this month, Deputy Administrator for Defense Nuclear

  1. ELECTROCHEMICAL DECONTAMINATION AND RECOVERY OF URANIUM VALUES

    DOE Patents [OSTI]

    McLaren, J.A.; Goode, J.H.

    1958-05-13

    An electrochemical process is described for separating uranium from fission products. The method comprises subjecting the mass of uranium to anodic dissolution in an electrolytic cell containing aqueous alkali bicarbonate solution as its electrolyte, thereby promoting a settling from the solution of a solid sludge from about the electrodes and separating the resulting electrolyte solution containing the anodically dissolved uranium from the sludge which contains the rare earth fission products.

  2. 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.

  3. 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.

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

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

    uranium, a vital national security asset. HEUMF is a massive concrete and steel structure that provides maximum security for the highly enriched uranium material that it protects. ...

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

    National Nuclear Security Administration (NNSA)

    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 ...

  6. Uranium Mining, Conversion, and Enrichment Industries

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

    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 1,600

  7. 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

  8. The Office of Environmental Management Uranium Enrichment D&D...

    Energy Savers [EERE]

    Uranium Enrichment D&D The Office of Environmental Management Uranium Enrichment D&D Microsoft Word - B996F741.doc (100.04 KB) More Documents & Publications Microsoft Word - PSRP ...

  9. EIS-0329: Proposed Construction, Operation, Decontamination/Decommissioning of Depleted Uranium Hexafluoride Conversion Facilities

    Broader source: Energy.gov [DOE]

    This EIS analyzes DOE's proposal to construct, operate, maintain, and decontaminate and decommission two depleted uranium hexafluoride (DUF 6) conversion facilities, at Portsmouth, Ohio, and Paducah, Kentucky.

  10. 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.

  11. 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.

  12. 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.

  13. 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

  14. 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

  15. 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.

  16. Highly Enriched Uranium Transparency Program | National Nuclear Security

    National Nuclear Security Administration (NNSA)

    Administration | (NNSA) 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

  17. 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.

  18. 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...

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

    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 ...

  20. 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...

  1. 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.

  2. 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

  3. 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.

  4. 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.

  5. Christenson Named Federal Project Director for Highly Enriched Uranium

    National Nuclear Security Administration (NNSA)

    Materials Facility | National Nuclear Security Administration | (NNSA) Christenson Named Federal Project Director for Highly Enriched Uranium Materials Facility November 20, 2008 Microsoft Office document icon NR-08-08 Christenson.doc

  6. 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...

  7. 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 ...

  8. US, Kazakhstan Cooperate to Eliminate Highly Enriched Uranium | National

    National Nuclear Security Administration (NNSA)

    Nuclear Security Administration | (NNSA) US, Kazakhstan Cooperate to Eliminate Highly Enriched Uranium January 07, 2015 WASHINGTON D.C - The Department of Energy's National Nuclear Security Administration (DOE/NNSA) announced today the removal of 36 kilograms (approximately 80 pounds) of highly enriched uranium (HEU) spent fuel from the Institute of Nuclear Physics (INP) in Almaty, Kazakhstan. The HEU was transported via two air shipments to a secure facility in Russia for permanent

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

    National Nuclear Security Administration (NNSA)

    Security Administration | (NNSA) 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

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. Uranium Mining, Conversion, and Enrichment Industries

    Broader source: Energy.gov (indexed) [DOE]

    ... At the fabrication facility, the enriched UF 6 is converted into UO 2 powder, and then formed into small ceramic pellets. These pellets are then loaded into metal tubes and ...

  17. H Canyon Moves Closer to Low Enriched Uranium Blend Down

    Broader source: Energy.gov [DOE]

    AIKEN, S.C. – The Savannah River Site’s (SRS) H Canyon has moved closer to restarting low enriched uranium (LEU) blend down by turning on the First Cycle unit operation for the first time in more than five years.

  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. 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.

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

    Office of Energy Efficiency and Renewable Energy (EERE)

    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.

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

    Energy Savers [EERE]

    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 ...

  2. 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.

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

    National Nuclear Security Administration (NNSA)

    National Nuclear Security Administration | (NNSA) 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

  4. NNSA Announces Elimination of Highly Enriched Uranium (HEU) from Indonesia

    National Nuclear Security Administration (NNSA)

    | National Nuclear Security Administration | (NNSA) Elimination of Highly Enriched Uranium (HEU) from Indonesia August 29, 2016 All of Southeast Asia Now HEU-Free (WASHINGTON, D.C.) - The U.S. Department of Energy's National Nuclear Security Administration (DOE/NNSA), Indonesian Nuclear Industry, LLC (PT INUKI), the National Nuclear Energy Agency (BATAN), and the Nuclear Energy Regulatory Agency (BAPETEN) of the Republic of Indonesia announced the completion of a collaborative effort to

  5. 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.

  6. PROCESS FOR DECONTAMINATING THORIUM AND URANIUM WITH RESPECT TO RUTHENIUM

    DOE Patents [OSTI]

    Meservey, A.A.; Rainey, R.H.

    1959-10-20

    The control of ruthenium extraction in solvent-extraction processing of neutron-irradiated thorium is presented. Ruthenium is rendered organic-insoluble by the provision of sulfite or bisulfite ions in the aqueous feed solution. As a result the ruthenium remains in the aqueous phase along with other fission product and protactinium values, thorium and uranium values being extracted into the organic phase. This process is particularly applicable to the use of a nitrate-ion-deficient aqueous feed solution and to the use of tributyl phosphate as the organic extractant.

  7. 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

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

    Office of Scientific and Technical Information (OSTI)

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

  9. 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...

  10. 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.

  11. Accelerating the Reduction of Excess Russian Highly Enriched Uranium

    SciTech Connect (OSTI)

    Benton, J; Wall, D; Parker, E; Rutkowski, E

    2004-02-18

    This paper presents the latest information on one of the Accelerated Highly Enriched Uranium (HEU) Disposition initiatives that resulted from the May 2002 Summit meeting between Presidents George W. Bush and Vladimir V. Putin. These initiatives are meant to strengthen nuclear nonproliferation objectives by accelerating the disposition of nuclear weapons-useable materials. The HEU Transparency Implementation Program (TIP), within the National Nuclear Security Administration (NNSA) is working to implement one of the selected initiatives that would purchase excess Russian HEU (93% 235U) for use as fuel in U.S. research reactors over the next ten years. This will parallel efforts to convert the reactors' fuel core from HEU to low enriched uranium (LEU) material, where feasible. The paper will examine important aspects associated with the U.S. research reactor HEU purchase. In particular: (1) the establishment of specifications for the Russian HEU, and (2) transportation safeguard considerations for moving the HEU from the Mayak Production Facility in Ozersk, Russia, to the Y-12 National Security Complex in Oak Ridge, TN.

  12. 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).

  13. 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.

  14. 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

  15. 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.

  16. 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.

  17. S. 2415: Title I may be cited as the Uranium Enrichment Act of 1990; Title II may be cited as the Uranium Security and Tailings Reclamation Act of 1989; and Title III may be cited as The Solar, Wind, Waste, and Geothermal Power Production Incentives Act of 1990, introduced in the Senate, One Hundred First Congress, Second Session, April 4, 1990

    SciTech Connect (OSTI)

    Not Available

    1990-01-01

    S. 2415 (which started out as a bill to encourage solar and geothermal power generation) now would amend the Atomic Energy Act of 1954 to redirect uranium enrichment enterprises to further the national interest, respond to competitive market forces, and to ensure the nation's common defense and security. It would establish a United States Enrichment Corporation for the following purposes: to acquire feed materials, enriched uranium, and enrichment facilities; to operate these facilities; to market enriched uranium for governmental purposes and qualified domestic and foreign persons; to conduct research into uranium enrichment; and to operate as a profitable, self-financing, reliable corporation and in a manner consistent with the health and safety of the public. The bill describes powers and duties of the corporation; the organization, finance, and management; decontamination and decommissioning. The second part of the bill would ensure an adequate supply of domestic uranium for defense and power production; provide assistance to the domestic uranium industry; and establish, facilitate, and expedite a comprehensive system for financing reclamation and remedial action at active uranium and thorium processing sites. The third part of the bill would remove the size limitations on power production facilities now part of the Public Utility Regulatory Policies Act of 1978. Solar, wind, waste, or geothermal power facilities would no longer have to be less than 80 MW to qualify as a small power production facility.

  18. 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.

  19. Secret Mission to Remove Highly Enriched Uranium Spent Nuclear Fuel from

    National Nuclear Security Administration (NNSA)

    Uzbekistan Successfully Completed | National Nuclear Security Administration | (NNSA) Secret Mission to Remove Highly Enriched Uranium Spent Nuclear Fuel from Uzbekistan Successfully Completed April 20, 2006 Four Shipments Have Been Sent to a Secure Facility in Russia WASHINGTON, D.C. -- The Department of Energy's National Nuclear Security Administration (NNSA) announced today that 63 kilograms (139 pounds) of highly enriched uranium (HEU) in spent nuclear fuel were safely and securely

  20. 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.

  1. 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

    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 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

  2. 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.

  3. 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.

  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. 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.

  6. 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

  7. 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.)

  8. System Chemical Decontamination Technology by the IF7 Gas

    SciTech Connect (OSTI)

    Ema, Akira; Sugitsue, Noritake; Zaitsu, Tomohisa

    2008-01-15

    When we decommission the equipments contaminated by the uranium, both the dismantling process of the equipments and the disposal process of the radioactive waste will be carried out. However, costs of these processes will be very expensive because the dismantling facilities are required in the dismantling process and many radioactive wastes are generated. Here, if it is possible to decontaminate the equipments below the clearance level, the amount of radioactive wastes is reduced very much. As a result, costs of the disposal will be able to reduce extremely. Moreover, if it is possible to decontaminate the equipments before dismantling, costs of the dismantling will be reduced, too. In other words, the total costs of the dismantling process and the disposal process will be minimized by decontaminating below the clearance level without dismantling. So, we have developed new decontamination technology to decontaminate below the clearance level without dismantling. In this paper, as an example, we tried to decontaminate the uranium enrichment plant in order to reveal the practicality of our technology. In conclusion: we proposed new decontamination technology by the IF7 gas. And, as an example, we tried to decontaminate the uranium enrichment plant. As a result, we could reveal that our technology has the high performance which can decontaminate below the assumed clearance level: 1.0 Bq/g in the limited time. Finally, we simulated the amount of the radioactive waste for decommissioning the uranium enrichment plant. As a result, though a lot of radioactive wastes will usually be generated, we obtained the prospect that it is possible to reduced 85% of the radioactive waste by using our technology.

  9. 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).

  10. 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.

  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, 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.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

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

    SciTech Connect (OSTI)

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

    1995-12-31

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

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

    SciTech Connect (OSTI)

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

    2010-10-01

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

  4. 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.

  5. Secretarial Determination Pursuant to USEC Privatization Act for the Sale or Transfer of Low-Enriched Uranium

    Office of Energy Efficiency and Renewable Energy (EERE)

    Secretarial determination regarding the potential impacts of the transfer by DOE of up to 48 metric tons of low-enriched uranium to USEC Inc. in exchange for DOE receiving approximately 409 metric...

  6. 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.

  7. 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,...

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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)

  13. 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

  14. 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

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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

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

    SciTech Connect (OSTI)

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

    2009-03-01

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

  3. 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

  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. 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.

  6. 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.

  7. 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

  8. 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.

  9. RUSSIAN-ORIGIN HIGHLY ENRICHED URANIUM SPENT NUCLEAR FUEL SHIPMENT FROM BULGARIA

    SciTech Connect (OSTI)

    Kelly Cummins; Igor Bolshinsky; Ken Allen; Tihomir Apostolov; Ivaylo Dimitrov

    2009-07-01

    In July 2008, the Global Threat Reduction Initiative and the IRT 2000 research reactor in Sofia, Bulgaria, operated by the Institute for Nuclear Research and Nuclear Energy (INRNE), safely shipped 6.4 kilograms of Russian origin highly enriched uranium (HEU) spent nuclear fuel (SNF) to the Russian Federation. The shipment, which resulted in the removal of all HEU from Bulgaria, was conducted by truck, barge, and rail modes of transport across two transit countries before reaching the final destination at the Production Association Mayak facility in Chelyabinsk, Russia. This paper describes the work, equipment, organizations, and approvals that were required to complete the spent fuel shipment and provides lessons learned that might assist other research reactor operators with their own spent nuclear fuel shipments.

  10. 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.

    2015-07-01

    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

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. Investigation of gas-phase decontamination of internally radioactively contaminated gaseous diffusion process equipment and piping

    SciTech Connect (OSTI)

    Bundy, R.D.; Munday, E.B.

    1991-01-01

    Construction of the gaseous diffusion plants (GDPs) was begun during World War 2 to produce enriched uranium for defense purposes. These plants, which utilized UF{sub 6} gas, were used primarily for this purpose through 1964. From 1959 through 1968, production shifted primarily to uranium enrichment to supply the nuclear power industry. Additional UF{sub 6}-handling facilities were built in feed and fuel-processing plants associated with the uranium enrichment process. Two of the five process buildings at Oak ridge were shut down in 1964. Uranium enrichment activities at Oak Ridge were discontinued altogether in 1985. In 1987, the Department of Energy (DOE) decided to proceed with a permanent shutdown of the Oak Ridge Gaseous Diffusion Plant (ORGDP). DOE intends to begin decommissioning and decontamination (D D) of ORGDP early in the next century. The remaining two GDPs are expected to be shut down during the next 10 to 40 years and will also require D D, as will the other UF{sub 6}-handling facilities. This paper presents an investigation of gas- phase decontamination of internally radioactively contaminated gaseous diffusion process equipment and piping using powerful fluorinating reagents that convert nonvolatile uranium compounds to volatile UF{sub 6}. These reagents include ClF{sub 3}, F{sub 2}, and other compounds. The scope of D D at the GDPs, previous work of gas-phase decontamination, four concepts for using gas-phase decontamination, plans for further study of gas-phase decontamination, and the current status of this work are discussed. 13 refs., 15 figs.

  18. 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

  19. 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

  20. EIS-0360: Depleted Uranium Oxide Conversion Product at the Portsmouth, Ohio

    Energy Savers [EERE]

    Site | Department of Energy 60: Depleted Uranium Oxide Conversion Product at the Portsmouth, Ohio Site EIS-0360: Depleted Uranium Oxide Conversion Product at the Portsmouth, Ohio Site Summary This site-specific EIS analyzes the construction, operation, maintenance, and decontamination and decommissioning of the proposed depleted uranium hexafluoride (DUF6) conversion facility at three alternative locations within the Portsmouth site; transportation of all cylinders (DUF6, enriched, and

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 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

  7. 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

  8. 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.

  9. 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.

  10. Environmental decontamination

    SciTech Connect (OSTI)

    Cristy, G.A.; Jernigan, H.C.

    1981-02-01

    The record of the proceedings of the workshop on environmental decontamination contains twenty-seven presentations. Emphasis is placed upon soil and surface decontamination, the decommissioning of nuclear facilities, and assessments of instrumentation and equipment used in decontamination. (DLS)

  11. 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.

  12. 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 world’s 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.

  13. 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.

  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. PYROCHEMICAL DECONTAMINATION METHOD FOR REACTOR FUEL

    DOE Patents [OSTI]

    Buyers, A.G.

    1959-06-30

    A pyro-chemical method is presented for decontaminating neutron irradiated uranium and separating plutonium therefrom by contact in the molten state with a metal chloride salt. Uranium trichloride and uranium tetrachloride either alone or in admixture with alkaline metal and alkaline eanth metal fluorides under specified temperature and specified phase ratio conditions extract substantially all of the uranium from the irradiated uranium fuel together with certain fission products. The phases are then separated leaving purified uranium metal. The uranium and plutonium in the salt phase can be reduced to forin a highly decontaminated uraniumplutonium alloy. The present method possesses advantages for economically decontaminating irradiated nuclear fuel elements since irradiated fuel may be proccessed immediately after withdrawal from the reactor and the uranium need not be dissolved and later reduced to the metallic form. Accordingly, the uranium may be economically refabricated and reinserted into the reactor.

  16. 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.

  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. 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

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

    SciTech Connect (OSTI)

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

    2009-07-04

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

  20. 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

  1. 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.

  2. 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.

  3. 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

  4. 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.; Novosel’tsev, 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.

  5. 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...

  6. 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.

  7. 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.

  8. 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.

  9. 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

  10. 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.

  11. 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.

  12. 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 Rež, 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.

  13. 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.

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

    SciTech Connect (OSTI)

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

    2015-12-15

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

  15. EA-1977: Acceptance and Disposition of Spent 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 spent nuclear fuel from the Federal Republic of Germany at DOE’s Savannah River Site (SRS) for processing and disposition. This spent 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.

  16. 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.

  17. 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.

  18. 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

  19. Uranium

    SciTech Connect (OSTI)

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

    1981-10-01

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

  20. 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.

  1. 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...

  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

  3. Uranium industry annual 1996

    SciTech Connect (OSTI)

    1997-04-01

    The Uranium Industry Annual 1996 (UIA 1996) provides current statistical data on the US uranium industry`s activities relating to uranium raw materials and uranium marketing. The UIA 1996 is prepared for use by the Congress, Federal and State agencies, the uranium and nuclear electric utility industries, and the public. Data on uranium raw materials activities for 1987 through 1996 including exploration activities and expenditures, EIA-estimated reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities for 1994 through 2006, including purchases of uranium and enrichment services, enrichment feed deliveries, uranium fuel assemblies, filled and unfilled market requirements, uranium imports and exports, and uranium inventories are shown in Chapter 2. A feature article, The Role of Thorium in Nuclear Energy, is included. 24 figs., 56 tabs.

  4. 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. Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

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

  6. 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.

  7. 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.

  8. 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.

  9. 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,...

  10. 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.

  11. Mobile worksystems for decontamination and dismantlement

    SciTech Connect (OSTI)

    Osborn, J.; Bares, L.C.; Thompson, B.R.

    1995-10-01

    Many DOE nuclear facilities have aged beyond their useful lifetimes. They need to be decommissioned in order to be safe for human presence in the short term, to eventually recover valuable materials they contain, and ultimately to be transitioned to alternative uses or green field conditions. Decontamination and dismantlement are broad classes of activities that will enable these changes to occur. Most of these facilities - uranium enrichment plants, weapons assembly plants, research and production reactors, and fuel recycling facilities - are dormant, though periodic inspection, surveillance and maintenance activities within them are on-going. DOE estimates that there are over 5000 buildings that require deactivation to reduce the costs of performing such work with manual labor. In the long term, 1200 buildings will be decommissioned, and millions of metric tons of metal and concrete will have to be recycled or disposed of. The magnitude of the problem calls for new approaches that are far more cost effective than currently available techniques. This paper describes a mobile workstation termed ROSIE, which provides remote work capabilities for D&D activities.

  12. 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

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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

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

    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

  18. 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.

  19. Planning, Preparation, and Transport of the High-Enriched Uranium Spent Nuclear Fuel from the Czech Republic to the Russian Federation

    SciTech Connect (OSTI)

    M. J. Tyacke; I. Bolshinsky; Frantisek Svitak

    2007-10-01

    The United States, Russian Federation, and the International Atomic Energy Agency have been working together on a program called the Russian Research Reactor Fuel Return (RRRFR) Program, which is part of the Global Threat Reduction Initiative. The purpose of this program is to return Soviet or Russian-supplied high-enriched uranium (HEU) fuel, currently stored at Russian-designed research reactors throughout the world, to Russia. In February 2003, the RRRFR Program began discussions with the Nuclear Research Institute (NRI) in Rež, Czech Republic, about returning their HEU spent nuclear fuel to the Russian Federation for reprocessing. In March 2005, the U.S. Department of Energy signed a contract with NRI to perform all activities needed for transporting their HEU spent nuclear fuel to Russia. After 2 years of intense planning, preparations, and coordination at NRI and with three other countries, numerous organizations and agencies, and a Russian facility, this shipment is scheduled for completion before the end of 2007. This paper will provide a summary of activities completed for making this international shipment. This paper contains an introduction and background of the RRRFR Program and the NRI shipment project. It summarizes activities completed in preparation for the shipment, including facility preparations at NRI in Rež and FSUE “Mayak” in Ozyorsk, Russia; a new transportation cask system; regulatory approvals; transportation planning and preparation in the Czech Republic, Slovakia, Ukraine, and the Russian Federation though completion of the Unified Project and Special Ecological Programs. The paper also describes fuel loading and cask preparations at NRI and final preparations/approvals for transporting the shipment across the Czech Republic, Slovakia, Ukraine, and the Russian Federation to FSUE Mayak where the HEU spent nuclear fuel will be processed, the uranium will be downblended and made into low-enriched uranium fuel for commercial reactor

  20. Y-12 Plant Decontamination and Decommissioning Technology Logic Diagram for Building 9201-4. Volume 1: Technology evaluation

    SciTech Connect (OSTI)

    1994-09-01

    During World War 11, the Oak Ridge Y-12 Plant was built as part of the Manhattan Project to supply enriched uranium for weapons production. In 1945, Building 9201-4 (Alpha-4) was originally used to house a uranium isotope separation process based on electromagnetic separation technology. With the startup of the Oak Ridge K-25 Site gaseous diffusion plant In 1947, Alpha-4 was placed on standby. In 1953, the uranium enrichment process was removed, and installation of equipment for the Colex process began. The Colex process--which uses a mercury solvent and lithium hydroxide as the lithium feed material-was shut down in 1962 and drained of process materials. Residual Quantities of mercury and lithium hydroxide have remained in the process equipment. Alpha-4 contains more than one-half million ft{sup 2} of floor area; 15,000 tons of process and electrical equipment; and 23,000 tons of insulation, mortar, brick, flooring, handrails, ducts, utilities, burnables, and sludge. Because much of this equipment and construction material is contaminated with elemental mercury, cleanup is necessary. The goal of the Y-12 Plant Decontamination and Decommissioning Technology Logic Diagram for Building 9201-4 is to provide a planning document that relates decontamination and decommissioning and waste management problems at the Alpha-4 building to the technologies that can be used to remediate these problems. The Y-12 Plant Decontamination and Decommissioning Technology Logic Diagram for Building 9201-4 builds on the methodology transferred by the U.S. Air Force to the Environmental Management organization with DOE and draws from previous technology logic diagram-efforts: logic diagrams for Hanford, the K-25 Site, and ORNL.

  1. 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.

  2. 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

  3. Reactive decontamination formulation

    DOE Patents [OSTI]

    Giletto, Anthony; White, William; Cisar, Alan J.; Hitchens, G. Duncan; Fyffe, James

    2003-05-27

    The present invention provides a universal decontamination formulation and method for detoxifying chemical warfare agents (CWA's) and biological warfare agents (BWA's) without producing any toxic by-products, as well as, decontaminating surfaces that have come into contact with these agents. The formulation includes a sorbent material or gel, a peroxide source, a peroxide activator, and a compound containing a mixture of KHSO.sub.5, KHSO.sub.4 and K.sub.2 SO.sub.4. The formulation is self-decontaminating and once dried can easily be wiped from the surface being decontaminated. A method for decontaminating a surface exposed to chemical or biological agents is also disclosed.

  4. Verification experiment on the downblending of high enriched uranium (HEU) at the Portsmouth Gaseous Diffusion Plant. Digital video surveillance of the HEU feed stations

    SciTech Connect (OSTI)

    Martinez, R.L.; Tolk, K.; Whiting, N.; Castleberry, K.; Lenarduzzi, R.

    1998-08-01

    As part of a Safeguards Agreement between the US and the International Atomic Energy Agency (IAEA), the Portsmouth Gaseous Diffusion Plant, Piketon, Ohio, was added to the list of facilities eligible for the application of IAEA safeguards. Currently, the facility is in the process of downblending excess inventory of HEU to low enriched uranium (LEU) from US defense related programs for commercial use. An agreement was reached between the US and the IAEA that would allow the IAEA to conduct an independent verification experiment at the Portsmouth facility, resulting in the confirmation that the HEU was in fact downblended. The experiment provided an opportunity for the DOE laboratories to recommend solutions/measures for new IAEA safeguards applications. One of the measures recommended by Sandia National Laboratories (SNL), and selected by the IAEA, was a digital video surveillance system for monitoring activity at the HEU feed stations. This paper describes the SNL implementation of the digital video system and its integration with the Load Cell Based Weighing System (LCBWS) from Oak Ridge National Laboratory (ORNL). The implementation was based on commercially available technology that also satisfied IAEA criteria for tamper protection and data authentication. The core of the Portsmouth digital video surveillance system was based on two Digital Camera Modules (DMC-14) from Neumann Consultants, Germany.

  5. Decontamination of process equipment using recyclable chelating solvent

    SciTech Connect (OSTI)

    Jevec, J.; Lenore, C.; Ulbricht, S.

    1995-10-01

    The Department of Energy (DOE) is now faced with the task of meeting decontamination and decommissioning obligations at numerous facilities by the year 2019. Due to the tremendous volume of material involved, innovative decontamination technologies are being sought that can reduce the volumes of contaminated waste materials and secondary wastes requiring disposal. This report describes the results of the performance testing of chelates and solvents for the dissolution of uranium.

  6. Uranium industry annual 1994

    SciTech Connect (OSTI)

    1995-07-05

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

  7. 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.

  8. RECOVERY OF URANIUM FROM PITCHBLENDE

    DOE Patents [OSTI]

    Ruehle, A.E.

    1958-06-24

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

  9. Decontamination and reuse of ORGDP aluminum scrap

    SciTech Connect (OSTI)

    Compere, A.L.; Griffith, W.L.; Hayden, H.W.; Wilson, D.F.

    1996-12-01

    The Gaseous Diffusion Plants, or GDPs, have significant amounts of a number of metals, including nickel, aluminum, copper, and steel. Aluminum was used extensively throughout the GDPs because of its excellent strength to weight ratios and good resistance to corrosion by UF{sub 6}. This report is concerned with the recycle of aluminum stator and rotor blades from axial compressors. Most of the stator and rotor blades were made from 214-X aluminum casting alloy. Used compressor blades were contaminated with uranium both as a result of surface contamination and as an accumulation held in surface-connected voids inside of the blades. A variety of GDP studies were performed to evaluate the amounts of uranium retained in the blades; the volume, area, and location of voids in the blades; and connections between surface defects and voids. Based on experimental data on deposition, uranium content of the blades is 0.3%, or roughly 200 times the value expected from blade surface area. However, this value does correlate with estimated internal surface area and with lengthy deposition times. Based on a literature search, it appears that gaseous decontamination or melt refining using fluxes specific for uranium removal have the potential for removing internal contamination from aluminum blades. A melt refining process was used to recycle blades during the 1950s and 1960s. The process removed roughly one-third of the uranium from the blades. Blade cast from recycled aluminum appeared to perform as well as blades from virgin material. New melt refining and gaseous decontamination processes have been shown to provide substantially better decontamination of pure aluminum. If these techniques can be successfully adapted to treat aluminum 214-X alloy, internal and, possibly, external reuse of aluminum alloys may be possible.

  10. 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

  11. Safeguards considerations related to the decontamination and decommissioning of former nuclear weapons facilities

    SciTech Connect (OSTI)

    Crawford, D.

    1995-12-31

    In response to the post-Cold War environment and the changes in the U. S. Department of Energy defense mission, many former nuclear operations are being permanently shut down. These operations include facilities where nuclear materials production, processing, and weapons manufacturing have occurred in support of the nation`s defense industry. Since defense-related operations have ceased, many of the classification and sensitive information concerns do not exist. However, nuclear materials found at these sites are of interest to the DOE from environmental, safety and health, and materials management perspectives. Since these facilities played a role in defense activities, the nuclear materials found at these facilities are considered special nuclear materials, primarily highly enriched uranium and/or plutonium. Consequently, these materials pose significant diversion, theft, and sabotage threats, and significant nuclear security issues exist that must be addressed. This paper focuses on the nuclear materials protection issues associated with facility decommissioning and decontamination, primarily safeguards.

  12. Paint decontamination kinetics

    SciTech Connect (OSTI)

    Thornton, E.W.

    1984-04-01

    Decontamination kinetics of a high-gloss polyurethane paint have been investigated using a novel flow cell experiment where the sample was counted in situ during decontamination. The /sup 134/Cs, /sup 137/Cs, and /sup 90/Y decontaminations follow a rate law that can be predicted theoretically for contaminant ion desorption from weakly heterogeneous random surface adsorption sites. Paint surfaces show the same decontamination kinetics after damage by abrasion or ultraviolet irradiation prior to contamination. The systems investigated exhibit Freundlich adsorption isotherm behavior during contamination; this is also characteristic of weakly heterogeneous random surfaces and is very commonly observed in ion adsorption studies at low concentrations.

  13. Portsmouth Decontamination & Decommissioning

    Broader source: Energy.gov [DOE]

    The Decontamination & Decommissioning (D&D) Program at the Portsmouth Site addresses potential future demolition and disposal of approximately 415 facilities (including buildings, utilities...

  14. 2015 Uranium Marketing Annual Survey

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

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

  15. 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

  16. 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...

  17. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

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

  18. Uranium industry annual 1995

    SciTech Connect (OSTI)

    1996-05-01

    The Uranium Industry Annual 1995 (UIA 1995) provides current statistical data on the U.S. uranium industry`s activities relating to uranium raw materials and uranium marketing. The UIA 1995 is prepared for use by the Congress, Federal and State agencies, the uranium and nuclear electric utility industries, and the public. It contains data for the period 1986 through 2005 as collected on the Form EIA-858, ``Uranium Industry Annual Survey``. Data collected on the ``Uranium Industry Annual Survey`` provide a comprehensive statistical characterization of the industry`s plans and commitments for the near-term future. Where aggregate data are presented in the UIA 1995, care has been taken to protect the confidentiality of company-specific information while still conveying accurate and complete statistical data. Data on uranium raw materials activities for 1986 through 1995 including exploration activities and expenditures, EIA-estimated reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities for 1994 through 2005, including purchases of uranium and enrichment services, enrichment feed deliveries, uranium fuel assemblies, filled and unfilled market requirements, uranium imports and exports, and uranium inventories are shown in Chapter 2. The methodology used in the 1995 survey, including data edit and analysis, is described in Appendix A. The methodologies for estimation of resources and reserves are described in Appendix B. A list of respondents to the ``Uranium Industry Annual Survey`` is provided in Appendix C. For the reader`s convenience, metric versions of selected tables from Chapters 1 and 2 are presented in Appendix D along with the standard conversion factors used. A glossary of technical terms is at the end of the report. 14 figs., 56 tabs.

  19. Audit Report: OAS-FS-09-06

    Broader source: Energy.gov [DOE]

    Uranium Enrichment Decontamination and Decommissioning Fund's Fiscal Year 2006 and 2005 Financial Statement Audit

  20. Audit Report: OAS-FS-13-02

    Office of Energy Efficiency and Renewable Energy (EERE)

    Management Letter on the Uranium Enrichment Decontamination and Decommissioning Fund's Fiscal Year 2011 Financial Statement Audit

  1. Long lasting decontamination foam

    DOE Patents [OSTI]

    Demmer, Ricky L.; Peterman, Dean R.; Tripp, Julia L.; Cooper, David C.; Wright, Karen E.

    2010-12-07

    Compositions and methods for decontaminating surfaces are disclosed. More specifically, compositions and methods for decontamination using a composition capable of generating a long lasting foam are disclosed. Compositions may include a surfactant and gelatin and have a pH of less than about 6. Such compositions may further include affinity-shifting chemicals. Methods may include decontaminating a contaminated surface with a composition or a foam that may include a surfactant and gelatin and have a pH of less than about 6.

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

    SciTech Connect (OSTI)

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

    1990-01-01

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

  3. 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.

  4. Lessons Learned from Decontamination Experiences

    SciTech Connect (OSTI)

    Sorensen, JH

    2000-11-16

    This interim report describes a DOE project currently underway to establish what is known about decontamination of buildings and people and the procedures and protocols used to determine when and how people or buildings are considered ''clean'' following decontamination. To fulfill this objective, the study systematically examined reported decontamination experiences to determine what procedures and protocols are currently employed for decontamination, the timeframe involved to initiate and complete the decontamination process, how the contaminants were identified, the problems encountered during the decontamination process, how response efforts of agencies were coordinated, and the perceived social psychological effects on people who were decontaminated or who participated in the decontamination process. Findings and recommendations from the study are intended to aid decision-making and to improve the basis for determining appropriate decontamination protocols for recovery planners and policy makers for responding to chemical and biological events.

  5. 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

  6. Portsmouth Decommissioning and Decontamination Project Director...

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

    Decommissioning and Decontamination Project Director's Final Findings and Order Portsmouth Decommissioning and Decontamination Project Director's Final Findings and Order...

  7. Oxidative Tritium Decontamination System

    DOE Patents [OSTI]

    Gentile, Charles A. , Guttadora, Gregory L. , Parker, John J.

    2006-02-07

    The Oxidative Tritium Decontamination System, OTDS, provides a method and apparatus for reduction of tritium surface contamination on various items. The OTDS employs ozone gas as oxidizing agent to convert elemental tritium to tritium oxide. Tritium oxide vapor and excess ozone gas is purged from the OTDS, for discharge to atmosphere or transport to further process. An effluent stream is subjected to a catalytic process for the decomposition of excess ozone to diatomic oxygen. One of two configurations of the OTDS is employed: dynamic apparatus equipped with agitation mechanism and large volumetric capacity for decontamination of light items, or static apparatus equipped with pressurization and evacuation capability for decontamination of heavier, delicate, and/or valuable items.

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

    Office of Environmental Management (EM)

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

  9. 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

  10. RUTHENIUM DECONTAMINATION METHOD

    DOE Patents [OSTI]

    Gresky, A.T.

    1960-07-19

    A liquid-liquid extraction method of separating uranium from fission products is given. A small amount of a low molecular weight ketone is added to an acidic aqueous solution containing neutron-irradiated uranium and its associated fission products. The resulting solution is digested and then contacted with an organic liquid that extracts uranium values. The purpose of the step of digesting the aqueous solution in the presence of the ketone is to suppress the extractability of ruthenium.

  11. Electrokinetic decontamination of concrete

    SciTech Connect (OSTI)

    Lomasney, H.

    1995-10-01

    The U.S. Department of Energy has assigned a priority to the advancement of technology for decontaminating concrete surfaces which have become contaminated with radionuclides, heavy metals, and toxic organics. This agency is responsible for decontamination and decommissioning of thousands of buildings. Electrokinetic extraction is one of the several innovative technologies which emerged in response to this initiative. This technique utilizes an electropotential gradient and the subsequent electrical transport mechanism to cause the controlled movement of ionics species, whereby the contaminants exit the recesses deep within the concrete. This report discusses the technology and use at the Oak Ridge k-25 plant.

  12. EIS-0359: Uranium Hexafluoride Conversion Facility at the Paducah, Kentucky

    Energy Savers [EERE]

    Site | Department of Energy 59: Uranium Hexafluoride Conversion Facility at the Paducah, Kentucky Site EIS-0359: Uranium Hexafluoride Conversion Facility at the Paducah, Kentucky Site Summary This site-specific EIS considers the construction, operation, maintenance, and decontamination and decommissioning of the proposed depleted uranium hexafluoride (DUF6) conversion facility at three locations within the Paducah site; transportation of depleted uranium conversion products and waste

  13. 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)

  14. 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.

  15. Decontaminating metal surfaces

    DOE Patents [OSTI]

    Childs, E.L.

    1984-01-23

    Radioactively contaminated surfaces can be electrolytically decontaminated with greatly increased efficiencies by using electrolytes containing higher than heretofore conventional amounts of nitrate, e.g., >600 g/1 of NaNO/sub 3/, or by using nitrate-containing electrolytes which are acidic, e.g., of a pH < 6.

  16. Decontaminating metal surfaces

    DOE Patents [OSTI]

    Childs, Everett L.

    1984-11-06

    Radioactively contaminated surfaces can be electrolytically decontaminated with greatly increased efficiencies by using electrolytes containing higher than heretofore conventional amounts of nitrate, e.g.,>600 g/l of NaNO.sub.3, or by using nitrate-containing electrolytes which are acidic, e.g., of a pH<6.

  17. 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.

  18. Laboratory Enrichment of Radioactive Assemblages and Estimation of Thorium and Uranium Radioactivity in Fractions Separated from Placer Sands in Southeast Bangladesh

    SciTech Connect (OSTI)

    Sasaki, Takayuki; Rajib, Mohammad; Akiyoshi, Masafumi; Kobayashi, Taishi; Takagi, Ikuji; Fujii, Toshiyuki; Zaman, Md. Mashrur

    2015-06-15

    The present study reports the likely first attempt of separating radioactive minerals for estimation of activity concentration in the beach placer sands of Bangladesh. Several sand samples from heavy mineral deposits located at the south-eastern coastal belt of Bangladesh were processed to physically upgrade their radioactivity concentrations using plant and laboratory equipment. Following some modified flow procedure, individual fractions were separated and investigated using gamma-ray spectrometry and powder-XRD analysis. The radioactivity measurements indicated contributions of the thorium and uranium radioactive series and of {sup 40}K. The maximum values of {sup 232}Th and {sup 238}U, estimated from the radioactivity of {sup 208}Tl and {sup 234}Th in secular equilibrium, were found to be 152,000 and 63,300 Bq/kg, respectively. The fraction of the moderately conductive part in electric separation contained thorium predominantly, while that of the non-conductive part was found to be uranium rich. The present arrangement of the pilot plant cascade and the fine tuning of setting parameters were found to be effective and economic separation process of the radioactive minerals from placer sands in Bangladesh. Probable radiological impacts and extraction potentiality of such radioactive materials are also discussed.

  19. 2015 Uranium Marketing Annual Report

    Gasoline and Diesel Fuel Update (EIA)

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

  20. Gas phase decontamination of gaseous diffusion process equipment

    SciTech Connect (OSTI)

    Bundy, R.D.; Munday, E.B.; Simmons, D.W.; Neiswander, D.W.

    1994-03-01

    D&D of the process facilities at the gaseous diffusion plants (GDPs) will be an enormous task. The EBASCO estimate places the cost of D&D of the GDP at the K-25 Site at approximately $7.5 billion. Of this sum, nearly $4 billion is associated with the construction and operation of decontamination facilities and the dismantlement and transport of contaminated process equipment to these facilities. In situ long-term low-temperature (LTLT) gas phase decontamination is being developed and demonstrated at the K-25 site as a technology that has the potential to substantially lower these costs while reducing criticality and safeguards concerns and worker exposure to hazardous and radioactive materials. The objective of gas phase decontamination is to employ a gaseous reagent to fluorinate nonvolatile uranium deposits to form volatile LJF6, which can be recovered by chemical trapping or freezing. The LTLT process permits the decontamination of the inside of gas-tight GDP process equipment at room temperature by substituting a long exposure to subatmospheric C1F for higher reaction rates at higher temperatures. This paper outlines the concept for applying LTLT gas phase decontamination, reports encouraging laboratory experiments, and presents the status of the design of a prototype mobile system. Plans for demonstrating the LTLT process on full-size gaseous diffusion equipment are also outlined briefly.

  1. Decontamination impacts on solidification

    SciTech Connect (OSTI)

    Piciulo, P.L.; Davis, M.S.

    1985-01-01

    The increased occupational exposure resulting from the accumulation of activated corrosion products in the primary system of LWRs has led to the development of chemical methods to remove the contamination. In the past, the problem of enhanced migration of radionuclides away from trenches used to dispose of low-level radioactive waste, has been linked to the presence, at the disposal unit, of chelating or complexing agents such as those used in decontamination processes. These agents have further been found to reduce the normal sorptive capacity of soils for radionuclides. The degree to which these agents inhibit the normal sorptive processes is dependent on the type of complexing agent, the radionuclide of concern, the soil properties and whether the nuclide is present as a complex or is already sorbed to the soil. Since the quantity of reagent employed in a full system decontamination is large (200 to 25,000 kg), the potential for enhanced migration of radionuclides from a site used to dispose of the decontamination wastes should be addressed and guidelines established for the safe disposal of these wastes.

  2. 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.

  3. Integrated decontamination process for metals

    DOE Patents [OSTI]

    Snyder, Thomas S.; Whitlow, Graham A.

    1991-01-01

    An integrated process for decontamination of metals, particularly metals that are used in the nuclear energy industry contaminated with radioactive material. The process combines the processes of electrorefining and melt refining to purify metals that can be decontaminated using either electrorefining or melt refining processes.

  4. Excess Uranium Management | Department of Energy

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

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

  5. Uranium at Y-12: Accountability | Y-12 National Security Complex

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

    ... Uranium at Y-12: Accountability Posted: July 22, 2013 - 3:37pm | Y-12 Report | Volume 10, Issue 1 | 2013 Accountability of enriched uranium is facilitated by the ability to put ...

  6. 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 ...

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

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

    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 ...

  8. Decontamination of plutonium from water with chitin

    DOE Patents [OSTI]

    Silver, Gary L.

    1978-01-01

    The invention relates to a process for decontaminating or removing radionuclides from aqueous solution.

  9. Granulated decontamination formulations

    DOE Patents [OSTI]

    Tucker, Mark D.

    2007-10-02

    A decontamination formulation and method of making that neutralizes the adverse health effects of both chemical and biological compounds, especially chemical warfare (CW) and biological warfare (BW) agents, and toxic industrial chemicals. The formulation provides solubilizing compounds that serve to effectively render the chemical and biological compounds, particularly CW and BW compounds, susceptible to attack, and at least one reactive compound that serves to attack (and detoxify or kill) the compound. The formulation includes at least one solubilizing agent, a reactive compound, a sorbent additive, and water. A highly adsorbent sorbent additive (e.g., amorphous silica, sorbitol, mannitol, etc.) is used to "dry out" one or more liquid ingredients into a dry, free-flowing powder that has an extended shelf life, and is more convenient to handle and mix in the field.

  10. Glovebox decontamination technology comparison

    SciTech Connect (OSTI)

    Quintana, D.M.; Rodriguez, J.B.; Cournoyer, M.E.

    1999-09-26

    Reconfiguration of the CMR Building and TA-55 Plutonium Facility for mission requirements will require the disposal or recycle of 200--300 gloveboxes or open front hoods. These gloveboxes and open front hoods must be decontaminated to meet discharge limits for Low Level Waste. Gloveboxes and open front hoods at CMR have been painted. One of the deliverables on this project is to identify the best method for stripping the paint from large numbers of gloveboxes. Four methods being considered are the following: conventional paint stripping, dry ice pellets, strippable coatings, and high pressure water technology. The advantages of each technology will be discussed. Last, cost comparisons between the technologies will be presented.

  11. Decontamination & decommissioning focus area

    SciTech Connect (OSTI)

    1996-08-01

    In January 1994, the US Department of Energy Office of Environmental Management (DOE EM) formally introduced its new approach to managing DOE`s environmental research and technology development activities. The goal of the new approach is to conduct research and development in critical areas of interest to DOE, utilizing the best talent in the Department and in the national science community. To facilitate this solutions-oriented approach, the Office of Science and Technology (EM-50, formerly the Office of Technology Development) formed five Focus AReas to stimulate the required basic research, development, and demonstration efforts to seek new, innovative cleanup methods. In February 1995, EM-50 selected the DOE Morgantown Energy Technology Center (METC) to lead implementation of one of these Focus Areas: the Decontamination and Decommissioning (D & D) Focus Area.

  12. Background Fact Sheet Transfer of Depleted Uranium and Subsequent...

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

    The Department's National Nuclear Security Administration (NNSA) requires U.S.-origin unobligated domestically enriched, domestic-origin uranium to support continued tritium ...

  13. 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). ...

  14. 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}).

  15. WIPP Begins Underground Decontamination Activities

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

    Yellow brattice cloth is suspended from the ceiling in this disposal room. It is rolled down to prevent air flow to the room. Brattice cloth also will serve as a barrier to decontaminate floors. WIPP UPDATE: March 13, 2015 WIPP Begins Underground Decontamination Activities Activities are underway in WIPP's underground facility to address the radioactive contamination that remains as a result of the February 14, 2014 event. Employees are using a modified piece of agricultural spraying equipment

  16. Large-bore pipe decontamination

    SciTech Connect (OSTI)

    Ebadian, M.A.

    1998-01-01

    The decontamination and decommissioning (D and D) of 1200 buildings within the US Department of Energy-Office of Environmental Management (DOE-EM) Complex will require the disposition of miles of pipe. The disposition of large-bore pipe, in particular, presents difficulties in the area of decontamination and characterization. The pipe is potentially contaminated internally as well as externally. This situation requires a system capable of decontaminating and characterizing both the inside and outside of the pipe. Current decontamination and characterization systems are not designed for application to this geometry, making the direct disposal of piping systems necessary in many cases. The pipe often creates voids in the disposal cell, which requires the pipe to be cut in half or filled with a grout material. These methods are labor intensive and costly to perform on large volumes of pipe. Direct disposal does not take advantage of recycling, which could provide monetary dividends. To facilitate the decontamination and characterization of large-bore piping and thereby reduce the volume of piping required for disposal, a detailed analysis will be conducted to document the pipe remediation problem set; determine potential technologies to solve this remediation problem set; design and laboratory test potential decontamination and characterization technologies; fabricate a prototype system; provide a cost-benefit analysis of the proposed system; and transfer the technology to industry. This report summarizes the activities performed during fiscal year 1997 and describes the planned activities for fiscal year 1998. Accomplishments for FY97 include the development of the applicable and relevant and appropriate regulations, the screening of decontamination and characterization technologies, and the selection and initial design of the decontamination system.

  17. 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.

  18. India's Worsening Uranium Shortage

    SciTech Connect (OSTI)

    Curtis, Michael M.

    2007-01-15

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

  19. U.S. Uranium Down-blending Activities: Fact Sheet | National Nuclear

    National Nuclear Security Administration (NNSA)

    Security Administration | (NNSA) Uranium Down-blending Activities: Fact Sheet March 23, 2012 The permanent disposition of Highly Enriched Uranium (HEU) permanently reduces nuclear security vulnerabilities. In 1996, the Department of Energy (DOE) announced plans to reduce stockpiles of surplus HEU by down-blending, or converting, it to low-enriched uranium (LEU). U.S. HEU Downblending The U.S. Surplus Highly Enriched Uranium (HEU) Disposition Program is making progress in disposing of surplus

  20. 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.

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

    National Nuclear Security Administration (NNSA)

    Administration | (NNSA) '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

  2. DECONTAMINATION TECHNOLOGIES FOR FACILITY REUSE

    SciTech Connect (OSTI)

    Bossart, Steven J.; Blair, Danielle M.

    2003-02-27

    As nuclear research and production facilities across the U.S. Department of Energy (DOE) nuclear weapons complex are slated for deactivation and decommissioning (D&D), there is a need to decontaminate some facilities for reuse for another mission or continued use for the same mission. Improved technologies available in the commercial sector and tested by the DOE can help solve the DOE's decontamination problems. Decontamination technologies include mechanical methods, such as shaving, scabbling, and blasting; application of chemicals; biological methods; and electrochemical techniques. Materials to be decontaminated are primarily concrete or metal. Concrete materials include walls, floors, ceilings, bio-shields, and fuel pools. Metallic materials include structural steel, valves, pipes, gloveboxes, reactors, and other equipment. Porous materials such as concrete can be contaminated throughout their structure, although contamination in concrete normally resides in the top quarter-inch below the surface. Metals are normally only contaminated on the surface. Contamination includes a variety of alpha, beta, and gamma-emitting radionuclides and can sometimes include heavy metals and organic contamination regulated by the Resource Conservation and Recovery Act (RCRA). This paper describes several advanced mechanical, chemical, and other methods to decontaminate structures, equipment, and materials.

  3. 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...

  4. 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...

  5. 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...

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

    National Nuclear Security Administration (NNSA)

    Congressmen tour Y-12 facilities During a recent visit to the Y-12 National Security Complex, Rep. Mike Simpson (R-Idaho), chairman of the House Energy and Water Appropriations ...

  7. 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. ...

  8. 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.

  9. Decontamination and decommissioning of the Mayaguez (Puerto Rico) facility

    SciTech Connect (OSTI)

    Jackson, P.K.; Freemerman, R.L.

    1989-11-01

    On February 6, 1987 the US Department of Energy (DOE) awarded the final phase of the decontamination and decommissioning of the nuclear and reactor facilities at the Center for Energy and Environmental Research (CEER), in Mayaguez, Puerto Rico. Bechtel National, Inc., was made the decontamination and decommissioning (D and D) contractor. The goal of the project was to enable DOE to proceed with release of the CEER facility for use by the University of Puerto Rico, who was the operator. This presentation describes that project and lesson learned during its progress. The CEER facility was established in 1957 as the Puerto Rico Nuclear Center, a part of the Atoms for Peace Program. It was a nuclear training and research institution with emphasis on the needs of Latin America. It originally consisted of a 1-megawatt Materials Testing Reactor (MTR), support facilities and research laboratories. After eleven years of operation the MTR was shutdown and defueled. A 2-megawatt TRIGA reactor was installed in 1972 and operated until 1976, when it woo was shutdown. Other radioactive facilities at the center included a 10-watt homogeneous L-77 training reactor, a natural uranium graphite-moderated subcritical assembly, a 200KV particle accelerator, and a 15,000 Ci Co-60 irradiation facility. Support facilities included radiochemistry laboratories, counting rooms and two hot cells. As the emphasis shifted to non-nuclear energy technology a name change resulted in the CEER designation, and plans were started for the decontamination and decommissioning effort.

  10. URANIUM ALLOYS

    DOE Patents [OSTI]

    Colbeck, E.W.

    1959-12-29

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

  11. Decontamination of steel by melt refining: A literature review

    SciTech Connect (OSTI)

    Ozturk, B.; Fruehan, R.J.

    1994-12-31

    It has been reported that a large amount of metal waste is produced annually by nuclear fuel processing and nuclear power plants. These metal wastes are contaminated with radioactive elements, such as uranium and plutonium. Current Department of Energy guidelines require retrievable storage of all metallic wastes containing transuranic elements above a certain level. Because of high cost, it is important to develop an effective decontamination and volume reduction method for low level contaminated metals. It has been shown by some investigators that a melt refining technique can be used for the processing of the contaminated metal wastes. In this process, contaminated metal is melted wit a suitable flux. The radioactive elements are oxidized and transferred to a slag phase. In order to develop a commercial process it is important to have information on the thermodynamics and kinetics of the removal. Therefore, a literature search was carried out to evaluate the available information on the decontamination uranium and transuranic-contaminated plain steel, copper and stainless steel by melt a refining technique. Emphasis was given to the thermodynamics and kinetics of the removal. Data published in the literature indicate that it is possible to reduce the concentration of radioactive elements to a very low level by the melt refining method. 20 refs.

  12. 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

  13. 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

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

    Office of Scientific and Technical Information (OSTI)

    In the end, it took the combined efforts of all three of these facilities to produce enough enriched uranium for the one and only uranium atomic bomb produced during the war. The ...

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

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

    Y-12 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

  17. Electrolytic decontamination of conductive materials

    SciTech Connect (OSTI)

    Nelson, T.O.; Campbell, G.M.; Parker, J.L.; Getty, R.H.; Hergert, T.R.; Lindahl, K.A.; Peppers, L.G.

    1993-10-01

    Using the electrolytic method, the authors have demonstrated removal of Pu from contaminated conductive material. At EG&G Rocky Flats, they electrolytically decontaminated stainless steel. Results from this work show removal of fixed contamination, including the following geometries: planar, large radius, bolt holes, glove ports, and protruding studs. More specifically, fixed contamination was reduced from levels ranging > 1,000,000 counts per minute (cpm) down to levels ranging from 1,500 to < 250 cpm with the electrolytic method. More recently, the electrolytic work has continued at LANL as a joint project with EG&G. Impressively, electrolytic decontamination experiments on removal of Pu from oralloy coupons have shown decreases in swipable contamination that initially ranged from 500,000 to 1,500,000 disintegrations per minute (dpm) down to 0--2 dpm.

  18. Portsmouth Decontamination and Decommissioning Record of Decision |

    Energy Savers [EERE]

    Department of Energy Decontamination and Decommissioning Record of Decision Portsmouth Decontamination and Decommissioning Record of Decision The Ohio Environmental Protection Agency (Ohio EPA) and the U.S. Department of Energy (DOE) have agreed upon a plan to demolish the large process buildings and other facilities undergoing decontamination and decommissioning (D&D) at the Portsmouth Gaseous Diffusion Plant. The Record of Decision (ROD) for the Process Buildings and Complex Facilities

  19. DECONTAMINATION DRESSDOWN AT A TRANSPORTATION ACCIDENT INVOLVING...

    Office of Environmental Management (EM)

    Video User' s Guide DECONTAMINATION DRESSDOWN AT A TRANSPORTATION ACCIDENT INVOLVING ... related to emergency response to a transportation accident involving radioactive material. ...

  20. Properties and solidification of decontamination wastes

    SciTech Connect (OSTI)

    Davis, M.S.; Piciulo, P.L.; Bowerman, B.S.; Adams, J.W.; Milian, L.

    1983-01-01

    LWRs will require one or more chemical decontaminations to achieve their designed lifetimes. Primary system decontamination is designed to lower radiation fields in areas where plant maintenance personnel must work. Chemical decontamination methods are either hard (concentrated chemicals, approximately 5 to 25 weight percent) or soft (dilute chemicals less than 1 percent by weight). These methods may have different chemical reagents, some tailor-made to the crud composition and many methods are and will be proprietary. One factor common to most commercially available processes is the presence of organic acids and chelates. These types of organic reagents are known to enhance the migration of radionuclides after disposal in a shallow land burial site. The NRC sponsors two programs at Brookhaven National Laboratory that are concerned with the management of decontamination wastes which will be generated by the full system decontamination of LWRs. These two programs focus on potential methods for degrading or converting decontamination wastes to more acceptable forms prior to disposal and the impact of disposing of solidified decontamination wastes. The results of the solidification of simulated decontamination resin wastes will be presented. Recent results on combustion of simulated decontamintion wastes will be described and procedures for evaluating the release of decontamination reagents from solidified wastes will be summarized.

  1. Joint Statement on Multinational Cooperation on High-Density Low-Enriched

    National Nuclear Security Administration (NNSA)

    Uranium Fuel Development | National Nuclear Security Administration | (NNSA) Joint Statement on Multinational Cooperation on High-Density Low-Enriched Uranium Fuel Development March 25, 2014 The White House Office of the Press Secretary Belgium, France, Germany, the Republic of Korea and the United States, the parties to this joint statement recognize that the ultimate goal of nuclear security is advanced by minimizing highly-enriched uranium (HEU) in civilian use, which is affirmed in the

  2. Proceedings of the 1988 International Meeting on Reduced Enrichment for Research and Test Reactors

    SciTech Connect (OSTI)

    Not Available

    1993-07-01

    The international effort to develop and implement new research reactor fuels utilizing low-enriched uranium, instead of highly- enriched uranium, continues to make solid progress. This effort is the cornerstone of a widely shared policy aimed at reducing, and possibly eliminating, international traffic in highly-enriched uranium and the nuclear weapon proliferation concerns associated with this traffic. To foster direct communication and exchange of ideas among the specialists in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the eleventh of a series which began 1978. Individual papers presented at the meeting have been cataloged separately.

  3. 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.

  4. 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.

  5. Uranium Processing Facility | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Uranium Processing Facility

  6. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

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

  7. 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)

  8. Uranium Isotopic Assay Instrument

    SciTech Connect (OSTI)

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

    2006-12-01

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

  9. Innovative Decontamination Technology for Use in Gaseous Diffusion Plant Decommissioning

    SciTech Connect (OSTI)

    Peters, M.J.; Norton, C.J.; Fraikor, G.B.; Potter, G.L.; Chang, K.C.

    2006-07-01

    The results of bench scale tests demonstrated that TechXtract{sup R} RadPro{sup TM} technology (hereinafter referred to as RadPro{sup R}) can provide 100% coverage of complex mockup gaseous diffusion plant (GDP) equipment and can decontaminate uranium (U) deposits with 98% to 99.99% efficiency. Deployment tests demonstrated RadPro{sup R} can be applied as foam, mist/fog, or steam, and fully cover the internal surfaces of complex mockup equipment, including large piping. Decontamination tests demonstrated that two formulations of RadPro{sup R}, one with neutron attenuators and one without neutron attenuators, could remove up to 99.99% of uranyl fluoride deposits, one of the most difficult to remove deposits in GDP equipment. These results were supplemented by results from previous tests conducted in 1994 that showed RadPro{sup R} could remove >97% of U and Tc-99 contamination from actual GDP components. Operational use of RadPro{sup R} at other DOE and commercial facilities also support these data. (authors)

  10. On the flexibility of high temperature reactor cores for high-and low-enriched fuel

    SciTech Connect (OSTI)

    Bzandes, S.; Lonhert, G.

    1982-07-01

    The operational flexibility of a high temperature reactor (HTR) is not restricted to either a low- or a high-enriched fuel cycle. Both fuel cycles are possible for the same core design. The fuel cycle cost is, however, penalized for low-enriched fuel; in addition, higher uranium consumption is required. Hence, an HTR is most economical to operate in the high-enriched thorium-uranium fuel cycle.

  11. Uranium industry annual 1997

    SciTech Connect (OSTI)

    1998-04-01

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

  12. 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.

  13. 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.

  14. Defense Program Awards of Excellence: Y-12 Uranium Mission Strategy...

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

    Making sure we can deliver enriched uranium for our missions is no easy task at a site ... necessary for success. The Enterprise's missions can't wait, so the team had to develop a ...

  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. ''Green'' Biopolymers for Improved Decontamination of Metals from Surfaces: Sorptive Characterization and Coating Properties

    SciTech Connect (OSTI)

    Davison, Brian H.; Kuritz, Tanya

    2000-06-01

    The proposed research aims to develop a fundamental understanding of important biological and physical chemical parameters for effective decontamination of metal surfaces using environmentally benign aqueous-based biopolymer solutions. Understanding how heavy metal-chelating biopolymers coat and interact with contaminated surfaces will benefit the development of novel, safe, easy-to-apply decontamination methodologies for removal of radionuclides and heavy metals. The benefits of these methodologies will include the following: (1) decreased exposure hazards for workers; (2) decreased secondary waste generation; (3) increased efficiency of decontamination; (4) positive public appeal and development of novel, nature-friendly business opportunities; and (5) lower cost of cleanup to the government. We propose to use aqueous biopolymer solutions to coat a contaminated metal surface (i.e., steel), solubilize the heavy metals (e.g., uranium) from the surface, and bind the heavy metals into the biopolymer. The biopolymer coating (containing the immobilized hazardous metal contaminants) will then be removed as a viscous film, as a dry powder, or by washing. This ''apply, wait, and remove'' procedure will reduce the amount of worker time spent in decontamination activities.

  17. ''Green'' Biopolymers for Improved Decontamination of Metals from Surfaces: Sorptive Characterization and Coating Properties

    SciTech Connect (OSTI)

    Davison, Brian H.

    2002-04-30

    The proposed research aimed to develop a fundamental understanding of important biological and physical chemical parameters for effective decontamination of metal surfaces using environmentally benign aqueous-based biopolymer solutions. Understanding how heavy metal-chelating biopolymers coat and interact with contaminated surfaces will benefit the development of novel, safe, easy-to-apply decontamination methodologies for removal of radionuclides and heavy metals. The benefits of these methodologies include the following: decreased exposure hazards for workers; decreased secondary waste generation; increased efficiency of decontamination; positive public appeal and development of novel, nature-friendly business opportunities; and lower cost of cleanup to the government. We proposed to use aqueous biopolymer solutions to coat a contaminated metal surface (i.e., steel), solubilize the heavy metals (e.g., uranium) from the surface, and bind the heavy metals into the biopolymer. The biopolymer coating (containing the immobilized hazardous metal contaminants) was to be removed as a viscous film, as a dry powder, or by washing. This ''apply, wait, and remove'' procedure will reduce the amount of worker time spent in decontamination activities.

  18. ''Green'' Biopolymers for Improved Decontamination of Metals from Surfaces: Sorptive Characterization and Coating Properties

    SciTech Connect (OSTI)

    Davison, Brian H.; Kurtiz,Tanya

    1999-06-01

    The proposed research aims to develop a fundamental understanding of important biological and physical chemical parameters for effective decontamination of metal surfaces using environmentally benign aqueous-based biopolymer solutions. Understanding how heavy metal-chelating biopolymers coat and interact with contaminated surfaces will benefit the development of novel, safe, easy-to-apply decontamination methodologies for removal of radionuclides and heavy metals. The benefits of these methodologies will include the following: (1) decreased exposure hazards for workers; (2) decreased secondary waste generation; (3) increased efficiency of decontamination; (4) positive public appeal and development of novel, nature-friendly business opportunities; and (5) lower cost of cleanup to the government. We propose to use aqueous biopolymer solutions to coat a contaminated metal surface (i.e., steel), solubilize the heavy metals (e.g., uranium) from the surface, and bind the heavy metals into the biopolymer. The biopolymer coating (containing the immobilized hazardous metal contaminants) will then be removed as a viscous film, as a dry powder, or by washing. This ''apply, wait, and remove'' procedure will reduce the amount of worker time spent in decontamination activities.

  19. Decontamination formulation with sorbent additive

    DOE Patents [OSTI]

    Tucker; Mark D. , Comstock; Robert H.

    2007-10-16

    A decontamination formulation and method of making that neutralizes the adverse health effects of both chemical and biological compounds, especially chemical warfare (CW) and biological warfare (BW) agents, and toxic industrial chemicals. The formulation provides solubilizing compounds that serve to effectively render the chemical and biological compounds, particularly CW and BW compounds, susceptible to attack, and at least one reactive compound that serves to attack (and detoxify or kill) the compound. The formulation includes at least one solubilizing agent, a reactive compound, a bleaching activator, a sorbent additive, and water. The highly adsorbent, water-soluble sorbent additive (e.g., sorbitol or mannitol) is used to "dry out" one or more liquid ingredients, such as the liquid bleaching activator (e.g., propylene glycol diacetate or glycerol diacetate) and convert the activator into a dry, free-flowing powder that has an extended shelf life, and is more convenient to handle and mix in the field.

  20. Savannah River Laboratory Decontamination Program

    SciTech Connect (OSTI)

    Rankin, W.N.

    1991-01-01

    Savannah River Laboratory (SRL) has had a Decontamination and Decommissioning (D D) Technology program since 1981. The objective of this program is to provide state-of-the-art technology for use in D D operations that will enable our customers to minimize waste generated and personal exposure, increase productivity and safety, and to minimize the potential for release and uptake of radioactive material. The program identifies and evaluates existing technology, develops new technology, and provides technical assistance to implement its use onsite. This program has impacted not only the Savannah River Site (SRS), but the entire Department of Energy (DOE) complex. To document and communicate the technology generated by this program, 28 papers have been presented at National and International meetings in the United States and Foreign Countries.

  1. Savannah River Laboratory Decontamination Program

    SciTech Connect (OSTI)

    Rankin, W.N.

    1991-12-31

    Savannah River Laboratory (SRL) has had a Decontamination and Decommissioning (D&D) Technology program since 1981. The objective of this program is to provide state-of-the-art technology for use in D&D operations that will enable our customers to minimize waste generated and personal exposure, increase productivity and safety, and to minimize the potential for release and uptake of radioactive material. The program identifies and evaluates existing technology, develops new technology, and provides technical assistance to implement its use onsite. This program has impacted not only the Savannah River Site (SRS), but the entire Department of Energy (DOE) complex. To document and communicate the technology generated by this program, 28 papers have been presented at National and International meetings in the United States and Foreign Countries.

  2. Microsoft Word - L15 01-22 Uranium Tranfers

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

    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

  3. Deactivation, Decontamination and Decommissioning Project Summaries

    SciTech Connect (OSTI)

    Peterson, David Shane; Webber, Frank Laverne

    2001-07-01

    This report is a compilation of summary descriptions of Deactivation, Decontamination and Decommissioning, and Surveillance and Maintenance projects planned for inactive facilities and sites at the INEEL from FY-2002 through FY-2010. Deactivations of contaminated facilities will produce safe and stable facilities requiring minimal surveillance and maintenance pending further decontamination and decommissioning. Decontamination and decommissioning actions remove contaminated facilities, thus eliminating long-term surveillance and maintenance. The projects are prioritized based on risk to DOE-ID, the public, and the environment, and the reduction of DOE-ID mortgage costs and liability at the INEEL.

  4. 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

  5. Radioactive Material or Multiple Hazardous Materials Decontamination

    Broader source: Energy.gov [DOE]

    The purpose of this procedure is to provide guidance for performing decontamination of individuals who have entered a “hot zone” during transportation incidents involving  radioactive.

  6. Urban Decontamination Experience at Pripyat Ukraine - 13526

    SciTech Connect (OSTI)

    Paskevych, Sergiy; Voropay, Dmitry; Schmieman, Eric

    2013-07-01

    This paper describes the efficiency of radioactive decontamination activities of the urban landscape in the town of Pripyat, Ukraine. Different methods of treatment for various urban infrastructure and different radioactive contaminants are assessed. Long term changes in the radiation condition of decontaminated urban landscapes are evaluated: 1. Decontamination of the urban system requires the simultaneous application of multiple methods including mechanical, chemical, and biological. 2. If a large area has been contaminated, decontamination of local areas of a temporary nature. Over time, there is a repeated contamination of these sites due to wind transport from neighboring areas. 3. Involvement of earth-moving equipment and removal of top soil by industrial method achieves 20-fold reduction in the level of contamination by radioactive substances, but it leads to large amounts of waste (up to 1500 tons per hectare), and leads to the re-contamination of treated areas due to scatter when loading, transport pollutants on the wheels of vehicles, etc.. (authors)

  7. 2015 Domestic Uranium Production Report

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

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

  8. Paducah Plant Begins Enrichment Operations after Five Parties Strike

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

    Agreement | Department of Energy Plant Begins Enrichment Operations after Five Parties Strike Agreement Paducah Plant Begins Enrichment Operations after Five Parties Strike Agreement May 1, 2012 - 12:00pm Addthis This cylinder hauler at Paducah’s Babcock & Wilcox Conversion Services plant delivers the first of DOE’s 14-ton depleted uranium cylinders to USEC for re-enrichment as part of a five-party agreement that is extending enrichment operations at the 60-year-old plant for

  9. PROCESS OF DECONTAMINATING MATERIAL CONTAMINATED WITH RADIOACTIVITY

    DOE Patents [OSTI]

    Overholt, D.C.; Peterson, M.D.; Acken, M.F.

    1958-09-16

    A process is described for decontaminating metallic objects, such as stainless steel equipment, which consists in contacting such objects with nltric acid in a concentration of 35 to 60% to remove the major portion of the contamination; and thereafter contacting the partially decontaminated object with a second solution containing up to 20% of alkali metal hydroxide and up to 20% sodium tartrate to remove the remaining radioactive contaminats.

  10. Nuclear reactor cooling system decontamination reagent regeneration

    DOE Patents [OSTI]

    Anstine, Larry D.; James, Dean B.; Melaika, Edward A.; Peterson, Jr., John P.

    1985-01-01

    An improved method for decontaminating the coolant system of water-cooled nuclear power reactors and for regenerating the decontamination solution. A small amount of one or more weak-acid organic complexing agents is added to the reactor coolant, and the pH is adjusted to form a decontamination solution which is circulated throughout the coolant system to dissolve metal oxides from the interior surfaces and complex the resulting metal ions and radionuclide ions. The coolant containing the complexed metal ions and radionuclide ions is passed through a strong-base anion exchange resin bed which has been presaturated with a solution containing the complexing agents in the same ratio and having the same pH as the decontamination solution. As the decontamination solution passes through the resin bed, metal-complexed anions are exchanged for the metal-ion-free anions on the bed, while metal-ion-free anions in the solution pass through the bed, thus removing the metal ions and regenerating the decontamination solution.

  11. 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.

  12. Calculated critical parameters for uranium-beryllium-water mixtures

    SciTech Connect (OSTI)

    Wetzel, L.L.

    1996-12-31

    Babcock & Wilcox recovers uranium from Sapphire material through chemical processing. Sapphire material consists of highly enriched uranium that contains various amounts of beryllium. Prior to the processing of Sapphire material, criticality safety analyses conservatively used uranium and water mixtures to model the solutions in the chemical processing operations. In the processing of Sapphire material, the presence of beryllium could change the safety limits. To determine the impact of the beryllium in the solution, critical parameters (mass or radius) for mixtures of uranium, beryllium, and water were calculated.

  13. 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

  14. Engineering assessment of inactive uranium mill tailings

    SciTech Connect (OSTI)

    Not Available

    1981-07-01

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

  15. 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.

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

    SciTech Connect (OSTI)

    N /A

    2003-11-28

    This document is a site-specific environmental impact statement (EIS) for construction and operation of a proposed depleted uranium hexafluoride (DUF{sub 6}) conversion facility at the U.S. Department of Energy (DOE) Portsmouth site in Ohio (Figure S-1). The proposed facility would convert the DUF{sub 6} stored at Portsmouth to a more stable chemical form suitable for use or disposal. The facility would also convert the DUF{sub 6} from the East Tennessee Technology Park (ETTP) site near Oak Ridge, Tennessee. In a Notice of Intent (NOI) published in the Federal Register on September 18, 2001 (Federal Register, Volume 66, page 48123 [66 FR 48123]), DOE announced its intention to prepare a single EIS for a proposal to construct, operate, maintain, and decontaminate and decommission two DUF{sub 6} conversion facilities at Portsmouth, Ohio, and Paducah, Kentucky, in accordance with the National Environmental Policy Act of 1969 (NEPA) (United States Code, Title 42, Section 4321 et seq. [42 USC 4321 et seq.]) and DOE's NEPA implementing procedures (Code of Federal Regulations, Title 10, Part 1021 [10 CFR Part 1021]). Subsequent to award of a contract to Uranium Disposition Services, LLC (hereafter referred to as UDS), Oak Ridge, Tennessee, on August 29, 2002, for design, construction, and operation of DUF{sub 6} conversion facilities at Portsmouth and Paducah, DOE reevaluated its approach to the NEPA process and decided to prepare separate site-specific EISs. This change was announced in a Federal Register Notice of Change in NEPA Compliance Approach published on April 28, 2003 (68 FR 22368); the Notice is included as Attachment B to Appendix C of this EIS. This EIS addresses the potential environmental impacts from the construction, operation, maintenance, and decontamination and decommissioning (D&D) of the proposed conversion facility at three alternative locations within the Portsmouth site; from the transportation of all ETTP cylinders (DUF{sub 6}, low-enriched UF6

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

    SciTech Connect (OSTI)

    N /A

    2003-11-28

    This document is a site-specific environmental impact statement (EIS) for construction and operation of a proposed depleted uranium hexafluoride (DUF{sub 6}) conversion facility at the U.S. Department of Energy (DOE) Paducah site in northwestern Kentucky (Figure S-1). The proposed facility would convert the DUF{sub 6} stored at Paducah to a more stable chemical form suitable for use or disposal. In a Notice of Intent (NOI) published in the ''Federal Register'' (FR) on September 18, 2001 (''Federal Register'', Volume 66, page 48123 [66 FR 48123]), DOE announced its intention to prepare a single EIS for a proposal to construct, operate, maintain, and decontaminate and decommission two DUF{sub 6} conversion facilities at Portsmouth, Ohio, and Paducah, Kentucky, in accordance with the National Environmental Policy Act of 1969 (NEPA) (''United States Code'', Title 42, Section 4321 et seq. [42 USC 4321 et seq.]) and DOE's NEPA implementing procedures (''Code of Federal Regulations'', Title 10, Part 1021 [10 CFR Part 1021]). Subsequent to award of a contract to Uranium Disposition Services, LLC (hereafter referred to as UDS), Oak Ridge, Tennessee, on August 29, 2002, for design, construction, and operation of DUF{sub 6} conversion facilities at Portsmouth and Paducah, DOE reevaluated its approach to the NEPA process and decided to prepare separate site-specific EISs. This change was announced in a ''Federal Register'' Notice of Change in NEPA Compliance Approach published on April 28, 2003 (68 FR 22368); the Notice is included as Attachment B to Appendix C of this EIS. This EIS addresses the potential environmental impacts from the construction, operation, maintenance, and decontamination and decommissioning (D&D) of the proposed conversion facility at three alternative locations within the Paducah site; from the transportation of depleted uranium conversion products to a disposal facility; and from the transportation, sale, use, or disposal of the fluoride

  18. Decontamination trade study for the Light Duty Utility Arm

    SciTech Connect (OSTI)

    Rieck, R.H.

    1994-09-29

    Various methods were evaluated for decontaminating the Light Duty Utility Arm (LDUA). Physical capabilities of each method were compared with the constraints and requirements for the LDUA Decontamination System. Costs were compared and a referred alternative was chosen.

  19. Radioactive scrap metal decontamination technology assessment report

    SciTech Connect (OSTI)

    Buckentin, J.M.; Damkroger, B.K.; Schlienger, M.E.

    1996-04-01

    Within the DOE complex there exists a tremendous quantity of radioactive scrap metal. As an example, it is estimated that within the gaseous diffusion plants there exists in excess of 700,000 tons of contaminated stainless steel. At present, valuable material is being disposed of when it could be converted into a high quality product. Liquid metal processing represents a true recycling opportunity for this material. By applying the primary production processes towards the material`s decontamination and re-use, the value of the strategic resource is maintained while drastically reducing the volume of material in need of burial. Potential processes for the liquid metal decontamination of radioactively contaminated metal are discussed and contrasted. Opportunities and technology development issues are identified and discussed. The processes compared are: surface decontamination; size reduction, packaging and burial; melting technologies; electric arc melting; plasma arc centrifugal treatment; air induction melting; vacuum induction melting; and vacuum induction melting and electroslag remelting.

  20. Experiences with decontaminating tritium-handling apparatus

    SciTech Connect (OSTI)

    Maienschein, J.L.; Garcia, F.; Garza, R.G.; Kanna, R.L.; Mayhugh, S.R.; Taylor, D.T. )

    1992-03-01

    Tritium-handling apparatus has been decontaminated as part of the downsizing of the LLNL Tritium Facility. Two stainless-steel glove boxes that had been used to process lithium deuteride-tritide (LiDT) slat were decontaminated using the Portable Cleanup System so that they could be flushed with room air through the facility ventilation system. In this paper the details on the decontamination operation are provided. A series of metal (palladium and vanadium) hydride storage beds have been drained of tritium and flushed with deuterium, in order to remove as much tritium as possible. The bed draining and flushing procedure is described, and a calculational method is presented which allows estimation of the tritium remaining in a bed after it has been drained and flushed. Data on specific bed draining and flushing are given.

  1. 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.

  2. Melting characteristics of the stainless steel generated from the uranium conversion plant

    SciTech Connect (OSTI)

    Choi, W.K.; Song, P.S.; Oh, W.Z.; Jung, C.H.; Min, B.Y.

    2007-07-01

    The partition ratio of cerium (Ce) and uranium (U) in the ingot, slag and dust phases has been investigated for the effect of the slag type, slag concentration and basicity in an electric arc melting process. An electric arc furnace (EAF) was used to melt the stainless steel wastes, simulated by uranium oxide and the real wastes from the uranium conversion plant in Korea Atomic Energy Research Institute (KAERI). The composition of the slag former used to capture the contaminants such as uranium, cerium, and cesium during the melt decontamination process generally consisted of silica (SiO{sub 2}), calcium oxide (CaO) and aluminum oxide (Al{sub 2}O{sub 3}). Also, Calcium fluoride (CaF{sub 2} ), nickel oxide (NiO), and ferric oxide (Fe{sub 2}O{sub 3}) were added to provide an increase in the slag fluidity and oxidative potential. Cerium was used as a surrogate for the uranium because the thermochemical and physical properties of cerium are very similar to those of uranium. Cerium was removed from the ingot phase to slag phase by up to 99% in this study. The absorption ratio of cerium was increased with an increase of the amount of the slag former. And the maximum removal of cerium occurred when the basicity index of the slag former was 0.82. The natural uranium (UO{sub 2}) was partitioned from the ingot phase to the slag phase by up to 95%. The absorption of the natural uranium was considerably dependent on the basicity index of the slag former and the composition of the slag former. The optimum condition for the removal of the uranium was about 1.5 for the basicity index and 15 wt% of the slag former. According to the increase of the amount of slag former, the absorption of uranium oxide in the slag phase was linearly increased due to an increase of its capacity to capture uranium oxide within the slag phase. Through experiments with various slag formers, we verified that the slag formers containing calcium fluoride (CaF{sub 2}) and a high amount of silica were more

  3. Decontamination of metals using chemical etching

    DOE Patents [OSTI]

    Lerch, Ronald E.; Partridge, Jerry A.

    1980-01-01

    The invention relates to chemical etching process for reclaiming contaminated equipment wherein a reduction-oxidation system is included in a solution of nitric acid to contact the metal to be decontaminated and effect reduction of the reduction-oxidation system, and includes disposing a pair of electrodes in the reduced solution to permit passage of an electrical current between said electrodes and effect oxidation of the reduction-oxidation system to thereby regenerate the solution and provide decontaminated equipment that is essentially radioactive contamination-free.

  4. DECONTAMINATION OF NEUTRON-IRRADIATED REACTOR FUEL

    DOE Patents [OSTI]

    Buyers, A.G.; Rosen, F.D.; Motta, E.E.

    1959-12-22

    A pyrometallurgical method of decontaminating neutronirradiated reactor fuel is presented. In accordance with the invention, neutron-irradiated reactor fuel may be decontaminated by countercurrently contacting the fuel with a bed of alkali and alkaine fluorides under an inert gas atmosphere and inductively melting the fuel and tracking the resulting descending molten fuel with induction heating as it passes through the bed. By this method, a large, continually fresh surface of salt is exposed to the descending molten fuel which enhances the efficiency of the scrubbing operation.

  5. COPPER COATED URANIUM ARTICLE

    DOE Patents [OSTI]

    Gray, A.G.

    1958-10-01

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

  6. Uranium Marketing Annual Report -

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

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

  7. Separation of uranium from technetium in recovery of spent nuclear fuel

    DOE Patents [OSTI]

    Friedman, Horace A.

    1985-01-01

    A method for decontaminating uranium product from the Purex process comprises addition of hydrazine to the product uranyl nitrate stream from the Purex process, which contains hexavalent (UO.sub.2.sup.2+) uranium and heptavalent technetium (TcO.sub.4 -). Technetium in the product stream is reduced and then complexed by the addition of oxalic acid (H.sub.2 C.sub.2 O.sub.4), and the Tc-oxalate complex is readily separated from the uranium by solvent extraction with 30 vol. % tributyl phosphate in n-dodecane.

  8. Separation of uranium from technetium in recovery of spent nuclear fuel

    DOE Patents [OSTI]

    Friedman, H.A.

    1984-06-13

    A method for decontaminating uranium product from the Purex 5 process comprises addition of hydrazine to the product uranyl nitrate stream from the Purex process, which contains hexavalent (UO/sub 2//sup 2 +/) uranium and heptavalent technetium (TcO/sub 4/-). Technetium in the product stream is reduced and then complexed by the addition of oxalic acid (H/sub 2/C/sub 2/O/sub 4/), and the Tc-oxalate complex is readily separated from the 10 uranium by solvent extraction with 30 vol % tributyl phosphate in n-dodecane.

  9. 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.

  10. 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.

  11. 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)

  12. 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.

  13. Decontamination and decommissioning focus area. Technology summary

    SciTech Connect (OSTI)

    1995-06-01

    This report presents details of the facility deactivation, decommissioning, and material disposition research for development of new technologies sponsored by the Department of Energy. Topics discussed include; occupational safety, radiation protection, decontamination, remote operated equipment, mixed waste processing, recycling contaminated metals, and business opportunities.

  14. Experiences with decontaminating tritium-handling apparatus

    SciTech Connect (OSTI)

    Maienschein, J.L.; Garcia, F.; Garza, R.G.; Kanna, R.L.; Mayhugh, S.R.; Taylor, D.T.

    1991-07-01

    Tritium-handling apparatus has been decontaminated as part of the shutdown of the LLNL Tritium Facility. Two stainless-steel gloveboxes that had been used to process lithium deuteride-tritide (LiDT) salt were decontaminated using the Portable Cleanup System so that they could be flushed with room air through the facility ventilation system. Further surface decontamination was performed by scrubbing the interior with paper towels and ethyl alcohol or Swish{trademark}. The surface contamination, as shown by swipe surveys, was reduced from 4{times}10{sup 4}--10{sup 6} disintegrations per minute (dpm)/cm{sup 2} to 2{times}10{sup 2}--4{times}10{sup 4} dpm/cm{sup 2}. Details on the decontamination operation are provided. A series of metal (palladium and vanadium) hydride storage beds have been drained of tritium and flushed with deuterium in order to remove as much tritium as possible. The bed draining and flushing procedure is described, and a calculational method is presented which allows estimation of the tritium remaining in a bed after it has been drained and flushed. Data on specific bed draining and flushing are given.

  15. Testing and evaluation of light ablation decontamination

    SciTech Connect (OSTI)

    Demmer, R.L.; Ferguson, R.L.

    1994-10-01

    This report details the testing and evaluation of light ablation decontamination. It details WINCO contracted research and application of light ablation efforts by Ames Laboratory. Tests were conducted with SIMCON (simulated contamination) coupons and REALCON (actual radioactive metal coupons) under controlled conditions to compare cleaning effectiveness, speed and application to plant process type equipment.

  16. INTEGRATED VERTICAL AND OVERHEAD DECONTAMINATION (IVOD) SYSTEM

    SciTech Connect (OSTI)

    M.A. Ebadian, Ph.D.

    2001-01-01

    The deactivation and decommissioning of 1200 buildings within the U.S. Department of Energy-Office of Environmental Management complex will require the disposition of a large quantity of contaminated concrete and metal surfaces. It has been estimated that 23 million cubic meters of concrete and over 600,000 tons of metal will need disposition. The disposition of such large quantities of material presents difficulties in the area of decontamination and characterization. The final disposition of this large amount of material will take time and money as well as risk to the D&D work force. A single automated system that would decontaminate and characterize surfaces in one step would not only reduce the schedule and decrease cost during D&D operations but would also protect the D&D workers from unnecessary exposures to contaminated surfaces. This report summarizes the activities performed during FY00 and describes the planned activities for FY01. Accomplishments for FY00 include the following: Development and field-testing of characterization system; Completion of Title III design of deployment platform and decontamination unit; In-house testing of deployment platform and decontamination unit; Completion of system integration design; Identification of deployment site; and Completion of test plan document for deployment of IVOD at Rancho Seco nuclear power facility.

  17. Microsoft Word - Transmittal Memo - FY 2009 D&D Fund Report

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

    Department of Energy's Uranium Enrichment Decontamination and Decommissioning Fund's (D&D Fund) Fiscal Year (FY) 2009 balance sheet and the related statements of net cost,...

  18. 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.

  19. 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.

  20. PRODUCTION OF URANIUM MONOCARBIDE

    DOE Patents [OSTI]

    Powers, R.M.

    1962-07-24

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

  1. Reduced enrichment for research and test reactors: Proceedings

    SciTech Connect (OSTI)

    Not Available

    1988-05-01

    The international effort to develop new research reactor fuel materials and designs based on the use of low-enriched uranium, instead of highly-enriched uranium, has made much progress during the eight years since its inception. To foster direct communication and exchange of ideas among the specialist in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the ninth of a series which began in 1978. All previous meetings of this series are listed on the facing page. The focus of this meeting was on the LEU fuel demonstration which was in progress at the Oak Ridge Research (ORR) reactor, not far from where the meeting was held. The visit to the ORR, where a silicide LEU fuel with 4.8 g A/cm/sup 3/ was by then in routine use, illustrated how far work has progressed.

  2. 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.

  3. 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, 2015 deliveries thousand pounds U3O8 equivalent; dollars per pound U3O8 equivalent Deliveries Uranium concentrate Natural UF6 Enriched UF6 Natural UF6 and Enriched UF6 Total U.S.-origin uranium Purchases 2,733 W W 686 3,419 Weighted-average price 46.23 W W 34.44 43.86 Foreign-origin uranium Purchases 28,179 W W 24,927 53,106 Weighted-average price 43.61 W W 44.77 44.14 Total

  4. NNSA Announces Arrival of Plutonium and Uranium from Japan's Fast

    National Nuclear Security Administration (NNSA)

    Critical Assembly at Savannah River Site and Y-12 National Security Complex | National Nuclear Security Administration | (NNSA) Arrival of Plutonium and Uranium from Japan's Fast Critical Assembly at Savannah River Site and Y-12 National Security Complex June 06, 2016 WASHINGTON (June 6, 2016) - A shipment of plutonium and highly enriched uranium (HEU) from Japan Atomic Energy Agency (JAEA)'s Fast Critical Assembly (FCA) reactor arrived safely at the Department of Energy's (DOE) Savannah

  5. 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.

  6. EIS-0111: Remedial Actions at the Former Vanadium Corporation of America Uranium Mill Site, Durango, La Plata County, Colorado

    Broader source: Energy.gov [DOE]

    The U.S. Department of Energy developed this statement to evaluate the environmental impacts of several scenarios for management and control of the residual radioactive wastes at the inactive Durango, Colorado, uranium processing site, including a no action alternative, an alternative to manage wastes on site, and three alternatives involving off-site management and decontamination of the Durango site.

  7. Decontamination and recovery of materials at nuclear facilites - operating history

    SciTech Connect (OSTI)

    Gillis, P.J. Jr.

    1994-12-31

    Non-Destructive Cleaning (NDC) Mobile CO{sub 2} Decontamination Facilities have more than 120 months of operational time conducting radioactive decontamination at Nuclear Power Stations and U.S. Department of Energy sites. During this time, we have compiled an extensive database on what has been decontaminated and the cost savings realized. The following are areas of interest: (1) how the CO{sub 2} decontamination process works; (2) how radioactive wastes are minimized and radioactive exposure to personnel is reduced with the use of the NDC Decontamination Facility; (3) how the self-contained Mobile Decontamination Facility works to provide adequate containment and control of the radioactive materials; (4) what kinds of items have been decontaminated, ranging from tools to underwater television cameras and from electric motors to lead shielding; (5) liquid radioactive waste volume reduction; (6) mixed-waste volume reduction; and (7) achievements in dose reduction to radiation levels that are as low as is reasonably achievable (ALARA) The design and operating features and performance of the Mobile Decontamination Facility, as well as the actual volumes of materials decontaminated, the decontamination factors achieved, the amounts and types of things that are free released, and the actual cost savings in all of these areas have been assessed. The data that was used is actual utility data and not the vendor`s data. All the experiences were from actual power plants.

  8. 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...

  9. Decontamination and Decommisioning Equipment Tracking System

    Energy Science and Technology Software Center (OSTI)

    1994-08-26

    DDETS is Relational Data Base Management System (RDBMS) which incorporates 1-D (code 39) and 2-D (PDF417) bar codes into its equipment tracking capabilities. DDETS is compatible with the Reportable Excess Automated Property System (REAPS), and has add, edit, delete and query capabilities for tracking equipment being decontaminated and decommissioned. In addition, bar code technology is utilized in the inventory tracking and shipping of equipment.

  10. Decontamination and Demolition at TA-21

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

    TA-21 Demolition Decontamination and Demolition at TA-21 On December 1, 2009, LANL began full-scale demolition at TA-21, the Cold War-era complex of buildings that once housed plutonium production and historic, nonweapons research. August 1, 2013 Water sprayed during demolition to protect air quality Water sprayed during demolition to protect air quality The buildings at TA-21 were built as long ago as the 1940s, replacing Manhattan-era facilities and housing labs, offices, and production

  11. Advanced robotics for decontamination and dismantlement

    SciTech Connect (OSTI)

    Hamel, W.R.; Haley, D.C.

    1994-06-01

    The decontamination and dismantlement (D&D) robotics technology application area of the US Department of Energy`s Robotics Technology Development Program is explained and described. D&D robotic systems show real promise for the reduction of human exposure to hazards, for improvement of productivity, and for the reduction of secondary waste generation. Current research and development pertaining to automated floor characterization, robotic equipment removal, and special inspection is summarized. Future research directions for these and emerging activities is given.

  12. Method for the decontamination of metallic surfaces

    DOE Patents [OSTI]

    Purohit, Ankur (Darien, IL); Kaminski, Michael D. (Lockport, IL); Nunez, Luis (Elmhurst, IL)

    2003-01-01

    A method of decontaminating a radioactively contaminated oxide on a surface. The radioactively contaminated oxide is contacted with a diphosphonic acid solution for a time sufficient to dissolve the oxide and subsequently produce a precipitate containing most of the radioactive values. Thereafter, the diphosphonic solution is separated from the precipitate. HEDPA is the preferred diphosphonic acid and oxidizing and reducing agents are used to initiate precipitation. SFS is the preferred reducing agent.

  13. Method for electrochemical decontamination of radioactive metal

    DOE Patents [OSTI]

    Ekechukwu, Amy A. (Augusta, GA)

    2008-06-10

    A decontamination method for stripping radionuclides from the surface of stainless steel or aluminum material comprising the steps of contacting the metal with a moderately acidic carbonate/bicarbonate electrolyte solution containing sodium or potassium ions and thereafter electrolytically removing the radionuclides from the surface of the metal whereby radionuclides are caused to be stripped off of the material without corrosion or etching of the material surface.

  14. Selective leaching of uranium from uranium-contaminated soils: Progress report 1

    SciTech Connect (OSTI)

    Francis, C.W.; Mattus, A.J.; Farr, L.L.; Elless, M.P.; Lee, S.Y.

    1993-02-01

    Three soils and a sediment contaminated with uranium were used to determine the effectiveness of sodium carbonate and citric acid leaching to decontaminated or remove uranium to acceptable regulatory levels. Two of the soils were surface soils from the DOE facility formerly called the Feed Materials Production Center (FMPC) at Fernald, Ohio. This facility is presently called the Femald Environmental Management Project (FEMP). Carbonate extractions generally removed from 70 to 90% of the uranium from the Fernald storage pad soil. Uranium was slightly more difficult to extract from the Fernald incinerator and the Y-12 landfarm soils. Very small amounts of uranium could be extracted from the storm sewer sediment. Extraction with carbonate at high solution-to-soil ratios were as effective as extractions at low solution-to-soil ratios, indicating attrition by the paddle mixer was not significantly different than that provided in a rotary extractor. Also, pretreatments such as milling or pulverizing the soil sample did not appear to increase extraction efficiency when carbonate extractions were carried out at elevated temperatures (60{degree}C) or long extraction times (23 h). Adding KMnO{sub 4} in the carbonate extraction appeared to increase extraction efficiency from the Fernald incinerator soil but not the Fernald storage pad soil. The most effective leaching rates (> 90 % from both Fernald soils) were obtained using a citrate/dithionite extraction procedure designed to remove amorphous (noncrystalline) iron/aluminum sesquioxides from surfaces of clay minerals. Citric acid also proved to be a very good extractant for uranium.

  15. Selective leaching of uranium from uranium-contaminated soils: Progress report 1

    SciTech Connect (OSTI)

    Francis, C.W.; Mattus, A.J.; Farr, L.L.; Elless, M.P.; Lee, S.Y.

    1993-02-01

    Three soils and a sediment contaminated with uranium were used to determine the effectiveness of sodium carbonate and citric acid leaching to decontaminated or remove uranium to acceptable regulatory levels. Two of the soils were surface soils from the DOE facility formerly called the Feed Materials Production Center (FMPC) at Fernald, Ohio. This facility is presently called the Femald Environmental Management Project (FEMP). Carbonate extractions generally removed from 70 to 90% of the uranium from the Fernald storage pad soil. Uranium was slightly more difficult to extract from the Fernald incinerator and the Y-12 landfarm soils. Very small amounts of uranium could be extracted from the storm sewer sediment. Extraction with carbonate at high solution-to-soil ratios were as effective as extractions at low solution-to-soil ratios, indicating attrition by the paddle mixer was not significantly different than that provided in a rotary extractor. Also, pretreatments such as milling or pulverizing the soil sample did not appear to increase extraction efficiency when carbonate extractions were carried out at elevated temperatures (60[degree]C) or long extraction times (23 h). Adding KMnO[sub 4] in the carbonate extraction appeared to increase extraction efficiency from the Fernald incinerator soil but not the Fernald storage pad soil. The most effective leaching rates (> 90 % from both Fernald soils) were obtained using a citrate/dithionite extraction procedure designed to remove amorphous (noncrystalline) iron/aluminum sesquioxides from surfaces of clay minerals. Citric acid also proved to be a very good extractant for uranium.

  16. 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.

  17. 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.

  18. Decontamination impacts on solidification and waste disposal

    SciTech Connect (OSTI)

    Kempf, C.R.; Soo, P.

    1988-01-01

    Research to determine chemical and physical conditions which could lead to thermal excursions, gas generation, and/or general degradation of decontamination-reagent-loaded resins has shown that IRN-78, IONAC A-365, and IRN-77 organic ion exchange resin moisture contents vary significantly depending on the counter ion loading.'' The extent/vigor of the reaction is very highly dependent on the degree of dewatering of the resins and on the method of solution addition. The heat generation may be due, in part, to the heat of neutralization. In studies of the long-term compatibility effects of decontamination waste resins in contact with waste package container materials in the presence of decontamination reagents, radiolysis products and gamma irradiation, it has been found that the corrosion of carbon steel and austenitic stainless steel in mixed bed resins is enhanced by gamma irradiation. However, cracking in high density polyethylene is essentially eliminated because of the rapid removal of oxygen from the environment by gamma-induced oxidation of the large resin mass. 13 refs., 10 figs., 3 tabs.

  19. Reports on investigations of uranium anomalies. National Uranium Resource Evaluation

    SciTech Connect (OSTI)

    Goodknight, C.S.; Burger, J.A.

    1982-10-01

    During the National Uranium Resource Evaluation (NURE) program, conducted for the US Department of Energy (DOE) by Bendix Field Engineering Corporation (BFEC), radiometric and geochemical surveys and geologic investigations detected anomalies indicative of possible uranium enrichment. Data from the Aerial Radiometric and Magnetic Survey (ARMS) and the Hydrogeochemical and Stream-Sediment Reconnaissance (HSSR), both of which were conducted on a national scale, yielded numerous anomalies that may signal areas favorable for the occurrence of uranium deposits. Results from geologic evaluations of individual 1/sup 0/ x 2/sup 0/ quadrangles for the NURE program also yielded anomalies, which could not be adequately checked during scheduled field work. Included in this volume are individual reports of field investigations for the following six areas which were shown on the basis of ARMS, HSSR, and (or) geologic data to be anomalous: (1) Hylas zone and northern Richmond basin, Virginia; (2) Sischu Creek area, Alaska; (3) Goodman-Dunbar area, Wisconsin; (4) McCaslin syncline, Wisconsin; (5) Mt. Withington Cauldron, Socorro County, New Mexico; (6) Lake Tecopa, Inyo County, California. Field checks were conducted in each case to verify an indicated anomalous condition and to determine the nature of materials causing the anomaly. The ultimate objective of work is to determine whether favorable conditions exist for the occurrence of uranium deposits in areas that either had not been previously evaluated or were evaluated before data from recent surveys were available. Most field checks were of short duration (2 to 5 days). The work was done by various investigators using different procedures, which accounts for variations in format in their reports. All papers have been abstracted and indexed.

  20. Process for electroslag refining of uranium and uranium alloys

    DOE Patents [OSTI]

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

    1975-07-22

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

  1. 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.

  2. 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.

  3. Volume plummets, restricted uranium below $10 per pound

    SciTech Connect (OSTI)

    1993-12-01

    This article is the November 1993 uranium market summary. The pace of deals slackened dramatically during this period, with only six deals occurring. Five were in the natural uranium spot market and one was in the conversion services market. Total spot concentrates volume came to just 994,000 lbs U3O8 equivalent. This compares to the 15 deals and 2.8 millions lbs volume during the previous reporting period. The bottom of the restricted uranium spot market price range dipped back below $10.00. In the unrestricted market, the range stayed the same. The same holds true for the enrichment services price range.

  4. Method of recovering uranium hexafluoride

    DOE Patents [OSTI]

    Schuman, S.

    1975-12-01

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

  5. Decontamination system study for the Tank Waste Retrieval System

    SciTech Connect (OSTI)

    Reutzel, T.; Manhardt, J.

    1994-05-01

    This report summarizes the findings of the Idaho National Engineering Laboratory`s decontamination study in support of the Tank Waste Retrieval System (TWRS) development program. Problems associated with waste stored in existing single shell tanks are discussed as well as the justification for the TWRS program. The TWRS requires a decontamination system. The subsystems of the TWRS are discussed, and a list of assumptions pertinent to the TWRS decontamination system were developed. This information was used to develop the functional and operational requirements of the TWRS decontamination system. The requirements were combined with a comprehensive review of currently available decontamination techniques to produced a set of evaluation criteria. The cleaning technologies and techniques were evaluated, and the CO{sub 2} blasting decontamination technique was chosen as the best technology for the TWRS.

  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. 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.

  8. 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.

  9. Uranium Marketing Annual Report -

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

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

  10. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

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

  11. Uranium Marketing Annual Report

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

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

  12. Uranium Marketing Annual Report -

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

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

  13. Uranium Marketing Annual Report

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

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

  14. Uranium Marketing Annual Report -

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

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

  15. 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.

  16. IAEA Verification Experiment at the Portsmouth Gaseous Diffusion Plant: Report on the Cascade Header Enrichment Monitor

    SciTech Connect (OSTI)

    P. L. Kerr; D. A. Close; W. S. Johnson; R. M. Kandarian; C. E. Moss; C. D. Romero

    1999-03-01

    The authors describe the Cascade Header Enrichment Monitor (CHEM) for the Portsmouth Gaseous Diffusion Plant at Piketon, Ohio, and present the calibration and measurement results. The US government has offered excess fissile material that is no longer needed for defense purposes for International Atomic Energy Agency (IAEA) inspection. Measurement results provided by the CHEM were used by the IAEA in a verification experiment to provide confidence that the US successfully blended excess highly enriched uranium (HEU) down to low enriched uranium (LEU). The CHEM measured the uranium enrichment in two cascade header pipes, a 20.32-cm HEU pipe and a 7.62-cm product LEU pipe. The CHEM determines the amount of {sup 235}U from the 185.7-keV gamma-ray photopeak and the amount of total uranium by x-ray fluorescence (XRF) of the 98.4-keV x-ray from uranium with a {sup 57}Co XRF source. The ratio yields the enrichment. The CHEM consists of a collimator assembly, an electromechanically cooled germanium detector, and a rack-mounted personal computer running commercial and custom software. The CHEM was installed in December 1997 and was used by the IAEA inspectors for announced and unannounced inspections on the HEU and LEU header pipes through October 1998. The equipment was sealed with tamper-indicating enclosures when the inspectors were not present.

  17. PROCESS OF PURIFYING URANIUM

    DOE Patents [OSTI]

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

    1958-12-23

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

  18. PREPARATION OF URANIUM HEXAFLUORIDE

    DOE Patents [OSTI]

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

    1959-10-01

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

  19. Toxic Substances Control Act Uranium Enrichment Federal Facility Compliance

    Energy Savers [EERE]

    Most Popular Energy Savers Web Pages of 2011 Top 10 Most Popular Energy Savers Web Pages of 2011 December 27, 2011 - 9:08am Addthis Chris Stewart Senior Communicator at DOE's National Renewable Energy Laboratory 2011 proved to be another successful year for consumers wanting to save money and energy at home. The Energy Savers website continues to be a great resource for readers interested in learning more about energy efficiency and tips for saving money and energy at home and on the road-and

  20. 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...

  1. 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...

  2. 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...

  3. 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...

  4. 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...

  5. 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.

  6. 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...

  7. NNSA Announces Elimination of Highly Enriched Uranium (HEU) from...

    National Nuclear Security Administration (NNSA)

    It also highlights NNSA's commitment to finding domestic disposition solutions for proliferation-sensitive nuclear material around the world." At the March 2016 Nuclear Security ...

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

    Office of Environmental Management (EM)

    provision herein shall require the obligation or payment of funds in violation of the Anti-Deficiency Act. In cases where payment would constitute a violation, the dates ...

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

    National Nuclear Security Administration (NNSA)

    Reactor Conversion is part of NNSA's Global Threat Reduction Initiative (GTRI) to develop and implement technologies to minimize and, to the extent possible, eliminate the civilian ...

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

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

    has been successfully removed from Mexico. | Photo courtesy of the NNSA. NNSA ... has been successfully removed from Mexico. | Photo courtesy of the NNSA. Michael ...

  11. 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...

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

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

    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

  13. 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.

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

    Energy Savers [EERE]

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

  15. Uranium Marketing Annual Report -

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

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

  16. 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.

  17. U.S. Uranium Reserves Estimates

    Gasoline and Diesel Fuel Update (EIA)

    Major U.S. Uranium Reserves

  18. Decontamination, decommissioning, and vendor advertorial issue, 2006

    SciTech Connect (OSTI)

    Agnihotri, Newal (ed.)

    2006-07-15

    The focus of the July-August issue is on Decontamination, decommissioning, and vendor advertorials. Major articles/reports in this issue include: NPP Krsko revised decommissioning program, by Vladimir Lokner and Ivica Levanat, APO d.o.o., Croatia, and Nadja Zeleznik and Irena Mele, ARAO, Slovenia; Supporting the renaissance, by Marilyn C. Kray, Exelon Nuclear; Outage world an engineer's delight, by Tom Chrisopher, Areva, NP Inc.; Optimizing refueling outages with R and D, by Ross Marcoot, GE Energy; and, A successful project, by Jim Lash, FirstEnergy.

  19. Decontamination formulations for disinfection and sterilization

    DOE Patents [OSTI]

    Tucker, Mark D.; Engler, Daniel E.

    2007-09-18

    Aqueous decontamination formulations that neutralize biological pathogens for disinfection and sterilization applications. Examples of suitable applications include disinfection of food processing equipment, disinfection of areas containing livestock, mold remediation, sterilization of medical instruments and direct disinfection of food surfaces, such as beef carcasses. The formulations include at least one reactive compound, bleaching activator, inorganic base, and water. The formulations can be packaged as a two-part kit system, and can have a pH value in the range of 7-8.

  20. Process for Descaling and Decontaminating Metals

    DOE Patents [OSTI]

    Baybarz, R. D.

    1961-04-25

    The oxide scale on the surface of stainless steels and similar metals is removed by contacting the metal under an inert atmosphere with a dilute H/sub 2/ SO/sub 4/ solution containing CrSO/sub 4/. The removed oxide scale is either dissolved or disintegrated into a slurry by the solution. Preferred reagent concentrations are 0.3 to 0.5 M CrSO/sub 4/ and 0.5 to 0.6 M H/sub 2/SO/sub 4/. The process is particularly applicable to decontamination of aqueous homogeneous nuclear reactor systems. (AEC)

  1. URANIUM RECOVERY PROCESS

    DOE Patents [OSTI]

    Yeager, J.H.

    1958-08-12

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

  2. URANIUM SEPARATION PROCESS

    DOE Patents [OSTI]

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

    1959-07-14

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

  3. PROCESS OF RECOVERING URANIUM

    DOE Patents [OSTI]

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

    1958-10-01

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

  4. PRODUCTION OF URANIUM

    DOE Patents [OSTI]

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

    1958-04-15

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

  5. 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.

  6. Final Environmental Impact Statement for the construction and operation of Claiborne Enrichment Center, Homer, Louisiana (Docket No. 70-3-70). Volume 2, Public comments and NRC response

    SciTech Connect (OSTI)

    Zeitoun, A.

    1994-08-01

    The Final Environmental Impact Statement (FEIS) (Volume 1), was prepared by the Nuclear Regulatory Commission (NRC) in accordance with regulation 10 CFR Part 51, which implements the National Environmental Policy Act (NEPA), to assess the potential environmental impacts for licensing the construction and operation of a proposed gaseous centrifuge enrichment facility to be built in Claiborne Parish, Louisiana by Louisiana Energy Services, L.P. (LES). The proposed facility would have a production capacity of about 866 metric tons annually of up to 5 weight percent enriched UF{sub 6}, using a proven centrifuge technology. Included in the assessment are co on, both normal operations and potential accidents (internal and external events), and the eventual decontamination and decommissioning of the site. In order to help assure that releases from the operation of the facility and potential impacts on the public are as low as reasonably achievable, an environmental monitoring program was developed by LES to detect significant changes in the background levels of uranium around the site. Other issues addressed include the purpose and need for the facility, the alternatives to the proposed action, potential disposition of the tails, the site selection process, and environmental justice. The NRC staff concludes that the facility can be constructed and operated with small and acceptable impacts on the public and the environment, and proposes to issue a license to the applicant, Louisiana Energy Services, to authorize construction and operation of the proposed facility. The letters in this Appendix have been divided into three sections. Section One contains letters to which the NRC responded by addressing specific comments. Section Two contains the letters that concerned the communities of Forest Grove and Center Springs. Section Three is composed of letters that required no response. These letters were generally in support of the facility.

  7. 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

  8. 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.

  9. Gas-phase decontamination demonstration on PORTS cell X-25-4-2. Final technology status report

    SciTech Connect (OSTI)

    Riddle, R.J.

    1997-09-01

    The Long-Term, Low Temperature (LTLT) process is a gas-phase in situ decontamination technique which has been tested by LMES/K-25 personnel on the laboratory scale with promising results. The purpose of the Gas-Phase Decontamination Demonstration at PORTS was to evaluate the LTLT process on an actual diffusion cascade cell at conditions similar to those used in the laboratory testing. The demonstration was conducted on PORTS diffusion cell X-25-4-2 which was one of the X-326 Building cells which was permanently shutdown as part of the Suspension of HEU Production at PORTS. The demonstration full-scale test consisted of rendering the cell leak-tight through the installation of Dresser seals onto the process seals, exposing the cell to the oxidants ClF{sub 3} and F{sub 2} for a period of 105 days and evaluating the effect of the clean-up treatment on cell samples and coupons representing the major diffusion cascade materials of construction. The results were extrapolated to determine the effectiveness of LTLT decontamination over the range of historical uranium isotope assays present in the diffusion complex. It was determined that acceptable surface contamination levels could be obtained in all of the equipment in the lower assay cascades which represents the bulk of the equipment contained in the diffusion complex.

  10. 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)

  11. Prices dip, activity increases in unrestricted uranium market. [Uranium market overview

    SciTech Connect (OSTI)

    Not Available

    1993-05-01

    April's activity in the restricted uranium market fluctuated in the same range as that observed in March. At the same time, NUKEM detects a weakening of prices in the unrestricted market to $7.45-$7.65. Unrestricted buyers seem to have detected lower prices as well; much of the new demand noted this month emerged in the unrestricted segment of the market. With this issue, NUKEM inaugurates a new market statistic. To better follow developments in the conversion market, we will report a spot price range for conversion services. This price measure will be derived in a manner analogous to NUKEM's other spot market price ranges. We will continue to publish the current NUKEM price range for new contracts for a few months. If you wish to retain the old conversion contract price range in future editions, please contact our US office. Four deals for near term delivery occurred in the uranium market in April, resulting in spot market transaction volume of 2.5 million lbs U3O8 equivalent. In the first week, a US non-utility purchased a small quantity of enriched uranium product from an intermediary in a spot transaction representing about 75,000 lbs U3O8. The second week saw the stealthy purchase of Portland General Electric's inventory of natural and enriched uranium. The buyer of PGE's 1.1 million lbs U3O8 equivalent has achieved an unusual degree of anonymity. Also during the second week, a US utility bought a small quantity of enriched uranium containing less than 25,000 lbs natural U3O8 equivalent.

  12. US Department of Energy Grand Junction Projects Office Remedial Action Project, final report of the decontamination and decommissioning of Building 36 at the Grand Junction Projects Office Facility

    SciTech Connect (OSTI)

    Widdop, M.R.

    1996-08-01

    The U.S. Department of Energy (DOE) Grand Junction Projects Office (GJPO) occupies a 61.7-acre facility along the Gunnison River near Grand Junction, Colorado. This site was contaminated with uranium ore and mill tailings during uranium refining activities of the Manhattan Engineer District and during pilot milling experiments conducted for the U.S. Atomic Energy Commission`s domestic uranium procurement program. The DOE Defense Decontamination and Decommissioning Program established the GJPO Remedial Action Project to clean up and restore the facility lands, improvements, and the underlying aquifer. The site contractor for the facility, Rust Geotech, also is the remedial action contractor. Building 36 was found to be radiologically contaminated and was demolished in 1996. The soil beneath the building was remediated in accordance with identified standards and can be released for unlimited exposure and unrestricted use. This document was prepared in response to a DOE request for an individual final report for each contaminated GJPO building.

  13. 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.)

  14. A remotely operated robot for decontamination tasks

    SciTech Connect (OSTI)

    Dudar, A.M.; Vandewalle, R.C.

    1994-02-01

    Engineers in the Robotics Development Group at the Westinghouse Savannah River Company (WSRC) have developed a robot which will be used to decontaminate a pipe gallery of a tank farm used for nuclear waste storage. Personnel access is required into this pipe gallery to inspect existing pipes and perform repairs to secondary containment walls around the tank farm. Presently, the pipe gallery is littered with debris of various sizes and its surface is contaminated with activity levels up to 2.5E6 DPM (disintegrations per minute) alpha and exposure levels as high as 20 Rad/hr. Cleaning up this pipe gallery win be the mission of an all-hydraulic robotic vehicle developed in-house at WSRC caged the ``Remote Decon`` robot. The Remote Decon is a tracked vehicle which utilizes skid steering and features a six-degree-of-freedom (DOF) manipulator arm, a five-DOF front end loader type bucket with a rotating brush for scrubbing and decontaminating surfaces, and a three-DOF pan/tilt mechanism with cameras and lights. The Remote Decon system is connected to a control console via a 200 foot tethered cable. The control console was designed with ergonomics and simplicity as the main design factors and features three joysticks, video monitors, LED panels, and audible alarms.

  15. FY 2005 Volume 5

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

    6 Volume 5 Environmental Management Environmental Management Defense Site Acceleration Completion Defense Site Acceleration Completion Defense Environmental Services Defense Environmental Services Non Non - - Defense Site Acceleration Completion Defense Site Acceleration Completion Non Non - - Defense Environmental Services Defense Environmental Services Uranium Enrichment Decontamination Uranium Enrichment Decontamination and Decommissioning Fund and Decommissioning Fund February 2004 Volume 5

  16. 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.

  17. Final report of the decontamination and decommissioning of the exterior land areas at the Grand Junction Projects Office facility

    SciTech Connect (OSTI)

    Widdop, M.R.

    1995-09-01

    The US Department of Energy (DOE) Grand Junction Projects Office (GJPO) facility occupies approximately 56.4 acres (22.8 hectares) along the Gunnison River near Grand Junction, Colorado. The site was contaminated with uranium ore and mill tailings during uranium-refining activities conducted by the Manhattan Engineer District and during pilot-milling experiments conducted for the US Atomic Energy Commission`s (AEC`s) domestic uranium procurement program. The GJPO facility was the collection and assay point for AEC uranium and vanadium oxide purchases until the early 1970s. The DOE Decontamination and Decommissioning Program sponsored the Grand Junction Projects Office Remedial Action Project (GJPORAP) to remediate the facility lands, site improvements, and the underlying aquifer. The site contractor, Rust Geotech, was the Remedial Action Contractor for GJPORAP. The exterior land areas of the facility assessed as contaminated have been remediated in accordance with identified standards and can be released for unrestricted use. Restoration of the aquifer will be accomplished through the natural flushing action of the aquifer during the next 50 to 80 years. The remediation of the DOE-GJPO facility buildings is ongoing and will be described in a separate report.

  18. 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

  19. 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)

  20. 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

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

  2. 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.

  3. Development of Novel Sorbents for Uranium Extraction from Seawater

    SciTech Connect (OSTI)

    Lin, Wenbin; Taylor-Pashow, Kathryn

    2014-01-08

    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.

  4. METHOD OF FABRICATING A URANIUM-ZIRCONIUM HYDRIDE REACTOR CORE

    DOE Patents [OSTI]

    Weeks, I.F.; Goeddel, W.V.

    1960-03-22

    A method is described of evenly dispersing uranlum metal in a zirconium hydride moderator to produce a fuel element for nuclear reactors. According to the invention enriched uranium hydride and zirconium hydride powders of 200 mesh particle size are thoroughly admixed to form a mixture containing 0.1 to 3% by weight of U/sup 235/ hydride. The mixed powders are placed in a die and pressed at 100 tons per square inch at room temperature. The resultant compacts are heated in a vacuum to 300 deg C, whereby the uranium hydride deoomposes into uranium metal and hydrogen gas. The escaping hydrogen gas forms a porous matrix of zirconium hydride, with uramum metal evenly dispersed therethrough. The advantage of the invention is that the porosity and uranium distribution of the final fuel element can be more closely determined and controlled than was possible using prior methods of producing such fuel ele- ments.

  5. Long-term decontamination engineering study. Volume 1

    SciTech Connect (OSTI)

    Geuther, W.J.

    1995-04-03

    This report was prepared by Westinghouse Hanford Company (WHC) with technical and cost estimating support from Pacific Northwest Laboratories (PNL) and Parsons Environmental Services, Inc. (Parsons). This engineering study evaluates the requirements and alternatives for decontamination/treatment of contaminated equipment at the Hanford Site. The purpose of this study is to determine the decontamination/treatment strategy that best supports the Hanford Site environmental restoration mission. It describes the potential waste streams requiring treatment or decontamination, develops the alternatives under consideration establishes the criteria for comparison, evaluates the alternatives, and draws conclusions (i.e., the optimum strategy for decontamination). Although two primary alternatives are discussed, this study does identify other alternatives that may warrant additional study. hanford Site solid waste management program activities include storage, special processing, decontamination/treatment, and disposal facilities. This study focuses on the decontamination/treatment processes (e.g., waste decontamination, size reduction, immobilization, and packaging) that support the environmental restoration mission at the Hanford Site.

  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. 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.

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

    DOE Patents [OSTI]

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

    1995-01-01

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

  9. Engineering evaluation/cost analysis for decontamination at the St. Louis Downtown Site, St. Louis, Missouri

    SciTech Connect (OSTI)

    Picel, M.H.; Hartmann, H.M.; Nimmagadda, M.R. ); Williams, M.J. )

    1991-05-01

    The US Department of Energy (DOE) is implementing a cleanup program for three groups of properties in the St. Louis, Missouri, area: the St. Louis Downtown Site (SLDS), the St. Louis Airport Site (SLAPS) and vicinity properties, and the Latty Avenue Properties, including the Hazelwood Interim Storage Site (HISS). The general location of these properties is shown in Figure 1; the properties are referred to collectively as the St. Louis Site. None of the properties are owned by DOE, but each property contains radioactive residues from federal uranium processing activities conducted at the SLDS during and after World War 2. The activities addressed in this environmental evaluation/cost analysis (EE/CA) report are being proposed as interim components of a comprehensive cleanup strategy for the St. Louis Site. As part of the Department's Formerly Utilized Sites Remedial Action Program (FUSRAP), DOE is proposing to conduct limited decontamination in support of proprietor-initiated activities at the SLDS, commonly referred to as the Mallinckrodt Chemical Works. The primary goal of FUSRAP activity at the SLDS is to eliminate potential environmental hazards associated with residual contamination resulting from the site's use for government-funded uranium processing activities. 17 refs., 3 figs., 5 tabs.

  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. Radioactive hot cell access hole decontamination machine

    DOE Patents [OSTI]

    Simpson, William E.

    1982-01-01

    Radioactive hot cell access hole decontamination machine. A mobile housing has an opening large enough to encircle the access hole and has a shielding door, with a door opening and closing mechanism, for uncovering and covering the opening. The housing contains a shaft which has an apparatus for rotating the shaft and a device for independently translating the shaft from the housing through the opening and access hole into the hot cell chamber. A properly sized cylindrical pig containing wire brushes and cloth or other disks, with an arrangement for releasably attaching it to the end of the shaft, circumferentially cleans the access hole wall of radioactive contamination and thereafter detaches from the shaft to fall into the hot cell chamber.

  12. 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

  13. 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.

  14. PRODUCTION OF URANIUM

    DOE Patents [OSTI]

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

    1957-11-12

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

  15. 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.

  16. 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.

  17. Domestic Uranium Production Report

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

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

  18. Decontamination Technologies, Task 3, Urban Remediation and Response Project

    SciTech Connect (OSTI)

    Heiser,J.; Sullivan, T.

    2009-06-30

    In the aftermath of a Radiological Dispersal Device (RDD, also known as a dirty bomb) it will be necessary to remediate the site including building exteriors and interiors, equipment, pavement, vehicles, personal items etc. Remediation will remove or reduce radioactive contamination from the area using a combination of removing and disposing of many assets (including possible demolition of buildings), decontaminating and returning to service other assets, and fixing in place or leaving in place contamination that is deemed 'acceptable'. The later will require setting acceptable dose standards, which will require negotiation with all involved parties and a balance of risk and cost to benefit. To accomplish the first two, disposal or decontamination, a combination of technologies will be deployed that can be loosely classified as: Decontamination; Equipment removal and size reduction; and Demolition. This report will deal only with the decontamination technologies that will be used to return assets to service or to reduce waste disposal. It will not discuss demolition, size reduction or removal technologies or equipment (e.g., backhoe mounted rams, rock splitter, paving breakers and chipping hammers, etc.). As defined by the DOE (1994), decontamination is removal of radiological contamination from the surfaces of facilities and equipment. Expertise in this field comes primarily from the operation and decommissioning of DOE and commercial nuclear facilities as well as a small amount of ongoing research and development closely related to RDD decontamination. Information related to decontamination of fields, buildings, and public spaces resulting from the Goiania and Chernobyl incidents were also reviewed and provide some meaningful insight into decontamination at major urban areas. In order to proceed with decontamination, the item being processed needs to have an intrinsic value that exceeds the cost of the cleaning and justifies the exposure of any workers during the

  19. 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.

  20. Occupational exposures to uranium: processes, hazards, and regulations

    SciTech Connect (OSTI)

    Stoetzel, G.A.; Fisher, D.R.; McCormack, W.D.; Hoenes, G.R.; Marks, S.; Moore, R.H.; Quilici, D.G.; Breitenstein, B.D.

    1981-04-01

    The United States Uranium Registry (USUR) was formed in 1978 to investigate potential hazards from occupational exposure to uranium and to assess the need for special health-related studies of uranium workers. This report provides a summary of Registry work done to date. The history of the uranium industry is outlined first, and the current commercial uranium industry (mining, milling, conversion, enrichment, and fuel fabrication) is described. This description includes information on basic processes and areas of greatest potential radiological exposure. In addition, inactive commercial facilities and other uranium operations are discussed. Regulation of the commercial production industry for uranium fuel is reported, including the historic development of regulations and the current regulatory agencies and procedures for each phase of the industry. A review of radiological health practices in the industry - facility monitoring, exposure control, exposure evaluation, and record-keeping - is presented. A discussion of the nonradiological hazards of the industry is provided, and the final section describes the tissue program developed as part of the Registry.

  1. Decontamination Systems Information and Research Program. Quarterly technical progress report, January 1--March 31, 1994

    SciTech Connect (OSTI)

    Not Available

    1994-05-01

    West Virginia University (WVU) and the US DOE Morgantown Energy Technology Center (METC) entered into a Cooperative Agreement on August 29, 1992 entitled ``Decontamination Systems Information and Research Programs.`` Stipulated within the Agreement is the requirement that WVU submit to METC a series of Technical Progress Reports on a quarterly basis. This report comprises the first Quarterly Technical Progress Report for Year 2 of the Agreement. This report reflects the progress and/or efforts performed on the sixteen (16) technical projects encompassed by the Year 2 Agreement for the period of January 1 through March 31, 1994. In situ bioremediation of chlorinated organic solvents; Microbial enrichment for enhancing in-situ biodegradation of hazardous organic wastes; Treatment of volatile organic compounds (VOCs) using biofilters; Drain-enhanced soil flushing (DESF) for organic contaminants removal; Chemical destruction of chlorinated organic compounds; Remediation of hazardous sites with steam reforming; Soil decontamination with a packed flotation column; Use of granular activated carbon columns for the simultaneous removal of organics, heavy metals, and radionuclides; Monolayer and multilayer self-assembled polyion films for gas-phase chemical sensors; Compact mercuric iodide detector technology development; Evaluation of IR and mass spectrometric techniques for on-site monitoring of volatile organic compounds; A systematic database of the state of hazardous waste clean-up technologies; Dust control methods for insitu nuclear and hazardous waste handling; Winfield Lock and Dam remediation; and Socio-economic assessment of alternative environmental restoration technologies.

  2. 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...

  3. 2015 Uranium Marketing Annual Report

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

    1 2015 Uranium Marketing Annual Report Release Date: May 24, 2016 Next Release Date: May ... 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

    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 ...

  5. 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 ...

  6. 2015 Uranium Marketing Annual Report

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

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

  7. 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 ...

  8. 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 ...

  9. 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.

  10. 2015 Uranium Market Annual Report

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

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

  11. 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 ...

  12. 2015 Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

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

  13. 2015 Domestic Uranium Production Report

    Gasoline and Diesel Fuel Update (EIA)

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

  14. URANIUM LEACHING AND RECOVERY PROCESS

    DOE Patents [OSTI]

    McClaine, L.A.

    1959-08-18

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

  15. Pilot-scale treatability testing -- Recycle, reuse, and disposal of materials from decontamination and decommissioning activities: Soda blasting demonstration

    SciTech Connect (OSTI)

    1995-08-01

    The US Department of Energy (DOE) is in the process of defining the nature and magnitude of decontamination and decommissioning (D and D) obligations at its sites. With disposal costs rising and available storage facilities decreasing, DOE is exploring and implementing new waste minimizing D and D techniques. Technology demonstrations are being conducted by LMES at a DOE gaseous diffusion processing plant, the K-25 Site, in Oak Ridge, Tennessee. The gaseous diffusion process employed at Oak Ridge separated uranium-235 from uranium ore for use in atomic weapons and commercial reactors. These activities contaminated concrete and other surfaces within the plant with uranium, technetium, and other constituents. The objective of current K-25 D and D research is to make available cost-effective and energy-efficient techniques to advance remediation and waste management methods at the K-25 Site and other DOE sites. To support this objective, O`Brien and Gere tested a decontamination system on K-25 Site concrete and steel surfaces contaminated with radioactive and hazardous waste. A scouring system has been developed that removes fixed hazardous and radioactive surface contamination and minimizes residual waste. This system utilizes an abrasive sodium bicarbonate medium that is projected at contaminated surfaces. It mechanically removes surface contamination while leaving the surface intact. Blasting residuals are captured and dissolved in water and treated using physical/chemical processes. Pilot-scale testing of this soda blasting system and bench and pilot-scale treatment of the generated residuals were conducted from December 1993 to September 1994.

  16. Preliminary investigations on the use of uranium silicide targets for fission Mo-99 production

    SciTech Connect (OSTI)

    Cols, H.; Cristini, P.; Marques, R.

    1997-08-01

    The National Atomic Energy Commission (CNEA) of Argentine Republic owns and operates an installation for production of molybdenum-99 from fission products since 1985, and, since 1991, covers the whole national demand of this nuclide, carrying out a program of weekly productions, achieving an average activity of 13 terabecquerel per week. At present they are finishing an enlargement of the production plant that will allow an increase in the volume of production to about one hundred of terabecquerel. Irradiation targets are uranium/aluminium alloy with 90% enriched uranium with aluminium cladding. In view of international trends held at present for replacing high enrichment uranium (HEU) for enrichment values lower than 20 % (LEU), since 1990 the authors are in contact with the RERTR program, beginning with tests to adapt their separation process to new irradiation target conditions. Uranium silicide (U{sub 3}Si{sub 2}) was chosen as the testing material, because it has an uranium mass per volume unit, so that it allows to reduce enrichment to a value of 20%. CNEA has the technology for manufacturing miniplates of uranium silicide for their purposes. In this way, equivalent amounts of Molybdenum-99 could be obtained with no substantial changes in target parameters and irradiation conditions established for the current process with Al/U alloy. This paper shows results achieved on the use of this new target.

  17. Decontamination of water using nitrate selective ion exchange resin

    DOE Patents [OSTI]

    Lockridge, James E.; Fritz, James S.

    1990-07-31

    A method for nitrate decontamination of water which involves passing the water through a bed of alkyl phosphonium anion exchange resin which has pendant alkyl groups of C.sub.3 or larger.

  18. Release of organic chelating agents from solidified decontamination wastes

    SciTech Connect (OSTI)

    Piciulo, P.L.; Adams, J.W.; Milian, L.W.

    1986-01-01

    In order to provide technical information needed by the US Nuclear Regulatory Commission to evaluate the adequacy of near-surface disposal of decontamination wastes, Brookhaven National Laboratory has measured the release of organic complexing agents from simulated decontamination resin wastes solidified in cement and vinyl ester-styrene. The simulated waste consisted of either mixed bed ion-exchange resins or anion exchange resins equilibrated with EDTA, oxalic acid, citric acid, picolinic acid, formic acid, simulated LOMI reagent or the LND-101A decontamination reagent. The standard procedure ANS 16.1 appeared to be adequate for determining a leachability index for organic acids for comparing the leach resistance of decontamination waste forms. Leachability indexes appeared to be specific for each organic acid. Further, the apparent diffusivities were generally less than those observed for Cs releases from cement wastes forms. The finder material and the composition of the simulated wastes affected the release of the reagents.

  19. Decontamination of water using nitrate selective ion exchange resin

    DOE Patents [OSTI]

    Lockridge, J.E.; Fritz, J.S.

    1990-07-31

    A method for nitrate decontamination of water which involves passing the water through a bed of alkyl phosphonium anion exchange resin which has pendant alkyl groups of C[sub 3] or larger.

  20. Portsmouth Decommissioning and Decontamination Project Director's Final Findings and Order

    Broader source: Energy.gov [DOE]

    Portsmouth Decommissioning and Decontamination (D&D) Project Director's Final Findings and Order defines the steps for identifying a range of technical alternatives for the D&D and waste...

  1. Enhanced toxic cloud knockdown spray system for decontamination applications

    DOE Patents [OSTI]

    Betty, Rita G.; Tucker, Mark D.; Brockmann, John E.; Lucero, Daniel A.; Levin, Bruce L.; Leonard, Jonathan

    2011-09-06

    Methods and systems for knockdown and neutralization of toxic clouds of aerosolized chemical or biological warfare (CBW) agents and toxic industrial chemicals using a non-toxic, non-corrosive aqueous decontamination formulation.

  2. Modification of the Decontamination Facility at the Kruemmel NPP - 13451

    SciTech Connect (OSTI)

    Klute, Stefan; Kupke, Peter

    2013-07-01

    In February 2009, Siempelkamp Nukleartechnik GmbH was awarded the contract for the design, manufacture, delivery and construction of a new Decontamination Facility in the controlled area for Kruemmel NPP. The new decontamination equipment has been installed according to the state of art of Kruemmel NPP. The existing space required the following modification, retrofitting and reconstruction works: - Demounting of the existing installation: to create space for the new facility it was necessary to dismantle the old facility. The concrete walls and ceilings were cut into sizes of no more than 400 kg for ease of handling. This enabled decontamination so largest possible amount could be released for recycling. All steel parts were cut into sizes fitting for iron-barred boxes, respecting the requirement to render the parts decontaminable and releasable. - Reconstructing a decontamination facility: Reconstruction of a decontamination box with separate air lock as access area for the decontamination of components and assemblies was conducted using pressurized air with abrasives (glass beads or steel shots). The walls were equipped with sound protection, the inner walls were welded gap-free to prevent the emergence of interstices and were equipped with changeable wear and tear curtains. Abrasive processing unit positioned underneath the dry blasting box adjacent to the two discharge hoppers. A switch has been installed for the separation of the glass beads and the steel shot. The glass beads are directed into a 200 l drum for the disposal. The steel shot was cleaned using a separator. The cleaned steel shot was routed via transportation devices to the storage container, making it available for further blasting operations. A decontamination box with separate air lock as access area for the decontamination of components and assemblies using high pressure water technology was provided by new construction. Water pressures between 160 bar and 800 bar can be selected. The inner

  3. Steam Generator Group Project. Task 6. Channel head decontamination

    SciTech Connect (OSTI)

    Allen, R.P.; Clark, R.L.; Reece, W.D.

    1984-08-01

    The Steam Generator Group Project utilizes a retired-from-service pressurized-water-reactor steam generator as a test bed and source of specimens for research. An important preparatory step to primary side research activities was reduction of the radiation field in the steam generator channel head. This task report describes the channel head decontamination activities. Though not a programmatic research objective it was judged beneficial to explore the use of dilute reagent chemical decontamination techniques. These techniques presented potential for reduced personnel exposure and reduced secondary radwaste generation, over currently used abrasive blasting techniques. Two techniques with extensive laboratory research and vendors prepared to offer commercial application were tested, one on either side of the channel head. As indicated in the report, both techniques accomplished similar decontamination objectives. Neither technique damaged the generator channel head or tubing materials, as applied. This report provides details of the decontamination operations. Application system and operating conditions are described.

  4. 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/.

  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. 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

  7. Evaluation of destructive methods for managing decontamination wastes

    SciTech Connect (OSTI)

    Piciulo, P.L.; Adams, J.W.

    1986-01-01

    Results are discussed of a laboratory evaluation of destructive methods for processing chemical decontamination wastes. Incineration, acid digestion and wet-air oxidation are capable of degrading decontamination reagents and organic ion-exchange resins. The extent of destruction as a function of operating parameters was waste specific. The reagents used in the testing were: EDTA, oxalic acid, citric acid, picolinic acid and LND-101A.

  8. METHOD AND COATING COMPOSITION FOR PROTECTING AND DECONTAMINATING SURFACES

    DOE Patents [OSTI]

    Overhold, D.C.; Peterson, M.D.

    1959-03-10

    A protective coating useful in the decontamination of surfaces exposed to radioactive substances is presented. This coating is placed on the surface before use and is soluble in waters allowing its easy removal in the event decontamination becomes necessary. Suitable coating compositions may be prepared by mixing a water soluble carbohydrate such as sucrose or dextrin, together with a hygroscopic agent such as calcium chloride or zinc chloride.

  9. Method and coating composition for protecting and decontaminating surfaces

    DOE Patents [OSTI]

    Overhold, D C; Peterson, M D

    1959-03-10

    A protective coating useful in the decontamination of surfaces exposed to radioactive substances is described. This coating is placed on the surface before use and is soluble in water, allowing its easy removal in the event decontamination becomes necessary. Suitable coating compositions may be prepared by mixing a water soluble carbohydrate such as sucrose or dextrin, together with a hygroscopic agent such as calcium chloride or zinc chloride.

  10. Use of Authorized Limits During Decontamination and Demolition Phase |

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

    Department of Energy Use of Authorized Limits During Decontamination and Demolition Phase Use of Authorized Limits During Decontamination and Demolition Phase Donald Dihel; Orville W. Cypret*; G. Vazquez; W. Alexander Williams Abstract: Protecting the public and the environment against undue risk from radiation associated with radiological activities conducted under the control of the Department of Energy (DOE) is the purpose of DOE environmental radiation protection programs. To meet this

  11. Uranium dioxide electrolysis

    SciTech Connect (OSTI)

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

    2009-12-29

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

  12. 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.

  13. 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.

  14. Y-12 Plant decontamination and decommissioning Technology Logic Diagram for Building 9201-4: Volume 3, Technology evaluation data sheets: Part B, Decontamination; robotics/automation; waste management

    SciTech Connect (OSTI)

    1994-09-01

    This volume consists of the Technology Logic Diagrams (TLDs) for the decontamination, robotics/automation, and waste management areas.

  15. A survey of decontamination processes applicable to DOE nuclear facilities

    SciTech Connect (OSTI)

    Chen, L.; Chamberlain, D.B.; Conner, C.; Vandegrift, G.F.

    1997-05-01

    The objective of this survey was to select an appropriate technology for in situ decontamination of equipment interiors as part of the decommissioning of U.S. Department of Energy nuclear facilities. This selection depends on knowledge of existing chemical decontamination methods. This report provides an up-to-date review of chemical decontamination methods. According to available information, aqueous systems are probably the most universally used method for decontaminating and cleaning metal surfaces. We have subdivided the technologies, on the basis of the types of chemical solvents, into acid, alkaline permanganate, highly oxidizing, peroxide, and miscellaneous systems. Two miscellaneous chemical decontamination methods (electrochemical processes and foam and gel systems) are also described. A concise technical description of various processes is given, and the report also outlines technical considerations in the choice of technologies, including decontamination effectiveness, waste handing, fields of application, and the advantages and limitations in application. On the basis of this survey, six processes were identified for further evaluation. 144 refs., 2 tabs.

  16. Anthrax Sampling and Decontamination: Technology Trade-Offs

    SciTech Connect (OSTI)

    Price, Phillip N.; Hamachi, Kristina; McWilliams, Jennifer; Sohn, Michael D.

    2008-09-12

    The goal of this project was to answer the following questions concerning response to a future anthrax release (or suspected release) in a building: 1. Based on past experience, what rules of thumb can be determined concerning: (a) the amount of sampling that may be needed to determine the extent of contamination within a given building; (b) what portions of a building should be sampled; (c) the cost per square foot to decontaminate a given type of building using a given method; (d) the time required to prepare for, and perform, decontamination; (e) the effectiveness of a given decontamination method in a given type of building? 2. Based on past experience, what resources will be spent on evaluating the extent of contamination, performing decontamination, and assessing the effectiveness of the decontamination in abuilding of a given type and size? 3. What are the trade-offs between cost, time, and effectiveness for the various sampling plans, sampling methods, and decontamination methods that have been used in the past?

  17. Nuclear reactor cooling system decontamination reagent regeneration. [PWR; BWR

    DOE Patents [OSTI]

    Anstine, L.D.; James, D.B.; Melaika, E.A.; Peterson, J.P. Jr.

    1980-06-06

    An improved method for decontaminating the coolant system of water-cooled nuclear power reactors and for regenerating the decontamination solution is described. A small amount of one or more weak-acid organic complexing agents is added to the reactor coolant, and the pH is adjusted to form a decontamination solution which is circulated throughout the coolant system to dissolve metal oxides from the interior surfaces and complex the resulting metal ions and radionuclide ions. The coolant containing the complexed metal ions and radionuclide ions is passed through a strong-base anion exchange resin bed which has been presaturated with a solution containing the complexing agents in the same ratio and having the same pH as the decontamination solution. As the decontamination solution passes through the resin bed, metal-complexed anions are exchanged for the metal-ion-free anions on the bed, while metal-ion-free anions in the solution pass through the bed, thus removing the metal ions and regenerating the decontamination solution.

  18. Demonstration recommendations for accelerated testing of concrete decontamination methods

    SciTech Connect (OSTI)

    Dickerson, K.S.; Ally, M.R.; Brown, C.H.; Morris, M.I.; Wilson-Nichols, M.J.

    1995-12-01

    A large number of aging US Department of Energy (DOE) surplus facilities located throughout the US require deactivation, decontamination, and decommissioning. Although several technologies are available commercially for concrete decontamination, emerging technologies with potential to reduce secondary waste and minimize the impact and risk to workers and the environment are needed. In response to these needs, the Accelerated Testing of Concrete Decontamination Methods project team described the nature and extent of contaminated concrete within the DOE complex and identified applicable emerging technologies. Existing information used to describe the nature and extent of contaminated concrete indicates that the most frequently occurring radiological contaminants are {sup 137}Cs, {sup 238}U (and its daughters), {sup 60}Co, {sup 90}Sr, and tritium. The total area of radionuclide-contaminated concrete within the DOE complex is estimated to be in the range of 7.9 {times} 10{sup 8} ft{sup 2}or approximately 18,000 acres. Concrete decontamination problems were matched with emerging technologies to recommend demonstrations considered to provide the most benefit to decontamination of concrete within the DOE complex. Emerging technologies with the most potential benefit were biological decontamination, electro-hydraulic scabbling, electrokinetics, and microwave scabbling.

  19. DECONTAMINATION OF ZIRCALOY SPENT FUEL CLADDING HULLS

    SciTech Connect (OSTI)

    Rudisill, T; John Mickalonis, J

    2006-09-27

    The reprocessing of commercial spent nuclear fuel (SNF) generates a Zircaloy cladding hull waste which requires disposal as a high level waste in the geologic repository. The hulls are primarily contaminated with fission products and actinides from the fuel. During fuel irradiation, these contaminants are deposited in a thin layer of zirconium oxide (ZrO{sub 2}) which forms on the cladding surface at the elevated temperatures present in a nuclear reactor. Therefore, if the hulls are treated to remove the ZrO{sub 2} layer, a majority of the contamination will be removed and the hulls could potentially meet acceptance criteria for disposal as a low level waste (LLW). Discard of the hulls as a LLW would result in significant savings due to the high costs associated with geologic disposal. To assess the feasibility of decontaminating spent fuel cladding hulls, two treatment processes developed for dissolving fuels containing zirconium (Zr) metal or alloys were evaluated. Small-scale dissolution experiments were performed using the ZIRFLEX process which employs a boiling ammonium fluoride (NH{sub 4}F)/ammonium nitrate (NH{sub 4}NO{sub 3}) solution to dissolve Zr or Zircaloy cladding and a hydrofluoric acid (HF) process developed for complete dissolution of Zr-containing fuels. The feasibility experiments were performed using Zircaloy-4 metal coupons which were electrochemically oxidized to produce a thin ZrO{sub 2} layer on the surface. Once the oxide layer was in place, the ease of removing the layer using methods based on the two processes was evaluated. The ZIRFLEX and HF dissolution processes were both successful in removing a 0.2 mm (thick) oxide layer from Zircaloy-4 coupons. Although the ZIRFLEX process was effective in removing the oxide layer, two potential shortcomings were identified. The formation of ammonium hexafluorozirconate ((NH{sub 4}){sub 2}ZrF{sub 6}) on the metal surface prior to dissolution in the bulk solution could hinder the decontamination

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

    SciTech Connect (OSTI)

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

    1995-11-30

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

  1. Final report of the decontamination and decommissioning of Building 39 at the Grand Junction Projects Office Facility

    SciTech Connect (OSTI)

    Widdop, M.R.

    1996-07-01

    The U.S. Department of Energy (DOE) Grand Junction Projects Office (GJPO) occupies a 61.7-acre facility along the Gunnison River near Grand Junction, Colorado. This site was contaminated with uranium ore and mill tailings during uranium refining activities of the Manhattan Engineer District and during pilot milling experiments conducted for the U.S. Atomic Energy Commission`s domestic uranium procurement program. The DOE Defense Decontamination and Decommissioning Program established the GJPO Remedial Action Project to clean up and restore the facility lands, improvements, and the underlying aquifer. The site contractor for the facility, Rust Geotech, is also the remedial action contractor. The soil beneath Building 39 was radiologically contaminated and the building was demolished in 1992. The soil area within the footprint of the building has been remediated in accordance with the identified standards and the area can be released for unlimited exposure and unrestricted use. This document was prepared in response to a DOE request for an individual final report for each contaminated GJPO building.

  2. Final report of the decontamination and decommissioning of Building 34 at the Grand Junction Projects Office Facility

    SciTech Connect (OSTI)

    Widdop, M.R.

    1996-08-01

    The U.S. Department of Energy (DOE) Grand Junction Projects Office (GJPO) occupies a 61.7 acre facility along the Gunnison River near Grand Junction, Colorado. This site was contaminated with uranium ore and mill tailings during uranium refining activities of the Manhattan Engineer District and during pilot milling experiments conducted for the U.S. Atomic Energy Commission`s domestic uranium procurement program. The DOE Defense Decontamination and Decommissioning Program established the Grand Junction Projects Office Remedial Action Project to clean up and restore the facility lands, improvements, and the underlying aquifer. The site contractor for the facility, Rust Geotech, was also the remedial action contractor. Building 34 was radiologically contaminated and the building was demolished in 1996. The soil area within the footprint of the building was analyzed and found to be not contaminated. The area can be released for unlimited exposure and unrestricted use. This document was prepared in response to a DOE request for an individual closeout report for each contaminated GJPO building.

  3. Final report of the decontamination and decommissioning of Building 6 at the Grand Junction Projects Office Facility

    SciTech Connect (OSTI)

    Widdop, M.R.

    1996-07-01

    The U.S. Department of Energy (DOE) Grand Junction Projects Office (GJPO) occupies a 61.7-acre facility along the Gunnison River near Grand Junction, Colorado. This site was contaminated with uranium ore and mill tailings during uranium refining activities of the Manhattan Engineer District and during pilot milling experiments conducted for the domestic uranium procurement program funded by the U.S. Atomic Energy Commission. The DOE Defense Decontamination and Decommissioning Program established the GJPO Remedial Action Project to clean up and restore the facility lands, improvements, and the underlying aquifer. The site contractor for the facility, Rust Geotech, is also the remedial action contractor. Radiological contamination was identified in Building 6, and the building was demolished in 1992. The soil area within the footprint of the building has been remediated in accordance with the identified standards and the area can be released for unlimited exposure and unrestricted use. This document was prepared in response to a DOE request for an individual final report for each contaminated GJPO building.

  4. Final report of the decontamination and decommissioning of Building 44 at the Grand Junction Projects Office Facility

    SciTech Connect (OSTI)

    Widdop, M.R.

    1996-07-01

    The U.S. Department of Energy (DOE) Junction Projects Office (GJPO) occupies a 61.7 acre facility along the Gunnison River near Grand Junction, Colorado. This site was contaminated with uranium ore and mill tailings during uranium refining activities of the Manhattan Engineer District and during pilot milling experiments conducted for the U.S. Atomic Energy Commission`s domestic uranium procurement program. The DOE Defense Decontamination and Decommissioning Program established the Grand Junction Projects Office Remedial Action Project to clean up and restore the facility lands, improvements, and the underlying aquifer. The site contractor for the facility, Rust Geotech, is also the remedial action contractor. Building 44 was radiologically contaminated and the building was demolished in 1994. The soil area within the footprint of the building was not contaminated; it complies with the identified standards and the area can be released for unlimited exposure and unrestricted use. This document was prepared in response to a DOE request for an individual final report for each contaminated GJPO building.

  5. Final report of the decontamination and decommissioning of Building 1 at the Grand Junction Projects Office Facility

    SciTech Connect (OSTI)

    Widdop, M.R.

    1996-08-01

    The U.S. Department of Energy (DOE) Grand Junction Projects Office (GJPO) occupies a 61.7-acre facility along the Gunnison River near Grand Junction, Colorado. This site was contaminated with uranium ore and mill tailings during uranium refining activities of the Manhattan Engineer District and during pilot milling experiments conducted for the U.S. Atomic Energy Commission`s domestic uranium procurement program. The DOE Defense Decontamination and Decommissioning Program established the GJPO Remedial Action Project to clean up and restore the facility lands, improvements, and the underlying aquifer. The site contractor for the facility, Rust Geotech, also is the remedial action contractor. Building 1 was found to be radiologically contaminated and was demolished in 1996. The soil beneath and adjacent to the building was remediated in accordance with identified standards and can be released for unlimited exposure and unrestricted use. This document was prepared in response to a DOE request for an individual final report for each contaminated GJPO building.

  6. Final report of the decontamination and decommission of Building 31 at the Grand Junction Projects Office Facility

    SciTech Connect (OSTI)

    Krabacher, J.E.

    1996-07-01

    The U.S. Department of Energy (DOE) Grand Junction Projects Office (GJPO) occupies a 61.7-acre facility along the Gunnison River near Grand Junction, Colorado. This site was contaminated with uranium ore and mill tailings during uranium refining activities of the Manhattan Engineer District and during pilot milling experiments conducted for the domestic uranium procurement program funded by the U.S. Atomic Energy Commission. The DOE Defense Decontamination and Decommissioning Program established the GJPO Remedial Action Project to clean up and restore the facility lands, improvements, and the underlying aquifer. The site contractor for the facility, Rust Geotech, also was the remedial action contractor. Radiological contamination was identified in Building 31 and the building was demolished in 1992. The soil area within the footprint of the building has been remediated in accordance with the identified standards and the area can be released for unlimited exposure and unrestricted use. This area was addressed in the summary final report of the remediation of the exterior areas of the GJPO facility. This document was prepared in response to a DOE request for an individual final report for each contaminated GJPO building.

  7. Final report of the decontamination and decommissioning of Building 18 at the Grand Junction Projects Office Facility

    SciTech Connect (OSTI)

    Widdop, M.R.

    1996-08-01

    The U.S. Department of Energy (DOE) Grand Junction Projects Office (GJPO) occupies a 61.7-acre facility along the Gunnison River near Grand Junction, Colorado. This site was contaminated with uranium ore and mill tailings during uranium refining activities of the Manhattan Engineer District and during pilot milling experiments conducted for the U.S. Atomic Energy Commission`s domestic uranium procurement program. The DOE Defense Decontamination and Decommissioning Program established the GJPO Remedial Action Project to clean up and restore the facility lands, improvements, and the underlying aquifer. The site contractor for the facility, Rust Geotech, also is the remedial action contractor. The soil beneath Building 18 was found to be radiologically contaminated; the building was not contaminated. The soil was remediated in accordance with identified standards. Building 18 and the underlying soil can be released for unlimited exposure and unrestricted use. This document was prepared in response to a DOE request for an individual final report for each contaminated GJPO building.

  8. Occurrence of Metastudtite (Uranium Peroxide Dihydrate) at a FUSRAP Site

    SciTech Connect (OSTI)

    Young, C.M.; Nelson, K.A.; Stevens, G.T.; Grassi, V.J.

    2006-07-01

    Uranium concentrations in groundwater in a localized area of a site exceed the USEPA Maximum Contaminant Level (MCL) by a factor of one thousand. Although the groundwater seepage velocity ranges up to 0.7 meters per day (m/day), data indicate that the uranium is not migrating in groundwater. We believe that the uranium is not mobile because of local geochemical conditions and the unstable nature of the uranium compound present at the site; uranium peroxide dihydrate (metastudtite). Metastudtite [UO{sub 4}.2(H{sub 2}O) or (U(O{sub 2})|O|(OH){sub 2}).3H{sub 2}O] has been identified at other sites as an alteration product in casks of spent nuclear fuel, but neither enriched nor depleted uranium were present at this site. Metastudtite was first identified as a natural mineral in 1983, although documented occurrences in the environment are uncommon. The U.S. Army Corps of Engineers (USACE) is conducting a remedial investigation at the DuPont Chambers Works in Deep water New Jersey under the Formerly Utilized Sites Remedial Action Program (FUSRAP) to evaluate radioactive contamination resulting from historical activities conducted in support of Manhattan Engineering District operations. From 1942 to 1947, Chambers Works converted uranium oxides to uranium tetrafluoride and uranium metal. More than half of the production at this facility resulted from the recovery process, where uranium-bearing dross and scrap were reacted with hydrogen peroxide to produce uranium peroxide dihydrate. The 280-hectare Chambers Works has produced some 600 products, including petrochemicals, aromatics, fluoro-chemicals, polymers, and elastomers. Contaminants resulting from these processes, including separate-phase petrochemicals, have also been detected within the boundaries of the FUSRAP investigation. USACE initiated remedial investigation field activities in 2002. The radionuclides of concern are natural uranium (U{sub nat}) and its short-lived progeny. Areas of impacted soil generally

  9. 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

  10. Waste Isolation Pilot Plant Salt Decontamination Testing

    SciTech Connect (OSTI)

    Rick Demmer; Stephen Reese

    2014-09-01

    On February 14, 2014, americium and plutonium contamination was released in the Waste Isolation Pilot Plant (WIPP) salt caverns. At the request of WIPP’s operations contractor, Idaho National Laboratory (INL) personnel developed several methods of decontaminating WIPP salt, using surrogate contaminants and also americium (241Am). The effectiveness of the methods is evaluated qualitatively, and to the extent possible, quantitatively. One of the requirements of this effort was delivering initial results and recommendations within a few weeks. That requirement, in combination with the limited scope of the project, made in-depth analysis impractical in some instances. Of the methods tested (dry brushing, vacuum cleaning, water washing, strippable coatings, and mechanical grinding), the most practical seems to be water washing. Effectiveness is very high, and it is very easy and rapid to deploy. The amount of wastewater produced (2 L/m2) would be substantial and may not be easy to manage, but the method is the clear winner from a usability perspective. Removable surface contamination levels (smear results) from the strippable coating and water washing coupons found no residual removable contamination. Thus, whatever is left is likely adhered to (or trapped within) the salt. The other option that shows promise is the use of a fixative barrier. Bartlett Nuclear, Inc.’s Polymeric Barrier System (PBS) proved the most durable of the coatings tested. The coatings were not tested for contaminant entrapment, only for coating integrity and durability.

  11. Kit systems for granulated decontamination formulations

    DOE Patents [OSTI]

    Tucker, Mark D.

    2010-07-06

    A decontamination formulation and method of making that neutralizes the adverse health effects of both chemical and biological compounds, especially chemical warfare (CW) and biological warfare (BW) agents, and toxic industrial chemicals. The formulation provides solubilizing compounds that serve to effectively render the chemical and biological compounds, particularly CW and BW compounds, susceptible to attack, and at least one reactive compound that serves to attack (and detoxify or kill) the compound. The formulation includes at least one solubilizing agent, a reactive compound, a sorbent additive, and water. A highly adsorbent sorbent additive (e.g., amorphous silica, sorbitol, mannitol, etc.) is used to "dry out" one or more liquid ingredients into a dry, free-flowing powder that has an extended shelf life, and is more convenient to handle and mix in the field. The formulation can be pre-mixed and pre-packaged as a multi-part kit system, where one or more of the parts are packaged in a powdered, granulated form for ease of handling and mixing in the field.

  12. Mobile workstation for decontamination and decommissioning operations

    SciTech Connect (OSTI)

    Whittaker, W.L.; Osborn, J.F.; Thompson, B.R.

    1993-10-01

    This project is an interdisciplinary effort to develop effective mobile worksystems for decontamination and decommissioning (D&D) of facilities within the DOE Nuclear Weapons Complex. These mobile worksystems will be configured to operate within the environmental and logistical constraints of such facilities and to perform a number of work tasks. Our program is designed to produce a mobile worksystem with capabilities and features that are matched to the particular needs of D&D work by evolving the design through a series of technological developments, performance tests and evaluations. The project has three phases. In this the first phase, an existing teleoperated worksystem, the Remote Work Vehicle (developed for use in the Three Mile Island Unit 2 Reactor Building basement), was enhanced for telerobotic performance of several D&D operations. Its ability to perform these operations was then assessed through a series of tests in a mockup facility that contained generic structures and equipment similar to those that D&D work machines will encounter in DOE facilities. Building upon the knowledge gained through those tests and evaluations, a next generation mobile worksystem, the RWV II, and a more advanced controller will be designed, integrated and tested in the second phase, which is scheduled for completion in January 1995. The third phase of the project will involve testing of the RWV II in the real DOE facility.

  13. Decontamination and decommissioning of Shippingport commercial reactor

    SciTech Connect (OSTI)

    Schreiber, J.

    1989-11-01

    To a certain degree, the decontamination and decommissioning (D and D) of the Shippingport reactor was a joint venture with Duquesne Light Company. The structures that were to be decommissioned were to be removed to at least three feet below grade. Since the land had been leased from Duquesne Light, there was an agreement with them to return the land to them in a radiologically safe condition. The total enclosure volume for the steam and nuclear containment systems was about 1.3 million cubic feet, more than 80% of which was below ground. Engineering plans for the project were started in July of 1980 and the final environmental impact statement (EIS) was published in May of 1982. The plant itself was shut down in October of 1982 for end-of-life testing and defueling. The engineering services portion of the decommissioning plans was completed in September of 1983. DOE moved onto the site and took over from the Navy in September of 1984. Actual physical decommissioning began after about a year of preparation and was completed about 44 months later in July of 1989. This paper describes the main parts of D and D.

  14. PROCESS OF RECOVERING URANIUM

    DOE Patents [OSTI]

    Kilner, S.B.

    1959-12-29

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

  15. 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.

  16. Depleted uranium management alternatives

    SciTech Connect (OSTI)

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

    1994-08-01

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

  17. A simplified model of decontamination by BWR steam suppression pools

    SciTech Connect (OSTI)

    Powers, D.A.

    1997-05-01

    Phenomena that can decontaminate aerosol-laden gases sparging through steam suppression pools of boiling water reactors during reactor accidents are described. Uncertainties in aerosol properties, aerosol behavior within gas bubbles, and bubble behavior in plumes affect predictions of decontamination by steam suppression pools. Uncertainties in the boundary and initial conditions that are dictated by the progression of severe reactor accidents and that will affect predictions of decontamination by steam suppression pools are discussed. Ten parameters that characterize boundary and initial condition uncertainties, nine parameters that characterize aerosol property and behavior uncertainties, and eleven parameters that characterize uncertainties in the behavior of bubbles in steam suppression pools are identified. Ranges for the values of these parameters and subjective probability distributions for parametric values within the ranges are defined. These uncertain parameters are used in Monte Carlo uncertainty analyses to develop uncertainty distributions for the decontamination that can be achieved by steam suppression pools and the size distribution of aerosols that do emerge from such pools. A simplified model of decontamination by steam suppression pools is developed by correlating features of the uncertainty distributions for total decontamination factor, DF(total), mean size of emerging aerosol particles, d{sub p}, and the standard deviation of the emerging aerosol size distribution, {sigma}, with pool depth, H. Correlations of the median values of the uncertainty distributions are suggested as the best estimate of decontamination by suppression pools. Correlations of the 10 percentile and 90 percentile values of the uncertainty distributions characterize the uncertainty in the best estimates. 295 refs., 121 figs., 113 tabs.

  18. 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.

  19. 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.

  20. URANIUM RECOVERY PROCESS

    DOE Patents [OSTI]

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

    1959-07-28

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