Sample records for reactors projections uranium

  1. International Atomic Energy Agency support of research reactor highly enriched uranium to low enriched uranium fuel conversion projects

    SciTech Connect (OSTI)

    Bradley, E.; Adelfang, P.; Goldman, I.N. [Research Reactors Unit, Division of Nuclear Fuel Cycle and Waste Technology, International Atomic Energy Agency, Wagramer Strasse 5, P.O. Box 100, A-1400 Vienna (Austria)

    2008-07-15T23:59:59.000Z

    The IAEA has been involved for more than twenty years in supporting international nuclear non- proliferation efforts associated with reducing the amount of highly enriched uranium (HEU) in international commerce. IAEA projects and activities have directly supported the Reduced Enrichment for Research and Test Reactors (RERTR) programme, as well as directly assisted efforts to convert research reactors from HEU to LEU fuel. HEU to LEU fuel conversion projects differ significantly depending on several factors including the design of the reactor and fuel, technical needs of the member state, local nuclear infrastructure, and available resources. To support such diverse endeavours, the IAEA tailors each project to address the relevant constraints. This paper presents the different approaches taken by the IAEA to address the diverse challenges involved in research reactor HEU to LEU fuel conversion projects. Examples of conversion related projects in different Member States are fully detailed. (author)

  2. Innovative design of uranium startup fast reactors

    E-Print Network [OSTI]

    Fei, Tingzhou

    2012-01-01T23:59:59.000Z

    Sodium Fast Reactors are one of the three candidates of GEN-IV fast reactors. Fast reactors play an important role in saving uranium resources and reducing nuclear wastes. Conventional fast reactors rely on transuranic ...

  3. Molten-Salt Depleted-Uranium Reactor

    E-Print Network [OSTI]

    Dong, Bao-Guo; Gu, Ji-Yuan

    2015-01-01T23:59:59.000Z

    The supercritical, reactor core melting and nuclear fuel leaking accidents have troubled fission reactors for decades, and greatly limit their extensive applications. Now these troubles are still open. Here we first show a possible perfect reactor, Molten-Salt Depleted-Uranium Reactor which is no above accident trouble. We found this reactor could be realized in practical applications in terms of all of the scientific principle, principle of operation, technology, and engineering. Our results demonstrate how these reactors can possess and realize extraordinary excellent characteristics, no prompt critical, long-term safe and stable operation with negative feedback, closed uranium-plutonium cycle chain within the vessel, normal operation only with depleted-uranium, and depleted-uranium high burnup in reality, to realize with fission nuclear energy sufficiently satisfying humanity long-term energy resource needs, as well as thoroughly solve the challenges of nuclear criticality safety, uranium resource insuffic...

  4. Uranium Mill Tailings Remedial Action Project surface project management plan

    SciTech Connect (OSTI)

    Not Available

    1994-09-01T23:59:59.000Z

    This Project Management Plan describes the planning, systems, and organization that shall be used to manage the Uranium Mill Tailings Remedial Action Project (UMTRA). US DOE is authorized to stabilize and control surface tailings and ground water contamination at 24 inactive uranium processing sites and associated vicinity properties containing uranium mill tailings and related residual radioactive materials.

  5. SHEEP MOUNTAIN URANIUM PROJECT CROOKS GAP, WYOMING

    E-Print Network [OSTI]

    ;PROJECT OVERVIEW ·Site Location·Site Location ·Fremont , Wyoming ·Existing Uranium Mine Permit 381C·Existing Uranium Mine Permit 381C ·Historical Operation ·Western Nuclear Crooks Gap Project ·Mined 1956 ­ 1988 and Open Pit Mining ·Current Mine Permit (381C) ·Updating POO, Reclamation Plan & Bond ·Uranium Recovery

  6. WISE Uranium Project - Fact Sheet

    E-Print Network [OSTI]

    Hazards From Depleted

    t in the depleted uranium. For this purpose, we first need to calculate the mass balance of the enrichment process. We then calculate the inhalation doses from the depleted uranium and compare the dose contributions from the nuclides of interest. Mass balance for uranium enrichment at Paducah [DOE_1984, p.35] Feed Product Tails Other Mass [st] 758002 124718 621894 11390 Mass fraction 100.00% 16.45% 82.04% 1.50% Concentration of plutonium in tails (depleted uranium) from enrichment of reprocessed uranium, assuming that all plutonium were transfered to the tails: Concentration of neptunium in tails from enrichment of reprocessed uranium uranium, assuming that all neptunium were transfered to the tails: - 2 - Schematic of historic uranium enrichment process at Paducah [DOE_1999b] - -7 For comparison, we first calculate the inhalation dose from depleted uranium produced from natural uranium. We assume that the short-lived decay products have reached secular equilibrium with th

  7. Liquid uranium alloy-helium fission reactor

    DOE Patents [OSTI]

    Minkov, V.

    1984-06-13T23:59:59.000Z

    This invention describes a nuclear fission reactor which has a core vessel and at least one tandem heat exchanger vessel coupled therewith across upper and lower passages to define a closed flow loop. Nuclear fuel such as a uranium alloy in its liquid phase fills these vessels and flow passages. Solid control elements in the reactor core vessel are adapted to be adjusted relative to one another to control fission reaction of the liquid fuel therein. Moderator elements in the other vessel and flow passages preclude fission reaction therein. An inert gas such as helium is bubbled upwardly through the heat exchanger vessel operable to move the liquid fuel upwardly therein and unidirectionally around the closed loop and downwardly through the core vessel. This helium gas is further directed to heat conversion means outside of the reactor vessels to utilize the heat from the fission reaction to generate useful output. The nuclear fuel operates in the 1200 to 1800/sup 0/C range, and even higher to 2500/sup 0/C.

  8. Liquid uranium alloy-helium fission reactor

    DOE Patents [OSTI]

    Minkov, Vladimir (Skokie, IL)

    1986-01-01T23:59:59.000Z

    This invention teaches a nuclear fission reactor having a core vessel and at least one tandem heat exchanger vessel coupled therewith across upper and lower passages to define a closed flow loop. Nuclear fuel such as a uranium alloy in its liquid phase fills these vessels and flow passages. Solid control elements in the reactor core vessel are adapted to be adjusted relative to one another to control fission reaction of the liquid fuel therein. Moderator elements in the other vessel and flow passages preclude fission reaction therein. An inert gas such as helium is bubbled upwardly through the heat exchanger vessel operable to move the liquid fuel upwardly therein and unidirectionally around the closed loop and downwardly through the core vessel. This helium gas is further directed to heat conversion means outside of the reactor vessels to utilize the heat from the fission reaction to generate useful output. The nuclear fuel operates in the 1200.degree.-1800.degree. C. range, and even higher to 2500.degree. C., limited only by the thermal effectiveness of the structural materials, increasing the efficiency of power generation from the normal 30-35% with 300.degree.-500.degree. C. upper limit temperature to 50-65%. Irradiation of the circulating liquid fuel, as contrasted to only localized irradiation of a solid fuel, provides improved fuel utilization.

  9. International Research Reactor Decommissioning Project

    SciTech Connect (OSTI)

    Leopando, Leonardo [Philippine Nuclear Research Institute, Quezon City (Philippines); Warnecke, Ernst [International Atomic Energy Agency, Vienna (Austria)

    2008-01-15T23:59:59.000Z

    Many research reactors have been or will be shut down and are candidates for decommissioning. Most of the respective countries neither have a decommissioning policy nor the required expertise and funds to effectively implement a decommissioning project. The IAEA established the Research Reactor Decommissioning Demonstration Project (R{sup 2}D{sup 2}P) to help answer this need. It was agreed to involve the Philippine Research Reactor (PRR-1) as model reactor to demonstrate 'hands-on' experience as it is just starting the decommissioning process. Other facilities may be included in the project as they fit into the scope of R{sup 2}D{sup 2}P and complement to the PRR-1 decommissioning activities. The key outcome of the R{sup 2}D{sup 2}P will be the decommissioning of the PRR-1 reactor. On the way to this final goal the preparation of safety related documents (i.e., decommissioning plan, environmental impact assessment, safety analysis report, health and safety plan, cost estimate, etc.) and the licensing process as well as the actual dismantling activities could provide a model to other countries involved in the project. It is expected that the R{sup 2}D{sup 2}P would initiate activities related to planning and funding of decommissioning activities in the participating countries if that has not yet been done.

  10. Uranium Mill Tailings Remedial Action (UMTRA) Project. [UMTRA project

    SciTech Connect (OSTI)

    Not Available

    1989-09-01T23:59:59.000Z

    The mission of the Uranium Mill Tailings Remedial Action (UMTRA) Project is explicitly stated and directed in the Uranium Mill Tailings Radiation Control Act of 1978, hereinafter referred to as the Act.'' Title I of the Act authorizes the Department of Energy (DOE) to undertake remedial action at designated inactive uranium processing sites (Attachment 1 and 2) and associated vicinity properties containing uranium mill tailings and other residual radioactive materials derived from the processing site. The purpose of the remedial actions is to stabilize and control such uranium mill tailings and other residual radioactive materials in a safe and environmentally sound manner to minimize radiation health hazards to the public. The principal health hazards and environmental concerns are: the inhalation of air particulates contaminated as a result of the emanation of radon from the tailings piles and the subsequent decay of radon daughters; and the contamination of surface and groundwaters with radionuclides or other chemically toxic materials. This UMTRA Project Plan identifies the mission and objectives of the project, outlines the technical and managerial approach for achieving them, and summarizes the performance, cost, and schedule baselines which have been established to guide operational activity. Estimated cost increases by 15 percent, or if the schedule slips by six months. 4 refs.

  11. Development of a low enrichment uranium core for the MIT reactor

    E-Print Network [OSTI]

    Newton, Thomas Henderson

    2006-01-01T23:59:59.000Z

    An investigation has been made into converting the MIT research reactor from using high enrichment uranium (HEU) to low enrichment uranium (LEU) with a newly developed fuel material. The LEU fuel introduces negative ...

  12. Thermal hydraulics analysis of the MIT research reactor in support of a low enrichment uranium (LEU) core conversion .

    E-Print Network [OSTI]

    Ko, Yu-Chih, Ph. D. Massachusetts Institute of Technology

    2008-01-01T23:59:59.000Z

    ??The MIT research reactor (MITR) is converting from the existing high enrichment uranium (HEU) core to a low enrichment uranium (LEU) core using a high-density (more)

  13. MANHATTAN PROJECT B REACTOR HANFORD WASHINGTON [HANFORD'S HISTORIC B REACTOR (12-PAGE BOOKLET)

    SciTech Connect (OSTI)

    GERBER MS

    2009-04-28T23:59:59.000Z

    The Hanford Site began as part of the United States Manhattan Project to research, test and build atomic weapons during World War II. The original 670-square mile Hanford Site, then known as the Hanford Engineer Works, was the last of three top-secret sites constructed in order to produce enriched uranium and plutonium for the world's first nuclear weapons. B Reactor, located about 45 miles northwest of Richland, Washington, is the world's first full-scale nuclear reactor. Not only was B Reactor a first-of-a-kind engineering structure, it was built and fully functional in just 11 months. Eventually, the shoreline of the Columbia River in southeastern Washington State held nine nuclear reactors at the height of Hanford's nuclear defense production during the Cold War era. The B Reactor was shut down in 1968. During the 1980's, the U.S. Department of Energy began removing B Reactor's support facilities. The reactor building, the river pumphouse and the reactor stack are the only facilities that remain. Today, the U.S. Department of Energy (DOE) Richland Operations Office offers escorted public access to B Reactor along a designated tour route. The National Park Service (NPS) is studying preservation and interpretation options for sites associated with the Manhattan Project. A draft is expected in summer 2009. A final report will recommend whether the B Reactor, along with other Manhattan Project facilities, should be preserved, and if so, what roles the DOE, the NPS and community partners will play in preservation and public education. In August 2008, the DOE announced plans to open B Reactor for additional public tours. Potential hazards still exist within the building. However, the approved tour route is safe for visitors and workers. DOE may open additional areas once it can assure public safety by mitigating hazards.

  14. Powder Metallurgy of Uranium Alloy Fuels for TRU-Burning Reactors Final Technical Report

    SciTech Connect (OSTI)

    Sean M. McDeavitt

    2011-04-29T23:59:59.000Z

    Overview Fast reactors were evaluated to enable the transmutation of transuranic isotopes generated by nuclear energy systems. The motivation for this was that TRU isotopes have high radiotoxicity and relatively long half-lives, making them unattractive for disposal in a long-term geologic repository. Fast reactors provide an efficient means to utilize the energy content of the TRUs while destroying them. An enabling technology that requires research and development is the fabrication metallic fuel containing TRU isotopes using powder metallurgy methods. This project focused upon developing a powder metallurgical fabrication method to produce U-Zr-transuranic (TRU) alloys at relatively low processing temperatures (500C to 600C) using either hot extrusion or alpha-phase sintering for charecterization. Researchers quantified the fundamental aspects of both processing methods using surrogate metals to simulate the TRU elements. The process produced novel solutions to some of the issues relating to metallic fuels, such as fuel-cladding chemical interactions, fuel swelling, volatility losses during casting, and casting mold material losses. Workscope There were two primary tasks associated with this project: 1. Hot working fabrication using mechanical alloying and extrusion Design, fabricate, and assemble extrusion equipment Extrusion database on DU metal Extrusion database on U-10Zr alloys Extrusion database on U-20xx-10Zr alloys Evaluation and testing of tube sheath metals 2. Low-temperature sintering of U alloys Design, fabricate, and assemble equipment Sintering database on DU metal Sintering database on U-10Zr alloys Liquid assisted phase sintering on U-20xx-10Zr alloys Appendices Outline Appendix A contains a Fuel Cycle Research & Development (FCR&D) poster and contact presentation where TAMU made primary contributions. Appendix B contains MSNE theses and final defense presentations by David Garnetti and Grant Helmreich outlining the beginning of the materials processing setup. Also included within this section is a thesis proposal by Jeff Hausaman. Appendix C contains the public papers and presentations introduced at the 2010 American Nuclear Society Winter Meeting. Appendix AMSNE theses of David Garnetti and Grant Helmreich and proposal by Jeff Hausaman A.1 December 2009 Thesis by David Garnetti entitled Uranium Powder Production Via Hydride Formation and Alpha Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications A.2 September 2009 Presentation by David Garnetti (same title as document in Appendix B.1) A.3 December 2010 Thesis by Grant Helmreich entitled Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications A.4 October 2010 Presentation by Grant Helmreich (same title as document in Appendix B.3) A.5 Thesis Proposal by Jeffrey Hausaman entitled Hot Extrusion of Alpha Phase Uranium-Zirconium Alloys for TRU Burning Fast Reactors Appendix BExternal presentations introduced at the 2010 ANS Winter Meeting B.1 J.S. Hausaman, D.J. Garnetti, and S.M. McDeavitt, Powder Metallurgy of Alpha Phase Uranium Alloys for TRU Burning Fast Reactors, Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.2 PowerPoint Presentation Slides from C.1 B.3 G.W. Helmreich, W.J. Sames, D.J. Garnetti, and S.M. McDeavitt, Uranium Powder Production Using a Hydride-Dehydride Process, Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.4. PowerPoint Presentation Slides from C.3 B.5 Poster Presentation from C.3 Appendix CFuel cycle research and development undergraduate materials and poster presentation C.1 Poster entitled Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys presented at the Fuel Cycle Technologies Program Annual Meeting C.2 April 2011 Honors Undergraduate Thesis by William Sames, Research Fellow, entitled Uranium Metal Powder Production, Particle Dis

  15. Final Scientific/Technical Report for Project entitled "Mechanism of Uranium Reduction by Shewanella oneidensis"

    SciTech Connect (OSTI)

    DiChristina, Thomas J. [Georgia Tech

    2013-04-30T23:59:59.000Z

    Final Scientific/Technical Report for Project entitled "Mechanism of Uranium Reduction by Shewanella oneidensis"

  16. Thermal hydraulic limits analysis for the MIT Research Reactor low enrichment uranium core conversion using statistical propagation of parametric uncertainties

    E-Print Network [OSTI]

    Chiang, Keng-Yen

    2012-01-01T23:59:59.000Z

    The MIT Research Reactor (MITR) is evaluating the conversion from highly enriched uranium (HEU) to low enrichment uranium (LEU) fuel. In addition to the fuel element re-design from 15 to 18 plates per element, a reactor ...

  17. Reactor physics project final report

    E-Print Network [OSTI]

    Driscoll, Michael J.

    1970-01-01T23:59:59.000Z

    This is the final report in an experimental and theoretical program to develop and apply single- and few-element methods for the determination of reactor lattice parameters. The period covered by the report is January 1, ...

  18. Uranium Mill Tailings Remedial Action Project 1993 Environmental Report

    SciTech Connect (OSTI)

    Not Available

    1994-10-01T23:59:59.000Z

    This annual report documents the Uranium Mill Tailing Remedial Action (UMTRA) Project environmental monitoring and protection program. The UMTRA Project routinely monitors radiation, radioactive residual materials, and hazardous constituents at associated former uranium tailings processing sites and disposal sites. At the end of 1993, surface remedial action was complete at 10 of the 24 designated UMTRA Project processing sites. In 1993 the UMTRA Project office revised the UMTRA Project Environmental Protection Implementation Plan, as required by the US DOE. Because the UMTRA Project sites are in different stages of remedial action, the breadth of the UMTRA environmental protection program differs from site to site. In general, sites actively undergoing surface remedial action have the most comprehensive environmental programs for sampling media. At sites where surface remedial action is complete and at sites where remedial action has not yet begun, the environmental program consists primarily of surface water and ground water monitoring to support site characterization, baseline risk assessments, or disposal site performance assessments.

  19. Uranium Mill Tailings Remedial Action (UMTRA) Surface Project: Project plan. Revision 1

    SciTech Connect (OSTI)

    Not Available

    1993-08-11T23:59:59.000Z

    The Uranium Mill Tailings Radiation Control Act of 1978 (UMTRCA) [Public Law (PL) 95-604, 42 United States Code (USC) 7901], hereinafter referred to as the ``Act,`` authorizes the US Department of Energy (DOE) to stabilize and control surface tailings and ground water contamination. To fulfill this mission, the DOE has established two projects under the Uranium Mill Tailings Remedial Action (UMTRA) Project Office. The Ground Water Project was established in April 1991 as a major project and a separate project plan will be prepared for that portion of the mission. This project plan covers the UMTRA Surface Project, a major system acquisition (MSA).

  20. Uranium mill tailings remedial action project real estate management plan

    SciTech Connect (OSTI)

    Not Available

    1994-09-01T23:59:59.000Z

    This plan summarizes the real estate requirements of the US Department of Energy`s (DOE) Uranium Mill Tailings Action (UMTRA) Project, identifies the roles and responsibilities of project participants involved in real estate activities, and describes the approaches used for completing these requirements. This document is intended to serve as a practical guide for all project participants. It is intended to be consistent with all formal agreements, but if a conflict is identified, the formal agreements will take precedence.

  1. Analysis of strategies for improving uranium utilization in pressurized water reactors

    E-Print Network [OSTI]

    Sefcik, Joseph A.

    1981-01-01T23:59:59.000Z

    Systematic procedures have been devised and applied to evaluate core design and fuel management strategies for improving uranium utilization in Pressurized Water Reactors operated on a once-through fuel cycle. A principal ...

  2. The selective use of thorium and heterogeneity in uranium-efficient pressurized water reactors

    E-Print Network [OSTI]

    Kamal, Altamash

    1982-01-01T23:59:59.000Z

    Systematic procedures have been developed and applied to assess the uranium utilization potential of a broad range of options involving the selective use of thorium in Pressurized Water Reactors (PWRs) operating on the ...

  3. Uranium Mill Tailings Remedial Action Project, Surface Project Management Plan. Revision 1

    SciTech Connect (OSTI)

    Not Available

    1994-12-01T23:59:59.000Z

    Title I of the Uranium Mill Tailings Radiation Control Act (UMTRCA) authorizes the US Department of Energy (DOE) to undertake remedial action at 24 designated inactive uranium processing sites and associated vicinity properties (VP) containing uranium mill tailings and related residual radioactive materials. The purpose of the Uranium Mill Tailings Remedial Action (UMTRA) Surface Project is to minimize or eliminate radiation health hazards to the public and the environment at the 24 sites and related VPs. This document describes the management organization, system, and methods used to manage the design, construction, and other activities required to clean up the designated sites and associated VPs, in accordance with the UMTRCA.

  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 [ORNL; Guida, Tracey [University of Pittsburgh

    2010-02-01T23:59:59.000Z

    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. Estimate of radiation release from MIT reactor with low enriched uranium (LEU) core during maximum hypothetical accident

    E-Print Network [OSTI]

    Plumer, Kevin E. (Kevin Edward)

    2011-01-01T23:59:59.000Z

    In accordance with a 1986 NRC ruling, the MIT Research Reactor (MITR) is planning on converting from the use of highly enriched uranium (HEU) to low enriched uranium (LEU) for fuel. A component of the conversion analysis ...

  6. Plutonium recovery from spent reactor fuel by uranium displacement

    DOE Patents [OSTI]

    Ackerman, John P. (Downers Grove, IL)

    1992-01-01T23:59:59.000Z

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

  7. Plutonium recovery from spent reactor fuel by uranium displacement

    DOE Patents [OSTI]

    Ackerman, J.P.

    1992-03-17T23:59:59.000Z

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

  8. UMTRA -- The US Uranium Mill Tailings Remedial Action Project

    SciTech Connect (OSTI)

    Lightner, R. [Dept. of Energy, Washington, DC (United States); Cormier, C. [Department of Energy, Albuquerque, NM (United States); Bierley, D. [Roy F. Weston, Inc., Albuquerque, NM (United States)

    1995-12-31T23:59:59.000Z

    In the late 1970s, the United States (US) established the first comprehensive regulatory structure for the management, disposal, and long-term care of wastes produced from its domestic uranium processing industry. This regulatory framework was established through the passage of the Uranium Mill Tailings Radiation Control Act of 1978, often referred to as UMTRCA. This legislation created the Uranium Mill Tailings Remedial Action (UMTRA) Project and assigned the US Department of Energy (DOE) the lead in conducting the required remedial action at 24 designated inactive uranium ore processing sites. With the majority of these 22 sites complete, the DOE`s UMTRA Project has established a distinguished reputation for safely and effectively remediating these low-level waste sites in a complex regulatory and socioeconomic environment. This paper describes the past accomplishments and current status of the UMTRA Project and discusses the DOE`s plans for addressing ground water contamination associated with these sites and its commitment to continuing the long-term care and management of these disposal cells.

  9. Criticality and Characteristic Neutronic Analysis of a Transient-State Shockwave in a Pulsed Spherical Gaseous Uranium-Hexafluoride Reactor.

    E-Print Network [OSTI]

    Boles, Jeremiah

    2013-01-01T23:59:59.000Z

    ??The purpose of this study is to analyze the theoretical criticality of a spherical uranium-hexafluoride reactor with a transient, pulsed shockwave emanating from the center (more)

  10. Manhattan Project: More Uranium Research, 1942

    Office of Scientific and Technical Information (OSTI)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary)morphinanInformation InInformation In closing,-- Energy, science,LinksCP-1 GOESF Reactor"Met

  11. Moab Uranium Mill Tailings Cleanup Project Steps into Spotlight at International Meeting in Vienna

    Broader source: Energy.gov [DOE]

    VIENNA The Moab Uranium Mill Tailings Remedial Action (UMTRA) Project has kept the United States at the forefront of characterization, remediation, and end-state reuse of uranium millsites around the world.

  12. E-Print Network 3.0 - australian uranium projects Sample Search...

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

    Powered by Explorit Topic List Advanced Search Sample search results for: australian uranium projects Page: << < 1 2 3 4 5 > >> 1 geology and Ranger 1 open-pit uranium mine in...

  13. Uranium Mill Tailings Remedial Action Project (UMTRAP) Public Participation Plan

    SciTech Connect (OSTI)

    NONE

    1981-05-01T23:59:59.000Z

    The purpose of this Public Participation Plan is to explain the Department of Energy`s plan for involving the public in the decision-making process related to the Uranium Mill Tailings Remedial Action (UMTRA) Project. This project was authorized by Congress in the Uranium Mill Tailings Radiation Control Act of 1978. The Act provides for a cooperative effort with affected states and Indian tribes for the eventual cleanup of abandoned or inactive uranium mill tailings sites, which are located in nine western states and in Pennsylvania. Section 111 of the Act states, ``in carrying out the provisions of this title, including the designation of processing sites, establishing priorities for such sites, the selection of remedial actions and the execution of cooperative agreements, the Secretary (of Energy), the Administrator (of the Environmental Protection Agency), and the (Nuclear Regulatory) Commission shall encourage public participation and, where appropriate, the Secretary shall hold public hearings relative to such matters in the States where processing sites and disposal sites are located.`` The objective of this document is to show when, where, and how the public will be involved in this project.

  14. UMTRA (Uranium Mill Tailings Remedial Action) Project site management manual

    SciTech Connect (OSTI)

    Not Available

    1990-10-01T23:59:59.000Z

    The purpose of this manual is to summarize the organizational interfaces and the technical approach used to manage the planning, design development, National Environmental Policy Act (NEPA) compliance, engineering, and remedial action required to stabilize and control the designated Uranium Mill Tailings Remedial Action (UMTRA) Project sites. This manual describes the Project's objective, participants' roles and responsibilities, technical approach for accomplishing the objective, and planning and managerial controls to be used in performing the site work. The narrative follows the flow of activities depicted in Figure 1.1, which provides the typical sequence of key Project activities. A list of acronyms used is presented at the end of the manual. The comparable manual for UMTRA Project vicinity properties is the Vicinity Properties Management and Implementation Manual'' (VPMIM) (UMTRA-DOE/AL-050601). Together, the two manuals cover the remedial action activities associated with UMTRA Project sites. The UMTRA Project's objective is to stabilize and control the uranium mill tailings, vicinity property materials, and other residual radioactive materials at the designated sites (Figure 1.2) in a safe and environmentally sound manner in order to minimize radiation health hazards to the public. 26 figs., 6 tabs.

  15. Uranium Mill Tailings Remedial Action Project 1994 environmental report

    SciTech Connect (OSTI)

    NONE

    1995-08-01T23:59:59.000Z

    This annual report documents the Uranium Mill Tailings Remedial Action (UMTRA) Project environmental monitoring and protection program. The UMTRA Project routinely monitors radiation, radioactive residual materials, and hazardous constituents at associated former uranium tailings processing sites and disposal sites. At the end of 1994, surface remedial action was complete at 14 of the 24 designated UMTRA Project processing sites: Canonsburg, Pennsylvania; Durango, Colorado; Grand Junction, Colorado; Green River Utah, Lakeview, Oregon; Lowman, Idaho; Mexican Hat, Utah; Riverton, Wyoming; Salt Lake City, Utah; Falls City, Texas; Shiprock, New Mexico; Spook, Wyoming, Tuba City, Arizona; and Monument Valley, Arizona. Surface remedial action was ongoing at 5 sites: Ambrosia Lake, New Mexico; Naturita, Colorado; Gunnison, Colorado; and Rifle, Colorado (2 sites). Remedial action has not begun at the 5 remaining UMTRA Project sites that are in the planning stage. Belfield and Bowman, North Dakota; Maybell, Colorado; and Slick Rock, Colorado (2 sites). The ground water compliance phase of the UMTRA Project started in 1991. Because the UMTRA Project sites are.` different stages of remedial action, the breadth of the UMTRA environmental protection program differs from site to site. In general, sites actively undergoing surface remedial action have the most comprehensive environmental programs for sampling media. At sites where surface remedial action is complete and at sites where remedial action has not yet begun, the environmental program consists primarily of surface water and ground water monitoring to support site characterization, baseline risk assessments, or disposal site performance assessments.

  16. Uranium Downblending and Disposition Project Technology Readiness

    Energy Savers [EERE]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of Inspector GeneralDepartment of EnergyofProject is on Track|Solar DecathlonManufacturing LoanMaterial

  17. Recovery and Blend-Down Uranium for Beneficial use in Commercial Reactors - 13373

    SciTech Connect (OSTI)

    Magoulas, Virginia [Savannah River National Laboratory, Savannah River Site, Aiken, SC 29808 (United States)] [Savannah River National Laboratory, Savannah River Site, Aiken, SC 29808 (United States)

    2013-07-01T23:59:59.000Z

    In April 2001 the Department of Energy (DOE) and the Tennessee Valley Authority (TVA) signed an Interagency Agreement to transfer approximately 33 MT of off-specification (off-spec) highly enriched uranium (HEU) from DOE to TVA for conversion to commercial reactor fuel. Since that time additional surplus off-spec HEU material has been added to the program, making the total approximately 46 MT off-spec HEU. The disposition path for approximately half (23 MT) of this 46 MT of surplus HEU material, was down blending through the H-canyon facility at the Savannah River Site (SRS). The HEU is purified through the H-canyon processes, and then blended with natural uranium (NU) to form low enriched uranium (LEU) solution with a 4.95% U-235 isotopic content. This material was then transported to a TVA subcontractor who converted the solution to uranium oxide and then fabricated into commercial light water reactor (LWR) fuel. This fuel is now powering TVA reactors and supplying electricity to approximately 1 million households in the TVA region. There is still in excess of approximately 10 to 14 MT of off-spec HEU throughout the DOE complex or future foreign and domestic research reactor returns that could be recovered and down blended for use in either currently designed light water reactors, ?5% enriched LEU, or be made available for use in subsequent advanced 'fast' reactor fuel designs, ?19% LEU. (authors)

  18. Uranium Mill Tailings Remedial Action Project Environmental Protection Implementation Plan

    SciTech Connect (OSTI)

    Not Available

    1992-10-01T23:59:59.000Z

    The Uranium Mill Tallings Remedial Action (UMTRA) Project Environmental Protection Implementation Plan (EPIP) has been prepared in accordance with the requirements of the US Department of Energy (DOE) Order 5400.1 (Chapter 3, paragraph 2). The UMTRA EPIP covers the time period of November 9, 1992, through November 8, 1993. It will be updated annually. Its purpose is to provide management direction to ensure that the UMTRA Project is operated and managed in a manner that will protect, maintain, and where necessary, restore environmental quality, minimize potential threats to public health and the environment, and comply with environmental regulations and DOE policies.

  19. Uranium Mill Tailings Remedial Action Project environmental protection implementation plan

    SciTech Connect (OSTI)

    Not Available

    1994-10-01T23:59:59.000Z

    The Uranium Mill Tailings Remedial Action (UMTRA) Project Environmental Protection Implementation Plan (EPIP) has been prepared in accordance with the requirements of the U.S. Department of Energy (DOE) Order 5400.1. The UMTRA EPIP is updated annually. This version covers the time period of 9 November 1994, through 8 November 1995. Its purpose is to provide management direction to ensure that the UMTRA Project is operated and managed in a manner that will protect, maintain, and where necessary, restore environmental quality, minimize potential threats to public health and the environment, and comply with environmental regulations and DOE policies.

  20. Uranium Mill Tailings Remedial Action Project. 1995 Environmental Report

    SciTech Connect (OSTI)

    NONE

    1996-06-01T23:59:59.000Z

    In accordance with U.S. Department of Energy (DOE) Order 23 1. 1, Environment, Safety and Health Reporting, the DOE prepares an annual report to document the activities of the Uranium Mill Tailings Remedial Action (UMTRA) Project environmental monitoring program. This monitoring must comply with appropriate laws, regulations, and standards, and it must identify apparent and meaningful trends in monitoring results. The results of all monitoring activities must be communicated to the public. The UMTRA Project has prepared annual environmental reports to the public since 1989.

  1. Thermal hydraulic limits analysis for the MIT Research Reactor low enrichment uranium core conversion using statistical propagation of parametric uncertainties ; Thermal hydraulic limits analysis for the Massachusetts Institute of Technology Research Reactor low enrichment uranium core conversion using statistical propagation of parametric uncertainties .

    E-Print Network [OSTI]

    Chiang, Keng-Yen

    2012-01-01T23:59:59.000Z

    ??The MIT Research Reactor (MITR) is evaluating the conversion from highly enriched uranium (HEU) to low enrichment uranium (LEU) fuel. In addition to the fuel (more)

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

    DOE Patents [OSTI]

    Travelli, A.

    1985-10-25T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Travelli, Armando (Hinsdale, IL)

    1988-01-01T23:59:59.000Z

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

  4. Recycled Uranium Mass Balance Project Y-12 National Security Complex Site Report

    SciTech Connect (OSTI)

    NONE

    2000-12-01T23:59:59.000Z

    This report has been prepared to summarize the findings of the Y-12 National Security Complex (Y-12 Complex) Mass Balance Project and to support preparation of associated U. S. Department of Energy (DOE) site reports. The project was conducted in support of DOE efforts to assess the potential for health and environmental issues resulting from the presence of transuranic (TRU) elements and fission products in recycled uranium (RU) processed by DOE and its predecessor agencies. The United States government used uranium in fission reactors to produce plutonium and tritium for nuclear weapons production. Because uranium was considered scarce relative to demand when these operations began almost 50 years ago, the spent fuel from U.S. fission reactors was processed to recover uranium for recycling. The estimated mass balance for highly enriched RU, which is of most concern for worker exposure and is the primary focus of this project, is summarized in a table. A discrepancy in the mass balance between receipts and shipments (plus inventory and waste) reflects an inability to precisely distinguish between RU and non-RU shipments and receipts involving the Y-12 Complex and Savannah River. Shipments of fresh fuel (non-RU) and sweetener (also non-RU) were made from the Y-12 Complex to Savannah River along with RU shipments. The only way to distinguish between these RU and non-RU streams using available records is by enrichment level. Shipments of {le}90% enrichment were assumed to be RU. Shipments of >90% enrichment were assumed to be non-RU fresh fuel or sweetener. This methodology using enrichment level to distinguish between RU and non-RU results in good estimates of RU flows that are reasonably consistent with Savannah River estimates. Although this is the best available means of distinguishing RU streams, this method does leave a difference of approximately 17.3 MTU between receipts and shipments. Slightly depleted RU streams received by the Y-12 Complex from ORGDP and PGDP are believed to have been returned to the shipping site or disposed of as waste on the Oak Ridge Reservation. No evidence of Y-12 Complex processing of this material was identified in the historical records reviewed by the Project Team.

  5. Uranium Mill Tailings Remedial Action Project Environmental Protection Implementation Plan

    SciTech Connect (OSTI)

    Vollmer, A.T.

    1993-10-01T23:59:59.000Z

    The Uranium Mill Tailings Remedial Action (UMTRA) Project Environmental Protection Implementation Plan (EPIP) has been prepared in accordance with the requirements of the US Department of Energy (DOE) Order 5400.1. The UMTRA EPIP covers the time period of November 9, 1993, through November 8, 1994. It will be updated annually. Its purpose is to provide management direction to ensure that the UMTRA Project is operated and managed in a manner that will protect, maintain, and where necessary, restore environmental quality, minimize potential threats to public health and the environment, and comply with environmental regulations and DOE policies. Contents of this report are: (1) general description of the UMTRA project environmental protection program; (2) notifications; (3) planning and reporting; (4) special programs; (5) environmental monitoring programs; (6) quality assurance and data verification; and (7) references.

  6. MODULAR PEBBLE BED REACTOR PROJECT UNIVERSITY RESEARCH CONSORTIUM

    E-Print Network [OSTI]

    Annual Report Page ii MODULAR PEBBLE BED REACTOR ABSTRACT This project is developing a fundamental. Publication of an archival journal article covering this work is being prepared. · Detailed gas reactor Abstract

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

    SciTech Connect (OSTI)

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

    2010-02-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Aji, Indarta Kuncoro [Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung (Indonesia); Waris, A., E-mail: awaris@fi.itb.ac.id [Nuclear Physics and Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jl. Ganesa No. 10 Bandung 40132 (Indonesia)

    2014-09-30T23:59:59.000Z

    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.

  9. Nuclear Criticality Control and Safety of Plutonium-Uranium Fuel Mixtures Outside Reactors

    SciTech Connect (OSTI)

    Biswas, D; Mennerdahl, D

    2008-06-23T23:59:59.000Z

    The ANSI/ANS 8.12 standard was first approved in July 1978. At that time, this edition was applicable to operations with plutonium-uranium oxide (MOX) fuel mixtures outside reactors and was limited to subcritical limits for homogeneous systems. The next major revision, ANSI/ANS-8.12-1987, included the addition of subcritical limits for heterogeneous systems. The standard was subsequently reaffirmed in February 1993. During late 1990s, substantial work was done by the ANS 8.12 Standard Working Group to re-examine the technical data presented in the standard using the latest codes and cross section sets. Calculations performed showed good agreement with the values published in the standard. This effort resulted in the reaffirmation of the standard in March 2002. The standard is currently in a maintenance mode. After 2002, activities included discussions to determine the future direction of the standard and to follow the MOX standard development by the International Standard Organization (ISO). In 2007, the Working Group decided to revise the standard to extend the areas of applicability by providing a wider range of subcritical data. The intent is to cover a wider domain of MOX fuel fabrication and operations. It was also decided to follow the ISO MOX standard specifications (related to MOX density and isotopics) and develop a new set of subcritical limits for homogeneous systems. This has resulted in the submittal (and subsequent approval) of the project initiation notification system form (PINS) in 2007.

  10. MIT PEBBLE BED REACTOR PROJECT ANDREW C. KADAK

    E-Print Network [OSTI]

    MIT PEBBLE BED REACTOR PROJECT ANDREW C. KADAK Nuclear Science and Engineering Department since approximately 1998, when a student design project concluded that in order to resurrect the nuclear

  11. Global terrestrial uranium supply and its policy implications : a probabilistic projection of future uranium costs

    E-Print Network [OSTI]

    Matthews, Isaac A

    2010-01-01T23:59:59.000Z

    An accurate outlook on long-term uranium resources is critical in forecasting uranium costresource relationships, and for energy policy planning as regards the development and deployment of nuclear fuel cycle alternatives. ...

  12. Plutonium Consumption Program, CANDU Reactor Project final report

    SciTech Connect (OSTI)

    Not Available

    1994-07-31T23:59:59.000Z

    DOE is investigating methods for long term dispositioning of weapons grade plutonium. One such method would be to utilize the plutonium in Mixed OXide (MOX) fuel assemblies in existing CANDU reactors. CANDU (Canadian Deuterium Uranium) reactors are designed, licensed, built, and supported by Atomic Energy of Canada Limited (AECL), and currently use natural uranium oxide as fuel. The MOX spent fuel assemblies removed from the reactor would be similar to the spent fuel currently produced using natural uranium fuel, thus rendering the plutonium as unattractive as that in the stockpiles of commercial spent fuel. This report presents the results of a study sponsored by the DOE for dispositioning the plutonium using CANDU technology. Ontario Hydro`s Bruce A was used as reference. The fuel design study defined the optimum parameters to disposition 50 tons of Pu in 25 years (or 100 tons). Two alternate fuel designs were studied. Safeguards, security, environment, safety, health, economics, etc. were considered. Options for complete destruction of the Pu were also studied briefly; CANDU has a superior ability for this. Alternative deployment options were explored and the potential impact on Pu dispositioning in the former Soviet Union was studied. An integrated system can be ready to begin Pu consumption in 4 years, with no changes required to the reactors other than for safe, secure storage of new fuel.

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

    SciTech Connect (OSTI)

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

    2010-10-01T23:59:59.000Z

    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.

  14. Using low-enriched uranium in research reactors: The RERTR program

    SciTech Connect (OSTI)

    Travelli, A.

    1994-05-01T23:59:59.000Z

    The goal of the RERTR program is to minimize and eventually eliminate use of highway enriched uranium (HEU) in research and test reactors. The program has been very successful, and has developed low-enriched uranium (LEU) fuel materials and designs which can be used effectively in approximately 90 percent of the research and test reactors which used HEU when the program began. This progress would not have been possible without active international cooperation among fuel developers, commercial vendors, and reactor operators. The new tasks which the RERTR program is undertaking at this time include development of new and better fuels that will allow use of LEU fuels in all research and test reactors; cooperation with Russian laboratories, which will make it possible to minimize and eventually eliminate use of HEU in research reactors throughout the world, irrespective of its origin; and development of an LEU-based process for the production of {sup 99}Mo. Continuation and intensification of international cooperation are essential to the achievement of the ultimate goals of the RERTR program.

  15. Program management plan for the Molten Salt Reactor Experiment Remediation Project at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    SciTech Connect (OSTI)

    NONE

    1996-09-01T23:59:59.000Z

    The primary mission of the Molten Salt Reactor Experiment (MSRE) Remediation Project is to effectively implement the risk-reduction strategies and technical plans to stabilize and prevent further migration of uranium within the MSRE facility, remove the uranium and fuel salts from the system, and dispose of the fuel and flush salts by storage in appropriate depositories to bring the facility to a surveillance and maintenance condition before decontamination and decommissioning. This Project Management Plan (PMP) for the MSRE Remediation Project details project purpose; technical objectives, milestones, and cost objectives; work plan; work breakdown structure (WBS); schedule; management organization and responsibilities; project management performance measurement planning, and control; conduct of operations; configuration management; environmental, safety, and health compliance; quality assurance; operational readiness reviews; and training.

  16. 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 [ORNL; Freels, James D [ORNL; Ilas, Germina [ORNL; Jolly, Brian C [ORNL; Miller, James Henry [ORNL; Primm, Trent [ORNL; Renfro, David G [ORNL; Sease, John D [ORNL; Pinkston, Daniel [ORNL

    2011-02-01T23:59:59.000Z

    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.

  17. TA-2 Water Boiler Reactor Decommissioning Project

    SciTech Connect (OSTI)

    Durbin, M.E. (ed.); Montoya, G.M.

    1991-06-01T23:59:59.000Z

    This final report addresses the Phase 2 decommissioning of the Water Boiler Reactor, biological shield, other components within the biological shield, and piping pits in the floor of the reactor building. External structures and underground piping associated with the gaseous effluent (stack) line from Technical Area 2 (TA-2) Water Boiler Reactor were removed in 1985--1986 as Phase 1 of reactor decommissioning. The cost of Phase 2 was approximately $623K. The decommissioning operation produced 173 m{sup 3} of low-level solid radioactive waste and 35 m{sup 3} of mixed waste. 15 refs., 25 figs., 3 tabs.

  18. Moab Project Disposes 2 Million Tons of Uranium Mill Tailings with Recovery Act Funds

    Broader source: Energy.gov [DOE]

    The Moab Uranium Mill Tailings Remedial Action Project reached its primary American Recovery and Reinvestment Act milestone ahead of schedule on Wednesday with the disposal of 2 million tons of...

  19. Reactor-specific spent fuel discharge projections, 1987-2020

    SciTech Connect (OSTI)

    Walling, R.C.; Heeb, C.M.; Purcell, W.L.

    1988-03-01T23:59:59.000Z

    The creation of five reactor-specific spent fuel data bases that contain information on the projected amounts of spent fuel to be discharged from U.S. commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water eactors (BWR), and one existing high temperature gas reactor (HTGR). The projections are based on individual reactor information supplied by the U.S. reactor owners. The basic information is adjusted to conform to Energy Information Administration (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: No New Orders (assumes increasing burnup), No New Orders with No Increased Burnup, Upper Reference (assumes increasing burnup), Upper Reference with No Increased Burnup, and Lower Reference (assumes increasing burnup). Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum at-reactor storage, and for storage requirements assuming maximum at-reactor storage plus intra-utility transshipment of spent fuel. 8 refs., 8 figs., 10 tabs.

  20. Destruction of plutonium using non-uranium fuels in pressurized water reactor peripheral assemblies

    SciTech Connect (OSTI)

    Chodak, P. III

    1996-05-01T23:59:59.000Z

    This thesis examines and confirms the feasibility of using non-uranium fuel in a pressurized water reactor (PWR) radial blanket to eliminate plutonium of both weapons and civilian origin. In the equilibrium cycle, the periphery of the PWR is loaded with alternating fresh and once burned non-uranium fuel assemblies, with the interior of the core comprised of conventional three batch UO{sub 2} assemblies. Plutonium throughput is such that there is no net plutonium production: production in the interior is offset by destruction in the periphery. Using this approach a 50 MT WGPu inventory could be eliminated in approximately 400 reactor years of operation. Assuming all other existing constraints were removed, the 72 operating US PWRs could disposition 50 MT of WGPu in 5.6 years. Use of a low fissile loading plutonium-erbium inert-oxide-matrix composition in the peripheral assemblies essentially destroys 100% of the {sup 239}Pu and {ge}90% {sub total}Pu over two 18 month fuel cycles. Core radial power peaking, reactivity vs EFPD profiles and core average reactivity coefficients were found to be comparable to standard PWR values. Hence, minimal impact on reload licensing is anticipated. Examination of potential candidate fuel matrices based on the existing experience base and thermo-physical properties resulted in the recommendation of three inert fuel matrix compositions for further study: zirconia, alumina and TRISO particle fuels. Objective metrics for quantifying the inherent proliferation resistance of plutonium host waste and fuel forms are proposed and were applied to compare the proposed spent WGPu non-uranium fuel to spent WGPu MOX fuels and WGPu borosilicate glass logs. The elimination disposition option spent non-uranium fuel product was found to present significantly greater barriers to proliferation than other plutonium disposal products.

  1. Reactor-specific spent fuel discharge projections: 1985 to 2020

    SciTech Connect (OSTI)

    Heeb, C.M.; Libby, R.A.; Walling, R.C.; Purcell, W.L.

    1986-09-01T23:59:59.000Z

    The creation of four spent-fuel data bases that contain information on the projected amounts of spent fuel to be discharged from US commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent-fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water reactors (BWR). The projections are based on individual reactor information supplied by the US reactor owners. The basic information is adjusted to conform to Energy Information Agency (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: (1) No New Orders with Extended Burnup, (2) No New Orders with Constant Burnup, (3) Middle Case with Extended Burnup, and (4) Middle Case with Constant Burnup. Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum-at-reactor storage, and for storage requirements assuming maximum-at-reactor plus intra-utility transshipment of spent fuel.

  2. Reactor physics project progress report no. 2

    E-Print Network [OSTI]

    Driscoll, Michael J.

    1969-01-01T23:59:59.000Z

    This is the second annual report in an experimental and theoretical program to develop and apply single and few element heterogeneous methods for the determination of reactor lattice parameters. During the period covered ...

  3. Minor actinide transmutation in thorium and uranium matrices in heavy water moderated reactors

    SciTech Connect (OSTI)

    Bhatti, Zaki; Hyland, B.; Edwards, G.W.R. [Atomic Energy of Canada Limited, Chalk River Laboratories, 1 Plant Road, Chalk River, Ontario, K0J 1J0 (Canada)

    2013-07-01T23:59:59.000Z

    The irradiation of Th{sup 232} breeds fewer of the problematic minor actinides (Np, Am, Cm) than the irradiation of U{sup 238}. This characteristic makes thorium an attractive potential matrix for the transmutation of these minor actinides, as these species can be transmuted without the creation of new actinides as is the case with a uranium fuel matrix. Minor actinides are the main contributors to long term decay heat and radiotoxicity of spent fuel, so reducing their concentration can greatly increase the capacity of a long term deep geological repository. Mixing minor actinides with thorium, three times more common in the Earth's crust than natural uranium, has the additional advantage of improving the sustainability of the fuel cycle. In this work, lattice cell calculations have been performed to determine the results of transmuting minor actinides from light water reactor spent fuel in a thorium matrix. 15-year-cooled group-extracted transuranic elements (Np, Pu, Am, Cm) from light water reactor (LWR) spent fuel were used as the fissile component in a thorium-based fuel in a heavy water moderated reactor (HWR). The minor actinide (MA) transmutation rates, spent fuel activity, decay heat and radiotoxicity, are compared with those obtained when the MA were mixed instead with natural uranium and taken to the same burnup. Each bundle contained a central pin containing a burnable neutron absorber whose initial concentration was adjusted to have the same reactivity response (in units of the delayed neutron fraction ?) for coolant voiding as standard NU fuel. (authors)

  4. Development of a core design optimization tool and analysis in support of the planned LEU conversion of the MIT Research Reactor (MITR-II) ; Development of a core design optimization tool and analysis in support of the planned low enriched uranium conversion of the MIT Research Reactor (MITR-II) .

    E-Print Network [OSTI]

    Connaway, Heather M. (Heather Moira)

    2012-01-01T23:59:59.000Z

    ??The MIT Research Reactor (MITR-II) is currently undergoing analysis for the planned conversion from high enriched uranium (HEU) to low enriched uranium (LEU), as part (more)

  5. Hydrothermal Testing of K Basin Sludge and N Reactor Fuel at Sludge Treatment Project Operating Conditions

    SciTech Connect (OSTI)

    Delegard, Calvin H.; Schmidt, Andrew J.; Thornton, Brenda M.

    2007-03-30T23:59:59.000Z

    The Sludge Treatment Project (STP), managed for the U. S. DOE by Fluor Hanford (FH), was created to design and operate a process to eliminate uranium metal from K Basin sludge prior to packaging for Waste Isolation Pilot Plant (WIPP). The STP process uses high temperature liquid water to accelerate the reaction, produce uranium dioxide from the uranium metal, and safely discharge the hydrogen. Under nominal process conditions, the sludge will be heated in pressurized water at 185C for as long as 72 hours to assure the complete reaction (corrosion) of up to 0.25-inch diameter uranium metal pieces. Under contract to FH, the Pacific Northwest National Laboratory (PNNL) conducted bench-scale testing of the STP hydrothermal process in November and December 2006. Five tests (~50 ml each) were conducted in sealed, un-agitated reaction vessels under the hydrothermal conditions (e.g., 7 to 72 h at 185C) of the STP corrosion process using radioactive sludge samples collected from the K East Basin and particles/coupons of N Reactor fuel also taken from the K Basins. The tests were designed to evaluate and understand the chemical changes that may be occurring and the effects that any changes would have on sludge rheological properties. The tests were not designed to evaluate engineering aspects of the process. The hydrothermal treatment affected the chemical and physical properties of the sludge. In each test, significant uranium compound phase changes were identified, resulting from dehydration and chemical reduction reactions. Physical properties of the sludge were significantly altered from their initial, as-settled sludge values, including, shear strength, settled density, weight percent water, and gas retention.

  6. Enhanced CANDU6: Reactor and fuel cycle options - Natural uranium and beyond

    SciTech Connect (OSTI)

    Ovanes, M.; Chan, P. S. W.; Mao, J.; Alderson, N.; Hopwood, J. M. [Candu Energy Inc., 2285 Speakman Drive, Mississauga, ON L5K 1B1 (Canada)

    2012-07-01T23:59:59.000Z

    The Enhanced CANDU 6{sup R} (ECo{sup R}) is the updated version of the well established CANDU 6 family of units incorporating improved safety characteristics designed to meet or exceed Generation III nuclear power plant expectations. The EC6 retains the excellent neutron economy and fuel cycle flexibility that are inherent in the CANDU reactor design. The reference design is based on natural uranium fuel, but the EC6 is also able to utilize additional fuel options, including the use of Recovered Uranium (RU) and Thorium based fuels, without requiring major hardware upgrades to the existing control and safety systems. This paper outlines the major changes in the EC6 core design from the existing C6 design that significantly enhance the safety characteristics and operating efficiency of the reactor. The use of RU fuel as a transparent replacement fuel for the standard 37-el NU fuel, and several RU based advanced fuel designs that give significant improvements in fuel burnup and inherent safety characteristics are also discussed in the paper. In addition, the suitability of the EC6 to use MOX and related Pu-based fuels will also be discussed. (authors)

  7. Comparative analysis of thorium and uranium fuel for transuranic recycle in a sodium cooled Fast Reactor

    SciTech Connect (OSTI)

    C. Fiorina; N. E. Stauff; F. Franceschini; M. T. Wenner; A. Stanculescu; T. K. Kim; A. Cammi; M. E. Ricotti; R. N. Hill; T. A. Taiwo; M. Salvatores

    2013-12-01T23:59:59.000Z

    The present paper compares the reactor physics and transmutation performance of sodium-cooled Fast Reactors (FRs) for TRansUranic (TRU) burning with thorium (Th) or uranium (U) as fertile materials. The 1000 MWt Toshiba-Westinghouse Advanced Recycling Reactor (ARR) conceptual core has been used as benchmark for the comparison. Both burner and breakeven configurations sustained or started with a TRU supply, and assuming full actinide homogeneous recycle strategy, have been developed. State-of-the-art core physics tools have been employed to establish fuel inventory and reactor physics performances for equilibrium and transition cycles. Results show that Th fosters large improvements in the reactivity coefficients associated with coolant expansion and voiding, which enhances safety margins and, for a burner design, can be traded for maximizing the TRU burning rate. A trade-off of Th compared to U is the significantly larger fuel inventory required to achieve a breakeven design, which entails additional blankets at the detriment of core compactness as well as fuel manufacturing and separation requirements. The gamma field generated by the progeny of U-232 in the U bred from Th challenges fuel handling and manufacturing, but in case of full recycle, the high contents of Am and Cm in the transmutation fuel impose remote fuel operations regardless of the presence of U-232.

  8. Calculation of the number of cancer deaths prevented by the Uranium Mill Tailings Remedial Action Project

    SciTech Connect (OSTI)

    Miller, M.L.; Pomatto, C.B. (Roy F. Weston, Inc., Albuquerque, NM (United States)); Cornish, R.E. (Dept. of Energy, Albuquerque, NM (United States). Albuquerque Operations Office)

    1999-05-01T23:59:59.000Z

    The Uranium Mill Tailings Remedial Action Project has completed remedial action at 22 uranium mill tailings sites and about 5,000 properties (vicinity properties) where tailings were used in construction, at a total cost of $1.45 billion. This paper uses existing data from Environmental Impact Statements and Environmental Assessments, and vicinity property calculations, to determine the total number of cancer deaths averted by the Uranium Mill Tailings Remedial Action Project. The cost-effectiveness of remediating each site, the vicinity properties, and the entire project is calculated. The cost per cancer death averted was four orders of magnitude higher at the least cost-effective site than at the most cost-effective site.

  9. Uranium 2014 resources, production and demand

    E-Print Network [OSTI]

    Organisation for Economic Cooperation and Development. Paris

    2014-01-01T23:59:59.000Z

    Published every other year, Uranium Resources, Production, and Demand, or the "Red Book" as it is commonly known, is jointly prepared by the OECD Nuclear Energy Agency and the International Atomic Energy Agency. It is the recognised world reference on uranium and is based on official information received from 43 countries. It presents the results of a thorough review of world uranium supplies and demand and provides a statistical profile of the world uranium industry in the areas of exploration, resource estimates, production and reactor-related requirements. It provides substantial new information from all major uranium production centres in Africa, Australia, Central Asia, Eastern Europe and North America. Long-term projections of nuclear generating capacity and reactor-related uranium requirements are provided as well as a discussion of long-term uranium supply and demand issues. This edition focuses on recent price and production increases that could signal major changes in the industry.

  10. Uranium 2005 resources, production and demand

    E-Print Network [OSTI]

    Organisation for Economic Cooperation and Development. Paris

    2006-01-01T23:59:59.000Z

    Published every other year, Uranium Resources, Production, and Demand, or the "Red Book" as it is commonly known, is jointly prepared by the OECD Nuclear Energy Agency and the International Atomic Energy Agency. It is the recognised world reference on uranium and is based on official information received from 43 countries. This 21st edition presents the results of a thorough review of world uranium supplies and demand as of 1st January 2005 and provides a statistical profile of the world uranium industry in the areas of exploration, resource estimates, production and reactor-related requirements. It provides substantial new information from all major uranium production centres in Africa, Australia, Central Asia, Eastern Europe and North America. Projections of nuclear generating capacity and reactor-related uranium requirements through 2025 are provided as well as a discussion of long-term uranium supply and demand issues. This edition focuses on recent price and production increases that could signal major c...

  11. Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel

    SciTech Connect (OSTI)

    Cowell, B.S.; Fisher, S.E.

    1999-02-01T23:59:59.000Z

    The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option.

  12. ADVANTAGES, DISADVANTAGES, AND LESSONS LEARNED FROM MULTI-REACTOR DECOMMISSIONING PROJECTS

    SciTech Connect (OSTI)

    Morton, M.R.; Nielson, R.R.; Trevino, R.A.

    2003-02-27T23:59:59.000Z

    This paper discusses the Reactor Interim Safe Storage (ISS) Project within the decommissioning projects at the Hanford Site and reviews the lessons learned from performing four large reactor decommissioning projects sequentially. The advantages and disadvantages of this multi-reactor decommissioning project are highlighted.

  13. [Uranium Mill Tailings Remedial Action Project Office Quality Assurance Program Plan]. Revision 4

    SciTech Connect (OSTI)

    Not Available

    1992-06-01T23:59:59.000Z

    The Uranium Mill Tailings Remedial Action (UMTRA) Project was established to accomplish remedial actions at inactive uranium mill tailings sites in accordance with Public Law 95-604, the Uranium Mill Tailings Radiation Control Act of 1978 (UMTRCA). The UMTRA Project`s mission is to stabilize and control the residual radioactive materials at designated sites in a safe and environmentally sound manner so as to minimize or eliminate radiation health hazards to the public. The US Department of Energy (DOE) UMTRA Project Office (UMTRA PO) directs the overall project. Since these efforts may involve possible risks to public health and safety, a quality assurance (QA) program that conforms to the applicable criteria (set forth in the reference documents) has been established to control the quality of the work. This document, the Quality Assurance Program Plan (QAPP), brings into one document the essential criteria to be applied on a selective basis, depending upon the nature of the activity being conducted, and describes how those criteria shall be applied to the UMTRA Project. The UMTRA PO shall require each Project contractor to prepare and submit for approval a more detailed QAPP that is based on the applicable criteria of this QAPP and the referenced documents. All QAPPs on the UMTRA Project shall fit within the framework of this plan.

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

    SciTech Connect (OSTI)

    Travelli, A.

    1988-01-19T23:59:59.000Z

    A nuclear fuel-containing plate structure for a nuclear reactor is described; such structure comprising a pair of malleable metallic non-fissionable matrix plates having confronting surfaces which are pressure bonded together and fully united to form a bonded surface, and elongated malleable wire-like fissionable fuel members separately confined and fully enclosed between the matrix plates along the interface to afford a high fuel density as well as structural integrity and effective retention of fission products. The plates have separate recesses formed in the confronting surfaces for closely receiving the wire-like fissionable fuel members. The wire-like fissionable fuel members are made of a maleable uranium alloy capable of being formed into elongated wire-like members and capable of withstanding pressure bonding. The wire-like fissionable fuel members are completely separated and isolated by fully united portions of the interface.

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

    SciTech Connect (OSTI)

    Ilas, Germina [ORNL; Primm, Trent [ORNL

    2011-05-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    2012-03-01T23:59:59.000Z

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

  17. [Uranium Mill Tailings Remedial Action Project Office Quality Assurance Program Plan

    SciTech Connect (OSTI)

    Not Available

    1992-06-01T23:59:59.000Z

    The Uranium Mill Tailings Remedial Action (UMTRA) Project was established to accomplish remedial actions at inactive uranium mill tailings sites in accordance with Public Law 95-604, the Uranium Mill Tailings Radiation Control Act of 1978 (UMTRCA). The UMTRA Project's mission is to stabilize and control the residual radioactive materials at designated sites in a safe and environmentally sound manner so as to minimize or eliminate radiation health hazards to the public. The US Department of Energy (DOE) UMTRA Project Office (UMTRA PO) directs the overall project. Since these efforts may involve possible risks to public health and safety, a quality assurance (QA) program that conforms to the applicable criteria (set forth in the reference documents) has been established to control the quality of the work. This document, the Quality Assurance Program Plan (QAPP), brings into one document the essential criteria to be applied on a selective basis, depending upon the nature of the activity being conducted, and describes how those criteria shall be applied to the UMTRA Project. The UMTRA PO shall require each Project contractor to prepare and submit for approval a more detailed QAPP that is based on the applicable criteria of this QAPP and the referenced documents. All QAPPs on the UMTRA Project shall fit within the framework of this plan.

  18. The International Reactor Physics Experiment Evaluation Project (IRPHEP)

    SciTech Connect (OSTI)

    J. Blair Briggs; Enrico Sartori; Lori Scott

    2006-09-01T23:59:59.000Z

    Since the beginning of the Nuclear Power industry, numerous experiments concerned with nuclear energy and technology have been performed at different research laboratories, worldwide. These experiments required a large investment in terms of infrastructure, expertise, and cost; however, many were performed without a high degree of attention to archival of results for future use. The degree and quality of documentation varies greatly. There is an urgent need to preserve integral reactor physics experimental data, including measurement methods, techniques, and separate or special effects data for nuclear energy and technology applications and the knowledge and competence contained therein. If the data are compromised, it is unlikely that any of these experiments will be repeated again in the future. The International Reactor Physics Evaluation Project (IRPhEP) was initiated, as a pilot activity in 1999 by the by the Organization of Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) Nuclear Science Committee (NSC). The project was endorsed as an official activity of the NSC in June of 2003. The purpose of the IRPhEP is to provide an extensively peer reviewed set of reactor physics related integral benchmark data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next generation reactors and establish the safety basis for operation of these reactors. A short history of the IRPhEP is presented and its purposes are discussed in this paper. Accomplishments of the IRPhEP, including the first publication of the IRPhEP Handbook, are highlighted and the future of the project outlined.

  19. The International Reactor Physics Experiment Evaluation Project (IRPhEP)

    SciTech Connect (OSTI)

    Blair Briggs, J. [Idaho National Laboratory, 2525 North Fremont, Idaho Falls, ID 83415-3860 (United States); Sartori, E. [OECD, Nuclear Energy Agency, Le Seine Saint-Germain, 12 Boulevard des Iles, F-92310 Issy-les-Moulineaux (France); Scott, L. [Cover to Cover, 1015 Cedar Hills Blvd., Belle Vernon, PA 15012 (United States)

    2006-07-01T23:59:59.000Z

    Since the beginning of the Nuclear Power industry, numerous experiments concerned with nuclear energy and technology have been performed at different research laboratories, worldwide. These experiments required a large investment in terms of infrastructure, expertise, and cost; however, many were performed without a high degree of attention to archival of results for future use. The degree and quality of documentation varies greatly. There is an urgent need to preserve integral reactor physics experimental data, including measurement methods, techniques, and separate or special effects data for nuclear energy and technology applications and the knowledge and competence contained therein. If the data are compromised, it is unlikely that any of these experiments will be repeated again in the future. The International Reactor Physics Evaluation Project (IRPhEP) was initiated, as a pilot activity in 1999 by the by the Organization of Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) Nuclear Science Committee (NSC). The project was endorsed as an official activity of the NSC in June of 2003. The purpose of the IRPhEP is to provide an extensively peer reviewed set of reactor physics related integral benchmark data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next generation reactors and establish the safety basis for operation of these reactors. A short history of the IRPhEP is presented and its purposes are discussed in this paper. Accomplishments of the IRPhEP, including the first publication of the IRPhEP Handbook, are highlighted and the future of the project outlined. (authors)

  20. Uranium Mill Tailings Remedial Action Project fiscal year 1997 annual report to stakeholders

    SciTech Connect (OSTI)

    NONE

    1997-12-31T23:59:59.000Z

    The fiscal year (FY) 1997 annual report is the 19th report on the status of the US Department of Energy`s (DOE) Uranium Mill Tailings Remedial Action (UMTRA) Project. In 1978, Congress directed the DOE to assess and clean up contamination at 24 designated former uranium processing sites. The DOE is also responsible for cleaning up properties in the vicinity of the sites where wind and water erosion deposited tailings or people removed them from the site for use in construction or landscaping. Cleanup has been undertaken in cooperation with state governments and Indian tribes within whose boundaries the sites are located. It is being conducted in two phases: the surface project and the groundwater project. This report addresses specifics about the UMTRA surface project.

  1. Fiscal year 1996 annual report to stakeholders, Uranium Mill Tailings Remedial Action Project

    SciTech Connect (OSTI)

    NONE

    1996-10-01T23:59:59.000Z

    This is the Fiscal Year (FY) 1996 annual report on the status of the US Department of Energy`s (DOE) Uranium Mill Tailings Remedial Action (UMTRA) Project. In 1978, Congress directed the DOE to assess and clean up contamination at 24 designated former uranium processing sites. The DOE is also responsible for cleaning up properties in the vicinity of the sites where wind and water erosion deposited tailings or people removed them from the site for use in construction of landscaping. Cleanup is being undertaken in cooperation with state governments and Indian tribes within whose boundaries the sites are located. It is being conducted in two phases: the surface project and the ground water project. This report addresses specifics about the surface phase of the UMTRA Project.

  2. Light Water Reactor Sustainability Constellation Pilot Project FY11 Summary Report

    SciTech Connect (OSTI)

    R. Johansen

    2011-09-01T23:59:59.000Z

    Summary report for Fiscal Year 2011 activities associated with the Constellation Pilot Project. The project is a joint effor between Constellation Nuclear Energy Group (CENG), EPRI, and the DOE Light Water Reactor Sustainability Program. The project utilizes two CENG reactor stations: R.E. Ginna and Nine Point Unit 1. Included in the report are activities associate with reactor internals and concrete containments.

  3. The feasibility study of small long-life gas cooled fast reactor with mixed natural Uranium/Thorium as fuel cycle input

    SciTech Connect (OSTI)

    Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Khairurrijal,; Monado, Fiber; Sekimoto, Hiroshi [Department of Physics Bandung Institute of Technology Jl. Ganesha 10, Bandung 40134, Physics Department, Sriwijaya University, Kampus Indralaya, Ogan Ilir, Sumatera Selatan (Indonesia); Department of Physics Bandung Institute of Technology Jl. Ganesha 10, Bandung 40134 (Indonesia); Department of Physics Bandung Institute of Technology Jl. Ganesha 10, Bandung 40134, Physics Department, Sriwijaya University, Kampus Indralaya, Ogan Ilir, Sumatera Selatan (Indonesia); Reserach of Laboratory for Nuclear Reactors, Tokyo Institute of Technology O-okayama, Meguro-ku, Tokyo 152 (Japan)

    2012-06-06T23:59:59.000Z

    A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE burn-up scheme has been performed. In this study, design GCFR with Helium coolant which can be continuously operated by supplying mixed Natural Uranium/Thorium without fuel enrichment plant or fuel reprocessing plant. The active reactor cores are divided into two region, Thorium fuel region and Uranium fuel region. Each fuel core regions are subdivided into ten parts (region-1 until region-10) with the same volume in the axial direction. The fresh Natural Uranium and Thorium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh natural Uranium/Thorium fuel. This concept is basically applied to all regions in both cores area, i.e. shifted the core of i{sup th} region into i+1 region after the end of 10 years burn-up cycle. For the next cycles, we will add only Natural Uranium and Thorium on each region-1. The calculation results show the reactivity reached by mixed Natural Uranium/Thorium with volume ratio is 4.7:1. This reactor can results power thermal 550 MWth. After reactor start-up the operation, furthermore reactor only needs Natural Uranium/Thorium supply for continue operation along 100 years.

  4. EIS-0198: Uranium Mill Tailings Remedial Action Groundwater Project

    Broader source: Energy.gov [DOE]

    This EISassesses the potential programmatic impacts of conducting the Ground Water Project, provides a method for determining the site-specific ground water compliance strategies, and provides...

  5. Uranium Mill Tailings Remedial Action Project, fiscal year 1995 annual report to stakeholders

    SciTech Connect (OSTI)

    NONE

    1995-09-30T23:59:59.000Z

    In 1978, Congress authorized the DOE to assess and clean up contamination at 24 designated former uranium processing sites. The DOE is also responsible for cleaning up properties in the vicinity of the sites where wind and water erosion deposited tailings or people removed them from the site for use in construction or landscaping projects. Cleanup is being undertaken in cooperation with state governments and Indian tribes within whose boundaries the sites are located. It is being conducted in two phases: the surface project and the ground water project. This report addresses specifics about both phases of the UMTRA Project. DOE`s UMTRA Project is the world`s largest materials management project ever undertaken to reduce or eliminate risk to the general public from exposure to potentially hazardous and radioactive materials. With an estimated cost at completion of nearly $2 billion for both phases of the UMTRA Project, and with the responsibility for encapsulating and isolating almost one-fourth of all the uranium mill tailings generated across the entire US (more than 44 million cubic yards), the UMTRA Project and its people have achieved a long record of safely and effectively completing its mission. It continually enhances its national reputation through its diligent process and cost efficiency as well as its international recognition for its technological innovation.

  6. Advanced Proliferation Resistant, Lower Cost, Uranium-Thorium Dioxide Fuels for Light Water Reactors (Progress report for work through June 2002, 12th quarterly report)

    SciTech Connect (OSTI)

    Mac Donald, Philip Elsworth

    2002-09-01T23:59:59.000Z

    The overall objective of this NERI project is to evaluate the potential advantages and disadvantages of an optimized thorium-uranium dioxide (ThO2/UO2) fuel design for light water reactors (LWRs). The project is led by the Idaho National Engineering and Environmental Laboratory (INEEL), with the collaboration of three universities, the University of Florida, Massachusetts Institute of Technology (MIT), and Purdue University; Argonne National Laboratory; and all of the Pressurized Water Reactor (PWR) fuel vendors in the United States (Framatome, Siemens, and Westinghouse). In addition, a number of researchers at the Korean Atomic Energy Research Institute and Professor Kwangheon Park at Kyunghee University are active collaborators with Korean Ministry of Science and Technology funding. The project has been organized into five tasks: Task 1 consists of fuel cycle neutronics and economics analysis to determine the economic viability of various ThO2/UO2 fuel designs in PWRs, Task 2 will determine whether or not ThO2/UO2 fuel can be manufactured economically, Task 3 will evaluate the behavior of ThO2/UO2 fuel during normal, off-normal, and accident conditions and compare the results with the results of previous UO2 fuel evaluations and U.S. Nuclear Regulatory Commission (NRC) licensing standards, Task 4 will determine the long-term stability of ThO2/UO2 high-level waste, and Task 5 consists of the Korean work on core design, fuel performance analysis, and xenon diffusivity measurements.

  7. Contents of environmental impact statements prepared for the Uranium Mill Tailings Remedial Action Project. [Uranium Mill Tailings Remedial Action (UMTRA) Project

    SciTech Connect (OSTI)

    Not Available

    1986-01-01T23:59:59.000Z

    This document presents two versions of the outline for the environmental impact statements (EISS) to be prepared for the Uranium Mill Tailings Remedial Action (UMTRA) Project. The first displays the basic structure of the statements; it lists only the titles of sections. The second is a guide to the contents of the statements which provides, under each title, a brief summary of contents. The outline is intended to comply with the planning requirements and the definitions of terms established by the Council on Environmental Quality as well as DOE Order 5440.lB (Implementation of the National Environmental Policy Act), and compliance with Floodplain/Wetlands Environmental Review Requirements. These requirements and definitions are implicity part of the outline. The outline presented in this document will guide the preparation of EISs Guidelines for preparation of environmental assessments for the UMTRA Project are available.

  8. Conceptual design study on very small long-life gas cooled fast reactor using metallic natural Uranium-Zr as fuel cycle input

    SciTech Connect (OSTI)

    Monado, Fiber, E-mail: fiber.monado@gmail.com [Nuclear Physics and Biophysics Research Group, Dept. of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung, Indonesia and Dept. of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University (Indonesia); Ariani, Menik [Dept. of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University (Indonesia); Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Permana, Sidik [Nuclear Physics and Biophysics Research Group, Dept. of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung (Indonesia); Aziz, Ferhat [National Nuclear Energy Agency of Indonesia (BATAN) (Indonesia); Sekimoto, Hiroshi [CRINES, Tokyo Institute of Technology, O-okoyama, Meguro-ku, Tokyo 152-8550 (Japan)

    2014-02-12T23:59:59.000Z

    A conceptual design study of very small 350 MWth Gas-cooled Fast Reactors with Helium coolant has been performed. In this study Modified CANDLE burn-up scheme was implemented to create small and long life fast reactors with natural Uranium as fuel cycle input. Such system can utilize natural Uranium resources efficiently without the necessity of enrichment plant or reprocessing plant. The core with metallic fuel based was subdivided into 10 regions with the same volume. The fresh Natural Uranium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh Natural Uranium fuel. This concept is basically applied to all axial regions. The reactor discharge burn-up is 31.8% HM. From the neutronic point of view, this design is in compliance with good performance.

  9. Uranium Mill Tailings remedial action project waste minimization and pollution prevention awareness program plan

    SciTech Connect (OSTI)

    Not Available

    1994-07-01T23:59:59.000Z

    The purpose of this plan is to establish a waste minimization and pollution prevention awareness (WM/PPA) program for the U.S. Department of Energy`s (DOE) Uranium Mill Tailings Remedial Action (UMTRA) Project. The program satisfies DOE requirements mandated by DOE Order 5400.1. This plan establishes planning objectives and strategies for conserving resources and reducing the quantity and toxicity of wastes and other environmental releases.

  10. Uranium 2009 resources, production and demand

    E-Print Network [OSTI]

    Organisation for Economic Cooperation and Development. Paris

    2010-01-01T23:59:59.000Z

    With several countries currently building nuclear power plants and planning the construction of more to meet long-term increases in electricity demand, uranium resources, production and demand remain topics of notable interest. In response to the projected growth in demand for uranium and declining inventories, the uranium industry the first critical link in the fuel supply chain for nuclear reactors is boosting production and developing plans for further increases in the near future. Strong market conditions will, however, be necessary to trigger the investments required to meet projected demand. The "Red Book", jointly prepared by the OECD Nuclear Energy Agency and the International Atomic Energy Agency, is a recognised world reference on uranium. It is based on information compiled in 40 countries, including those that are major producers and consumers of uranium. This 23rd edition provides a comprehensive review of world uranium supply and demand as of 1 January 2009, as well as data on global ur...

  11. RERTR program activities related to the development and application of new LEU fuels. [Reduced Enrichment Research and Test Reactor; low-enriched uranium

    SciTech Connect (OSTI)

    Travelli, A.

    1983-01-01T23:59:59.000Z

    The statue of the U.S. Reduced Enrichment Research and Test Reactor (RERTR) Program is reviewed. After a brief outline of RERTR Program objectives and goals, program accomplishments are discussed with emphasis on the development, demonstration and application of new LEU fuels. Most program activities have proceeded as planned, and a combination of two silicide fuels (U/sub 3/Si/sub 2/-Al and U/sub 3/Si-Al) holds excellent promise for achieving the long-term program goals. Current plans and schedules project the uranium density of qualified RERTR fuels for plate-type reactors to grow by approximately 1 g U/cm/sup 3/ each year, from the current 1.7 g U/cm/sup 3/ to the 7.0 g U/cm/sup 3/ which will be reached in late 1988. The technical needs of research and test reactors for HEU exports are also forecasted to undergo a gradual but dramatic decline in the coming years.

  12. Policy and procedures for classification of Class III groundwater at UMTRA Project sites. [Uranium Mill Tailings Remedial Action (UMTRA) Project

    SciTech Connect (OSTI)

    Not Available

    1989-03-01T23:59:59.000Z

    The US Environmental Protection Agency (EPA) has recently proposed groundwater regulations for the US Department of Energy's )DOE) Uranium Mill Tailings Remedial Action (UMTRA) Project. These regulations allow the application of supplemental standards at UMTRA Project sites in specific situations. The designation of groundwater as Class III permits the application of supplemental standards. This document discusses a final UMTRA Project policy and procedures for identifying Class III groundwater, including identification of a review area, definition of water quality, quantification of aquifer yield, and identification of methods reasonably employed for public water supply systems. These items, either individually or collectively, need to be investigated in order to determine if groundwaters at UMTRA Project sites are Class III. This document provides a framework for the DOE to determine Class III groundwaters.

  13. 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. [ORNL] [ORNL; Ellis, R. J. [ORNL] [ORNL; Gehin, J. C. [ORNL] [ORNL; Clarno, K. T. [ORNL] [ORNL; Williams, K. A. [ORNL] [ORNL; Moses, D. L. [ORNL] [ORNL

    2006-11-01T23:59:59.000Z

    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.

  14. Review and analysis of proposed EPA groundwater standards for the UMTRA Project. [Uranium Mill Tailings Remedial Action (UMTRA) Project

    SciTech Connect (OSTI)

    Not Available

    1987-10-01T23:59:59.000Z

    The Title I groundwater standards for inactive uranium mill tailings sites, which were promulgated on January 5, 1983, by the US Environmental Protection Agency (EPA) for the Uranium Mill Tailings Remedial Action (UMTRA) Project, were remanded to the EPA on September 3, 1985, by the US Tenth Circuit Court of Appeals. The Court instructed the EPA to compile general groundwater standards for all sites. On September 24, 1987, the EPA published proposed standards in response to the remand. This Summary Report includes an evaluation of the potential effects of the proposed EPA groundwater standards on the UMTRA Project as well as a discussion of the DOE's position on the proposed standards. This report is accompanied by a detailed Technical Report and Appendices which provide supporting information and analyses. This Summary Report results from a study undertaken to: determine the impact of the proposed standards on the UMTRA Project; and recommend provisions for the implementation of the final standards that will minimize adverse impact to the conduct of the UMTRA Project while ensuring protection of human health and the environment. Specifically, the following were considered: the flexibility of the proposed standards; interpretations of the proposed standards; the extent of aquifer restoration that may be required to implement the proposed standards at each site; the costs of aquifer restoration; and design changes necessary to meet the standards.

  15. Separation and Recovery of Uranium Metal from Spent Light Water Reactor Fuel via Electrolytic Reduction and Electrorefining

    SciTech Connect (OSTI)

    S. D. Herrmann; S. X. Li

    2010-09-01T23:59:59.000Z

    A series of bench-scale experiments was performed in a hot cell at Idaho National Laboratory to demonstrate the separation and recovery of uranium metal from spent light water reactor (LWR) oxide fuel. The experiments involved crushing spent LWR fuel to particulate and separating it from its cladding. Oxide fuel particulate was then converted to metal in a series of six electrolytic reduction runs that were performed in succession with a single salt loading of molten LiCl 1 wt% Li2O at 650 C. Analysis of salt samples following the series of electrolytic reduction runs identified the diffusion of select fission products from the spent fuel to the molten salt electrolyte. The extents of metal oxide conversion in the post-test fuel were also quantified, including a nominal 99.7% conversion of uranium oxide to metal. Uranium metal was then separated from the reduced LWR fuel in a series of six electrorefining runs that were performed in succession with a single salt loading of molten LiCl-KCl-UCl3 at 500 C. Analysis of salt samples following the series of electrorefining runs identified additional partitioning of fission products into the molten salt electrolyte. Analyses of the separated uranium metal were performed, and its decontamination factors were determined.

  16. Management and overview Quality Assurance Program Plan. [Uranium Mill Tailings Remedial Action (UMTRA) Project

    SciTech Connect (OSTI)

    Not Available

    1986-08-01T23:59:59.000Z

    The Uranium Mill Tailings Remedial Action (UMTRA) Project Office (DOE/ UMTRA-PO) is the US Department of Energy (DOE) Albuquerque Operations Office (AL) organization charged with the responsibility of managing and coordinating the activities of the various participating organizations and support contractors working on the UMTRA Project. This Quality Assurance Program Plan (QAPP) describes how the DOE/UMTRA-PO, as assisted by the Technical Assistance Contractor (TAC), performs the quality assurance (QA) aspects of managing and coordinating UMTRA Project activities. This QAPP was developed to comply with DOE Order 5700.6A, August, 1981, and AL Order 5700.6B, April, 1984, which contain the criteria applicable to Project QA activities.

  17. Using flow through reactors to study the non-reductive biomineralization of uranium phosphate minerals.

    E-Print Network [OSTI]

    Williams, Anna Rachel

    2012-01-01T23:59:59.000Z

    ??Uranium contaminations of the subsurface in the vicinity of nuclear materials processing sites pose a health risk as the uranyl ion in its oxidized state, (more)

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

    SciTech Connect (OSTI)

    Michael A. Pope

    2012-07-01T23:59:59.000Z

    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.

  19. An economic analysis of a light and heavy water moderated reactor synergy: burning americium using recycled uranium

    SciTech Connect (OSTI)

    Wojtaszek, D.; Edwards, G. [Atomic Energy of Canada Ltd., Chalk River Laboratories, Chalk River, Ontario (Canada)

    2013-07-01T23:59:59.000Z

    An economic analysis is presented for a proposed synergistic system between 2 nuclear utilities, one operating light water reactors (LWR) and another running a fleet of heavy water moderated reactors (HWR). Americium is partitioned from LWR spent nuclear fuel (SNF) to be transmuted in HWRs, with a consequent averted disposal cost to the LWR operator. In return, reprocessed uranium (RU) is supplied to the HWRs in sufficient quantities to support their operation both as power generators and americium burners. Two simplifying assumptions have been made. First, the economic value of RU is a linear function of the cost of fresh natural uranium (NU), and secondly, plutonium recycling for a third utility running a mixed oxide (MOX) fuelled reactor fleet has been already taking place, so that the extra cost of americium recycling is manageable. We conclude that, in order for this scenario to be economically attractive to the LWR operator, the averted disposal cost due to partitioning americium from LWR spent fuel must exceed 214 dollars per kg, comparable to estimates of the permanent disposal cost of the high level waste (HLW) from reprocessing spent LWR fuel. (authors)

  20. Determination of aerosol size distributions at uranium mill tailings remedial action project sites

    SciTech Connect (OSTI)

    Newton, G.J.; Reif, R.H. [CWM Federal Environmental Services, Inc., Albuquerque, NM (United States); Hoover, M.D.

    1994-11-01T23:59:59.000Z

    The U.S. Department of Energy (DOE) has an ongoing program, the Uranium Mill Tailings Remedial Action (UMTRA) Project, to stabilize piles of uranium mill tailings in order to reduce the potential radiological hazards to the public. Protection of workers and the general public against airborne radioactivity during remedial action is a top priority at the UMTRA Project. The primary occupational radionuclides of concern are {sup 230}Th, {sup 226}Ra, {sup 210}Pb, {sup 210}Po, and the short-lived decay products of {sup 222}Rn with {sup 230}Th causing the majority of the committed effective dose equivalent (CEDE) from inhaling uranium mill tailings. Prior to this study, a default particle size of 1.0 {mu}m activity median aerodynamic diameter (AMAD) was assumed for airborne radioactive tailings dust. Because of recent changes in DOE requirements, all DOE operations are now required to use the CEDE methodology, instead of the annual effective dose equivalent (AEDE) methodology, to evaluate internal radiation exposures. Under the transition from AEDE to CEDE, with a 1.0 {mu}m AMAD particle size, lower bioassay action levels would be required for the UMTRA Project. This translates into an expanded internal dosimetry program where significantly more bioassay monitoring would be required at the UMTRA Project sites. However, for situations where the particle size distribution is known to differ significantly from 1.0 {mu}m AMAD, the DOE allows for corrections to be made to both the estimated dose to workers and the derived air concentration (DAC) values. For particle sizes larger than 1.0 {mu}m AMAD, the calculated CEDE from inhaling tailings would be relatively lower.

  1. A fission-fusion hybrid reactor in steady-state L-mode tokamak configuration with natural uranium

    SciTech Connect (OSTI)

    Reed, Mark; Parker, Ronald R.; Forget, Benoit [Department of Nuclear Science and Engineering, Massachusetts Institute of Technology 77 Massachusetts Avenue, Cambridge, MA 02139 (United States)

    2012-06-19T23:59:59.000Z

    This work develops a conceptual design for a fusion-fission hybrid reactor operating in steady-state L-mode tokamak configuration with a subcritical natural or depleted uranium pebble bed blanket. A liquid lithium-lead alloy breeds enough tritium to replenish that consumed by the D-T fusion reaction. The fission blanket augments the fusion power such that the fusion core itself need not have a high power gain, thus allowing for fully non-inductive (steady-state) low confinement mode (L-mode) operation at relatively small physical dimensions. A neutron transport Monte Carlo code models the natural uranium fission blanket. Maximizing the fission power gain while breeding sufficient tritium allows for the selection of an optimal set of blanket parameters, which yields a maximum prudent fission power gain of approximately 7. A 0-D tokamak model suffices to analyze approximate tokamak operating conditions. This fission blanket would allow the fusion component of a hybrid reactor with the same dimensions as ITER to operate in steady-state L-mode very comfortably with a fusion power gain of 6.7 and a thermal fusion power of 2.1 GW. Taking this further can determine the approximate minimum scale for a steady-state L-mode tokamak hybrid reactor, which is a major radius of 5.2 m and an aspect ratio of 2.8. This minimum scale device operates barely within the steady-state L-mode realm with a thermal fusion power of 1.7 GW. Basic thermal hydraulic analysis demonstrates that pressurized helium could cool the pebble bed fission blanket with a flow rate below 10 m/s. The Brayton cycle thermal efficiency is 41%. This reactor, dubbed the Steady-state L-mode non-Enriched Uranium Tokamak Hybrid (SLEUTH), with its very fast neutron spectrum, could be superior to pure fission reactors in terms of breeding fissile fuel and transmuting deleterious fission products. It would likely function best as a prolific plutonium breeder, and the plutonium it produces could actually be more proliferation-resistant than that bred by conventional fast reactors. Furthermore, it can maintain constant total hybrid power output as burnup proceeds by varying the neutron source strength.

  2. Summary of the radiological assessment of the fuel cycle for a thorium-uranium carbide-fueled fast breeder reactor

    SciTech Connect (OSTI)

    Tennery, V.J.; Bomar, E.S.; Bond, W.D.; Meyer, H.R.; Morse, L.E.; Till, J.E.; Yalcintas, M.G.

    1980-01-01T23:59:59.000Z

    A large fraction of the potential fuel for nuclear power reactors employing fissionable materials exists as ores of thorium. In addition, certain characteristics of a fuel system based on breeding of the fissionable isotope {sup 233}U from thorium offer the possibility of a greater resistance to the diversion of fissionable material for the fabrication of nuclear weapons. This report consolidates into a single source the principal content of two previous reports which assess the radiological environmental impact of mining and milling of thorium ore and of the reprocessing and refabrication of spent FBR thorium-uranium carbide fuel.

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

    SciTech Connect (OSTI)

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

    2007-11-01T23:59:59.000Z

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

  4. Assumptions and Criteria for Performing a Feasability Study of the Conversion of the High Flux Isotope Reactor Core to Use Low-Enriched Uranium Fuel

    SciTech Connect (OSTI)

    Primm, R.T., III; Ellis, R.J.; Gehin, J.C.; Moses, D.L.; Binder, J.L.; Xoubi, N. (U. of Cincinnati)

    2006-02-01T23:59:59.000Z

    A computational study will be initiated during fiscal year 2006 to examine the feasibility of converting the High Flux Isotope Reactor from highly enriched uranium fuel to low-enriched uranium. The study will be limited to steady-state, nominal operation, reactor physics and thermal-hydraulic analyses of a uranium-molybdenum alloy that would be substituted for the current fuel powder--U{sub 3}O{sub 8} mixed with aluminum. The purposes of this document are to (1) define the scope of studies to be conducted, (2) define the methodologies to be used to conduct the studies, (3) define the assumptions that serve as input to the methodologies, (4) provide an efficient means for communication with the Department of Energy and American research reactor operators, and (5) expedite review and commentary by those parties.

  5. Industrial opportunities on the International Thermonuclear Experimental Reactor (ITER) project

    SciTech Connect (OSTI)

    Ellis, W.R. [Raytheon Engineers and Constructors, New York, NY (United States)

    1996-12-01T23:59:59.000Z

    Industry has been a long-term contributor to the magnetic fusion program, playing a variety of important roles over the years. Manufacturing firms, engineering-construction companies, and the electric utility industry should all be regarded as legitimate stakeholders in the fusion energy program. In a program focused primarily on energy production, industry`s future roles should follow in a natural way, leading to the commercialization of the technology. In a program focused primarily on science and technology, industry`s roles, in the near term, should be, in addition to operating existing research facilities, largely devoted to providing industrial support to the International Thermonuclear Experimental Reactor (ITER) Project. Industrial opportunities on the ITER Project will be guided by the amount of funding available to magnetic fusion generally, since ITER is funded as a component of that program. The ITER Project can conveniently be discussed in terms of its phases, namely, the present Engineering Design Activities (EDA) phase, and the future (as yet not approved) construction phase. 2 refs., 3 tabs.

  6. Print this article Close This Window EU OKs India joining ITER nuclear reactor project

    E-Print Network [OSTI]

    Print this article Close This Window EU OKs India joining ITER nuclear reactor project Fri Dec 2-billion-euro project to build an experimental nuclear fusion reactor that in the long-run could provide virtually unlimited, cheap and clean energy. The EU's willingness to work with India on a civil nuclear

  7. Preplanning guidance document for groundwater restoration. [Uranium Mill Tailings Remedial Action (UMTRA) Project

    SciTech Connect (OSTI)

    Not Available

    1991-06-01T23:59:59.000Z

    This document is intended to present decision makers on the Uranium Mill Tailings Remedial Action (UMTRA) Project with a strategy for evaluating the need for and accomplishing groundwater restoration at those UMTRA Project sites that will require restoration as specified in Subpart B of 40 CFR 192. A synopsis of the Uranium Mill Tailings Restoration Control Act (UMTRCA) and a discussion of the proposed US Environmental Protection Agency (EPA) groundwater cleanup standards in 40 CFR 192 are provided to define the regulatory basis of groundwater cleanup. Once the EPA groundwater standards are finalized, this document may be revised, depending on the changes in the final standards. A procedure for determining the need for groundwater restoration is outlined and a cost-effective strategy for selecting, designing, implementing, and evaluating appropriate restoration procedures is presented. The determination of the need for groundwater restoration is based on the EPA groundwater cleanup standards and must be compatible with and complementary to the National Environmental Policy Act (NEPA) process. Therefore, this document describes a programmatic approach for compliance with the NEPA process.

  8. Contents of environmental impact statements prepared for the Uranium Mill Tailings Remedial Action Project

    SciTech Connect (OSTI)

    Not Available

    1986-01-01T23:59:59.000Z

    This document presents two versions of the outline for the environmental impact statements (EISS) to be prepared for the Uranium Mill Tailings Remedial Action (UMTRA) Project. The first displays the basic structure of the statements; it lists only the titles of sections. The second is a guide to the contents of the statements which provides, under each title, a brief summary of contents. The outline is intended to comply with the planning requirements and the definitions of terms established by the Council on Environmental Quality as well as DOE Order 5440.lB (Implementation of the National Environmental Policy Act), and compliance with Floodplain/Wetlands Environmental Review Requirements. These requirements and definitions are implicity part of the outline. The outline presented in this document will guide the preparation of EISs Guidelines for preparation of environmental assessments for the UMTRA Project are available.

  9. Electrochemical separation of aluminum from uranium for research reactor spent nuclear fuel applications.

    SciTech Connect (OSTI)

    Slater, S. A.; Willit, J. L.; Gay, E. C.; Chemical Engineering

    1999-01-01T23:59:59.000Z

    Researchers at Argonne National Laboratory (ANL) are developing an electrorefining process to treat aluminum-based spent nuclear fuel by electrochemically separating aluminum from uranium. The aluminum electrorefiner is modeled after the high-throughput electrorefiner developed at ANL. Aluminum is electrorefined, using a fluoride salt electrolyte, in a potential range of -0.1 V to -0.2 V, while uranium is electrorefined in a potential range of -0.3 V to -0.4 V; therefore, aluminum can be selectively separated electrochemically from uranium. A series of laboratory-scale experiments was performed to demonstrate the aluminum electrorefining concept. These experiments involved selecting an electrolyte (determining a suitable fluoride salt composition); selecting a crucible material for the electrochemical cell; optimizing the operating conditions; determining the effect of adding alkaline and rare earth elements to the electrolyte; and demonstrating the electrochemical separation of aluminum from uranium, using a U-Al-Si alloy as a simulant for aluminum-based spent nuclear fuel. Results of the laboratory-scale experiments indicate that aluminum can be selectively electrotransported from the anode to the cathode, while uranium remains in the anode basket.

  10. 100-KE REACTOR CORE REMOVAL PROJECT ALTERNATIVE ANALYSIS WORKSHOP REPORT

    SciTech Connect (OSTI)

    HARRINGTON RA

    2010-01-15T23:59:59.000Z

    On December 15-16, 2009, a 100-KE Reactor Core Removal Project Alternative Analysis Workshop was conducted at the Washington State University Consolidated Information Center, Room 214. Colburn Kennedy, Project Director, CH2M HILL Plateau Remediation Company (CHPRC) requested the workshop and Richard Harrington provided facilitation. The purpose of the session was to select the preferred Bio Shield Alternative, for integration with the Thermal Shield and Core Removal and develop the path forward to proceed with project delivery. Prior to this workshop, the S.A. Robotics (SAR) Obstruction Removal Alternatives Analysis (565-DLV-062) report was issued, for use prior to and throughout the session, to all the team members. The multidisciplinary team consisted ofrepresentatives from 100-KE Project Management, Engineering, Radcon, Nuclear Safety, Fire Protection, Crane/Rigging, SAR Project Engineering, the Department of Energy Richland Field Office, Environmental Protection Agency, Washington State Department of Ecology, Defense Nuclear Facility Safety Board, and Deactivation and Decommission subject matter experts from corporate CH2M HILL and Lucas. Appendix D contains the workshop agenda, guidelines and expectations, opening remarks, and attendance roster going into followed throughout the workshop. The team was successful in selecting the preferred alternative and developing an eight-point path forward action plan to proceed with conceptual design. Conventional Demolition was selected as the preferred alternative over two other alternatives: Diamond Wire with Options, and Harmonic Delamination with Conventional Demolition. The teams preferred alternative aligned with the SAR Obstruction Removal Alternative Analysis report conclusion. However, the team identified several Path Forward actions, in Appendix A, which upon completion will solidify and potentially enhance the Conventional Demolition alternative with multiple options and approaches to achieve project delivery. In brief, the Path Forward was developed to reconsider potential open air demolition areas; characterize to determine if any zircaloy exists, evaluate existing concrete data to determine additional characterization needs, size the new building to accommodate human machine interface and tooling, consider bucket thumb and use ofshape-charges in design, and finally to utilize complex-wide and industry explosive demolition lessons learned in the design approach. Appendix B documents these results from the team's use ofValue Engineering process tools entitled Weighted Analysis Alternative Matrix, Matrix Conclusions, Evaluation Criteria, and Alternative Advantages and Disadvantages. These results were further supported with the team's validation of parking-lot information sheets: memories (potential ideas to consider), issues/concerns, and assumptions, contained in Appendix C. Appendix C also includes the recorded workshop flipchart notes taken from the SAR Alternatives and Project Overview presentations. The SAR workshop presentations, including a 3-D graphic illustration demonstration video have been retained in the CHPRC project file, and were not included in this report due to size limitations. The workshop concluded with a round robin close-out where each member was engaged for any last minute items and meeting utility. In summary, the team felt the session was value added and looked forward to proceeding with the recommended actions and conceptual design.

  11. U.S. Department of Energy Uranium Mill Tailings Remedial Action Ground Water Project: Project plan

    SciTech Connect (OSTI)

    Not Available

    1994-09-01T23:59:59.000Z

    The scope of the Project is to develop and implement a ground water compliance strategy for all 24 UMTRA Project processing sites. The compliance strategy for the processing sites must satisfy the proposed EPA ground water cleanup standards in 40 CFR Part 192, Subparts B and C (1987). This scope of work will entail the following activities on a site-specific basis: Develop a compliance strategy based on modification of the UMTRA Surface Project RAPs or develop Ground Water Project RAPs with NRC concurrence on the RAP and full participation of the affected states and tribes. Implement the RAP to include institutional controls, where appropriate, as an interim measure until compliance with the standards is achieved. Institute long-term verification monitoring for transfer to a separate long-term surveillance program on or before the Project end date. Prepare certification or confirmation reports and modify the long-term surveillance plan (LTSP), where needed, on those sites completed prior to the Project end date.

  12. Hydro-mechanical analysis of low enriched uranium fuel plates for University of Missouri Research Reactor .

    E-Print Network [OSTI]

    Kennedy, John C.

    2012-01-01T23:59:59.000Z

    ??As part of the Global Threat Reduction Initiative (GTRI) Reactor Conversion program, work is underway to analyze and validate a new fuel assembly for the (more)

  13. Uranium-mill-tailings remedial-action project (UMTRAP) cover and liner technology development project

    SciTech Connect (OSTI)

    Hartley, J.N.; Gee, G.W.; Freeman, H.D.; Cline, J.F.; Beedlow, P.A.; Buelt, J.L.; Relyea, J.R.; Tamura, T.

    1982-01-01T23:59:59.000Z

    Cover and liner systems for uranium mill tailings in the United States must satisfy stringent requirements regarding long-term stability, radon control, and radionuclide and hazardous chemical migration. The cover placed over a tailings pile serves three basic purposes: (1) to reduce the release of radon, (2) to prevent the intrusion of plant roots and burrowing animals into the tailings, and (3) to limit surface erosion. The liner placed under a tailings pile prevents the migration of radionuclides and hazardous chemicals to groundwater. Pacific Northwest Laboratory is developing and evaluating cover and liner systems that meet these objectives and conform to federal standards. The cover and liner technology discussed in this paper involves: (1) single and multilayer earthen cover systems, (2) asphalt emulsion radon barrier systems, (3) biobarrier systems, (4) revegetation and rock covers, and (5) asphalt, clay, and synthetic liner systems. These systems have been tested at the Grand Junction, Colorado, tailings pile, where they have been shown to effectively reduce radon releases and radionuclide and chemical migration.

  14. Action Memorandum for the Engineering Test Reactor under the Idaho Cleanup Project

    SciTech Connect (OSTI)

    A. B. Culp

    2007-01-26T23:59:59.000Z

    This Action Memorandum documents the selected alternative for decommissioning of the Engineering Test Reactor at the Idaho National Laboratory under the Idaho Cleanup Project. Since the missions of the Engineering Test Reactor Complex have been completed, an engineering evaluation/cost analysis that evaluated alternatives to accomplish the decommissioning of the Engineering Test Reactor Complex was prepared adn released for public comment. The scope of this Action Memorandum is to encompass the final end state of the Complex and disposal of the Engineering Test Reactor vessol. The selected removal action includes removing and disposing of the vessel at the Idaho CERCLA Disposal Facility and demolishing the reactor building to ground surface.

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

    SciTech Connect (OSTI)

    NONE

    1995-11-30T23:59:59.000Z

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

  16. Vegetative covers: Special study. [Uranium Mill Tailings Remedial Action (UMTRA) Project

    SciTech Connect (OSTI)

    Not Available

    1988-11-01T23:59:59.000Z

    This report describes the findings of a special study on the use of vegetative covers to stabilize tailings piles for the Uranium Mill Tailings Remedial Action (UMTRA) Project. The principal rationale for using plants would be to establish a dynamic system for controlling water balance. Specifically, vegetation would be used to intercept and transpire precipitation to the atmosphere, rather than allowing water to drain into the tailings and mobilize contaminants. This would facilitate compliance with groundwater standards proposed for the UMTRA Project by the Environmental Protection Agency. The goals of the study were to (1) evaluate the feasibility of using vegetative covers on UMTRA Project piles, (2) define the advantages and disadvantages of vegetative covers, and (3) develop general guidelines for their use when such use seems reasonable. The principal method for the study was to analyze and apply to the UMTRA Project the results of research programs on vegetative covers at other US Department of Energy (DOE) waste management facilities. The study also relied upon observations made of existing stabilized piles at UMTRA Project sites (Shiprock, New Mexico; Burrell, Pennsylvania; and Clive, Utah) where natural vegetation is growing on the rock-covered surfaces. Water balance and erosion models were also used to quantify the long-term performance of vegetative covers planned for the topslopes of stabilized piles at Grand Junction and Durango, Colorado, two UMTRA Project sites where the decision was made during the course of this special study to use vegetative covers. Elements in the design and construction of the vegetative covers at these two sites are discussed in the report, with explanations of the differing features that reflect differing environmental conditions.

  17. Mitigation and monitoring plan for impacted wetlands at the Gunnison UMTRA Project site, Gunnison, Colorado. [Uranium Mill Tailings Remedial Action (UMTRA) Project

    SciTech Connect (OSTI)

    Not Available

    1992-06-01T23:59:59.000Z

    The U.S Department of Energy (DOE) administers the Uranium Mill Tailings Remedial Action (UMTRA) Project. The UMTRA Project is the result of the Uranium Mill Tailings Radiation Control Act(UMTRA) which was passed in response to the public's concern over the potential public health hazards related to uranium mill tailings and associated contaminated material at abandoned or otherwise uncontrolled inactive processing sites throughout the United States. The Gunnison, Colorado abandoned uranium mill site is one of the sites slated for cleanup by the DOE under authority of UMTRA. The contaminated material at this site will be transported to a disposal site on US Bureau of Land Management (BLM) land east of Gunnison. Remedial action activities will temporarily disturb 0.8 acre and permanently eliminate 5.1 acres of wetlands. This report describes the proposed mitigation plan for the 5.9 acres of impacted wetlands. In conjunction with the mitigation of the permanently impacted wetlands through the enhancement of wetland and adjacent riparian areas, impacts to wildlife as a result of this project will also be mitigated. However, wildlife mitigation is not the focus of this document and is covered in relevant BLM permits for this project. This plan proposes the enhancement of a 3:1 ratio of impacted wetlands in accordance with US Environmental Protection Agency guidelines, plus the enhancement of riparian areas for wildlife mitigation. Included in this mitigation plan is a monitoring plan to ensure that the proposed measures are working and being maintained.

  18. Groundwater protection management program plan. [Uranium Mill Tailings Remedial Action (UMTRA) Project

    SciTech Connect (OSTI)

    Not Available

    1992-06-01T23:59:59.000Z

    US Department of Energy (DOE) Order 5400.1 requires the establishment of a groundwater protection management program to ensure compliance with DOE requirements and applicable Federal, state, and local laws and regulations. The Uranium Mill Tailings Remedial Action (UMTRA) Project Office has prepared a Groundwater Protection Management Program Plan'' (groundwater protection plan) of sufficient scope and detail to reflect the program's significance and address the seven activities required in DOE Order 5400.1, Chapter 3, for special program planning. The groundwater protection plan highlights the methods designed to preserve, protect, and monitor groundwater resources at UMTRA Project processing and disposal sites. The plan includes an overview of the remedial action status at the 24 designated processing sites and identifies project technical guidance documents and site-specific documents for the UMTRA groundwater protection management program. In addition, the groundwater protection plan addresses the general information required to develop a water resources protection strategy at the permanent disposal sites. Finally, the plan describes ongoing activities that are in various stages of development at UMTRA sites (long-term care at disposal sites and groundwater restoration at processing sites). This plan will be reviewed annually and updated every 3 years in accordance with DOE Order 5400.1.

  19. Selection of water treatment processes special study. [Uranium Mill Tailings Remedial Action (UMTRA) Project

    SciTech Connect (OSTI)

    Not Available

    1991-11-01T23:59:59.000Z

    Characterization of the level and extent of groundwater contamination in the vicinity of Title I mill sites began during the surface remedial action stage (Phase 1) of the Uranium Mill Tailings Remedial Action (UMTRA) Project. Some of the contamination in the aquifer(s) at the abandoned sites is attributable to milling activities during the years the mills were in operation. The restoration of contaminated aquifers is to be undertaken in Phase II of the UMTRA Project. To begin implementation of Phase II, DOE requested that groundwater restoration methods and technologies be investigated by the Technical Assistance Contractor (TAC). and that the results of the TAC investigations be documented in special study reports. Many active and passive methods are available to clean up contaminated groundwater. Passive groundwater treatment includes natural flushing, geochemical barriers, and gradient manipulation by stream diversion or slurry walls. Active groundwater.cleanup techniques include gradient manipulation by well extraction or injection. in-situ biological or chemical reclamation, and extraction and treatment. Although some or all of the methods listed above may play a role in the groundwater cleanup phase of the UMTRA Project, the extraction and treatment (pump and treat) option is the only restoration alternative discussed in this report. Hence, all sections of this report relate either directly or indirectly to the technical discipline of process engineering.

  20. Economic impact study of the Uranium Mill Tailings Remedial Action Project in Colorado: Colorado state fiscal year 1993

    SciTech Connect (OSTI)

    Not Available

    1993-12-01T23:59:59.000Z

    The Colorado economic impact study summarizes employment and economic benefits to the state from activities associated with the Uranium Mill Tailings Remedial Action (UMTRA) Project during Colorado state fiscal year 1993 (July 1, 1992, through June 30, 1993). To capture employment benefits, a questionnaire was distributed to subcontractor employees at the active UMTRA Project sites of Grand Junction, Rifle, and Gunnison, Colorado. An estimated 52 percent of the employees working on the UMTRA Project responded to this information request. Economic data were requested from each site prime subcontractor, as well as from the Remedial Action Contractor. The most significant benefits associated with the UMTRA Project in Colorado are summarized.

  1. Microbiological, Geochemical and Hydrologic Processes Controlling Uranium Mobility: An Integrated Field-Scale Subsurface Research Challenge Site at Rifle, Colorado, Quality Assurance Project Plan

    SciTech Connect (OSTI)

    Fix, N. J.

    2008-01-07T23:59:59.000Z

    The U.S. Department of Energy (DOE) is cleaning up and/or monitoring large, dilute plumes contaminated by metals, such as uranium and chromium, whose mobility and solubility change with redox status. Field-scale experiments with acetate as the electron donor have stimulated metal-reducing bacteria to effectively remove uranium [U(VI)] from groundwater at the Uranium Mill Tailings Site in Rifle, Colorado. The Pacific Northwest National Laboratory and a multidisciplinary team of national laboratory and academic collaborators has embarked on a research proposed for the Rifle site, the object of which is to gain a comprehensive and mechanistic understanding of the microbial factors and associated geochemistry controlling uranium mobility so that DOE can confidently remediate uranium plumes as well as support stewardship of uranium-contaminated sites. This Quality Assurance Project Plan provides the quality assurance requirements and processes that will be followed by the Rifle Integrated Field-Scale Subsurface Research Challenge Project.

  2. 105-H Reactor Interim Safe Storage Project Final Report

    SciTech Connect (OSTI)

    E.G. Ison

    2008-11-08T23:59:59.000Z

    The following information documents the decontamination and decommissioning of the 105-H Reactor facility, and placement of the reactor core into interim safe storage. The D&D of the facility included characterization, engineering, removal of hazardous and radiologically contaminated materials, equipment removal, decontamination, demolition of the structure, and restoration of the site. The ISS work also included construction of the safe storage enclosure, which required the installation of a new roofing system, power and lighting, a remote monitoring system, and ventilation components.

  3. Economic impact study of the Uranium Mill Tailings Remedial Action project in Colorado: Colorado state fiscal year 1995

    SciTech Connect (OSTI)

    NONE

    1995-12-01T23:59:59.000Z

    This Colorado economic impact study summarizes employment and economic benefits to the state from activities associated with the Uranium Mill Tailings Remedial Action (UMTRA) Project during Colorado state fiscal year (FY) 1995 (1 July 1994 through 30 June 1995). To capture employment information, a questionnaire was distributed to subcontractor employees at the active UMTRA Project sites of Grand Junction, Gunnison, Maybell, Naturita, Rifle, and Slick Rock, Colorado. Economic data were requested from the Remedial Action Contractor (RAC), the Technical Assistance Contractor (TAC) and the US Department of Energy (DOE). The most significant benefits associated with the UMTRA Project in Colorado are summarized.

  4. On the possibility of using uranium-beryllium oxide fuel in a VVER reactor

    SciTech Connect (OSTI)

    Kovalishin, A. A.; Prosyolkov, V. N.; Sidorenko, V. D. [National Research Center Kurchatov Institute (Russian Federation); Stogov, Yu. V., E-mail: YVStogov@mephi.ru [National Research Nuclear University MEPhI (Russian Federation)

    2014-12-15T23:59:59.000Z

    The possibility of using UO{sub 2}-BeO fuel in a VVER reactor is considered with allowance for the thermophysical properties of this fuel. Neutron characteristics of VVER fuel assemblies with UO{sub 2}-BeO fuel pellets are estimated.

  5. 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

    2008-07-01T23:59:59.000Z

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

  6. THE ENERGY SPECTRA OF URANIUM ATOMS SPUTTERED FROM URANIUM METAL AND URANIUM DIOXIDE TARGETS

    E-Print Network [OSTI]

    Winfree, Erik

    THE ENERGY SPECTRA OF URANIUM ATOMS SPUTTERED FROM URANIUM METAL AND URANIUM DIOXIDE TARGETS Thesis. I have benefitted from conversations with many persons w~ile engaged in this project. I would like

  7. Mobile water treatment plant special study. Uranium Mill Tailings Remedial Action Project

    SciTech Connect (OSTI)

    Not Available

    1992-12-01T23:59:59.000Z

    Characterization of the level and extent of groundwater contamination in the vicinity of Title I mill sites began during the surface remedial action stage (Phase 1) of the Uranium Mill Tailings Remedial Action (UMTRA) Project. Some of the contamination in the aquifer(s) at the abandoned sites is attributable to milling activities during the years the mills were in operation. To begin implementation of Phase 11 groundwater remediation, the US Department of Energy (DOE) requested that (1) the Technical Assistance Contractor (TAC) conduct a study to provide for the design of a mobile water treatment plant to treat groundwater extracted during site characterization studies at completed Phase I UMTRA sites, and (2) the results of the TAC investigations be documented in a special study report. This special study develops the design criteria for a water treatment plant that can be readily transported from one UMTRA site to another and operated as a complete treatment system. The 1991 study provides the basis for selecting a mobile water treatment system to meet the operating requirements recommended in this special study. The scope of work includes the following: Determining contaminants, flows, and loadings. Setting effluent quality criteria. Sizing water treatment unit(s). Evaluating non-monetary aspects of alternate treatment processes. Comparing costs of alternate treatment processes. Recommending the mobile water treatment plant design criteria.

  8. Study on Evaluation of Project Management Data for Decommissioning of Uranium Refining and Conversion Plant - 12234

    SciTech Connect (OSTI)

    Usui, Hideo; Izumo, Sari; Tachibana, Mitsuo [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki, 319-1195 (Japan); Shibahara, Yuji [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki, 319-1195 (Japan); University of Fukui, Fukui-shi, Fukui, 910-8507 (Japan); Morimoto, Yasuyuki; Tokuyasu, Takashi; Takahashi, Nobuo; Tanaka, Yoshio; Sugitsue, Noritake [Japan Atomic Energy Agency, Kagamino-cho, Tomata-gun, Okayama, 708-0698 (Japan)

    2012-07-01T23:59:59.000Z

    Some of nuclear facilities that would no longer be required have been decommissioned in JAEA (Japan Atomic Energy Agency). A lot of nuclear facilities have to be decommissioned in JAEA in near future. To implement decommissioning of nuclear facilities, it was important to make a rational decommissioning plan. Therefore, project management data evaluation system for dismantling activities (PRODIA code) has been developed, and will be useful for making a detailed decommissioning plan for an object facility. Dismantling of dry conversion facility in the uranium refining and conversion plant (URCP) at Ningyo-toge began in 2008. During dismantling activities, project management data such as manpower and amount of waste generation have been collected. Such collected project management data has been evaluated and used to establish a calculation formula to calculate manpower for dismantling equipment of chemical process and calculate manpower for using a green house (GH) which was a temporary structure for preventing the spread of contaminants during dismantling. In the calculation formula to calculate project management data related to dismantling of equipment, the relation of dismantling manpower to each piece of equipment was evaluated. Furthermore, the relation of dismantling manpower to each chemical process was evaluated. The results showed promise for evaluating dismantling manpower with respect to each chemical process. In the calculation formula to calculate project management data related to use of the GH, relations of GH installation manpower and removal manpower to GH footprint were evaluated. Furthermore, the calculation formula for secondary waste generation was established. In this study, project management data related to dismantling of equipment and use of the GH were evaluated and analyzed. The project management data, manpower for dismantling of equipment, manpower for installation and removal of GH, and secondary waste generation from GH were considered. Establishment of the calculation formula for dismantling of each kind of equipment makes it possible to evaluate manpower for dismantling the whole facility. However, it is not easy to prepare calculation formula for all kinds of equipment that exist in the facility. Therefore, a simpler evaluation method was considered to calculate manpower based on facility characteristics. The results showed promise for evaluating dismantling manpower with respect to each chemical process. For dismantling of contaminated equipment, a GH has been used for protection of the spread of contamination. The use of a GH increases manpower for installation and removal of GH etc. Moreover, structural materials of the GH such as plastic sheets, adhesive tape become a burnable secondary waste. To create an effective dismantling plan, it is necessary to carefully consider use of a GH preliminarily. Thus, an evaluation method of project management data such as manpower and secondary waste generation was considered. The results showed promise for evaluating project management data of GH by using established calculation formula. (authors)

  9. Opportunities to reduce consumption of natural uranium in reactor SVBR-75/100 when changing over to the closed fuel cycle

    SciTech Connect (OSTI)

    Toshinsky, G.I.; Komlev, O.G.; Mel'nikov, K.G.; Novikova, N.N. [FSUE SSC RF-IPPE, 1, Bondarenko sq., Obninsk, Kaluga rg., 249033 (Russian Federation)

    2007-07-01T23:59:59.000Z

    The design of reactor SVBR-75/100 allows it to operate using different types of fuel and in different fuel cycles without changing its design and deteriorating its safety characteristics. Fuel-at-once refueling adopted in the design (lack of partial refueling) makes it possible to change the core content at each refueling by using the type of fuel that is the most economically effective at the current stage of nuclear power (NP) development. In the nearest future use of mastered oxide uranium fuel and operating in the opened fuel cycle with postponed reprocessing will be the most economically effective. Changeover to the mixed uranium-plutonium fuel and closed nuclear fuel cycle (NFC) will be economically effective in an event of increase of natural uranium costs when the expenditures for construction of the enterprises on reprocessing the spent nuclear fuel (SNF), re-fabrication of new fuel with plutonium and their operating are less than the corresponding costs of natural uranium, its enrichment costs, the costs of manufacturing fresh uranium fuel and long temporary storage of SNF. At this, it is possible to use both MOX fuel with weapon or reactor plutonium and mixed nitride fuel in case its usage is more profitable. As fast reactors (FR) using uranium fuel and operating in the opened NFC consume much more natural uranium in comparison with thermal reactors (TR), and at the expected high paces of NP development the cheap resources of natural uranium will be exhausted prior to the middle of the century that will cause increase in the uranium cost, the period of FRs operating in the opened NFC must be maximally reduced. However, it should be mentioned that it is difficult to forecast reliably the date when because of the increased cost of natural uranium the NP will lose its competitiveness with electric power using fossil fuel. This is conditioned by the fact that the cost of the NPP produced electricity is less sensitive to the cost of natural uranium in contrast to the cost of electricity produced by thermal power plants using fossil fuel. At the same time, the available resources of natural uranium are increasing progressively with increase of its cost. The expenditure caused by changeover to the closed NFC will be less, if plutonium extracted from the own SNF of uranium loads is used in fabrication of the first MOX fuel loads. If the oxide uranium fuel is used, by the end of the lifetime a comparatively high breeding ratio (BR) ({approx}0.84) provides a sufficiently high content of plutonium in the SNF that may be used in the next fuel lifetimes when organizing the closed fuel cycle. Moreover, the own SNF of starting loads from oxide uranium fuel contains large quantity of unburned uranium-235 that is expedient to use for forming load for the next lifetime. From the very beginning of realization of the extended program on implementation of reactors SVBR-75/100 in the NP, use of plutonium extracted from the TRs' SNF for forming the starting loads of those reactors for the purpose of total elimination of natural uranium consumption will be more expensive as compared with the considered variant of changeover from the opened NFC to the closed NFC. This is conditioned by the fact that for the plutonium extracted from the TRs' SNF, the plutonium cost determined by a volume of SNF reprocessing per ton of plutonium will be several times higher as compared with its cost in case of using the own SNF because of considerably less content of plutonium in the TRs' SNF. It should be taken into account that the organization of the enterprise on large-scale reprocessing of TRs' SNF and MOX fuel fabrication must precede the construction of NPPs with FRs. Thus, the demands in investments are increased. At the same time, for the proposed changeover from the opened NFC to the closed one the construction of the closed NFC enterprise may be long postponed from FR launching that reduces the investment demands. At this, as the assessments have revealed, the investment fund for construction of such enterprise could be formed during abo ut t

  10. Contents of environmental assessments prepared for the Uranium Mill Tailings Remedial Action Project

    SciTech Connect (OSTI)

    Not Available

    1986-01-01T23:59:59.000Z

    This document presents two versions of the outline for the environmental assessments (EAS) to be prepared for the Uranium Mill Tailings Remedial Action (UMTRA) Project. The first displays the basic structure of the assessments; it lists only the titles of sections. The second is a guide to the contents of the assessments which provides, under each title, a brief summary of contents. The outline is intended to comply with the planning requirements (40 CFR Part 1501) and the definitions of terms (40-' CFR Part 1508) established by the Council on Environmental Quality as well as DOE order 5440.lB (Implementation of the National Environmental Policy Act), and compliance with Floodplain/Wetlands Environmental Review Requirements (10 CFR Part 1022). These requirements and definitions are implicitly part of the outline. The outline presented in this document will guide the preparation of EAs. The UMTRA Project EAs will be used in determining whether the DOE should prepare an environmental impact statement or a finding of no significant impact for the actions at each of the sites. If no impact statement is necessary, the environmental assessment for that site will aid the DOE in complying with the National Environmental Policy Act before beginning remedial actions. If an impact statement is needed, the assessment will aid its preparation. These purposes, established by the Council on Environmental Quality in 40 CFR Part 1508.9(a), have guided the construction of the outline presented in this document. Remedial actions at each site will include the cleanup of properties in the vicinity of the tailings sites that have been contaminated by the tailings.

  11. Contents of environmental assessments prepared for the Uranium Mill Tailings Remedial Action Project

    SciTech Connect (OSTI)

    Not Available

    1986-01-01T23:59:59.000Z

    This document presents two versions of the outline for the environmental assessments (EAS) to be prepared for the Uranium Mill Tailings Remedial Action (UMTRA) Project. The first displays the basic structure of the assessments; it lists only the titles of sections. The second is a guide to the contents of the assessments which provides, under each title, a brief summary of contents. The outline is intended to comply with the planning requirements (40 CFR Part 1501) and the definitions of terms (40-` CFR Part 1508) established by the Council on Environmental Quality as well as DOE order 5440.lB (Implementation of the National Environmental Policy Act), and compliance with Floodplain/Wetlands Environmental Review Requirements (10 CFR Part 1022). These requirements and definitions are implicitly part of the outline. The outline presented in this document will guide the preparation of EAs. The UMTRA Project EAs will be used in determining whether the DOE should prepare an environmental impact statement or a finding of no significant impact for the actions at each of the sites. If no impact statement is necessary, the environmental assessment for that site will aid the DOE in complying with the National Environmental Policy Act before beginning remedial actions. If an impact statement is needed, the assessment will aid its preparation. These purposes, established by the Council on Environmental Quality in 40 CFR Part 1508.9(a), have guided the construction of the outline presented in this document. Remedial actions at each site will include the cleanup of properties in the vicinity of the tailings sites that have been contaminated by the tailings.

  12. INDEPENDENT VERIFICATION SURVEY OF THE HIGH FLUX BEAM REACTOR DECOMMISSIONING PROJECT OUTSIDE AREAS BROOKHAVEN NATIONAL LABORATORY UPTON, NEW YORK

    SciTech Connect (OSTI)

    P.C. Weaver

    2010-12-15T23:59:59.000Z

    5098-SR-03-0 FINAL REPORT- INDEPENDENT VERIFICATION SURVEY OF THE HIGH FLUX BEAM REACTOR DECOMMISSIONING PROJECT OUTSIDE AREAS, BROOKHAVEN NATIONAL LABORATORY

  13. LETTER REPORT INDEPENDENT VERIFICATION OF THE HIGH FLUX BEAM REACTOR DECOMMISSIONING PROJECT FAN HOUSE, BUILDING 704 BNL

    SciTech Connect (OSTI)

    P.C. Weaver

    2010-10-22T23:59:59.000Z

    5098-LR-01-0 -LETTER REPORT INDEPENDENT VERIFICATION OF THE HIGH FLUX BEAM REACTOR DECOMMISSIONING PROJECT FAN HOUSE, BUILDING 704 BROOKHAVEN NATIONAL LABORATORY

  14. Preparation and Characterization of Uranium Oxides in Support of the K Basin Sludge Treatment Project

    SciTech Connect (OSTI)

    Sinkov, Sergey I.; Delegard, Calvin H.; Schmidt, Andrew J.

    2008-07-08T23:59:59.000Z

    Uraninite (UO2) and metaschoepite (UO32H2O) are the uranium phases most frequently observed in K Basin sludge. Uraninite arises from the oxidation of uranium metal by anoxic water and metaschoepite arises from oxidation of uraninite by atmospheric or radiolytic oxygen. Studies of the oxidation of uraninite by oxygen to form metaschoepite were performed at 21C and 50C. A uranium oxide oxidation state characterization method based on spectrophotometry of the solution formed by dissolving aqueous slurries in phosphoric acid was developed to follow the extent of reaction. This method may be applied to determine uranium oxide oxidation state distribution in K Basin sludge. The uraninite produced by anoxic corrosion of uranium metal has exceedingly fine particle size (6 nm diameter), forms agglomerates, and has the formula UO2.0040.007; i.e., is practically stoichiometric UO2. The metaschoepite particles are flatter and wider when prepared at 21C than the particles prepared at 50C. These particles are much smaller than the metaschoepite observed in prolonged exposure of actual K Basin sludge to warm moist oxidizing conditions. The uraninite produced by anoxic uranium metal corrosion and the metaschoepite produced by reaction of uraninite aqueous slurries with oxygen may be used in engineering and process development testing. A rapid alternative method to determine uranium metal concentrations in sludge also was identified.

  15. GIF sodium fast reactor project R and D on safety and operation

    SciTech Connect (OSTI)

    Vasile, A.; Sofu, T.; Jeong, H. Y.; Sakai, T. [CEA DEN Cadarache, DER, 13108 Saint-Paul-Lez-Durance (France)

    2012-07-01T23:59:59.000Z

    The 'Safety and Operation' project is started in 2009 within the framework of Generation-IV International Forum (GIF) Sodium Fast Reactor (SFR) research and development program. In the safety area, the project involves R and D activities on phenomenological model development and experimental programs, conceptual studies in support of the design of safety provisions, preliminary assessment of safety systems, framework and methods for analysis of safety architecture. In the operation area, the project involves R and D activities on fast reactors safety tests and analysis of reactor operations, feedback from decommissioning, in-service inspection technique development, under-sodium viewing and sodium chemistry. This paper presents a summary of such activities and the main achievements. (authors)

  16. Continuous fiber ceramic composite cladding for commercial water reactor fuel; Final Project

    SciTech Connect (OSTI)

    Herbert Feinroth

    2001-04-30T23:59:59.000Z

    This project is a research effort to develop and demonstrate the feasibility of an improved ceramics-based cladding material for water reactor fuel, which will be significantly more resistant to structural damage during a LOCA accident than the current Zircaloy cladding material. Specifically, the goal of this NERI project is to determine, via engineering type tests, the feasibility of substituting such advanced ceramic materials for the Zircaloy cladding now in use. This report presents the project research and development activities, which included prototype material design, fabrication, characterization, LOCA type of thermal shock testing, and in-reactor irradiation/corrosion testing. The report also presents the technical finding and discussions of results. The technical task were performed in collaboration with four subcontractors: The Advanced Materials Section of McDermott Technology Incorporated (MTI), the Nuclear Reactor Laboratory of Massachusetts Institute of Technology (MTI), Swales Aerospace Inc., and the Thin Film Laboratory of Northwestern University.

  17. Final report for the Light Water Breeder Reactor proof-of-breeding analytical support project

    SciTech Connect (OSTI)

    Graczyk, D.G.; Hoh, J.C.; Martino, F.J.; Nelson, R.E.; Osudar, J.; Levitz, N.M.

    1987-05-01T23:59:59.000Z

    The technology of breeding /sup 233/U from /sup 232/Th in a light water reactor is being developed and evaluated by the Westinghouse Bettis Atomic Power Laboratory (BAPL) through operation and examination of the Shippingport Light Water Breeder Reactor (LWBR). Bettis is determining the end-of-life (EOL) inventory of fissile uranium in the LWBR core by nondestructive assay of a statistical sample comprising approximately 500 EOL fuel rods. This determination is being made with an irradiated-fuel assay gauge based on neutron interrogation and detection of delayed neutrons from each rod. The EOL fissile inventory will be compared with the beginning-of-life fissile loading of the LWBR to determine the extent of breeding. In support of the BAPL proof-of-breeding (POB) effort, Argonne National Laboratory (ANL) carried out destructive physical, chemical, and radiometric analyses on 17 EOL LWBR fuel rods that were previously assayed with the nondestructive gauge. The ANL work included measurements on the intact rods; shearing of the rods into pre-designated contiguous segments; separate dissolution of each of the more than 150 segments; and analysis of the dissolver solutions to determine each segment's uranium content, uranium isotopic composition, and loading of selected fission products. This report describes the facilities in which this work was carried out, details operations involved in processing each rod, and presents a comprehensive discussion of uncertainties associated with each result of the ANL measurements. Most operations were carried out remotely in shielded cells. Automated equipment and procedures, controlled by a computer system, provided error-free data acquisition and processing, as well as full replication of operations with each rod. Despite difficulties that arose during processing of a few rod segments, the ANL destructive-assay results satisfied the demanding needs of the parent LWBR-POB program.

  18. 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

    2012-09-01T23:59:59.000Z

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

  19. 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.

  20. Control of structure and reactivity by ligand design : applications to small molecule activation by low-valent uranium complexes

    E-Print Network [OSTI]

    Lam, Oanh Phi

    2010-01-01T23:59:59.000Z

    coordination chemistry is depleted uranium, a by-product innuclear reactors. Depleted uranium Figure 1-1. The periodic

  1. Final report on LDRD project : biodiesel production from vegetable oils using slit-channel reactors.

    SciTech Connect (OSTI)

    Kalu, E. Eric (FAMU-FSU College of Engineering, Tallahassee, FL); Chen, Ken Shuang

    2008-01-01T23:59:59.000Z

    This report documents work done for a late-start LDRD project, which was carried out during the last quarter of FY07. The objective of this project was to experimentally explore the feasibility of converting vegetable (e.g., soybean) oils to biodiesel by employing slit-channel reactors and solid catalysts. We first designed and fabricated several slit-channel reactors with varying channel depths, and employed them to investigate the improved performance of slit-channel reactors over traditional batch reactors using a NaOH liquid catalyst. We then evaluated the effectiveness of several solid catalysts, including CaO, ZnO, MgO, ZrO{sub 2}, calcium gluconate, and heteropolyacid or HPA (Cs{sub 2.5}H{sub 0.5}PW{sub 12}O{sub 40}), for catalyzing the soybean oil-to-biodiesel transesterification reaction. We found that the slit-channel reactor performance improves as channel depth decreases, as expected; and the conversion efficiency of a slit-channel reactor is significantly higher when its channel is very shallow. We further confirmed CaO as having the highest catalytic activity among the solid catalysts tested, and we demonstrated for the first time calcium gluconate as a promising solid catalyst for converting soybean oil to biodiesel, based on our preliminary batch-mode conversion experiments.

  2. Thoriumbased fuel cycles : saving uranium in a 200 MWth pebble bed high temperature reactor / S.K. Gintner.

    E-Print Network [OSTI]

    Gintner, Stephan Konrad

    2010-01-01T23:59:59.000Z

    ??The predominant nuclear fuel used globally at present is uranium which is a finite resource. Thorium has been identified as an alternative nuclear fuel source (more)

  3. Design of a boiling water reactor equilibrium core using thorium-uranium fuel

    SciTech Connect (OSTI)

    Francois, J-L.; Nunez-Carrera, A.; Espinosa-Paredes, G.; Martin-del-Campo, C.

    2004-10-06T23:59:59.000Z

    In this paper the design of a Boiling Water Reactor (BWR) equilibrium core using thorium is presented; a heterogeneous blanket-seed core arrangement concept was adopted. The design was developed in three steps: in the first step two different assemblies were designed based on the integrated blanket-seed concept, they are the blanket-dummy assembly and the blanket-seed assembly. The integrated blanketseed concept comes from the fact that the blanket and the seed rods are located in the same assembly, and are burned-out in a once-through cycle. In the second step, a core design was developed to achieve an equilibrium cycle of 365 effective full power days in a standard BWR with a reload of 104 fuel assemblies designed with an average 235U enrichment of 7.5 w/o in the seed sub-lattice. The main operating parameters, like power, linear heat generation rate and void distributions were obtained as well as the shutdown margin. It was observed that the analyzed parameters behave like those obtained in a standard BWR. The shutdown margin design criterion was fulfilled by addition of a burnable poison region in the assembly. In the third step an in-house code was developed to evaluate the thorium equilibrium core under transient conditions. A stability analysis was also performed. Regarding the stability analysis, five operational states were analyzed; four of them define the traditional instability region corner of the power-flow map and the fifth one is the operational state for the full power condition. The frequency and the boiling length were calculated for each operational state. The frequency of the analyzed operational states was similar to that reported for BWRs; these are close to the unstable region that occurs due to the density wave oscillation phenomena in some nuclear power plants. Four transient analyses were also performed: manual SCRAM, recirculation pumps trip, main steam isolation valves closure and loss of feed water. The results of these transients are similar to those obtained with the traditional UO2 nuclear fuel.

  4. In Situ NDA Conformation Measurements Performed at Auxiliary Charcoal Bed and Other Main Charcoal Beds After Uranium Removal from Molten Salt Reactor Experiment ACB at Oak Ridge National Laboratory

    SciTech Connect (OSTI)

    Haghighi, M. H.; Kring, C. T.; McGehee, J. T.; Jugan, M. R.; Chapman, J.; Meyer, K. E.

    2002-02-26T23:59:59.000Z

    The Molten Salt Reactor Experiment (MSRE) site is located in Tennessee, on the U.S. Department of Energy (DOE) Oak Ridge Reservation (ORR). The MSRE was run by Oak Ridge National Laboratory (ORNL) to demonstrate the desirable features of the molten-salt concept in a practical reactor that could be operated safely and reliably. It introduced the idea of a homogeneous reactor using fuel salt media and graphite moderation for power and breeder reactors. The MSRE reactor and associated components are located in cells beneath the floor in the high-bay area of Building 7503. The reactor was operated from June 1965 to December 1969. When the reactor was shut down, fuel salt was drained from the reactor circuit to two drain tanks. A ''clean'' salt was then circulated through the reactor as a decontamination measure and drained to a third drain tank. When operations ceased, the fuel and flush salts were allowed to cool and solidify in the drain tanks. At shutdown, the MSRE facility complex was placed in a surveillance and maintenance program. Beginning in 1987, it was discovered that gaseous uranium (U-233/U-232) hexafluoride (UF6) had moved throughout the MSRE process systems. The UF6 had been generated when radiolysis in the fluorine salts caused the individual constituents to dissociate to their component atoms, including free fluorine. Some of the free fluorine combined with uranium fluorides (UF4) in the salt to produce UF6. UF6 is gaseous at slightly above ambient temperatures; thus, periodic heating of the fuel salts (which was intended to remedy the radiolysis problems) and simple diffusion had allowed the UF6 to move out of the salt and into the process systems of MSRE. One of the systems that UF6 migrated into due to this process was the offgas system which is vented to the MSRE main charcoal beds and MSRE auxiliary charcoal bed (ACB). Recently, the majority of the uranium laden-charcoal material residing within the ACB was safely and successfully removed using the uranium deposit removal system and equipment. After removal a series of NDA measurements was performed to determine the amount of uranium material remaining in the ACB, the amount of uranium material removed from the ACB, and the amount of uranium material remaining in the uranium removal equipment due to removal activities.

  5. Final programmatic environmental impact statement for the uranium mill tailings remedial action ground water project. Volume I

    SciTech Connect (OSTI)

    None

    1996-10-01T23:59:59.000Z

    This programmatic environmental impact statement (PElS) was prepared for the Uranium Mill Tailings Remedial Action (UMTRA) Ground Water Project to comply with the National Environmental Policy Act (NEPA). This PElS provides an analysis of the potential impacts of the alternatives and ground water compliance strategies as well as potential cumulative impacts. On November 8, 1978, Congress enacted the Uranium Mill Tailings Radiation Control Act (UMTRCA) of 1978, Public Law, codified at 42 USC 7901 et seq. Congress found that uranium mill tailings " ... may pose a potential and significant radiation health hazard to the public, and that every reasonable effort should be made to provide for stabilization, disposal, and control in a safe, and environmentally sound manner of such tailings in order to prevent or minimize other environmental hazards from such tailings." Congress authorized the Secretary of Energy to designate inactive uranium processing sites for remedial action by the U.S. Department of Energy (DOE). Congress also directed the U.S. Environmental Protection Agency (EPA) to set the standards to be followed by the DOE for this process of stabilization, disposal, and control. On January 5, 1983, EPA published standards (40 CFR Part 192) for the disposal and cleanup of residual radioactive materials. On September 3, 1985, the U.S. Court of Appeals for the Tenth Circuit set aside and remanded to EPA the ground water provisions of the standards. The EPA proposed new standards to replace remanded sections and changed other sections of 40 CFR Part 192. These proposed standards were published in the Federal Register on September 24, 1987 (52 FR 36000). Section 108 of the UMTRCA requires that DOE comply with EPA's proposed standards in the absence of final standards. The Ground Water Project was planned under the proposed standards. On January 11, 1995, EPA published the final rule, with which the DOE must now comply. The PElS and the Ground Water Project are in accordance with the final standards. The EPA reserves the right to modify the ground water standards, if necessary, based on changes in EPA drinking water standards. Appendix A contains a copy of the 1983 EPA ground water compliance standards, the 1987 proposed changes to the standards, and the 1995 final rule. Under UMTRA, DOE is responsible for bringing the designated processing sites into compliance with the EPA ground water standards and complying with all other applicable standards and requirements. The U.S. Nuclear Regulatory Commission (NRC) must concur with DOE's actions. States are full participants in the process. The DOE also must consult with any affected Indian tribes and the Bureau of Indian Affairs. Uranium processing activities at most of the inactive mill sites resulted in the contamination of ground water beneath and, in some cases, downgradient of the sites. This contaminated ground water often has elevated levels of constituents such as but not limited to uranium and nitrates. The purpose of the UMTRA Ground Water Project is to eliminate or reduce to acceptable levels the potential health and environmental consequences of milling activities by meeting the EPA ground water standards.

  6. Considerations Associated with Reactor Technology Selection for the Next Generation Nuclear Plant Project

    SciTech Connect (OSTI)

    L.E. Demick

    2010-09-01T23:59:59.000Z

    At the inception of the Next Generation Nuclear Plant Project and during predecessor activities, alternative reactor technologies have been evaluated to determine the technology that best fulfills the functional and performance requirements of the targeted energy applications and market. Unlike the case of electric power generation where the reactor performance is primarily expressed in terms of economics, the targeted energy applications involve industrial applications that have specific needs in terms of acceptable heat transport fluids and the associated thermodynamic conditions. Hence, to be of interest to these industrial energy applications, the alternative reactor technologies are weighed in terms of the reactor coolant/heat transport fluid, achievable reactor outlet temperature, and practicality of operations to achieve the very high reliability demands associated with the petrochemical, petroleum, metals and related industries. These evaluations have concluded that the high temperature gas-cooled reactor (HTGR) can uniquely provide the required ranges of energy needs for these target applications, do so with promising economics, and can be commercialized with reasonable development risk in the time frames of current industry interest i.e., within the next 10-15 years.

  7. Cleanup protocols when encountering thorium-230 at U.S. DOE Uranium Mill Tailings Remedial Action (UMTRA) Project sites

    SciTech Connect (OSTI)

    Miller, M.L.; Hylko, J.M.; Cornish, R.E.

    1995-12-31T23:59:59.000Z

    The passage of the Uranium Mill Tailings Radiation Control Act (UMTRCA) of 1978, established the regulatory framework, under which the US EPA charged with developing standards for the cleanup and disposal of tailings at 24 designated inactive uranium processing sites located in 10 states. 40 CFR 192.12 requires that the concentration of Ra-226 in land averaged over any area of 100 square meters shall not exceed the background level by more than 5 pCi/g, averaged over the first 15 cm of soil below the surface, 15 pCi/g, averaged over 15-cm-thick layers of soils more than 15 cm below the surface. However, Th-230 is not specifically addressed by the EPA in 40 CFR 192.12, which naturally decays with a half-life of 77,000 years to form Ra-226. Consequently, the cleanup of the initial Ra-226 contamination according to the standards will not necessarily mitigate against the eventual ingrowth of residual Ra-226 with time, due to the radioactive decay of residual Th-230. Therefore, to direct the excavation of residual Th-230, four generic protocols are being used at Uranium Mill Tailings Remedial Action (UMTRA) Project sites, as follows: Determining the allowable remaining concentration of Th-230 in surface and subsurface soils; Encountering Th-230 contamination in the unsaturated subsurface soil; Encountering Th-230 contamination in the saturated zone; and Verification sampling. The four generic protocols, developed in conjunction with the supplemental standards provision, ensure protection of the general public by reducing exposures to levels that are As Low As Reasonably Achievable, while considering practical measures necessary to excavate Th-230 under conditions encountered at the UMTRA Project site.

  8. Economic impact study of the Uranium Mill Tailings Remedial Action Project in Colorado: Colorado State fiscal year 1994. Revision 1

    SciTech Connect (OSTI)

    Not Available

    1994-12-01T23:59:59.000Z

    The Colorado economic impact study summarizes employment and economic benefits to the state from activities associated with the Uranium Mill Tailings Remedial Action (UMTRA) Project during Colorado state fiscal year 1994 (1 July 1993 through 30 June 1994). To capture employment information, a questionnaire was distributed to subcontractor employees at the active UMTRA Project sites of Grand Junction, Naturita, Gunnison, and Rifle, Colorado. Economic data were requested from each site prime subcontractor, as well as from the Remedial Action Contractor. Information on wages, taxes, and subcontract expenditures in combination with estimates and economic multipliers is used to estimate the dollar economic benefits to Colorado during the state fiscal year. Finally, the fiscal year 1994 estimates are compared to fiscal year 1993 employment and economic information.

  9. Uranium industry annual 1994

    SciTech Connect (OSTI)

    NONE

    1995-07-05T23:59:59.000Z

    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.

  10. Economic impact study of the Uranium Mill Tailings Remedial Action Project in Colorado: Colorado state fiscal year 1994

    SciTech Connect (OSTI)

    Not Available

    1994-11-01T23:59:59.000Z

    The Colorado economic impact study summarizes employment and economic benefits to the state from activities associated with the Uranium Mill Tailings Remedial Action (UMTRA) Project during Colorado state fiscal year 1994. To capture employment information, a questionnaire was distributed to subcontractor employees at the active UMTRA Project sites of Grand Junction, Naturita, Gunnison, and Rifle, Colorado. Economic data were requested from each site prime subcontractor, as well as from the Remedial Action Contractor. The most significant benefits associated with the UMTRA Project in Colorado are summarized. This study assesses benefits associated with the Grand Junction, Gunnison, Naturita, and Rifle UMTRA Projects sites for the 1-year period under study. Work at the Naturita site was initiated in April 1994 and involved demolition of buildings at the processing site. Actual start-up of remediation of Naturita is planned to begin in the spring of 1995. Work at the Slick Rock and Maybell sites is expected to begin in 1995. The only current economic benefits associated with these sites are related to UMTRA Project support work.

  11. Fuel channel analysis for a large-break loss-of-coolant accident in a Canada deuterium uranium reactor loaded with CANFLEX fuel bundles

    SciTech Connect (OSTI)

    Oh, D.J.; Lim, H.S.; Ohn, M.Y.; Lee, K.M.; Suk, H.C. [Korea Atomic Energy Research Inst., Taejon (Korea, Republic of)

    1996-06-01T23:59:59.000Z

    The CATHENA ``slave`` channel model is used for fuel channel analysis of a 30% reactor inlet header break in a Canada deuterium uranium (CANDU)-6 reactor loaded with 43-element bundles of advanced CANDU [CANDU flexible fueling (CANFLEX)] fuel. The predicted results are compared with those for the reactor loaded with standard 37-element bundles. The maximum fuel centerline and sheath temperatures for the CANFLEX bundle are lower by 388 and 128C, respectively, than those for the standard bundle because of the lower maximum linear power of the CANFLEX bundle. The pressure tube (PT)/calandria tube (CT) contact for the CANFLEX bundle occurs 2 s later than that for the standard bundle. The PT/CT contact temperature for the CANFLEX bundle is 7C lower than that for the standard bundle. These provide the CANFLEX bundle with a slightly enhanced safety margin for fuel channel integrity in the CANDU-6 reactor, compared with the standard bundle. The effect of bearing pad (BP)/PT contact on the PT temperature predictions is assessed. A BP/PT contact conductance of 3 kW/m{sup 2} {center_dot} K after PT ballooning does not create any hot spot because it gives the contacted PT sector approximately the same heat transfer as convective heating by the hot coolant for the adjacent sector. The assumed BP/PT contact conductance does not threaten the fuel channel integrity.

  12. Decommissioning Small Research and Training Reactors; Experience on Three Recent University Projects - 12455

    SciTech Connect (OSTI)

    Gilmore, Thomas [LVI Services Inc. (United States); DeWitt, Corey; Miller, Dustin; Colborn, Kurt [Enercon Services, Inc. (United States)

    2012-07-01T23:59:59.000Z

    Decommissioning small reactors within the confines of an active University environment presents unique challenges. These range from the radiological protection of the nearby University population and grounds, to the logistical challenges of working in limited space without benefit of the established controlled, protected, and vital areas common to commercial facilities. These challenges, and others, are discussed in brief project histories of three recent (calendar year 2011) decommissioning activities at three University training and research reactors. These facilities include three separate Universities in three states. The work at each of the facilities addresses multiple phases of the decommissioning process, from initial characterization and pre-decommissioning waste removal, to core component removal and safe storage, through to complete structural dismantlement and site release. The results of the efforts at each University are presented, along with the challenges that were either anticipated or discovered during the decommissioning efforts, and results and lessons learned from each of the projects. (authors)

  13. High Temperature Gas-cooled Reactor Projected Markets and Scoping Economics

    SciTech Connect (OSTI)

    Larry Demick

    2010-08-01T23:59:59.000Z

    The NGNP Project has the objective of developing the high temperature gas-cooled reactor (HTGR) technology to supply high temperature process heat to industrial processes as a substitute for burning of fossil fuels, such as natural gas. Applications of the HTGR technology that have been evaluated by the NGNP Project for supply of process heat include supply of electricity, steam and high-temperature gas to a wide range of industrial processes, and production of hydrogen and oxygen for use in petrochemical, refining, coal to liquid fuels, chemical, and fertilizer plants.

  14. 1992 North Dakota Economic Impact Study for the Uranium Mill Tailings Remedial Action Project, Belfield and Bowman, North Dakota

    SciTech Connect (OSTI)

    Not Available

    1993-01-01T23:59:59.000Z

    The goal of the Uranium Mill Tailings Remedial Action (UMTRA) Project in North Dakota is to improve the environment and reduce the negative health effects associated with residual radioactive material (RRM) from the inactive processing sites at Belfield and Bowman, North Dakota. A secondary benefit of the UMTRA Project is economic gain. The 1992 North Dakota Economic Impact Study (NDEIS) analyzes the impact of the remedial actions at the inactive Belfield and Bowman processing sites and their associated vicinity properties. This analysis is based on the assumption that the state of North Dakota will provide 10 percent of the funding required for remediation. For every dollar the state of North Dakota invests in the Belfield and Bowman onsite portion of the UMTRA Project, it will realize $5.04 in gross labor income (i.e., gross labor income divided by the state's total funding requirement). For every dollar the state of North Dakota invests, it will realize a net return of $3.04 (i.e., net benefit divided by the state's total funding requirement). This reflects only labor expenditure and employment impact. ff state and local non-labor tax benefits were considered in the net economic benefit, North Dakota could receive significantly more than $3.04 for each dollar it invests. The UMTRA Project work at Belfield and Bowman will benefit the state of North Dakota. Benefits include a reduction in the negative health effects caused by low-level RRM, an improvement in the environment, and increased economic growth.

  15. Economic impact study of the Uranium Mill Tailings Remedial Action project in Colorado: Colorado state fiscal year 1995. Revision 1

    SciTech Connect (OSTI)

    NONE

    1995-12-01T23:59:59.000Z

    As required by the Romer-Twining Agreement of 1990, the US Department of Energy (DOE) has prepared this annual economic impact study for the state of Colorado. This report assesses the economic impacts related to the DOE Uranium Mill Tailings Remedial Action (UMTRA) Project in Colorado during the state fiscal year (FY) between 1 July 1994 and 30 June 1995. To estimate net economic benefit, employment, salaries and wages, and other related economic benefits are discussed, quantified, and then compared to the state`s 10 percent share of the remedial action costs. Actual data obtained from sites currently undergoing remedial action were used as the basis for analyses. If data were not available, estimates were used to derive economic indicators. This study describes the types of employment associated with the UMTRA Project and estimates of the numbers of people employed by UMTRA Project subcontractors in Colorado during state FY 1995. Employment totals are reported in estimated average annual jobs; however, the actual number of workers at the site fluctuates depending on weather and on the status of remedial action activities. In addition, the actual number of people employed on the Project during the year may be higher than the average annual employment reported due to the temporary nature of some of the jobs.

  16. Fiver years of uranium mine and mill decommissioning in Germany: Progress of the Wismut environmental remediation project

    SciTech Connect (OSTI)

    Mager, D. [German Federal Ministry of Economics, Bonn (Germany)

    1996-12-31T23:59:59.000Z

    With the end of the Cold War and the fall of the iron curtain Germany inherited in 1990 by its reunification the legacy of 45 years of unrestricted, intensive uranium mining in the eastern part of Germany. The environmental damages and risks related to widespread soil and groundwater contamination, caused by huge tailings ponds and numerous waste rock piles in one of Germany`s most densely populated areas, made it necessary to implement one of the world`s largest environmental clean-up programs. 13 billion German Marks (8.7 billion US-$) will be spent within approx. 20 years for decommissioning and environmental restoration efforts. Five years after the start-up of the Wismut project considerable results have been achieved. Conceptual work, even regarding technically difficult issues of the program, is far advanced, and about one third of the physical work program has been completed. At the former mining and milling locations significant improvements of the environmental situation have been achieved. The further on-schedule progress of the Wismut program is an important prerequisite for the socio-economical development in the former uranium mining districts.

  17. 1991 New Mexico economic impact study for the Uranium Mill Tailings Remedial Action Project, Ambrosia Lake, New Mexico, site

    SciTech Connect (OSTI)

    Not Available

    1991-06-01T23:59:59.000Z

    The University of New Mexico Bureau of Business and Economic Research completed an abbreviated cost-benefit analysis of the income and employment impact of the US Department of Energy (DOE) and contractor offices in Albuquerque. Since the Project Office will have a significant positive impact on the State`s economy (shown on Table 8), the impact is combined with the impact of remedial actions at the Ambrosia Lake site to highlight the cost-benefit of the entire Uranium Mill Tailings Remedial Action (UMTRA) Project. The UMTRA Project at the Ambrosia Lake site will generate $12.509 million in gross labor income in New Mexico between 1989 and 1994. This includes $1.161 million in federal tax revenue, $1.015 million in State personal income tax revenue, and seven thousand in local tax revenue. The UMTRA Project will generate the equivalent of 84 full-time jobs during the peak year of remedial action at Ambrosia Lake site. New Mexico`s total funding requirement for the UMTRA Project is estimated to be $2.963 million. The net economic benefit of the Ambrosia Lake portion of the UMTRA Project to New Mexico after the State`s share of the project`s cost, the federal income tax, and the $0.936 million income impact of the alternate use of the State funding are subtracted, will be $7.451 million between 1990 and 1994. In Fiscal Year 1990 the UMTRA Project DOE and contractor offices in Albuquerque directly employed 163 people. Another 78 jobs were also maintained in support of the industry sector and 166 jobs were also maintained in other sections of the New Mexico economy. It is estimated that $19 million dollars of income was generated and 1.949 million of State and local taxes were collected. The University of New Mexico study shows that for every dollar the State of New Mexico invests in the UMTRA Project, it will realize $95.05 in gross labor income. This corresponds to a net return on the States investment in the Project of $97.20 for every dollar invested.

  18. Colorado economic impact study on the Uranium Mill Tailings Remedial Action Project in Colorado: Colorado state fiscal year 1993

    SciTech Connect (OSTI)

    Not Available

    1993-11-12T23:59:59.000Z

    The Colorado economic impact study summarizes employment and economic benefits to the state from activities associated with the Uranium Mill Tailings Remedial Action (UMTRA) Project during Colorado state fiscal year (FY) 1993. To capture employment benefits, a questionnaire was distributed to subcontractor employees at the active UMTRA Project sites of Grand Junction, Rifle, and Gunnison, Colorado. An estimated 52 percent of the employees working on the UMTRA Project responded to this information request. Economic data were requested from each prime subcontractor, as well as from the Remedial Action Contractor. The most significant benefits associated with the UMTRA Project in Colorado are: Direct employment was estimated at 894 workers; An estimated 89 percent of all direct employment was local; Secondary employment resulting from remedial action at the active Colorado UMTRA Project sites and the Grand Junction vicinity property program is estimated at 546 workers. Total employment (direct and secondary) is estimated at 1440 workers for the period of study (July 1, 1992, to June 30, 1993). An estimated $24.1 million was paid in wages to UMTRA workers in Colorado during FY1993; Direct and secondary wage earnings were estimated at $39.9 million; Income tax payments to the state of Colorado were estimated at $843,400 during FY1993; The gross economic impact of UMTRA Project activities in the state of Colorado is estimated at $70 million during the 1-year study period; and the net economic benefit to the state of Colorado was estimated at $57.5 million, or $5.90 per dollar of funding provided by Colorado. This figure includes both direct and secondary benefits but does not include the impact of alternative uses of the state funding.

  19. Uranium and cesium diffusion in fuel cladding of electrogenerating channel

    SciTech Connect (OSTI)

    Vasilev, I. V., E-mail: fnti@mail.ru; Ivanov, A. S.; Churin, V. A. [National Research Center Kurchatov Institute (Russian Federation)

    2014-12-15T23:59:59.000Z

    The results of reactor tests of a carbonitride fuel in a single-crystal cladding from a molybdenum-based alloy can be used in substantiating the operational reliability of fuels in developing a project of a megawatt space nuclear power plant. The results of experimental studies of uranium and cesium penetration into the single-crystal cladding of fuel elements with a carbonitride fuel are interpreted. Those fuel elements passed nuclear power tests in the Ya-82 pilot plant for 8300 h at a temperature of about 1500C. It is shown that the diffusion coefficients for uranium diffusion into the cladding are virtually coincident with the diffusion coefficients measured earlier for uranium diffusion into polycrystalline molybdenum. It is found that the penetration of uranium into the cladding is likely to occur only in the case of a direct contact between the cladding and fuel. The experimentally observed nonmonotonic uranium-concentration profiles are explained in terms of predominant uranium diffusion along grain boundaries. It is shown that a substantially nonmonotonic behavior observed in our experiment for the uranium-concentration profile may be explained by the presence of a polycrystalline structure of the cladding in the surface region from its inner side. The diffusion coefficient is estimated for the grain-boundary diffusion of uranium. The diffusion coefficients for cesium are estimated on the basis of experimental data obtained in the present study.

  20. US Department of Energy Uranium Mill Tailings Remedial Action ground water Project. Revision 1, Version 1: Final project plan

    SciTech Connect (OSTI)

    Not Available

    1993-12-21T23:59:59.000Z

    The scope of the Project is to develop and implement a ground water compliance strategy for all 24 UMTRA processing sites. The compliance strategy for the processing sites must satisfy requirements of the proposed EPA ground water cleanup standards in 40 CFR Part 192, Subparts B and C (1988). This scope of work will entail the following activities, on a site-specific basis: Development of a compliance strategy based upon modification of the UMTRA Surface Project remedial action plans (RAP) or development of Ground Water Project RAPs with NRC and state or tribal concurrence on the RAP; implementation of the RAP to include establishment of institutional controls, where appropriate; institution of long-term verification monitoring for transfer to a separate DOE program on or before the Project end date; and preparation of completion reports and final licensing on those sites that will be completed prior to the Project end date.

  1. Environmental assessment of remedial action at the Gunnison Uranium Mill Tailings Site, Gunnison, Colorado. [UMTRA Project

    SciTech Connect (OSTI)

    Bachrach, A.; Hoopes, J.; Morycz, D. (Jacobs Engineering Group, Inc., Pasadena, CA (USA)); Bone, M.; Cox, S.; Jones, D.; Lechel, D.; Meyer, C.; Nelson, M.; Peel, R.; Portillo, R.; Rogers, L.; Taber, B.; Zelle, P. (Weston (Roy F.), Inc., Washington, DC (USA)); Rice, G. (Sergent, Hauskins and Beckwith (USA))

    1984-12-01T23:59:59.000Z

    This document assesses and compares the environmental impacts of various alternatives for remedial action at the Gunnison uranium of mill tailings site located 0.5 miles south of Gunnison, Colorado. The site covers 56 acres and contains 35 acres of tailings, 2 of the original mill buildings and a water tower. The Uranium Mill Tailings Radiation Control of Act of 1978 (UMTRCA), Public Law 95-604, authorizes the US Department of Energy to clean up the site to reduce the potential health impacts associated with the residual radioactive materials remaining at the site and at associated (vicinity) properties off the site. The US Environmental Protection Agency promulgated standards for the remedial actions (40 CFR 192). Remedial actions must be performed in accordance with these standards and with the occurrence of the Nuclear Regulatory Commission. Four alternatives have been addressed in this document. The first alternative is to consolidate the tailings and associated contaminated soils into a recontoured pile on the southern portion of the existing site. A radon barrier of silty clay would be constructed over the pile and various erosion control measures would be taken to assure the long-term integrity of the pile. Two other alternatives which involve moving the tailings to new locations are assessed in this document. These alternatives generally involve greater short-term impacts and are more costly but would result in the tailings being stabilized in a location farther from the city of Gunnison. The no action alternative is also assessed.

  2. Analysis of infiltration through a clay radon barrier at an UMTRA disposal cell. Uranium Mill Tailings Remedial Action Project

    SciTech Connect (OSTI)

    Not Available

    1991-01-01T23:59:59.000Z

    An infiltration study was initiated in January 1988 to assess the percent saturation in, and infiltration through, clay radon barriers of typical Uranium Mill Tailings Remedial Action (UMTRA) Project disposal cells. Predicting infiltration through the radon barrier is necessary to evaluate whether the disposal cell will comply with the proposed US Environmental Protection Agency (EPA) groundwater protection standards (40 CFR 192). The groundwater standards require demonstrating that tailings seepage will not cause background concentrations or maximum concentration limits (MCLs) to be exceeded at the downgradient edge of the disposal facility (the point of compliance, or POC). This demonstration generally consists of incorporating the predicted seepage flux and the concentration of the specific hazardous constituents into a contaminant transport model, and predicting the resultant concentrations at the POC. The infiltration study consisted of a field investigation to evaluate moisture conditions in the radon barrier of the completed Shiprock, New Mexico, UMTRA Project disposal cell and previously completed UMTRA Project disposal cells at Clive, Utah, and Burrell, Pennsylvania. Coring was conducted to measure percent saturation profiles in the radon barriers at these disposal cells. In addition, a detailed investigation of the Shiprock radon barrier was conducted to establish the effects of meteorological stresses on moisture conditions in the filter layer and radon barrier. The Shiprock infiltration study was also intended to characterize hydraulic gradients and operational unsaturated hydraulic conductivities in the radon barrier.

  3. GROWTH OF THE INTERNATIONAL CRITICALITY SAFETY AND REACTOR PHYSICS EXPERIMENT EVALUATION PROJECTS

    SciTech Connect (OSTI)

    J. Blair Briggs; John D. Bess; Jim Gulliford

    2011-09-01T23:59:59.000Z

    Since the International Conference on Nuclear Criticality Safety (ICNC) 2007, the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) have continued to expand their efforts and broaden their scope. Eighteen countries participated on the ICSBEP in 2007. Now, there are 20, with recent contributions from Sweden and Argentina. The IRPhEP has also expanded from eight contributing countries in 2007 to 16 in 2011. Since ICNC 2007, the contents of the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments1' have increased from 442 evaluations (38000 pages), containing benchmark specifications for 3955 critical or subcritical configurations to 516 evaluations (nearly 55000 pages), containing benchmark specifications for 4405 critical or subcritical configurations in the 2010 Edition of the ICSBEP Handbook. The contents of the Handbook have also increased from 21 to 24 criticality-alarm-placement/shielding configurations with multiple dose points for each, and from 20 to 200 configurations categorized as fundamental physics measurements relevant to criticality safety applications. Approximately 25 new evaluations and 150 additional configurations are expected to be added to the 2011 edition of the Handbook. Since ICNC 2007, the contents of the 'International Handbook of Evaluated Reactor Physics Benchmark Experiments2' have increased from 16 different experimental series that were performed at 12 different reactor facilities to 53 experimental series that were performed at 30 different reactor facilities in the 2011 edition of the Handbook. Considerable effort has also been made to improve the functionality of the searchable database, DICE (Database for the International Criticality Benchmark Evaluation Project) and verify the accuracy of the data contained therein. DICE will be discussed in separate papers at ICNC 2011. The status of the ICSBEP and the IRPhEP will be discussed in the full paper, selected benchmarks that have been added to the ICSBEP Handbook will be highlighted, and a preview of the new benchmarks that will appear in the September 2011 edition of the Handbook will be provided. Accomplishments of the IRPhEP will also be highlighted and the future of both projects will be discussed. REFERENCES (1) International Handbook of Evaluated Criticality Safety Benchmark Experiments, NEA/NSC/DOC(95)03/I-IX, Organisation for Economic Co-operation and Development-Nuclear Energy Agency (OECD-NEA), September 2010 Edition, ISBN 978-92-64-99140-8. (2) International Handbook of Evaluated Reactor Physics Benchmark Experiments, NEA/NSC/DOC(2006)1, Organisation for Economic Co-operation and Development-Nuclear Energy Agency (OECD-NEA), March 2011 Edition, ISBN 978-92-64-99141-5.

  4. The Advanced Neutron Source (ANS) project: A world-class research reactor facility

    SciTech Connect (OSTI)

    Thompson, P.B. [Martin Marietta Energy Systems, Inc., Oak Ridge, TN (US); Meek, W.E. [Gilbert/Commonwealth, Inc., Pittsburgh, PA (US)

    1993-07-01T23:59:59.000Z

    This paper provides an overview of the Advanced Neutron Source (ANS), a new research facility being designed at Oak Ridge National Laboratory. The facility is based on a 330 MW, heavy-water cooled and reflected reactor as the neutron source, with a thermal neutron flux of about 7.5{times}10{sup 19}m{sup {minus}2}{center_dot}sec{sup {minus}1}. Within the reflector region will be one hot source which will serve 2 hot neutron beam tubes, two cryogenic cold sources serving fourteen cold neutron beam tubes, two very cold beam tubes, and seven thermal neutron beam tubes. In addition there will be ten positions for materials irradiation experiments, five of them instrumented. The paper touches on the project status, safety concerns, cost estimates and scheduling, a description of the site, the reactor, and the arrangements of the facilities.

  5. Decontamination and decommissioning of the Experimental Boiling Water Reactor (EBWR): Project final report, Argonne National Laboratory

    SciTech Connect (OSTI)

    Fellhauer, C.R.; Boing, L.E. [Argonne National Lab., IL (United States); Aldana, J. [NES, Inc., Danbury, CT (United States)

    1997-03-01T23:59:59.000Z

    The Final Report for the Decontamination and Decommissioning (D&D) of the Argonne National Laboratory - East (ANL-E) Experimental Boiling Water Reactor (EBWR) facility contains the descriptions and evaluations of the activities and the results of the EBWR D&D project. It provides the following information: (1) An overall description of the ANL-E site and EBWR facility. (2) The history of the EBWR facility. (3) A description of the D&D activities conducted during the EBWR project. (4) A summary of the final status of the facility, including the final and confirmation surveys. (5) A summary of the final cost, schedule, and personnel exposure associated with the project, including a summary of the total waste generated. This project report covers the entire EBWR D&D project, from the initiation of Phase I activities to final project closeout. After the confirmation survey, the EBWR facility was released as a {open_quotes}Radiologically Controlled Area,{close_quotes} noting residual elevated activity remains in inaccessible areas. However, exposure levels in accessible areas are at background levels. Personnel working in accessible areas do not need Radiation Work Permits, radiation monitors, or other radiological controls. Planned use for the containment structure is as an interim transuranic waste storage facility (after conversion).

  6. Crosscutting Requirements in the International Project on Innovative Reactors and Fuel Cycles (INPRO)

    SciTech Connect (OSTI)

    Steur, Ronald; Lyubenov Yaven, Yanko; Gueorguiev, Boris; Mahadeva, Rao; Shen, Wenquan [International Atomic Energy Agency - IAEA, P.O. Box 100, Wagramer Strasse 5, A-1400 Vienna (Austria)

    2002-07-01T23:59:59.000Z

    There are two categories of requirements: (i) user requirements that need to be met by the designers and manufacturers of innovative reactors and fuel cycles, and (ii) a wide spectrum of requirements that need to be met by countries, willing to successfully deploy innovative nuclear reactors for energy production. This part of the International Project on Innovative Reactors and Fuel Cycles will mainly deal with the second category of requirements. Both categories of requirements will vary depending on the institutional development, infrastructure availability and social attitude in any given country. Out of the need for sustainable development requirements will also more specific in the future. Over a 50-year time frame both categories of requirements will evolve with social and economic development as nuclear technology develops further. For example, the deployment of innovative reactors in countries with marginal or non-existing nuclear infrastructures would be possible only if the reactors are built, owned and operated by an international nuclear utility or if they are inherently safe and can be delivered as a 'black box - nuclear battery'. A number of issues will need to be addressed and conditions and requirements developed if this is going to become a reality. One general requirement for wider utilization of innovative nuclear power will be the public and environmental considerations, which will play a role in the decision making processes. Five main clusters of topics will be handled: - Infra-structural aspects, typology and consequences for nuclear development. - Industrial requirements for the different innovative concepts. - Institutional developments and requirements for future deployment of nuclear energy. (National as well as international) - Socio-political aspects, a.o. public acceptance and role of governments. - Sustainability: requirements following the need for sustainability Analysis will be made of the evolution of national and international social, institutional and infrastructure requirements for the deployment of innovative nuclear technology through 2050 and beyond and requirements will be identified following the need for (authors)

  7. India's Worsening Uranium Shortage

    SciTech Connect (OSTI)

    Curtis, Michael M.

    2007-01-15T23:59:59.000Z

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

  8. Safety Design Strategy for the Advanced Test Reactor Diesel Bus (E-3) and Switchgear Replacement Project

    SciTech Connect (OSTI)

    Noel Duckwitz

    2011-06-01T23:59:59.000Z

    In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3B, Program and Project Management for the Acquisition of Capital Assets, safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3B and DOE Order 420.1B, Facility Safety, and the expectations of DOE-STD-1189-2008, Integration of Safety into the Design Process, provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Advanced Test Reactor Reliability Sustainment Project. While this project does not introduce new hazards to the ATR, it has the potential for significant impacts to safety-related systems, structures, and components that are credited in the ATR safety basis and are being replaced. Thus the project has been determined to meet the definition of a major modification and is being managed accordingly.

  9. Safety Design Strategy for the Advanced Test Reactor Emergency Firewater Injection System Replacement Project

    SciTech Connect (OSTI)

    Noel Duckwitz

    2011-06-01T23:59:59.000Z

    In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3B, Program and Project Management for the Acquisition of Capital Assets, safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3B and DOE Order 420.1B, Facility Safety, and the expectations of DOE-STD-1189-2008, Integration of Safety into the Design Process, provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Advanced Test Reactor Reliability Sustainment Project. While this project does not introduce new hazards to the ATR, it has the potential for significant impacts to safety-related systems, structures, and components that are credited in the ATR safety basis and are being replaced. Thus the project has been determined to meet the definition of a major modification and is being managed accordingly.

  10. Safety Design Strategy for the Advanced Test Reactor Primary Coolant Pump and Motor Replacement Project

    SciTech Connect (OSTI)

    Noel Duckwitz

    2011-06-01T23:59:59.000Z

    In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3B, Program and Project Management for the Acquisition of Capital Assets, safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3B and DOE Order 420.1B, Facility Safety, and the expectations of DOE-STD-1189-2008, Integration of Safety into the Design Process, provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Advanced Test Reactor Reliability Sustainment Project. While this project does not introduce new hazards to the ATR, it has the potential for significant impacts to safety-related systems, structures, and components that are credited in the ATR safety basis and are being replaced. Thus the project has been determined to meet the definition of a major modification and is being managed accordingly.

  11. Simulation of Low-Enriched Uranium (LEU) Burnup in Russian VVER Reactors with the HELIOS Code Package

    SciTech Connect (OSTI)

    Murphy, B.D.; Kravchenko, J.; Lazarenko, A.; Pavlovitchev, A.; Sidorenko, V.; Chetverikov, A.

    2000-03-01T23:59:59.000Z

    The HELIOS reactor-physics computer program system was used to simulate the burnup of UO{sub 2} fuel in three VVER reactors. The manner in which HELIOS was used in these simulations is described. Predictions of concentrations for actinides up to {sup 244}Cm and for isotopes of neodymium were compared with laboratory-measured values. Reasonable agreement between calculated and measured values was seen for experimental samples from a fuel rod in the interior of an assembly.

  12. IAEA coordinated research project on thermal-hydraulics of Supercritical Water-Cooled Reactors (SCWRs)

    SciTech Connect (OSTI)

    Yamada, K. [Vienna International Centre, P.O. Box 100, 1400 Vienna (Austria); Aksan, S. N. [International Atomic Energy Agency, 1400 Vienna (Austria)

    2012-07-01T23:59:59.000Z

    The Supercritical Water-Cooled Reactor (SCWR) is an innovative water-cooled reactor concept, which uses supercritical pressure water as reactor coolant. It has been attracting interest of many researchers in various countries mainly due to its benefits of high thermal efficiency and simple primary systems, resulting in low capital cost. The IAEA started in 2008 a Coordinated Research Project (CRP) on Thermal-Hydraulics of SCWRs as a forum to foster the exchange of technical information and international collaboration in research and development. This paper summarizes the activities and current status of the CRP, as well as major progress achieved to date. At present, 15 institutions closely collaborate in several tasks. Some organizations have been conducting thermal-hydraulics experiments and analysing the data, and others have been participating in code-to-test and/or code-to-code benchmark exercises. The expected outputs of the CRP are also discussed. Finally, the paper introduces several IAEA activities relating to or arising from the CRP. (authors)

  13. Remedial Action Plan and site design for stabilization of the inactive uranium mill tailings site at Falls City, Texas. [Uranium Mill Tailings Remedial Action (UMTRA) Project

    SciTech Connect (OSTI)

    Chernoff, A.R. (USDOE Albuquerque Field Office, NM (United States). Uranium Mill Tailings Remedial Action Project Office); Lacker, D.K. (Texas State Dept. of Health, Austin, TX (United States). Bureau of Radiation Control)

    1992-09-01T23:59:59.000Z

    The uranium processing site near Falls City, Texas, was one of 24 inactive uranium mill sites designated to be remediated by the US Department of Energy (DOE) under Title I of the Uranium Mill Tailings Radiation Control Act of 1978 (UMTRCA). The UMTRCA requires that the US Nuclear Regulatory Commission (NRC) concur with the DOE's remedial action plan (RAP) and certify that the remedial action conducted at the site complies with the standards promulgated by the US Environmental Protection Agency (EPA). The RAP, which includes this summary remedial action selection report (RAS), serves a two-fold purpose. First, it describes the activities proposed by the DOE to accomplish long-term stabilization and control of the residual radioactive materials at the inactive uranium processing site near Falls City, Texas. Second, this document and the remainder of the RAP, upon concurrence and execution by the DOE, the State of Texas, and the NRC, becomes Appendix B of the Cooperative Agreement between the DOE and the State of Texas.

  14. Remedial action plan and site design for stabilization of the inactive uranium mill tailings site at Tuba City, Arizona. [Uranium Mill Tailings Remedial Action (UMTRA) Project

    SciTech Connect (OSTI)

    Not Available

    1987-05-01T23:59:59.000Z

    This appendix assesses the present conditions and data for the inactive uranium mill site near Tuba City, Arizona. It consolidates available engineering, radiological, geotechnical, hydrological, meterological, and other information pertinent to the design of the Remedial Action Plan (RAP). The data characterize conditions at the mill and tailings site so that the Remedial Action Contractor (RAC) may complete final designs of the remedial actions.

  15. 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...

  16. Prototype Tests for the Recovery and Conversion of UF6Chemisorbed in NaF Traps for the Molten Salt Reactor Remediation Project

    SciTech Connect (OSTI)

    Del Cul, G.D.

    2000-06-07T23:59:59.000Z

    The remediation of the Molten Salt Reactor Experiment (MSRE) site includes the removal of about 37 kg of uranium. Of that inventory, about 23 kg have already been removed from the piping system and chemisorbed in 25 NaF traps. This material is being stored in Building 3019. The planned recovery of {approx}11 kg of uranium from the fuel salt will generate another 15 to 19 NaF traps. The remaining 2 to 3 kg of uranium are present in activated charcoal beds, which are also scheduled to be removed from the reactor site. Since all of these materials (NaF traps and the uranium-laden charcoal) are not suitable for long-term storage, they will be converted to a chemical form [uranium oxide (U{sub 3}O{sub 8})], which is suitable for long-term storage. This document describes the process that will be used to recover and convert the uranium in the NaF traps into a stable oxide for long-term storage. Included are a description of the process, equipment, test results, and lessons learned. The process was developed for remote operation in a hot cell. Lessons learned from the prototype testing were incorporated into the process design.

  17. Prototype Tests for the Recovery and Conversion of UF6 Chemisorbed in NaF Traps for the Molten Salt Reactor Remediation Project

    SciTech Connect (OSTI)

    Del Cul, G.D.; Icenhour, A.S.; Simmons, D.W.

    2000-04-01T23:59:59.000Z

    The remediation of the Molten Salt Reactor Experiment (MSRE) site includes the removal of about 37 kg of uranium. Of that inventory, about 23 kg have already been removed from the piping system and chemisorbed in 25 NaF traps. This material is being stored in Building 3019. The planned recovery of -11 kg of uranium from the fuel salt will generate another 15 to 19 NaF traps. The remaining 2 to 3 kg of uranium are present in activated charcoal beds, which are also scheduled to be removed from the reactor site. Since all of these materials (NaF traps and the uranium-laden charcoal) are not suitable for long-term storage, they will be converted to a chemical form [uranium oxide], which is suitable for long-term storage. This document describes the process that will be used to recover and convert the uranium in the NaF traps into a stable oxide for long-term storage. Included are a description of the process, equipment, test results, and lessons learned. The process was developed for remote operation in a hot cell. Lessons learned from the prototype testing were incorporated into the process design.

  18. Reactor Physics and Criticality Benchmark Evaluations for Advanced Nuclear Fuel - Final Technical Report

    SciTech Connect (OSTI)

    William Anderson; James Tulenko; Bradley Rearden; Gary Harms

    2008-09-11T23:59:59.000Z

    The nuclear industry interest in advanced fuel and reactor design often drives towards fuel with uranium enrichments greater than 5 wt% 235U. Unfortunately, little data exists, in the form of reactor physics and criticality benchmarks, for uranium enrichments ranging between 5 and 10 wt% 235U. The primary purpose of this project is to provide benchmarks for fuel similar to what may be required for advanced light water reactors (LWRs). These experiments will ultimately provide additional information for application to the criticality-safety bases for commercial fuel facilities handling greater than 5 wt% 235U fuel.

  19. Plutonium Consumption Program, CANDU Reactor Project: Feasibility of BNFP Site as MOX Fuel Supply Facility. Final report

    SciTech Connect (OSTI)

    NONE

    1995-06-30T23:59:59.000Z

    An evaluation was made of the technical feasibility, cost, and schedule for converting the existing unused Barnwell Nuclear Fuel Facility (BNFP) into a Mixed Oxide (MOX) CANDU fuel fabrication plant for disposition of excess weapons plutonium. This MOX fuel would be transported to Ontario where it would generate electricity in the Bruce CANDU reactors. Because CANDU MOX fuel operates at lower thermal load than natural uranium fuel, the MOX program can be licensed by AECB within 4.5 years, and actual Pu disposition in the Bruce reactors can begin in 2001. Ontario Hydro will have to be involved in the entire program. Cost is compared between BNFP and FMEF at Hanford for converting to a CANDU MOX facility.

  20. Uranium dioxide electrolysis

    DOE Patents [OSTI]

    Willit, James L. (Batavia, IL); Ackerman, John P. (Prescott, AZ); Williamson, Mark A. (Naperville, IL)

    2009-12-29T23:59:59.000Z

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

  1. PROJECT-SPECIFIC TYPE A VERIFICATION FOR THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PHASE 3 TRENCH 1, BROOKHAVEN NATIONAL LABORATORY UPTON, NEW YORK

    SciTech Connect (OSTI)

    E.M. Harpenau

    2010-12-15T23:59:59.000Z

    5098-SR-05-0 PROJECT-SPECIFIC TYPE A VERIFICATION FOR THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PHASE 3 TRENCH 1 BROOKHAVEN NATIONAL LABORATORY

  2. PROJECT-SPECIFIC TYPE A VERIFICATION FOR THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PHASE 3 TRENCH 5, BROOKHAVEN NATIONAL LABORATORY UPTON, NEW YORK

    SciTech Connect (OSTI)

    P.C. Weaver

    2010-11-03T23:59:59.000Z

    5098-SR-04-0 PROJECT-SPECIFIC TYPE A VERIFICATION FOR THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PHASE 3 TRENCH 5, BROOKHAVEN NATIONAL LABORATORY

  3. Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

    SciTech Connect (OSTI)

    John D. Bess; Thomas L. Maddock; Margaret A. Marshall; Leland M. Montierth

    2013-03-01T23:59:59.000Z

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has been evaluated as an acceptable benchmark experiment. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has also been evaluated as an acceptable benchmark experiment. Calculated eigenvalues differ significantly (~1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  4. Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

    SciTech Connect (OSTI)

    John D. Bess; Thomas L. Maddock; Margaret A. Marshall; Leland M. Montierth

    2011-03-01T23:59:59.000Z

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has been evaluated as an acceptable benchmark experiment. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has not been evaluated as it is very similar to the evaluated core configuration. The benchmark eigenvalue is 1.0012 0.0029. Calculated eigenvalues differ significantly (~1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  5. Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

    SciTech Connect (OSTI)

    John D. Bess; Thomas L. Maddock; Margaret A. Marshall; Leland M. Montierth

    2014-03-01T23:59:59.000Z

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has been evaluated as an acceptable benchmark experiment. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has not been evaluated as it is very similar to the evaluated core configuration. The benchmark eigenvalue is 1.0012 0.0029. Calculated eigenvalues differ significantly (~1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

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

    SciTech Connect (OSTI)

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

    1986-06-01T23:59:59.000Z

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

  7. Uranium Mill Tailings Management

    SciTech Connect (OSTI)

    Nelson, J.D.

    1982-01-01T23:59:59.000Z

    This book presents the papers given at the Fifth Symposium on Uranium Mill Tailings Management. Advances made with regard to uranium mill tailings management, environmental effects, regulations, and reclamation are reviewed. Topics considered include tailings management and design (e.g., the Uranium Mill Tailings Remedial Action Project, environmental standards for uranium mill tailings disposal), surface stabilization (e.g., the long-term stability of tailings, long-term rock durability), radiological aspects (e.g. the radioactive composition of airborne particulates), contaminant migration (e.g., chemical transport beneath a uranium mill tailings pile, the interaction of acidic leachate with soils), radon control and covers (e.g., radon emanation characteristics, designing surface covers for inactive uranium mill tailings), and seepage and liners (e.g., hydrologic observations, liner requirements).

  8. Crystal Chemistry of Early Actinides (Thorium, Uranium, and Neptunium) and Uranium Mesoporous Materials.

    E-Print Network [OSTI]

    Sigmon, Ginger E.

    2010-01-01T23:59:59.000Z

    ??Despite their considerable global importance, the structural chemistry of actinides remains understudied. Thorium and uranium fuel cycles are used in commercial nuclear reactors in India (more)

  9. On-Site Oxy-Lance Size Reduction of South Texas Project Reactor Vessel Heads - 12324

    SciTech Connect (OSTI)

    Posivak, Edward [WMG, inc. (United States); Keeney, Gilbert; Wheeler, Dean [Shaw Group (United States)

    2012-07-01T23:59:59.000Z

    On-Site Oxy-Lance size reduction of mildly radioactive large components has been accomplished at other operating plants. On-Site Oxy-Lance size reduction of more radioactive components like Reactor Vessel Heads had previously been limited to decommissioning projects. Building on past decommissioning and site experience, subcontractors for South Texas Project Nuclear Operating Company (STPNOC) developed an innovative integrated system to control smoke, radioactive contamination, worker dose, and worker safety. STP's innovative, easy to use CEDM containment that provided oxy lance access, smoke control, and spatter/contamination control was the key to successful segmentation for cost-effective and ALARA packaging and transport for disposal. Relative to CEDM milling, STP oxy-lance segmentation saved approximately 40 person- REM accrued during 9,000 hours logged into the radiological controlled area (RCA) during more than 3,800 separate entries. Furthermore there were no personnel contamination events or respiratory uptakes of radioactive material during the course of the entire project. (authors)

  10. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2013

    SciTech Connect (OSTI)

    David W. Nigg

    2013-09-01T23:59:59.000Z

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance, and to some extent, experiment management, are inconsistent with the state of modern nuclear engineering practice, and are difficult, if not impossible, to verify and validate (V&V) according to modern standards. Furthermore, the legacy staff knowledge required for effective application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In late 2009, the Idaho National Laboratory (INL) initiated a focused effort, the ATR Core Modeling Update Project, to address this situation through the introduction of modern high-fidelity computational software and protocols. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF).

  11. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2010

    SciTech Connect (OSTI)

    Rahmat Aryaeinejad; Douglas S. Crawford; Mark D. DeHart; George W. Griffith; D. Scott Lucas; Joseph W. Nielsen; David W. Nigg; James R. Parry; Jorge Navarro

    2010-09-01T23:59:59.000Z

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance and, to some extent, experiment management are obsolete, inconsistent with the state of modern nuclear engineering practice, and are becoming increasingly difficult to properly verify and validate (V&V). Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In 2009 the Idaho National Laboratory (INL) initiated a focused effort to address this situation through the introduction of modern high-fidelity computational software and protocols, with appropriate V&V, within the next 3-4 years via the ATR Core Modeling and Simulation and V&V Update (or Core Modeling Update) Project. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF).

  12. High Temperature Gas-Cooled Reactor Projected Markets and Preliminary Economics

    SciTech Connect (OSTI)

    Larry Demick

    2011-08-01T23:59:59.000Z

    This paper summarizes the potential market for process heat produced by a high temperature gas-cooled reactor (HTGR), the environmental benefits reduced CO2 emissions will have on these markets, and the typical economics of projects using these applications. It gives examples of HTGR technological applications to industrial processes in the typical co-generation supply of process heat and electricity, the conversion of coal to transportation fuels and chemical process feedstock, and the production of ammonia as a feedstock for the production of ammonia derivatives, including fertilizer. It also demonstrates how uncertainties in capital costs and financial factors affect the economics of HTGR technology by analyzing the use of HTGR technology in the application of HTGR and high temperature steam electrolysis processes to produce hydrogen.

  13. United States Department of Energy projects related to reactor pressure vessel annealing optimization

    SciTech Connect (OSTI)

    Rosinski, S.T.; Nakos, J.T.

    1993-09-01T23:59:59.000Z

    Light water reactor pressure vessel (RPV) material properties reduced by long-term exposure to neutron irradiation can be recovered through a thermal annealing treatment. This technique to extend RPV life, discussed in this report, provides a complementary approach to analytical methodologies to evaluate RPV integrity. RPV annealing has been successfully demonstrated in the former Soviet Union and on a limited basis by the US (military applications only). The process of demonstrating the technical feasibility of annealing commercial US RPVs is being pursued through a cooperative effort between the nuclear industry and the US Department of Energy (USDOE) Plant Lifetime Improvement (PLIM) Program. Presently, two projects are under way through the USDOE PLIM Program to demonstrate the technical feasibility of annealing commercial US RPVS, (1) annealing re-embrittlement data base development and (2) heat transfer boundary condition experiments.

  14. Plutonium partitioning in uranium and plutonium co-recovery system for fast reactor fuel recycling with enhanced nuclear proliferation resistance

    SciTech Connect (OSTI)

    Nakahara, Masaumi; Koma, Yoshikazu; Nakajima, Yasuo [Japan Atomic Energy Agency: 4-33 Muramatsu, Naka-gun, Tokai-mura, Ibaraki 319-1194 (Japan)

    2013-07-01T23:59:59.000Z

    For enhancement of nuclear proliferation resistance, a 'co-processing' method for U and Pu co-recovery was studied. Two concepts, no U scrubbing and no Pu reduction partitioning, were employed to formulate two types of flow sheets by using a calculation code. Their process performance was demonstrated using radioactive solutions derived from an irradiated fast reactor fuel. These experimental results indicated that U and Pu were co-recovered in the U/Pu product, and the Pu content in the U/Pu product increased approximately 2.3 times regardless of using reductant. The proposed no U scrubbing and no Pu reductant flow sheet is applicable to fast reactor fuel reprocessing and enhances its resistance to nuclear proliferation. (authors)

  15. A strategy for transition from a uranium fueled, open cycle SFR to a transuranic fueled, closed cycle sodium cooled fast reactor

    E-Print Network [OSTI]

    Richard, Joshua (Joshua Glenn)

    2012-01-01T23:59:59.000Z

    Reactors utilizing a highly energetic neutron spectrum, often termed fast reactors, offer large fuel utilization improvements over the thermal reactors currently used for nuclear energy generation. Conventional fast reactor ...

  16. 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

    2012-07-31T23:59:59.000Z

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

  17. Uranium Ore Uranium is extracted

    E-Print Network [OSTI]

    Milling of Uranium Ore Uranium is extracted from ore with strong acids or bases. The uranium is concentrated in a solid substance called"yellowcake." Chemical Conversion Plants convert the uranium in yellowcake to uranium hexafluoride (UF6 ), a compound that can be made into nuclear fuel. Enrichment

  18. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2011

    SciTech Connect (OSTI)

    David W. Nigg; Devin A. Steuhm

    2011-09-01T23:59:59.000Z

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance and, to some extent, experiment management are obsolete, inconsistent with the state of modern nuclear engineering practice, and are becoming increasingly difficult to properly verify and validate (V&V). Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In 2009 the Idaho National Laboratory (INL) initiated a focused effort to address this situation through the introduction of modern high-fidelity computational software and protocols, with appropriate V&V, within the next 3-4 years via the ATR Core Modeling and Simulation and V&V Update (or 'Core Modeling Update') Project. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF). The ATR Core Modeling Update Project, targeted for full implementation in phase with the anticipated ATR Core Internals Changeout (CIC) in the 2014 time frame, began during the last quarter of Fiscal Year 2009, and has just completed its first full year. Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (SCALE, KENO-6, HELIOS, NEWT, and ATTILA) have been installed at the INL under various permanent sitewide license agreements and corresponding baseline models of the ATR and ATRC are now operational, demonstrating the basic feasibility of these code packages for their intended purpose. Furthermore, a capability for rigorous sensitivity analysis and uncertainty quantification based on the TSUNAMI system is being implemented and initial computational results have been obtained. This capability will have many applications in 2011 and beyond as a tool for understanding the margins of uncertainty in the new models as well as for validation experiment design and interpretation. Finally we note that although full implementation of the new computational models and protocols will extend over a period 3-4 years as noted above, interim applications in the much nearer term have already been demonstrated. In particular, these demonstrations included an analysis that was useful for understanding the cause of some issues in December 2009 that were triggered by a larger than acceptable discrepancy between the measured excess core reactivity and a calculated value that was based on the legacy computational methods. As the Modeling Update project proceeds we anticipate further such interim, informal, applications in parallel with formal qualification of the system under the applicable INL Quality Assurance procedures and standards.

  19. Fabrication and Characterization of Uranium-based High Temperature...

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

    Fabrication and Characterization of Uranium-based High Temperature Reactor Fuel June 01, 2013 The Uranium Fuel Development Laboratory is a modern R&D scale lab for the fabrication...

  20. Final environmental assessment for the U.S. Department of Energy, Oak Ridge Operations receipt and storage of uranium materials from the Fernald Environmental Management Project site

    SciTech Connect (OSTI)

    NONE

    1999-06-01T23:59:59.000Z

    Through a series of material transfers and sales agreements over the past 6 to 8 years, the Fernald Environmental Management Project (FEMP) has reduced its nuclear material inventory from 14,500 to approximately 6,800 metric tons of uranium (MTU). This effort is part of the US Department of energy`s (DOE`s) decision to change the mission of the FEMP site; it is currently shut down and the site is being remediated. This EA focuses on the receipt and storage of uranium materials at various DOE-ORO sites. The packaging and transportation of FEMP uranium material has been evaluated in previous NEPA and other environmental evaluations. A summary of these evaluation efforts is included as Appendix A. The material would be packaged in US Department of Transportation-approved shipping containers and removed from the FEMP site and transported to another site for storage. The Ohio Field Office will assume responsibility for environmental analyses and documentation for packaging and transport of the material as part of the remediation of the site, and ORO is preparing this EA for receipt and storage at one or more sites.

  1. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2012

    SciTech Connect (OSTI)

    David W. Nigg, Principal Investigator; Kevin A. Steuhm, Project Manager

    2012-09-01T23:59:59.000Z

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance, and to some extent, experiment management, are inconsistent with the state of modern nuclear engineering practice, and are difficult, if not impossible, to properly verify and validate (V&V) according to modern standards. Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In late 2009, the Idaho National Laboratory (INL) initiated a focused effort, the ATR Core Modeling Update Project, to address this situation through the introduction of modern high-fidelity computational software and protocols. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF). The ATR Core Modeling Update Project, targeted for full implementation in phase with the next anticipated ATR Core Internals Changeout (CIC) in the 2014-2015 time frame, began during the last quarter of Fiscal Year 2009, and has just completed its third full year. Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (HELIOS, KENO6/SCALE, NEWT/SCALE, ATTILA, and an extended implementation of MCNP5) has been installed at the INL under various licensing arrangements. Corresponding models of the ATR and ATRC are now operational with all five codes, demonstrating the basic feasibility of the new code packages for their intended purpose. Of particular importance, a set of as-run core depletion HELIOS calculations for all ATR cycles since August 2009, Cycle 145A through Cycle 151B, was successfully completed during 2012. This major effort supported a decision late in the year to proceed with the phased incorporation of the HELIOS methodology into the ATR Core Safety Analysis Package (CSAP) preparation process, in parallel with the established PDQ-based methodology, beginning late in Fiscal Year 2012. Acquisition of the advanced SERPENT (VTT-Finland) and MC21 (DOE-NR) Monte Carlo stochastic neutronics simulation codes was also initiated during the year and some initial applications of SERPENT to ATRC experiment analysis were demonstrated. These two new codes will offer significant additional capability, including the possibility of full-3D Monte Carlo fuel management support capabilities for the ATR at some point in the future. Finally, a capability for rigorous sensitivity analysis and uncertainty quantification based on the TSUNAMI system has been implemented and initial computational results have been obtained. This capability will have many applications as a tool for understanding the margins of uncertainty in the new models as well as for validation experiment design and interpretation.

  2. 1992 Colorado Economic Impact Study for the US Department of Energy and Colorado Department of Health Uranium Mill Tailings Remedial Action (UMTRA) Project

    SciTech Connect (OSTI)

    Not Available

    1991-10-22T23:59:59.000Z

    The findings of the 1992 Colorado Economic Impact Study (CEIS) for the Uranium Mill Tailings Remedial Action (UMTRA) Project are outlined below. All dollar amounts used in the study are in year-of-expenditure dollars. The total funding requirement for the State of Colorado for the UMTRA Project is estimated to be $66.8 million, or 10 percent of the remedial action costs for the UMTRA Project in Colorado. The UMTRA Project will generate $487.5 million in gross labor income in Colorado between 1983 and 1996. This includes $54.4 million in state and local tax revenues and $41.2 million in federal individual income tax revenues. The net economic benefit of the UMTRA Project to Colorado is $355.1 million. For every dollar the State of Colorado invests in the UMTRA Project, it will realize $5.32 in gross labor income. The employment impact to the Western Slope region is significant. The UMTRA Project will create a total employment impact of 13,749 fulltime equivalents (FTES) spread over. a period of 13 years in seven site areas. Nearly 100 percent of the labor will be drawn from the local communities. The State of Colorado's Western Slope is anticipated to be minimally impacted by the phaseout of the UMTRA Project. Unlike industries that shut down operations without warning, the UMTRA Project workers, local government, and businesses know the schedule for completion and can consider and prepare for the impact of UMTRA Project conclusion. Further, because the majority of the work force is local, there has not been a significant investment in each community's infrastructure. Any small increases in the infrastructure will not be abandoned at the end of the UMTRA Project due to a marked increase in migration out of the local community.

  3. 1992 Colorado Economic Impact Study for the US Department of Energy and Colorado Department of Health Uranium Mill Tailings Remedial Action (UMTRA) Project. Preliminary final

    SciTech Connect (OSTI)

    Not Available

    1991-10-22T23:59:59.000Z

    The findings of the 1992 Colorado Economic Impact Study (CEIS) for the Uranium Mill Tailings Remedial Action (UMTRA) Project are outlined below. All dollar amounts used in the study are in year-of-expenditure dollars. The total funding requirement for the State of Colorado for the UMTRA Project is estimated to be $66.8 million, or 10 percent of the remedial action costs for the UMTRA Project in Colorado. The UMTRA Project will generate $487.5 million in gross labor income in Colorado between 1983 and 1996. This includes $54.4 million in state and local tax revenues and $41.2 million in federal individual income tax revenues. The net economic benefit of the UMTRA Project to Colorado is $355.1 million. For every dollar the State of Colorado invests in the UMTRA Project, it will realize $5.32 in gross labor income. The employment impact to the Western Slope region is significant. The UMTRA Project will create a total employment impact of 13,749 fulltime equivalents (FTES) spread over. a period of 13 years in seven site areas. Nearly 100 percent of the labor will be drawn from the local communities. The State of Colorado`s Western Slope is anticipated to be minimally impacted by the phaseout of the UMTRA Project. Unlike industries that shut down operations without warning, the UMTRA Project workers, local government, and businesses know the schedule for completion and can consider and prepare for the impact of UMTRA Project conclusion. Further, because the majority of the work force is local, there has not been a significant investment in each community`s infrastructure. Any small increases in the infrastructure will not be abandoned at the end of the UMTRA Project due to a marked increase in migration out of the local community.

  4. International Atomic Energy Agency (IAEA) Coordinated Research Projects on Structural Integrity of Reactor Pressure Vessels

    SciTech Connect (OSTI)

    Server, W. L. [ATI Consulting, Pinehurst, NC; Nanstad, Randy K [ORNL

    2009-01-01T23:59:59.000Z

    The International Atomic Energy Agency (IAEA) has conducted a series of Coordinated Research Projects (CRPs) that have focused on irradiated reactor pressure vessel (RPV) steel fracture toughness properties and approaches for assuring structural integrity of RPVs throughout operating life. A series of nine CRPs have been sponsored by the IAEA, starting in the early 1970s, focused on neutron radiation effects on RPV steels. The purpose of the CRPs was to develop comparisons and correlations to test the uniformity of irradiated results through coordinated international research studies and data sharing. Consideration of dose rate effects, effects of alloying (nickel, manganese, silicon, etc.) and residual elements (eg., copper and phosphorus), and drop in upper shelf toughness are also important for assessing neutron embrittlement effects. The ultimate use of embrittlement understanding is assuring structural integrity of the RPV under current and future operation and accident conditions. Material fracture toughness is the key ingredient needed for this assessment, and many of the CRPs have focused on measurement and application of irradiated fracture toughness. This paper presents an overview of the progress made since the inception of the CRPs in the early 1970s. The chronology and importance of each CRP have been reviewed and put into context for continued and long-term safe operation of RPVs.

  5. Heterogeneous Recycling in Fast Reactors

    SciTech Connect (OSTI)

    Dr. Benoit Forget; Michael Pope; Piet, Steven J.; Michael Driscoll

    2012-07-30T23:59:59.000Z

    Current sodium fast reactor (SFR) designs have avoided the use of depleted uranium blankets over concerns of creating weapons grade plutonium. While reducing proliferation risks, this restrains the reactor design space considerably. This project will analyze various blanket and transmutation target configurations that could broaden the design space while still addressing the non-proliferation issues. The blanket designs will be assessed based on the transmutation efficiency of key minor actinide (MA) isotopes and also on mitigation of associated proliferation risks. This study will also evaluate SFR core performance under different scenarios in which depleted uranium blankets are modified to include minor actinides with or without moderators (e.g. BeO, MgO, B4C, and hydrides). This will be done in an effort to increase the sustainability of the reactor and increase its power density while still offering a proliferation resistant design with the capability of burning MA waste produced from light water reactors (LWRs). Researchers will also analyze the use of recycled (as opposed to depleted) uranium in the blankets. The various designs will compare MA transmutation efficiency, plutonium breeding characteristics, proliferation risk, shutdown margins and reactivity coefficients with a current reference sodium fast reactor design employing homogeneous recycling. The team will also evaluate the out-of-core accumulation and/or burn-down rates of MAs and plutonium isotopes on a cycle-by-cycle basis. This cycle-by-cycle information will be produced in a format readily usable by the fuel cycle systems analysis code, VISION, for assessment of the sustainability of the deployment scenarios.

  6. Flowsheets and source terms for radioactive waste projections

    SciTech Connect (OSTI)

    Forsberg, C.W. (comp.)

    1985-03-01T23:59:59.000Z

    Flowsheets and source terms used to generate radioactive waste projections in the Integrated Data Base (IDB) Program are given. Volumes of each waste type generated per unit product throughput have been determined for the following facilities: uranium mining, UF/sub 6/ conversion, uranium enrichment, fuel fabrication, boiling-water reactors (BWRs), pressurized-water reactors (PWRs), and fuel reprocessing. Source terms for DOE/defense wastes have been developed. Expected wastes from typical decommissioning operations for each facility type have been determined. All wastes are also characterized by isotopic composition at time of generation and by general chemical composition. 70 references, 21 figures, 53 tables.

  7. Next Generation Nuclear Plant Project Technology Development Roadmaps: The Technical Path Forward for 750800C Reactor Outlet Temperature

    SciTech Connect (OSTI)

    John Collins

    2009-08-01T23:59:59.000Z

    This document presents the NGNP Critical PASSCs and defines their technical maturation path through Technology Development Roadmaps (TDRMs) and their associated Technology Readiness Levels (TRLs). As the critical PASSCs advance through increasing levels of technical maturity, project risk is reduced and the likelihood of within-budget and on-schedule completion is enhanced. The current supplier-generated TRLs and TDRMs for a 750800C reactor outlet temperature (ROT) specific to each supplier are collected in Appendix A.

  8. Light Water Reactor Sustainability Constellation Pilot Project FY13 Summary Report

    SciTech Connect (OSTI)

    R. Johansen

    2013-09-01T23:59:59.000Z

    Summary report for Light Water Reactor Sustainability (LWRS) activities related to the R. E. Ginna and Nine Mile Point Unit 1 for FY13.

  9. Light Water Reactor Sustainability Constellation Pilot Project FY12 Summary Report

    SciTech Connect (OSTI)

    R. Johansen

    2012-09-01T23:59:59.000Z

    Summary report for Light Water Reactor Sustainability (LWRS) activities related to the R. E. Ginna and Nine Mile Point Unit 1 for FY12.

  10. Stability of uranium incorporated into Fe(hydr)oxides under fluctuating redox conditions

    E-Print Network [OSTI]

    Stewart, B.D.

    2009-01-01T23:59:59.000Z

    uranium deposit, Northern Australia - Lessons from the Alligator Rivers analogue project. Physics and Chemistry

  11. Growth and Expansion of the International Criticality Safety Benchmark Evaluation Project and the Newly Organized International Reactor Physics Experiment Evaluation Project

    SciTech Connect (OSTI)

    J. Blair Briggs; Lori Scott; Yolanda Rugama; Enrico Satori

    2007-05-01T23:59:59.000Z

    Since ICNC 2003, the International Criticality Safety Benchmark Evaluation Project (ICSBEP) has continued to expand its efforts and broaden its scope. Criticality-alarm / shielding type benchmarks and fundamental physics measurements that are relevant to criticality safety applications are not only included in the scope of the project, but benchmark data are also included in the latest version of the handbook. A considerable number of improvements have been made to the searchable database, DICE and the criticality-alarm / shielding benchmarks and fundamental physics measurements have been included in the database. There were 12 countries participating on the ICSBEP in 2003. That number has increased to 18 with recent contributions of data and/or resources from Brazil, Czech Republic, Poland, India, Canada, and China. South Africa, Germany, Argentina, and Australia have been invited to participate. Since ICNC 2003, the contents of the International Handbook of Evaluated Criticality Safety Benchmark Experiments have increased from 350 evaluations (28,000 pages) containing benchmark specifications for 3070 critical or subcritical configurations to 442 evaluations (over 38,000 pages) containing benchmark specifications for 3957 critical or subcritical configurations, 23 criticality-alarm-placement / shielding configurations with multiple dose points for each, and 20 configurations that have been categorized as fundamental physics measurements that are relevant to criticality safety applications in the 2006 Edition of the ICSBEP Handbook. Approximately 30 new evaluations and 250 additional configurations are expected to be added to the 2007 Edition of the Handbook. Since ICNC 2003, a reactor physics counterpart to the ICSBEP, The International Reactor Physics Experiment Evaluation Project (IRPhEP) was initiated. Beginning in 1999, the IRPhEP was conducted as a pilot activity by the by the Organization of Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) Nuclear Science Committee (NSC). The project was endorsed as an official activity of the NSC in June of 2003. The IRPhEP is patterned after its predecessor, the ICSBEP, but focuses on other integral measurements such as buckling, spectral characteristics, reactivity effects, reactivity coefficients, kinetics measurements, reaction-rate and power distributions, nuclide compositions and other miscellaneous types of measurements in addition to the critical configuration. The two projects are closely coordinated to avoid duplication of effort and to leverage limited resources to achieve a common goal. The purpose of the IRPhEP is to provide an extensively peer reviewed set of reactor physics related integral benchmark data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next generation reactors and establish the safety basis for operation of these reactors. While coordination and administration of the IRPhEP takes place at an international level, each participating country is responsible for the administration, technical direction, and priorities of the project within their respective countries. The work of the IRPhEP is documented in an OECD NEA Handbook entitled, International Handbook of Evaluated Reactor Physics Benchmark Experiments. The first edition of this Handbook, the 2006 Edition spans over 2000 pages and contains data from 16 different experimental series that were

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

    SciTech Connect (OSTI)

    Marsh, Terence L.

    2013-07-30T23:59:59.000Z

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

  13. URANIUM MILL TAILINGS RADON FLUX CALCULATIONS

    E-Print Network [OSTI]

    URANIUM MILL TAILINGS RADON FLUX CALCULATIONS PI?ON RIDGE PROJECT MONTROSE COUNTY, COLORADO Inc. (Golder) was commissioned by EFRC to evaluate the operations of the uranium mill tailings storage in this report were conducted using the WISE Uranium Mill Tailings Radon Flux Calculator, as updated on November

  14. Development of Nuclear Reactor remote Monitoring software (NRM) for the Star project

    E-Print Network [OSTI]

    Gautier, Vincent Charles

    2002-01-01T23:59:59.000Z

    As a response to the needs of developing countries to meet their rapidly growing energy requirements, the Safe, Transportable, Autonomous Reactor (STAR) program originated. This concept relies on small, passively safe, and highly autonomous nuclear...

  15. Report of the ANS Project Feasibility Workshop for a High Flux Isotope Reactor-Center for Neutron Research Facility

    SciTech Connect (OSTI)

    Peretz, F.J.; Booth, R.S. [comp.

    1995-07-01T23:59:59.000Z

    The Advanced Neutron Source (ANS) Conceptual Design Report (CDR) and its subsequent updates provided definitive design, cost, and schedule estimates for the entire ANS Project. A recent update to this estimate of the total project cost for this facility was $2.9 billion, as specified in the FY 1996 Congressional data sheet, reflecting a line-item start in FY 1995. In December 1994, ANS management decided to prepare a significantly lower-cost option for a research facility based on ANS which could be considered during FY 1997 budget deliberations if DOE or Congressional planners wished. A cost reduction for ANS of about $1 billion was desired for this new option. It was decided that such a cost reduction could be achieved only by a significant reduction in the ANS research scope and by maximum, cost-effective use of existing High Flux Isotope Reactor (HFIR) and ORNL facilities to minimize the need for new buildings. However, two central missions of the ANS -- neutron scattering research and isotope production-were to be retained. The title selected for this new option was High Flux Isotope Reactor-Center for Neutron Research (HFIR-CNR) because of the project`s maximum use of existing HFIR facilities and retention of selected, central ANS missions. Assuming this shared-facility requirement would necessitate construction work near HFIR, it was specified that HFIR-CNR construction should not disrupt normal operation of HFIR. Additional objectives of the study were that it be highly credible and that any material that might be needed for US Department of Energy (DOE) and Congressional deliberations be produced quickly using minimum project resources. This requirement made it necessary to rely heavily on the ANS design, cost, and schedule baselines. A workshop methodology was selected because assessment of each cost and/or scope-reduction idea required nearly continuous communication among project personnel to ensure that all ramifications of propsed changes.

  16. Preliminary results of calculations for heavy-water nuclear-power-plant reactors employing {sup 235}U, {sup 233}U, and {sup 232}Th as a fuel and meeting requirements of a nonproliferation of nuclear weapons

    SciTech Connect (OSTI)

    Ioffe, B. L.; Kochurov, B. P. [Institute of Theoretical and Experimental Physics (Russian Federation)

    2012-02-15T23:59:59.000Z

    A physical design is developed for a gas-cooled heavy-water nuclear reactor intended for a project of a nuclear power plant. As a fuel, the reactor would employ thorium with a small admixture of enriched uranium that contains not more than 20% of {sup 235}U. It operates in the open-cycle mode involving {sup 233}U production from thorium and its subsequent burnup. The reactor meets the conditions of a nonproliferation of nuclear weapons: the content of fissionable isotopes in uranium at all stages of the process, including the final one, is below the threshold for constructing an atomic bomb, the amount of product plutonium being extremely small.

  17. Uranium deposits of Brazil

    SciTech Connect (OSTI)

    NONE

    1991-09-01T23:59:59.000Z

    Brazil is a country of vast natural resources, including numerous uranium deposits. In support of the country`s nuclear power program, Brazil has developed the most active uranium industry in South America. Brazil has one operating reactor (Angra 1, a 626-MWe PWR), and two under construction. The country`s economic challenges have slowed the progress of its nuclear program. At present, the Pocos de Caldas district is the only active uranium production. In 1990, the Cercado open-pit mine produced approximately 45 metric tons (MT) U{sub 3}O{sub 8} (100 thousand pounds). Brazil`s state-owned uranium production and processing company, Uranio do Brasil, announced it has decided to begin shifting its production from the high-cost and nearly depleted deposits at Pocos de Caldas, to lower-cost reserves at Lagoa Real. Production at Lagoa Real is schedules to begin by 1993. In addition to these two districts, Brazil has many other known uranium deposits, and as a whole, it is estimated that Brazil has over 275,000 MT U{sub 3}O{sub 8} (600 million pounds U{sub 3}O{sub 8}) in reserves.

  18. Uranium hexafluoride handling. Proceedings

    SciTech Connect (OSTI)

    Not Available

    1991-12-31T23:59:59.000Z

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

  19. Remedial action plan and site design for stabilization of the inactive uranium mill tailings site at Tuba City, Arizona: Phase 2, Construction, Subcontract documents: Appendix E, final report. [Uranium Mill Tailings Remedial Action (UMTRA) Project

    SciTech Connect (OSTI)

    Not Available

    1989-08-01T23:59:59.000Z

    This appendix discusses Phase II construction and subcontract documents uranium mill site near Tuba City, Arizona. It contains the bid schedule, special conditions, specifications, and subcontract drawings.

  20. The Multi-Scale Mass Transfer Processes Controlling Natural Attenuation and Engineered Remediation: An IFC Focused on Hanfords 300 Area Uranium Plume Quality Assurance Project Plan

    SciTech Connect (OSTI)

    Fix, N. J.

    2008-01-31T23:59:59.000Z

    The purpose of the project is to conduct research at an Integrated Field-Scale Research Challenge Site in the Hanford Site 300 Area, CERCLA OU 300-FF-5 (Figure 1), to investigate multi-scale mass transfer processes associated with a subsurface uranium plume impacting both the vadose zone and groundwater. The project will investigate a series of science questions posed for research related to the effect of spatial heterogeneities, the importance of scale, coupled interactions between biogeochemical, hydrologic, and mass transfer processes, and measurements/approaches needed to characterize a mass-transfer dominated system. The research will be conducted by evaluating three (3) different hypotheses focused on multi-scale mass transfer processes in the vadose zone and groundwater, their influence on field-scale U(VI) biogeochemistry and transport, and their implications to natural systems and remediation. The project also includes goals to 1) provide relevant materials and field experimental opportunities for other ERSD researchers and 2) generate a lasting, accessible, and high-quality field experimental database that can be used by the scientific community for testing and validation of new conceptual and numerical models of subsurface reactive transport.

  1. 10 CFR 830 Major Modification Determination for Advanced Test Reactor LEU Fuel Conversion

    SciTech Connect (OSTI)

    Boyd D. Christensen; Michael A. Lehto; Noel R. Duckwitz

    2012-05-01T23:59:59.000Z

    The Advanced Test Reactor (ATR), located in the ATR Complex of the Idaho National Laboratory (INL), was constructed in the 1960s for the purpose of irradiating reactor fuels and materials. Other irradiation services, such as radioisotope production, are also performed at ATR. The ATR is fueled with high-enriched uranium (HEU) matrix (UAlx) in an aluminum sandwich plate cladding. The National Nuclear Security Administration Global Threat Reduction Initiative (GTRI) strategic mission includes efforts to reduce and protect vulnerable nuclear and radiological material at civilian sites around the world. Converting research reactors from using HEU to low-enriched uranium (LEU) was originally started in 1978 as the Reduced Enrichment for Research and Test Reactors (RERTR) Program under the U.S. Department of Energy (DOE) Office of Science. Within this strategic mission, GTRI has three goals that provide a comprehensive approach to achieving this mission: The first goal, the driver for the modification that is the subject of this determination, is to convert research reactors from using HEU to LEU. Thus the mission of the ATR LEU Fuel Conversion Project is to convert the ATR and Advanced Test Reactor Critical facility (ATRC) (two of the six U.S. High-Performance Research Reactors [HPRR]) to LEU fuel by 2017. The major modification criteria evaluation of the project pre-conceptual design identified several issues that lead to the conclusion that the project is a major modification.

  2. 1992 North Dakota Economic Impact Study for the Uranium Mill Tailings Remedial Action Project, Belfield and Bowman, North Dakota. Preliminary final report

    SciTech Connect (OSTI)

    Not Available

    1993-01-01T23:59:59.000Z

    The goal of the Uranium Mill Tailings Remedial Action (UMTRA) Project in North Dakota is to improve the environment and reduce the negative health effects associated with residual radioactive material (RRM) from the inactive processing sites at Belfield and Bowman, North Dakota. A secondary benefit of the UMTRA Project is economic gain. The 1992 North Dakota Economic Impact Study (NDEIS) analyzes the impact of the remedial actions at the inactive Belfield and Bowman processing sites and their associated vicinity properties. This analysis is based on the assumption that the state of North Dakota will provide 10 percent of the funding required for remediation. For every dollar the state of North Dakota invests in the Belfield and Bowman onsite portion of the UMTRA Project, it will realize $5.04 in gross labor income (i.e., gross labor income divided by the state`s total funding requirement). For every dollar the state of North Dakota invests, it will realize a net return of $3.04 (i.e., net benefit divided by the state`s total funding requirement). This reflects only labor expenditure and employment impact. ff state and local non-labor tax benefits were considered in the net economic benefit, North Dakota could receive significantly more than $3.04 for each dollar it invests. The UMTRA Project work at Belfield and Bowman will benefit the state of North Dakota. Benefits include a reduction in the negative health effects caused by low-level RRM, an improvement in the environment, and increased economic growth.

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

    SciTech Connect (OSTI)

    Alekseev, N. I.; Kalugin, M. A.; Kulakov, A. S.; Novoseltsev, A. P.; Sergeev, G. S.; Shkarovskiy, D. A.; Yudkevich, M. S., E-mail: umark@adis.vver.kiae.ru [National Research Center Kurchatov Institute (Russian Federation)

    2014-12-15T23:59:59.000Z

    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.

  4. Uranium Mill Tailings Remedial Action Project Environmental Line Management Audit Action Plan. Final report. Audit, October 26, 1992--November 6, 1992

    SciTech Connect (OSTI)

    NONE

    1993-07-01T23:59:59.000Z

    This Action Plan contains responses, planned actions, and estimated costs for addressing the findings discovered in the Environmental Management Audit conducted for the U.S. Department of Energy (DOE) Uranium Mill Tailings Remedial Action Project (UMTRA), October 26 through November 6, 1992. This document should be read in conjunction with the Audit Report to ensure the findings addressed in this document are fully understood. The scope of the UMTRA Environmental Management Audit was comprehensive and encompassed all areas of environmental management except environmental programs pertaining to the National Environmental Policy Act (NEPA) compliance. The Audit Report listed 18 findings: 11 were identified as compliance findings, and the remaining 7 were best management practice findings. Root cause analysis was performed on all the findings. The results of the analysis as well as planned corrective actions are summarized in Section 5.0. All planned actions were prioritized using the Tiger Team Assessment Corrective Action Plan system. Based on assigned priorities, all planned actions were costed by fiscal year. This Action Plan contains a description of the organizational and management structures to be used to implement the Action Plan, a brief discussion of root cause analysis and funding, followed by the responses and planned actions for each finding. A member of the UMTRA Project Office (PO) has been assigned responsibility for tracking the progress on each of the findings. The UMTRA PO staff wrote and/or approved all of the corrective actions recorded in this Action Plan.

  5. Idaho National Engineering and Environmental Laboratory Site Report on the Production and Use of Recycled Uranium

    SciTech Connect (OSTI)

    L. C. Lewis; D. C. Barg; C. L. Bendixsen; J. P. Henscheid; D. R. Wenzel; B. L. Denning

    2000-09-01T23:59:59.000Z

    Recent allegations regarding radiation exposure to radionuclides present in recycled uranium sent to the gaseous diffusion plants prompted the Department of Energy to undertake a system-wide study of recycled uranium. Of particular interest, were the flowpaths from site to site operations and facilities in which exposure to plutonium, neptunium and technetium could occur, and to the workers that could receive a significant radiation dose from handling recycled uranium. The Idaho National Engineering and Environmental Laboratory site report is primarily concerned with two locations. Recycled uranium was produced at the Idaho Chemical Processing Plant where highly enriched uranium was recovered from spent fuel. The other facility is the Specific Manufacturing Facility (SMC) where recycled, depleted uranium is manufactured into shapes for use by their customer. The SMC is a manufacturing facility that uses depleted uranium metal as a raw material that is then rolled and cut into shapes. There are no chemical processes that might concentrate any of the radioactive contaminant species. Recyclable depleted uranium from the SMC facility is sent to a private metallurgical facility for recasting. Analyses on the recast billets indicate that there is no change in the concentrations of transuranics as a result of the recasting process. The Idaho Chemical Processing Plant was built to recover high-enriched uranium from spent nuclear fuel from test reactors. The facility processed diverse types of fuel which required uniquely different fuel dissolution processes. The dissolved fuel was passed through three cycles of solvent extraction which resulted in a concentrated uranyl nitrate product. For the first half of the operating period, the uranium was shipped as the concentrated solution. For the second half of the operating period the uranium solution was thermally converted to granular, uranium trioxide solids. The dose reconstruction project has evaluated work exposure and exposure to the public as the result of normal operations and accidents that occurred at the INEEL. As a result of these studies, the maximum effective dose equivalent from site activities did not exceed seventeen percent of the natural background in Eastern Idaho. There was no year in which the radiation dose to the public exceeded the applicable limits for that year. Worker exposure to recycled uranium was minimized by engineering features that reduced the possibility of direct exposure.

  6. Nuclear power fleets and uranium resources recovered from phosphates

    SciTech Connect (OSTI)

    Gabriel, S.; Baschwitz, A.; Mathonniere, G. [CEA, DEN/DANS/I-tese, F-91191 Gif-sur-Yvette (France)

    2013-07-01T23:59:59.000Z

    Current light water reactors (LWR) burn fissile uranium, whereas some future reactors, as Sodium fast reactors (SFR) will be capable of recycling their own plutonium and already-extracted depleted uranium. This makes them a feasible solution for the sustainable development of nuclear energy. Nonetheless, a sufficient quantity of plutonium is needed to start up an SFR, with the plutonium already being produced in light water reactors. The availability of natural uranium therefore has a direct impact on the capacity of the reactors (both LWR and SFR) that we can build. It is therefore important to have an accurate estimate of the available uranium resources in order to plan for the world's future nuclear reactor fleet. This paper discusses the correspondence between the resources (uranium and plutonium) and the nuclear power demand. Sodium fast reactors will be built in line with the availability of plutonium, including fast breeders when necessary. Different assumptions on the global uranium resources are taken into consideration. The largely quoted estimate of 22 Mt of uranium recovered for phosphate rocks can be seriously downscaled. Based on our current knowledge of phosphate resources, 4 Mt of recoverable uranium already seems to be an upper bound value. The impact of the downscaled estimate on the deployment of a nuclear fleet is assessed accordingly. (authors)

  7. DIRECT ENERGY CONVERSION (DEC) FISSION REACTORS - A U.S. NERI PROJECT

    SciTech Connect (OSTI)

    D. BELLER; G. POLANSKY; ET AL

    2000-11-01T23:59:59.000Z

    The direct conversion of the electrical energy of charged fission fragments was examined early in the nuclear reactor era, and the first theoretical treatment appeared in the literature in 1957. Most of the experiments conducted during the next ten years to investigate fission fragment direct energy conversion (DEC) were for understanding the nature and control of the charged particles. These experiments verified fundamental physics and identified a number of specific problem areas, but also demonstrated a number of technical challenges that limited DEC performance. Because DEC was insufficient for practical applications, by the late 1960s most R&D ceased in the US. Sporadic interest in the concept appears in the literature until this day, but there have been no recent programs to develop the technology. This has changed with the Nuclear Energy Research Initiative that was funded by the U.S. Congress in 1999. Most of the previous concepts were based on a fission electric cell known as a triode, where a central cathode is coated with a thin layer of nuclear fuel. A fission fragment that leaves the cathode with high kinetic energy and a large positive charge is decelerated as it approaches the anode by a charge differential of several million volts, it then deposits its charge in the anode after its kinetic energy is exhausted. Large numbers of low energy electrons leave the cathode with each fission fragment; they are suppressed by negatively biased on grid wires or by magnetic fields. Other concepts include magnetic collimators and quasi-direct magnetohydrodynamic generation (steady flow or pulsed). We present the basic principles of DEC fission reactors, review the previous research, discuss problem areas in detail and identify technological developments of the last 30 years relevant to overcoming these obstacles. A prognosis for future development of direct energy conversion fission reactors will be presented.

  8. Computerization upgrade project for the Rocky Flats Plant Critical Mass Laboratory Reactor Control Console

    SciTech Connect (OSTI)

    Bachman, H.C.; Miles, R.E.; Sachs, R.D.

    1987-01-01T23:59:59.000Z

    This report discusses present and planned future work on computerization of the Rocky Flats Plant (RFP) Critical Mass Laboratory (CML) Nuclear Reactor Control Console. No computerized control functions are planned or anticipated at this time. The scope of this computerization effort is limited to Data Acquisition and Analysis. In this work an IBM-PC will be connected to four (4) Nuclear Safety channels, and two (2) nonnuclear safety channels. Programming is being done in interpretive advanced BASIC. At the present time only two channels, Linear Picoammeters 1 and 2, are having their signals processed by the IBM-PC.

  9. 1 hour, 25 minutes ago Japan will ask the European Union to declare it a "joint host" of a revolutionary nuclear energy project even if the reactor is located in

    E-Print Network [OSTI]

    " of a revolutionary nuclear energy project even if the reactor is located in France, a newspaper said. The Nihon to 'joint host' revolutionary nuclear reactor 6/6/05 8:23 AMPrint Story: Japan to ask EU to 'joint host' revolutionary nuclear reactor on Yahoo! News Page 1 of 1http://news.yahoo.com/s/afp/20050606/sc

  10. Direct energy conversion in fission reactors: A U.S. NERI project

    SciTech Connect (OSTI)

    SLUTZ,STEPHEN A.; SEIDEL,DAVID B.; POLANSKY,GARY F.; ROCHAU,GARY E.; LIPINSKI,RONALD J.; BESENBRUCH,G.; BROWN,L.C.; PARISH,T.A.; ANGHAIE,S.; BELLER,D.E.

    2000-05-30T23:59:59.000Z

    In principle, the energy released by a fission can be converted directly into electricity by using the charged fission fragments. The first theoretical treatment of direct energy conversion (DEC) appeared in the literature in 1957. Experiments were conducted over the next ten years, which identified a number of problem areas. Research declined by the late 1960's due to technical challenges that limited performance. Under the Nuclear Energy Research Initiative the authors are determining if these technical challenges can be overcome with todays technology. The authors present the basic principles of DEC reactors, review previous research, discuss problem areas in detail, and identify technological developments of the last 30 years that can overcome these obstacles. As an example, the fission electric cell must be insulated to avoid electrons crossing the cell. This insulation could be provided by a magnetic field as attempted in the early experiments. However, from work on magnetically insulated ion diodes they know how to significantly improve the field geometry. Finally, a prognosis for future development of DEC reactors will be presented .

  11. Uranium industry annual 1997

    SciTech Connect (OSTI)

    NONE

    1998-04-01T23:59:59.000Z

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

  12. URANIUM IN ALKALINE ROCKS

    E-Print Network [OSTI]

    Murphy, M.

    2011-01-01T23:59:59.000Z

    Greenland," in Uranium Exploration Geology, Int. AtomicOklahoma," 1977 Nure Geology Uranium Symposium, Igneous HostMcNeil, M. , 1977. "Geology of Brazil's Uranium and Thorium

  13. REACTOR DOSIMETRY STUDY OF THE RHODE ISLAND NUCLEAR SCIENCE CENTER.

    SciTech Connect (OSTI)

    HOLDEN, N.E.,; RECINIELLO, R.N.; HU, J.-P.

    2005-05-08T23:59:59.000Z

    The Rhode Island Nuclear Science Center (RINSC), located on the Narragansett Bay Campus of the University of Rhode Island, is a state-owned and US NRC-licensed nuclear facility constructed for educational and industrial applications. The main building of RINSC houses a two-megawatt (2 MW) thermal power critical reactor immersed in demineralized water within a shielded tank. As its original design in 1958 by the Rhode Island Atomic Energy Commission focused on the teaching and research use of the facility, only a minimum of 3.85 kg fissile uranium-235 was maintained in the fuel elements to allow the reactor to reach a critical state. In 1986 when RINSC was temporarily shutdown to start US DOE-directed core conversion project for national security reasons, all the U-Al based Highly-Enriched Uranium (HEU, 93% uranium-235 in the total uranium) fuel elements were replaced by the newly developed U{sub 3}Si{sub 2}-Al based Low Enriched Uranium (LEU, {le}20% uranium-235 in the total uranium) elements. The reactor first went critical after the core conversion was achieved in 1993, and feasibility study on the core upgrade to accommodate Boron Neutron-Captured Therapy (BNCT) was completed in 2000 [3]. The 2-MW critical reactor at RINSC which includes six beam tubes, a thermal column, a gamma-ray experimental station and two pneumatic tubes has been extensive utilized as neutron-and-photon dual source for nuclear-specific research in areas of material science, fundamental physics, biochemistry, and radiation therapy. After the core conversion along with several major system upgrade (e.g. a new 3-MW cooling tower, a large secondary piping system, a set of digitized power-level instrument), the reactor has become more compact and thus more effective to generate high beam flux in both the in-core and ex-core regions for advance research. If not limited by the manpower and operating budget in recent years, the RINSC built ''in concrete'' structure and control systems should have been systematically upgraded to a 5 Mw power facility to further enhance its experimental capability while still maintaining its safe margin as designed.

  14. France gets nuclear fusion plant France will get to host the project to build a 10bn-euro (6.6bn) nuclear fusion reactor, in

    E-Print Network [OSTI]

    ) nuclear fusion reactor, in the face of strong competition from Japan. The International Thermonuclear division, which is responsible for the UK's thermonuclear fusion programme, said the decisionFrance gets nuclear fusion plant France will get to host the project to build a 10bn-euro (£6.6bn

  15. OECD MCCI project enhancing instrumentation for reactor materials experiments, Rev. 0 September 3, 2002.

    SciTech Connect (OSTI)

    Lomperski, S.; Basu, S. (Nuclear Engineering Division); (NRC)

    2011-05-23T23:59:59.000Z

    Reactor safety experiments for studying the reactions of a molten core (corium) with water and/or concrete involve materials at extremely high temperature. Such high temperature severely restricts the types of sensors that can be employed to measure characteristics of the corium itself. Yet there is great interest in improving instrumentation so that the state of the melt can be established with more precision. In particular, it would be beneficial to increase both the upper range limit and accuracy of temperature measurements. The poor durability of thermocouples at high temperature is also an important issue. For experiments involving a water-quenched melt, direct measurements of the growth rate of the crust separating the melt and water would be of great interest. This is a key element in determining the nature of heat transfer between the melt and coolant. Despite its importance, no one has been able to directly measure the crust thickness during such tests. This paper considers three specialized sensors that could be introduced to enhance melt characterization: (1) A commercially fabricated, single point infrared temperature measurement with the footprint of a thermowell. A lens assembly and fiber optic cable linked to a receiver and amplifier measures the temperature at the base of a tungsten thermowell. The upper range limit is 3000 C and accuracy is {+-}0.25% of the reading. (2) In-house development of an ultrasonic temperature sensor that would provide multipoint measurements at temperatures up to {approx}3000 C. The sensors are constructed from tungsten rods and have a high temperature durability that is superior to that of thermocouples. (3) In-house development of an ultrasonic probe to measure the growth rate of the corium crust. This ultrasonic sensor would include a tungsten waveguide that transmits ultrasonic pulses up through the corium melt towards the crust and detects reflections from the melt/crust interface. A measurement of the echo time delay would provide the location of the interface. These three sensors would provide a considerable upgrade of the instrumentation used in our reactor materials tests. The infracouple is a commercial product that could provide an immediate improvement in temperature measurements. The sensor could also serve to corroborate thermocouple data by providing a measurement based upon a different physical principle. The ultrasonic temperature sensor would involve a greater investment and longer time frame than the infracouple, but offers all the advantages of the infracouple along with miniaturization and the ability to measure at multiple locations. In addition, the UTS is the platform from which we would begin development of the crust detector. Of the three sensors, the crust detector requires the most effort and entails the greatest uncertainty. However, a real-time crust thickness measurement has never before been made and such data would be unique and of great benefit to reactor materials experiments.

  16. Composition of the U.S. DOE Depleted Uranium Inventory

    E-Print Network [OSTI]

    Concentration Of Less

    about 2.75 wt% U-235. For further enrichment, the material was shipped to the Oak Ridge and Portsmouth plants. In addition to natural uranium, also uranium recycled from spent fuel was fed into the Paducah enrichment cascade (Table 2 and Fig. 2). The recycled uranium introduced various isotopes not found in natural uranium into the cascade: fission products, such as Technetium-99; transuranics, such as Neptunium-237 and Plutonium-239; and the artificial uranium isotope of Uranium-236. The spent fuel, from which uranium was recycled, originated from the Hanford and Savannah River military plutonium production reactors. This uranium was recycled, although its assay of U-235 was somewhat lower than in natural uranium (Table 2). This obviously must be seen in the context of the Cold War era, when uranium was a scarce resource. Due to the low burn-up of the military reactors, concentrations of artificial U-236 are comparatively low in this recycled uranium. The recycled uranium represents

  17. Fracture mechanics models developed for piping reliability assessment in light water reactors: piping reliability project

    SciTech Connect (OSTI)

    Harris, D.O.; Lim, E.Y.; Dedhia, D.D.; Woo, H.H.; Chou, C.K.

    1982-06-01T23:59:59.000Z

    The efforts concentrated on modifications of the stratified Monte Carlo code called PRAISE (Piping Reliability Analysis Including Seismic Events) to make it more widely applicable to probabilistic fracture mechanics analysis of nuclear reactor piping. Pipe failures are considered to occur as the result of crack-like defects introduced during fabrication, that escape detection during inspections. The code modifications allow the following factors in addition to those considered in earlier work to be treated: other materials, failure criteria and subcritical crack growth characteristic; welding residual and vibratory stresses; and longitudinal welds (the original version considered only circumferential welds). The fracture mechanics background for the code modifications is included, and details of the modifications themselves provided. Additionally, an updated version of the PRAISE user's manual is included. The revised code, known as PRAISE-B was then applied to a variety of piping problems, including various size lines subject to stress corrosion cracking and vibratory stresses. Analyses including residual stresses and longitudinal welds were also performed.

  18. Implementation of UMTRA Project Environmental Audit Action Plan status report for period ending September 30, 1992. [Uranium Mill Tailings Remedial Action (UMTRA) Project

    SciTech Connect (OSTI)

    Not Available

    1992-09-01T23:59:59.000Z

    This report provides the status of implementation of corrective actions for findings made in an Environmental Audit conducted by DOE Headquarters, Office of Environmental Audit, in June 1991. An Action Plan, dated December 1991, was developed to address the findings. The Action Plan was approved by DOE Headquarters, Office of Environment, Safety and Health, in July 1992. This report provides status for each activity listed in the approved Action Plan. Of 48 findings identified in the August 1991 Environmental Audit Report, 4 required no action, 5 were combined with others and actions to correct 19 are complete. Although it appears no progress has been made since the last status report was issued, UMTRA has completed 89% of the findings identified, compared to 72% identified in the last status report. The table below lists the 20 findings where actions are still underway, the current projected completion date, the organization(s) responsible for taking action on the finding, and the UMTRA Project Off ice staff member assigned responsibility for the finding.

  19. Uranium Oxide Aerosol Transport in Porous Graphite

    SciTech Connect (OSTI)

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

    2012-01-23T23:59:59.000Z

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

  20. The Windscale Advanced Gas Cooled Reactor (WAGR) Decommissioning Project A Close Out Report for WAGR Decommissioning Campaigns 1 to 10 - 12474

    SciTech Connect (OSTI)

    Halliwell, Chris [Sellafield Ltd, Sellafield (United Kingdom)

    2012-07-01T23:59:59.000Z

    The reactor core of the Windscale Advanced Gas-Cooled Reactor (WAGR) has been dismantled as part of an ongoing decommissioning project. The WAGR operated until 1981 as a development reactor for the British Commercial Advanced Gas cooled Reactor (CAGR) power programme. Decommissioning began in 1982 with the removal of fuel from the reactor core which was completed in 1983. Subsequently, a significant amount of engineering work was carried out, including removal of equipment external to the reactor and initial manual dismantling operations at the top of the reactor, in preparation for the removal of the reactor core itself. Modification of the facility structure and construction of the waste packaging plant served to provide a waste route for the reactor components. The reactor core was dismantled on a 'top-down' basis in a series of 'campaigns' related to discrete reactor components. This report describes the facility, the modifications undertaken to facilitate its decommissioning and the strategies employed to recognise the successful decommissioning of the reactor. Early decommissioning tasks at the top of the reactor were undertaken manually but the main of the decommissioning tasks were carried remotely, with deployment systems comprising of little more than crane like devices, intelligently interfaced into the existing structure. The tooling deployed from the 3 tonne capacity (3te) hoist consisted either purely mechanical devices or those being electrically controlled from a 'push-button' panel positioned at the operator control stations, there was no degree of autonomy in the 3te hoist or any of the tools deployed from it. Whilst the ATC was able to provide some tele-robotic capabilities these were very limited and required a good degree of driver input which due to the operating philosophy at WAGR was not utilised. The WAGR box proved a successful waste package, adaptable through the use of waste box furniture specific to the waste-forms generated throughout the various decommissioning campaigns. The use of low force compaction for insulation and soft wastes provided a simple, robust and cost effective solution as did the direct encapsulation of LLW steel components in the later stages of reactor decommissioning. Progress through early campaigns was good, often bettering the baseline schedule, especially when undertaking the repetitive tasks seen during Neutron Shield and Graphite Core decommissioning, once the operators had become experienced with the equipment, though delays became more pronounced, mainly as a result of increased failures due to the age and maintainability of the RDM and associated equipment. Extensive delays came about as a result of the unsupported insulation falling away from the pressure vessel during removal and the inability of the ventilation system to manage the sub micron particulate generated during IPOPI cutting operations, though the in house development of revised and new methodologies ultimately led to the successful completion of PV and I removal. In a programme spanning over 12 years, the decommissioning of the reactor pressure vessel and core led to the production 110 ILW and 75 LLW WAGR boxes, with 20 LLW ISO freight containers of primary reactor wastes, resulting in an overall packaged volume of approximately 2500 cubic metres containing the estimated 460 cubic metres of the reactor structure. (authors)

  1. Electrolytic process for preparing uranium metal

    DOE Patents [OSTI]

    Haas, Paul A. (Knoxville, TN)

    1990-01-01T23:59:59.000Z

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

  2. Probabilistic evaluation of main coolant pipe break indirectly induced by earthquakes: Savannah River Project L and P Reactors

    SciTech Connect (OSTI)

    Short, S.A.; Wesley, D.A.; Awadalla, N.G.; Kennedy, R.P. (Impell Corp., Mission Viejo, CA (USA); Westinghouse Savannah River Co., Aiken, SC (USA); Structural Mechanics Consulting, Inc., Yorba Linda, CA (USA))

    1989-01-01T23:59:59.000Z

    A probabilistic evaluation of seismically-induced indirect pipe break for the Savannah River Project (SRP) L- and P-Reactor main coolant (process water) piping has been conducted. Seismically-induced indirect pipe break can result primarily from: (1) failure of the anchorage of one or more of the components to which the pipe is anchored; or (2) failure of the pipe due to collapse of the structure. The potential for both types of seismically-induced indirect failures was identified during a seismic walkdown of the main coolant piping. This work involved: (1) identifying components or structures whose failure could result in pipe failure; (2) developing seismic capacities or fragilities of these components; (3) combining component fragilities to develop plant damage state fragilities; and (4) convolving the plant seismic fragilities with a probabilistic seismic hazard estimate for the site in order to obtain estimates of seismic risk in terms of annual probability of seismic-induced indirect pipe break. 6 refs., 5 figs., 2 tabs.

  3. Ssessment methodology for proliferation resistant fast breeder reactor

    E-Print Network [OSTI]

    Singh, Mohit, S.M. Massachusetts Institute of Technology

    2014-01-01T23:59:59.000Z

    Due to perceived proliferation risks, current US fast reactor designs have avoided the use of uranium blankets. While reducing the amount of plutonium produced, this omission also restrains the reactor design space and has ...

  4. A parametric study of the breeding ratio in sodium cooled fast breeder reactors

    E-Print Network [OSTI]

    Sobey, Thomas Milburn

    1969-01-01T23:59:59.000Z

    in Radial Blanket of Pancake Core 20 25 CHAPTER I INTRODUCTION In recent years projected figures have been arrived at that clearly indicate the need for a breeder reactor if nuclear power development is going to proceed at a rate that will enable... it to compete with the fossil fuel plants presently in use. A large demand in the nuclear power development program is for a continu- ing supply of low cost fuel, either uranium or plutonium. It i. s for this reason that t'ne nuclear reactor development...

  5. Examination of the conversion of the U.S. submarine fleet from highly enriched uranium to low enriched uranium

    E-Print Network [OSTI]

    McCord, Cameron (Cameron Liam)

    2014-01-01T23:59:59.000Z

    The nuclear reactors used by the U.S. Navy for submarine propulsion are currently fueled by highly enriched uranium (HEU), but HEU brings administrative and political challenges. This issue has been studied by the Navy ...

  6. EIS-0291: High Flux Beam Reactor (HFBR) Transition Project at the Brookhaven National Laboratory, Upton, New York

    Broader source: Energy.gov [DOE]

    The EIS evaluates the range of reasonable alternatives and their impacts regarding the future management of the High Flux Beam Reactor (HFBR) at Brookhaven National Laboratory (BNL).

  7. Uranium industry annual 1996

    SciTech Connect (OSTI)

    NONE

    1997-04-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Not Available

    1980-01-01T23:59:59.000Z

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

  9. Preserving Ultra-Pure Uranium-233

    SciTech Connect (OSTI)

    Krichinsky, Alan M [ORNL; Goldberg, Dr. Steven A. [DOE SC - Chicago Office; Hutcheon, Dr. Ian D. [Lawrence Livermore National Laboratory (LLNL)

    2011-10-01T23:59:59.000Z

    Uranium-233 ({sup 233}U) is a synthetic isotope of uranium formed under reactor conditions during neutron capture by natural thorium ({sup 232}Th). At high purities, this synthetic isotope serves as a crucial reference material for accurately quantifying and characterizing uranium-bearing materials assays and isotopic distributions for domestic and international nuclear safeguards. Separated, high purity {sup 233}U is stored in vaults at Oak Ridge National Laboratory (ORNL). These materials represent a broad spectrum of {sup 233}U from the standpoint of isotopic purity - the purest being crucial for precise analyses in safeguarding uranium. All {sup 233}U at ORNL is currently scheduled to be disposed of by down-blending with depleted uranium beginning in 2015. This will reduce safety concerns and security costs associated with storage. Down-blending this material will permanently destroy its potential value as a certified reference material for use in uranium analyses. Furthermore, no credible options exist for replacing {sup 233}U due to the lack of operating production capability and the high cost of restarting currently shut down capabilities. A study was commissioned to determine the need for preserving high-purity {sup 233}U. This study looked at the current supply and the historical and continuing domestic need for this crucial isotope. It examined the gap in supplies and uses to meet domestic needs and extrapolated them in the context of international safeguards and security activities - superimposed on the recognition that existing supplies are being depleted while candidate replacement material is being prepared for disposal. This study found that the total worldwide need by this projection is at least 850 g of certified {sup 233}U reference material over the next 50 years. This amount also includes a strategic reserve. To meet this need, 18 individual items totaling 959 g of {sup 233}U were identified as candidates for establishing a lasting supply of certified reference materials (CRM), all having an isotopic purity of at least 99.4% {sup 233}U and including materials up to 99.996% purity. Current plans include rescuing the purest {sup 233}U materials during a 3-year project beginning in FY 2012 in three phases involving preparations, handling preserved materials, and cleanup. The first year will involve preparations for handling the rescued material for sampling, analysis, distribution, and storage. Such preparations involve modifying or developing work control documents and physical preparations in the laboratory, which include preparing space for new material-handling equipment and procuring and (in some cases) refurbishing equipment needed for handling {sup 233}U or qualifying candidate CRM. Once preparations are complete, an evaluation of readiness will be conducted by independent reviewers to verify that the equipment, work controls, and personnel are ready for operations involving handling radioactive materials with nuclear criticality safety as well as radiological control requirements. The material-handling phase will begin in FY 2013 and be completed early in FY 2014, as currently scheduled. Material handling involves retrieving candidate CRM items from the ORNL storage facility and shipping them to another laboratory at ORNL; receiving and handling rescued items at the laboratory (including any needed initial processing, acquisition and analysis of samples from each item, and preparation for shipment); and shipping bulk material to destination labs or to a yet-to-be-designated storage location. There are seven groups of {sup 233}U identified for handling based on isotopic purity that require the utmost care to prevent cross-contamination. The last phase, cleanup, also will be completed in 2014. It involves cleaning and removing the equipment and material-handling boxes and characterizing, documenting, and disposing of waste. As part of initial planning, the cost of rescuing candidate {sup 233}U items was estimated roughly. The annualized costs were found to be $1,228K in FY 2012, $1,375K in FY 2013,

  10. Thermal Hydraulics of the Very High Temperature Gas Cooled Reactor

    SciTech Connect (OSTI)

    Chang Oh; Eung Kim; Richard Schultz; Mike Patterson; Davie Petti

    2009-10-01T23:59:59.000Z

    The U.S Department of Energy (DOE) is conducting research on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core will be either a prismatic graphite block type core or a pebble bed core. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during reactor core-accidents. The objectives of the NGNP Project are to: Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission, and Demonstrate safe and economical nuclear-assisted production of hydrogen and electricity. The DOE laboratories, led by the INL, perform research and development (R&D) that will be critical to the success of the NGNP, primarily in the areas of: High temperature gas reactor fuels behavior High temperature materials qualification Design methods development and validation Hydrogen production technologies Energy conversion. This paper presents current R&D work that addresses fundamental thermal hydraulics issues that are relevant to a variety of possible NGNP designs.

  11. Examination of the conversion of the U.S. submarine fleet from highly enriched uranium to low enriched uranium ; Examination of the conversion of the United States submarine fleet from HEU to low LEU .

    E-Print Network [OSTI]

    McCord, Cameron (Cameron Liam)

    2014-01-01T23:59:59.000Z

    ??The nuclear reactors used by the U.S. Navy for submarine propulsion are currently fueled by highly enriched uranium (HEU), but HEU brings administrative and political (more)

  12. SUMMARY AND RESULTS LETTER REPORT INDEPENDENT VERIFICATION OF THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PROJECT, PHASE 3: TRENCHES 2, 3, AND 4 BROOKHAVEN NATIONAL LABORATORY UPTON, NEW YORK

    SciTech Connect (OSTI)

    E.M. Harpenau

    2010-11-15T23:59:59.000Z

    5098-LR-02-0 SUMMARY AND RESULTS LETTER REPORT INDEPENDENT VERIFICATION OF THE HIGH FLUX BEAM REACTOR UNDERGROUND UTILITIES REMOVAL PROJECT, PHASE 3 TRENCHES 2, 3, AND 4 BROOKHAVEN NATIONAL LABORATORY

  13. State of Advancement of the International REVE Project: Computational Modelling of Irradiation-Induced Hardening in Reactor Pressure Vessel Steels and Relevant Experimental Validation Programme

    SciTech Connect (OSTI)

    Malerba, Lorenzo; Van Walle, Eric [SCK.CEN, Boeretang 200, 2400 Mol (Belgium); Domain, Christophe; Jumel, Stephanie; Van Duysen, Jean-Claude [EDR R and D (France)

    2002-07-01T23:59:59.000Z

    The REVE (Reactor for Virtual Experiments) project is an international joint effort aimed at developing multi-scale modelling computational toolboxes capable of simulating the behaviour of materials under irradiation at different time and length scales. Well grounded numerical techniques such as molecular dynamics (MD) and Monte Carlo (MC) algorithms, as well as rate equation (RE) and dislocation-defect interaction theory, form the basis on which the project is built. The goal is to put together a suite of integrated codes capable of deducing the changes in macroscopic properties starting from a detailed simulation of the microstructural changes produced by irradiation in materials. To achieve this objective, several European laboratories are closely collaborating, while exchanging data with American and Japanese laboratories currently pursuing similar approaches. The material chosen for the first phase of this project is reactor pressure vessel (RPV) steel, the target macroscopic magnitude to be predicted being the yield strength increase ({delta}{sigma}y) due, essentially, to irradiation-enhanced formation of intragranular solute atom precipitates or clouds, as well as irradiation induced defects in the matrix, such as point defect clusters and dislocation loops. A description of the methodological approach used in the project and its current state is given in the paper. The development of the simulation tools requires a continuous feedback from ad hoc experimental data. In the framework of the REVE project SCK EN has therefore performed a neutron irradiation campaign of model alloys of growing complexity (from pure Fe to binary and ternary systems and a real RPV steel) in the Belgian test reactor BR2 and is currently carrying on the subsequent materials characterisation using its hot cell facilities. The paper gives the details of this experimental programme - probably the first large-scale one devoted to the validation of numerical simulation tools - and presents and discusses the first available results, with a view to their use as feedback for the improvement of the computational modelling. (authors)

  14. Monitoring Uranium Transformations Determined by the Evolution of Biogeochemical Processes: Design of Mixed Batch Reactor and Column Studies at Oak Ridge National Laboratory

    SciTech Connect (OSTI)

    Criddle, Craig S.; Wu, Weimin

    2013-04-17T23:59:59.000Z

    With funds provided by the US DOE, Argonne National Laboratory subcontracted the design of batch and column studies to a Stanford University team with field experience at the ORNL IFRC, Oak Ridge, TN. The contribution of the Stanford group ended in 2011 due to budget reduction in ANL. Over the funded research period, the Stanford research team characterized ORNL IFRC groundwater and sediments and set up microcosm reactors and columns at ANL to ensure that experiments were relevant to field conditions at Oak Ridge. The results of microcosm testing demonstrated that U(VI) in sediments was reduced to U(IV) with the addition of ethanol. The reduced products were not uraninite but were instead U(IV) complexes associated with Fe. Fe(III) in solid phase was only partially reduced. The Stanford team communicated with the ANL team members through email and conference calls and face to face at the annual ERSP PI meeting and national meetings.

  15. Nuclear Safeguards Considerations For The Pebble Bed Modular Reactor (PBMR)

    SciTech Connect (OSTI)

    Phillip Casey Durst; David Beddingfield; Brian Boyer; Robert Bean; Michael Collins; Michael Ehinger; David Hanks; David L. Moses; Lee Refalo

    2009-10-01T23:59:59.000Z

    High temperature reactors (HTRs) have been considered since the 1940s, and have been constructed and demonstrated in the United Kingdom (Dragon), United States (Peach Bottom and Fort Saint Vrain), Japan (HTTR), Germany (AVR and THTR-300), and have been the subject of conceptual studies in Russia (VGM). The attraction to these reactors is that they can use a variety of reactor fuels, including abundant thorium, which upon reprocessing of the spent fuel can produce fissile U-233. Hence, they could extend the stocks of available uranium, provided the fuel is reprocessed. Another attractive attribute is that HTRs typically operate at a much higher temperature than conventional light water reactors (LWRs), because of the use of pyrolytic carbon and silicon carbide coated (TRISO) fuel particles embedded in ceramic graphite. Rather than simply discharge most of the unused heat from the working fluid in the power plant to the environment, engineers have been designing reactors for 40 years to recover this heat and make it available for district heating or chemical conversion plants. Demonstrating high-temperature nuclear energy conversion was the purpose behind Fort Saint Vrain in the United States, THTR-300 in Germany, HTTR in Japan, and HTR-10 and HTR-PM, being built in China. This resulted in nuclear reactors at least 30% or more thermodynamically efficient than conventional LWRs, especially if the waste heat can be effectively utilized in chemical processing plants. A modern variant of high temperature reactors is the Pebble Bed Modular Reactor (PBMR). Originally developed in the United States and Germany, it is now being redesigned and marketed by the Republic of South Africa and China. The team examined historical high temperature and high temperature gas reactors (HTR and HTGR) and reviewed safeguards considerations for this reactor. The following is a preliminary report on this topic prepared under the ASA-100 Advanced Safeguards Project in support of the NNSA Next Generation Safeguards Initiative (NGSI).

  16. Corrosion Evaluation of RERTR Uranium Molybdenum Fuel

    SciTech Connect (OSTI)

    A K Wertsching

    2012-09-01T23:59:59.000Z

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

  17. Radial power flattening in sodium fast reactors

    E-Print Network [OSTI]

    Krentz-Wee, Rebecca (Rebecca Elizabeth)

    2012-01-01T23:59:59.000Z

    In order to improve a new design for a uranium startup sodium cooled fast reactor which was proposed at MIT, this thesis evaluated radial power flattening by varying the fuel volume fraction at a fixed U-235 enrichment of ...

  18. Uranium Industry Annual, 1992

    SciTech Connect (OSTI)

    Not Available

    1993-10-28T23:59:59.000Z

    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.

  19. Independent Oversight Assessment, Portsmouth/Paducah Project...

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

    Project Office - May 2012 May 2012 Assessment of the PortsmouthPaducah Project Office Conduct of Operations Oversight of the Depleted Uranium Hexafluoride Conversion Plants...

  20. Depleted uranium disposal options evaluation

    SciTech Connect (OSTI)

    Hertzler, T.J.; Nishimoto, D.D.; Otis, M.D. [Science Applications International Corp., Idaho Falls, ID (United States). Waste Management Technology Div.

    1994-05-01T23:59:59.000Z

    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.

  1. PROJECT-SPECIFIC TYPE A VERIFICATION FOR THE BROOKHAVEN GRAPHITE RESEARCH REACTOR ENGINEERED CAP, BROOKHAVEN NATIONAL LABORATORY UPTON, NEW YORK DCN 5098-SR-07-0

    SciTech Connect (OSTI)

    Evan Harpenau

    2011-07-15T23:59:59.000Z

    The Oak Ridge Institute for Science and Education (ORISE) has reviewed the project documentation and data for the Brookhaven Graphite Research Reactor (BGRR) Engineered Cap at Brookhaven National Laboratory (BNL) in Upton, New York. The Brookhaven Science Associates (BSA) have completed removal of affected soils and performed as-left surveys by BSA associated with the BGRR Engineered Cap. Sample results have been submitted, as required, to demonstrate that remediation efforts comply with the cleanup goal of {approx}15 mrem/yr above background to a resident in 50 years (BNL 2011a).

  2. Uranium Management - Preservation of a National Asset

    SciTech Connect (OSTI)

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

    2002-02-27T23:59:59.000Z

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

  3. Imaging Fukushima Daiichi reactors with muons

    SciTech Connect (OSTI)

    Miyadera, Haruo; Borozdin, Konstantin N.; Greene, Steve J.; Milner, Edward C.; Morris, Christopher L. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Lukic, Zarija [Computational Cosmology Center, Lawrence Berkeley National Laboratory, Berkeley, CA 94720 (United States); Masuda, Koji [University of New Mexico, Albuquerque, NM 87131 (United States); Perry, John O. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); University of New Mexico, Albuquerque, NM 87131 (United States)

    2013-05-15T23:59:59.000Z

    A study of imaging the Fukushima Daiichi reactors with cosmic-ray muons to assess the damage to the reactors is presented. Muon scattering imaging has high sensitivity for detecting uranium fuel and debris even through thick concrete walls and a reactor pressure vessel. Technical demonstrations using a reactor mockup, detector radiation test at Fukushima Daiichi, and simulation studies have been carried out. These studies establish feasibility for the reactor imaging. A few months of measurement will reveal the spatial distribution of the reactor fuel. The muon scattering technique would be the best and probably the only way for Fukushima Daiichi to make this determination in the near future.

  4. Evaluation of nuclear facility decommissioning projects. Three Mile Island Unit 2 reactor building decontamination. Summary status report. Volume 2

    SciTech Connect (OSTI)

    Doerge, D.H.; Miller, R.L.; Scotti, K.S.

    1986-05-01T23:59:59.000Z

    This document summarizes information relating to decontamination of the Three Mile Island Unit 2 (TMI-2) reactor building. The report covers activities for the period of June 1, 1979 through March 29, 1985. The data collected from activity reports, reactor containment entry records, and other sources were entered into a computerized data system which permits extraction/manipulation of specific information which can be used in planning for recovery from an accident similar to that experienced at TMI-2 on March 28, 1979. This report contains summaries of man-hours, manpower, and radiation exposures incurred during decontamination of the reactor building. Support activities conducted outside of radiation areas are excluded from the scope of this report. Computerized reports included in this document are: a chronological summary listing work performed relating to reactor building decontamination for the period specified; and summary reports for each major task during the period. Each task summary is listed in chronological order for zone entry and subtotaled for the number of personnel entries, exposures, and man-hours. Manually-assembled table summaries are included for: labor and exposures by department and labor and exposures by major activity.

  5. The New Generation of Uranium In Situ Recovery Facilities: Design Improvements Should Reduce Radiological Impacts Relative to First Generation Uranium Solution Mining Plants

    SciTech Connect (OSTI)

    Brown, S.H. [CHP, SHB INC., Centennial, Colorado (United States)

    2008-07-01T23:59:59.000Z

    In the last few years, there has been a significant increase in the demand for Uranium as historical inventories have been consumed and new reactor orders are being placed. Numerous mineralized properties around the world are being evaluated for Uranium recovery and new mining / milling projects are being evaluated and developed. Ore bodies which are considered uneconomical to mine by conventional methods such as tunneling or open pits, can be candidates for non-conventional recovery techniques, involving considerably less capital expenditure. Technologies such as Uranium In Situ Leaching / In Situ Recovery (ISL / ISR - also referred to as 'solution mining'), have enabled commercial scale mining and milling of relatively small ore pockets of lower grade, and are expected to make a significant contribution to overall world wide uranium supplies over the next ten years. Commercial size solution mining production facilities have operated in the US since the mid 1970's. However, current designs are expected to result in less radiological wastes and emissions relative to these 'first' generation plants (which were designed, constructed and operated through the 1980's). These early designs typically used alkaline leach chemistries in situ including use of ammonium carbonate which resulted in groundwater restoration challenges, open to air recovery vessels and high temperature calcining systems for final product drying vs the 'zero emissions' vacuum dryers as typically used today. Improved containment, automation and instrumentation control and use of vacuum dryers in the design of current generation plants are expected to reduce production of secondary waste byproduct material, reduce Radon emissions and reduce potential for employee exposure to uranium concentrate aerosols at the back end of the milling process. In Situ Recovery in the U.S. typically involves the circulation of groundwater, fortified with oxidizing (gaseous oxygen e.g) and complexing agents (carbon dioxide, e.g) into an ore body, solubilizing the uranium in situ, and then pumping the solutions to the surface where they are fed to a processing plant ( mill). Processing involves ion exchange and may also include precipitation, drying or calcining and packaging operations depending on facility specifics. This paper presents an overview of the ISR process and the health physics monitoring programs developed at a number of commercial scale ISL / ISR Uranium recovery and production facilities as a result of the radiological character of these processes. Although many radiological aspects of the process are similar to that of conventional mills, conventional-type tailings as such are not generated. However, liquid and solid byproduct materials may be generated and impounded. The quantity and radiological character of these by products are related to facility specifics. Some special monitoring considerations are presented which are required due to the manner in which radon gas is evolved in the process and the unique aspects of controlling solution flow patterns underground. The radiological character of these processes are described using empirical data collected from many operating facilities. Additionally, the major aspects of the health physics and radiation protection programs that were developed at these first generation facilities are discussed and contrasted to circumstances of the current generation and state of the art of uranium ISR technologies and facilities. In summary: This paper has presented an overview of in situ Uranium recovery processes and associated major radiological aspects and monitoring considerations. Admittedly, the purpose was to present an overview of those special health physics considerations dictated by the in situ Uranium recovery technology, to point out similarities and differences to conventional mill programs and to contrast these alkaline leach facilities to modern day ISR designs. As evidenced by the large number of ISR projects currently under development in the U.S. and worldwide, non conventional Uranium recovery techniques

  6. Simulated Verification of Fuel Element Inventory in a Small Reactor Core Using the Nuclear Materials Identification System (NMIS)

    SciTech Connect (OSTI)

    Grogan, Brandon R [ORNL; Mihalczo, John T [ORNL

    2009-01-01T23:59:59.000Z

    The International Panel on Climate Change projects that by 2050 the world energy demand may double. Although the primary focus for new nuclear power plants in industrialized nations is on large plants in the 1000-1600 MWe range, there is an increasing demand for small and medium reactors (SMRs). About half of the innovative SMR concepts are for small (<300 MWe) reactors with a 5-30 year life without on-site refueling. This type of reactor is also known as a battery-type reactor. These reactors are particularly attractive to countries with small power grids and for non-electrical purposes such as heating, hydrogen production, and seawater desalination. Traditionally, this type of reactor has been used in a nautical propulsion role. This type of reactor is designed as a permanently sealed unit to prevent the diversion of the uranium in the core by the user. However, after initial fabrication it will be necessary to verify that the newly fabricated reactor core contains the quantity of uranium that initially entered the fuel fabrication plant. In most instances, traditional inspection techniques can be used to perform this verification, but in certain situations the core design will be considered sensitive. Non-intrusive verification techniques must be utilized in these situations. The Nuclear Materials Identification System (NMIS) with imaging uses active interrogation and a fast time correlation processor to characterize fissile material. The MCNP-PoliMi computer code was used to simulate NMIS measurements of a small, sealed reactor core. Because most battery-type reactor designs are still in the early design phase, a more traditional design based on a Russian icebreaker core was used in the simulations. These simulations show how the radiography capabilities of the NMIS could be used to detect the diversion of fissile material by detecting void areas in the assembled core where fuel elements have been removed.

  7. Annual report to the Advisory Council on Historic Preservation and the Colorado State Historic Preservation Officer on the US Department of Energy's cultural resource activities at Colorado UMTRA Project sites, January--December 1991. [Uranium Mill Tailings Remedial Action (UMTRA) Project

    SciTech Connect (OSTI)

    Not Available

    1992-04-01T23:59:59.000Z

    This report is a summary of the US Department of Energy's (DOE) cultural resource investigations for the Uranium Mill Tailings Remedial Action (UMTRA) Project sites in Colorado. This report is intended to fulfill the DOE's obligation for an annual report as stated in the Programmatic Memorandum of Agreement executed between the DOE, the Advisory Council on Historic Preservation, and the Colorado State Historic Preservation Officer in December 1984. Summaries of the cultural resource surveys and identified resources are provided for the UMTRA Project sites in the vicinities of Durango, Grand Junction, Gunnison, Maybell, Naturita, Rifle, and Slick Rock. This report covers all UMTRA Project cultural resource activities in Colorado from January through December 1991.

  8. Developing fuel management capabilities based on coupled Monte Carlo depletion in support of the MIT Research Reactor (MITR) conversion

    E-Print Network [OSTI]

    Romano, Paul K. (Paul Kollath)

    2009-01-01T23:59:59.000Z

    Pursuant to a 1986 NRC ruling, the MIT Reactor (MITR) is planning on converting from the use of highly enriched uranium (HEU) to low enriched uranium (LEU) for fuel. Prior studies have shown that the MITR will be able to ...

  9. Breeding nuclear fuels with accelerators: replacement for breeder reactors

    SciTech Connect (OSTI)

    Grand, P.; Takahashi, H.

    1984-01-01T23:59:59.000Z

    One application of high energy particle accelerators has been, and still is, the production of nuclear fuel for the nuclear energy industry; tantalizing because it would create a whole new industry. This approach to producing fissile from fertile material was first considered in the early 1950's in the context of the nuclear weapons program. A considerable development effort was expended before discovery of uranium ore in New Mexico put an end to the project. Later, US commitment to the Liquid Metal Fast Breeder Reactors (LMFBR) killed any further interest in pursuing accelerator breeder technology. Interest in the application of accelerators to breed nuclear fuels, and possibly burn nuclear wastes, revived in the late 1970's, when the LMFBR came under attack during the Carter administration. This period gave the opportunity to revisit the concept in view of the present state of the technology. This evaluation and the extensive calculational modeling of target designs that have been carried out are promising. In fact, a nuclear fuel cycle of Light Water Reactors and Accelerator Breeders is competitive to that of the LMFBR. At this time, however, the relative abundance of uranium reserves vs electricity demand and projected growth rate render this study purely academic. It will be for the next generation of accelerator builders to demonstate the competitiveness of this technology versus that of other nuclear fuel cycles, such as LMFBR's or Fusion Hybrid systems. 22 references, 1 figure, 5 tables.

  10. Removal of uranium from uranium-contaminated soils -- Phase 1: Bench-scale testing. Uranium in Soils Integrated Demonstration

    SciTech Connect (OSTI)

    Francis, C. W.

    1993-09-01T23:59:59.000Z

    To address the management of uranium-contaminated soils at Fernald and other DOE sites, the DOE Office of Technology Development formed the Uranium in Soils Integrated Demonstration (USID) program. The USID has five major tasks. These include the development and demonstration of technologies that are able to (1) characterize the uranium in soil, (2) decontaminate or remove uranium from the soil, (3) treat the soil and dispose of any waste, (4) establish performance assessments, and (5) meet necessary state and federal regulations. This report deals with soil decontamination or removal of uranium from contaminated soils. The report was compiled by the USID task group that addresses soil decontamination; includes data from projects under the management of four DOE facilities [Argonne National Laboratory (ANL), Los Alamos National Laboratory (LANL), Oak Ridge National Laboratory (ORNL), and the Savannah River Plant (SRP)]; and consists of four separate reports written by staff at these facilities. The fundamental goal of the soil decontamination task group has been the selective extraction/leaching or removal of uranium from soil faster, cheaper, and safer than current conventional technologies. The objective is to selectively remove uranium from soil without seriously degrading the soil`s physicochemical characteristics or generating waste forms that are difficult to manage and/or dispose of. Emphasis in research was placed more strongly on chemical extraction techniques than physical extraction techniques.

  11. Study of Pu consumption in Advanced Light Water Reactors. Evaluation of GE Advanced Boiling Water Reactor plants

    SciTech Connect (OSTI)

    Not Available

    1993-05-13T23:59:59.000Z

    Timely disposal of the weapons plutonium is of paramount importance to permanently safeguarding this material. GE`s 1300 MWe Advanced Boiling Water Reactor (ABWR) has been designed to utilize fill] core loading of mixed uranium-plutonium oxide fuel. Because of its large core size, a single ABWR reactor is capable of disposing 100 metric tons of plutonium within 15 years of project inception in the spiking mode. The same amount of material could be disposed of in 25 years after the start of the project as spent fuel, again using a single reactor, while operating at 75 percent capacity factor. In either case, the design permits reuse of the stored spent fuel assemblies for electrical energy generation for the remaining life of the plant for another 40 years. Up to 40 percent of the initial plutonium can also be completely destroyed using ABWRS, without reprocessing, either by utilizing six ABWRs over 25 years or by expanding the disposition time to 60 years, the design life of the plants and using two ABWRS. More complete destruction would require the development and testing of a plutonium-base fuel with a non-fertile matrix for an ABWR or use of an Advanced Liquid Metal Reactor (ALMR). The ABWR, in addition, is fully capable of meeting the tritium target production goals with already developed target technology.

  12. Evaluation of nuclear facility decommissioning projects. Three Mile Island Unit 2 reactor defueling and disassembly. Summary status report. Volume 3

    SciTech Connect (OSTI)

    Doerge, D.H.; Miller, R.L.; Scotti, K.S.

    1986-05-01T23:59:59.000Z

    This document summarizes information relating to the preparations for defueling the Three Mile Island Unit 2 (TMI-2) reactor and disassembly activities being performed concurrently with decontamination of the facility. Data have been collected from activity reports, reactor containment entry records, and other sources and entered in a computerized data sysem which permits extraction/manipulation of specific data which can be used in planning for recovery from a loss of coolant event similar to that experienced at TMI-2 on March 28, 1979. This report contains summaries of man-hours, manpower, and radiation exposures incurred during the period of April 23, 1979 to April 16, 1985, in the completion of activities related to preparation for reactor defueling. Support activities conducted outside of radiation areas are not included within the scope of this report. Computerized reports included in this document are: A chronological summary listing work performed for the period; and summary reports for each major task undertaken in connection with the specific scope of this report. Presented in chronological order for the referenced time period. Manually-assembled table summaries are included for: Labor and exposures by department; and labor and exposures by major activity.

  13. Laboratory Directed Research and Development (LDRD) on Mono-uranium Nitride Fuel Development for SSTAR and Space Applications

    SciTech Connect (OSTI)

    Choi, J; Ebbinghaus, B; Meiers, T; Ahn, J

    2006-02-09T23:59:59.000Z

    The US National Energy Policy of 2001 advocated the development of advanced fuel and fuel cycle technologies that are cleaner, more efficient, less waste-intensive, and more proliferation resistant. The need for advanced fuel development is emphasized in on-going DOE-supported programs, e.g., Global Nuclear Energy Initiative (GNEI), Advanced Fuel Cycle Initiative (AFCI), and GEN-IV Technology Development. The Directorates of Energy & Environment (E&E) and Chemistry & Material Sciences (C&MS) at Lawrence Livermore National Laboratory (LLNL) are interested in advanced fuel research and manufacturing using its multi-disciplinary capability and facilities to support a design concept of a small, secure, transportable, and autonomous reactor (SSTAR). The E&E and C&MS Directorates co-sponsored this Laboratory Directed Research & Development (LDRD) Project on Mono-Uranium Nitride Fuel Development for SSTAR and Space Applications. In fact, three out of the six GEN-IV reactor concepts consider using the nitride-based fuel, as shown in Table 1. SSTAR is a liquid-metal cooled, fast reactor. It uses nitride fuel in a sealed reactor vessel that could be shipped to the user and returned to the supplier having never been opened in its long operating lifetime. This sealed reactor concept envisions no fuel refueling nor on-site storage of spent fuel, and as a result, can greatly enhance proliferation resistance. However, the requirement for a sealed, long-life core imposes great challenges to research and development of the nitride fuel and its cladding. Cladding is an important interface between the fuel and coolant and a barrier to prevent fission gas release during normal and accidental conditions. In fabricating the nitride fuel rods and assemblies, the cladding material should be selected based on its the coolant-side corrosion properties, the chemical/physical interaction with the nitride fuel, as well as their thermal and neutronic properties. The US NASA space reactor, the SP-100 was designed to use mono-uranium nitride fuel. Although the SP-100 reactor was not commissioned, tens of thousand of nitride fuel pellets were manufactured and lots of them, cladded in Nb-1-Zr had been irradiated in fast test reactors (FFTF and EBR-II) with good irradiation results. The Russian Naval submarines also use nitride fuel with stainless steel cladding (HT-9) in Pb-Bi coolant. Although the operating experience of the Russian submarine is not readily available, such combination of fuel, cladding and coolant has been proposed for a commercial-size liquid-metal cooled fast reactor (BREST-300). Uranium mono-nitride fuel is studied in this LDRD Project due to its favorable properties such as its high actinide density and high thermal conductivity. The thermal conductivity of mono-nitride is 10 times higher than that of oxide (23 W/m-K for UN vs. 2.3 W/m-K for UO{sub 2} at 1000 K) and its melting temperature is much higher than that of metal fuel (2630 C for UN vs. 1132 C for U metal). It also has relatively high actinide density, (13.51 gU/cm{sup 3} in UN vs. 9.66 gU/cm{sup 3} in UO{sub 2}) which is essential for a compact reactor core design. The objective of this LDRD Project is to: (1) Establish a manufacturing capability for uranium-based ceramic nuclear fuel, (2) Develop a computational capability to analyze nuclear fuel performance, (3) Develop a modified UN-based fuel that can support a compact long-life reactor core, and (4) Collaborate with the Nuclear Engineering Department of UC Berkeley on nitride fuel reprocessing and disposal in a geologic repository.

  14. RERTR Program: goals, progress and plans. [Reduced Enrichment Research and Test Reactor

    SciTech Connect (OSTI)

    Travelli, A.

    1984-09-25T23:59:59.000Z

    The status of the US Reduced Enrichment Research and Test Reactor (RERTR) Program is reviewed. After a brief outline of RERTR Program objectives and goals, program accomplishments are discussed with emphasis on the development, demonstration and application of new LEU fuels. Most program activities have proceeded as planned, and a combination of two silicide fuels (U/sub 3/Si/sub 2/-Al and U/sub 3/Si-Al) holds excellent promise for achieving the long-term program goals. Current plans and schedules project the uranium density of qualified RERTR fuels for plate-type reactors to grow by approximately 1 g U/cm/sup 3/ each year, from the nearly null value of 1982 to the 7.0 g U/cm/sup 3/ which will be reached in early 1989. The technical needs of research reactors for HEU exports are also estimated to undergo a gradual but dramatic decline in the coming years.

  15. Depleted uranium hexafluoride: Waste or resource?

    SciTech Connect (OSTI)

    Schwertz, N.; Zoller, J.; Rosen, R.; Patton, S. [Lawrence Livermore National Lab., CA (United States); Bradley, C. [USDOE Office of Nuclear Energy, Science, Technology, Washington, DC (United States); Murray, A. [SAIC (United States)

    1995-07-01T23:59:59.000Z

    the US Department of Energy is evaluating technologies for the storage, disposal, or re-use of depleted uranium hexafluoride (UF{sub 6}). This paper discusses the following options, and provides a technology assessment for each one: (1) conversion to UO{sub 2} for use as mixed oxide duel, (2) conversion to UO{sub 2} to make DUCRETE for a multi-purpose storage container, (3) conversion to depleted uranium metal for use as shielding, (4) conversion to uranium carbide for use as high-temperature gas-cooled reactor (HTGR) fuel. In addition, conversion to U{sub 3}O{sub 8} as an option for long-term storage is discussed.

  16. In situ remediation of uranium contaminated groundwater

    SciTech Connect (OSTI)

    Dwyer, B.P.; Marozas, D.C. [Sandia National Labs., Albuquerque, NM (United States)

    1997-12-31T23:59:59.000Z

    In an effort to develop cost-efficient techniques for remediating uranium contaminated groundwater at DOE Uranium Mill Tailing Remedial Action (UMTRA) sites nationwide, Sandia National Laboratories (SNL) deployed a pilot scale research project at an UMTRA site in Durango, CO. Implementation included design, construction, and subsequent monitoring of an in situ passive reactive barrier to remove Uranium from the tailings pile effluent. A reactive subsurface barrier is produced by emplacing a reactant material (in this experiment - various forms of metallic iron) in the flow path of the contaminated groundwater. Conceptually the iron media reduces and/or adsorbs uranium in situ to acceptable regulatory levels. In addition, other metals such as Se, Mo, and As have been removed by the reductive/adsorptive process. The primary objective of the experiment was to eliminate the need for surface treatment of tailing pile effluent. Experimental design, and laboratory and field preliminary results are discussed with regard to other potential contaminated groundwater treatment applications.

  17. In situ remediation of uranium contaminated groundwater

    SciTech Connect (OSTI)

    Dwyer, B.P.; Marozas, D.C.

    1997-02-01T23:59:59.000Z

    In an effort to develop cost-efficient techniques for remediating uranium contaminated groundwater at DOE Uranium Mill Tailing Remedial Action (UMTRA) sites nationwide, Sandia National Laboratories (SNL) deployed a pilot scale research project at an UMTRA site in Durango, CO. Implementation included design, construction, and subsequent monitoring of an in situ passive reactive barrier to remove Uranium from the tailings pile effluent. A reactive subsurface barrier is produced by emplacing a reactant material (in this experiment various forms of metallic iron) in the flow path of the contaminated groundwater. Conceptually the iron media reduces and/or adsorbs uranium in situ to acceptable regulatory levels. In addition, other metals such as Se, Mo, and As have been removed by the reductive/adsorptive process. The primary objective of the experiment was to eliminate the need for surface treatment of tailing pile effluent. Experimental design, and laboratory and field results are discussed with regard to other potential contaminated groundwater treatment applications.

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

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

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

  19. Depleted Uranium Technical Brief

    E-Print Network [OSTI]

    Depleted Uranium Technical Brief United States Environmental Protection Agency Office of Air and Radiation Washington, DC 20460 EPA-402-R-06-011 December 2006 #12;#12;Depleted Uranium Technical Brief EPA of Radiation and Indoor Air Radiation Protection Division ii #12;iii #12;FOREWARD The Depleted Uranium

  20. DETERMINING OPTICAL CONSTANTS OF URANIUM NITRIDE THIN FILMS IN THE EXTREME

    E-Print Network [OSTI]

    Hart, Gus

    DETERMINING OPTICAL CONSTANTS OF URANIUM NITRIDE THIN FILMS IN THE EXTREME ULTRAVIOLET (1.6-35 NM deposition and characterization of reactively-sputtered uranium nitride thin films. I also report optical.1 Application 1 1.2 Optical Constants 2 1.3 Project Focus 7 2 Uranium Nitride Thin Films 8 2.1 Sputtering 8 2

  1. High Purity Germanium Gamma-PHA Assay of Uranium Storage Pigs for 321-M Facility

    SciTech Connect (OSTI)

    Dewberry, R.A.

    2001-09-18T23:59:59.000Z

    The Analytical Development Section of SRTC was requested by the Facilities Disposition Division (FDD) to determine the holdup of enriched uranium in the 321-M facility as part of an overall deactivation project of the facility. The 321-M facility was used to fabricate enriched uranium fuel assemblies, lithium-aluminum target tubes, neptunium assemblies, and miscellaneous components for the production reactors. The facility also includes the 324-M storage building and the passageway connecting it to 321-M. The results of the holdup assays are essential for determining compliance with the Solid Waste's Waste Acceptance Criteria, Material Control and Accountability, and to meet criticality safety controls. This report describes and documents the use of a portable HPGe detector and EG and G Dart system that contains a high voltage power supply, signal processing electronics, a personal computer with Gamma-Vision software, and space to store and manipulate multiple 4096-channel g-ray spectra to assay for 235U content in 268 uranium shipping and storage pigs. This report includes a description of three efficiency calibration configurations and also the results of the assay. A description of the quality control checks is included as well.

  2. E-Print Network 3.0 - assessing uranium bioremediation Sample...

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

    Mountain Project, US EPA Collection: Environmental Sciences and Ecology 75 Mathematical Geology, Vol. 33, No. 1, 2001 Modeling Uranium Transport in Koongarra, Summary: waste...

  3. Developing fuel management capabilities based on coupled Monte Carlo depletion in support of the MIT Research Reactor (MITR) conversion .

    E-Print Network [OSTI]

    Romano, Paul K. (Paul Kollath)

    2009-01-01T23:59:59.000Z

    ??Pursuant to a 1986 NRC ruling, the MIT Reactor (MITR) is planning on converting from the use of highly enriched uranium (HEU) to low enriched (more)

  4. P Reactor Grouting

    SciTech Connect (OSTI)

    None

    2010-01-01T23:59:59.000Z

    Filling the P Reactor with grout. This seals the radioactive material and reduces the environmental footprint left from the Cold War. Project sponsored by the Recovery Act at the Savannah River Site.

  5. Documented Safety Analysis Addendum for the Neutron Radiography Reactor Facility Core Conversion

    SciTech Connect (OSTI)

    Boyd D. Christensen

    2009-05-01T23:59:59.000Z

    The Neutron Radiography Reactor Facility (NRAD) is a Training, Research, Isotope Production, General Atomics (TRIGA) reactor which was installed in the Idaho National Laboratory (INL) Hot Fuels Examination Facility (HFEF) at the Materials and Fuels Complex (MFC) in the mid 1970s. The facility provides researchers the capability to examine both irradiated and non-irradiated materials in support of reactor fuel and components programs through non-destructive neutron radiography examination. The facility has been used in the past as one facet of a suite of reactor fuels and component examination facilities available to researchers at the INL and throughout the DOE complex. The facility has also served various commercial research activities in addition to the DOE research and development support. The reactor was initially constructed using Fuel Lifetime Improvement Program (FLIP)- type highly enriched uranium (HEU) fuel obtained from the dismantled Puerto Rico Nuclear Center (PRNC) reactor. In accordance with international non-proliferation agreements, the NRAD core will be converted to a low enriched uranium (LEU) fuel and will continue to utilize the PRNC control rods, control rod drives, startup source, and instrument console as was previously used with the HEU core. The existing NRAD Safety Analysis Report (SAR) was created and maintained in the preferred format of the day, combining sections of both DOE-STD-3009 and Nuclear Regulatory Commission Regulatory Guide 1.70. An addendum was developed to cover the refueling and reactor operation with the LEU core. This addendum follows the existing SAR format combining required formats from both the DOE and NRC. This paper discusses the project to successfully write a compliant and approved addendum to the existing safety basis documents.

  6. DOE Selects Contractor for Depleted Hexafluoride Conversion Project...

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

    to the DOE Portsmouth Paducah Project Office (PPPO) in Lexington, Kentucky and the Depleted Uranium Hexafluoride (DUF6) Conversion Project in Paducah, Kentucky and...

  7. Method for converting uranium oxides to uranium metal

    DOE Patents [OSTI]

    Duerksen, Walter K. (Norris, TN)

    1988-01-01T23:59:59.000Z

    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.

  8. Optical Constants ofOptical Constants of Uranium Nitride Thin FilmsUranium Nitride Thin Films

    E-Print Network [OSTI]

    Hart, Gus

    Optical Constants ofOptical Constants of Uranium Nitride Thin FilmsUranium Nitride Thin FilmsDelta--Beta Scatter Plot at 220 eVBeta Scatter Plot at 220 eV #12;Why Uranium Nitride?Why Uranium Nitride? UraniumUranium, uranium,Bombard target, uranium, with argon ionswith argon ions Uranium atoms leaveUranium atoms leave

  9. The use of burnable poison to improve uranium utilization in PWRs

    E-Print Network [OSTI]

    Loh, Wee Tee

    1982-01-01T23:59:59.000Z

    A methodology based on the linear reactivity model of core behavior has been developed and employed to evaluate fuel management tactics for improving uranium utilization in Pressurized Water Reactors in a once-through fuel ...

  10. Portsmouth Paducah Project Office

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

    being returned to inventory for reuse or resale. As part of the project, cylinders of uranium hexafluoride (UF6) feed material were inspected and pre-sampled to determine Tc-99...

  11. http://arXiv.org/physics/0507088 Teaching About Nature's Nuclear Reactors

    E-Print Network [OSTI]

    Learned, John

    http://arXiv.org/physics/0507088 Teaching About Nature's Nuclear Reactors J. Marvin Herndon reactors existed in uranium deposits on Earth long before Enrico Fermi built the first man-made nuclear reactors. The subject of planetocentric nuclear fission reactors can be a jumping off point for stimulating

  12. Expanding and optimizing fuel management and data analysis capabilities of MCODE-FM in support of MIT research reactor (MITR-II) LEU conversion

    E-Print Network [OSTI]

    Horelik, Nicholas E. (Nicholas Edward)

    2012-01-01T23:59:59.000Z

    Studies are underway in support of the MIT research reactor (MITR-II) conversion from high enriched Uranium (HEU) to low enriched Uranium (LEU), as required by recent non-proliferation policy. With the same core configuration ...

  13. Welding of uranium and uranium alloys

    SciTech Connect (OSTI)

    Mara, G.L.; Murphy, J.L.

    1982-03-26T23:59:59.000Z

    The major reported work on joining uranium comes from the USA, Great Britain, France and the USSR. The driving force for producing this technology base stems from the uses of uranium as a nuclear fuel for energy production, compact structures requiring high density, projectiles, radiation shielding, and nuclear weapons. This review examines the state-of-the-art of this technology and presents current welding process and parameter information. The welding metallurgy of uranium and the influence of microstructure on mechanical properties is developed for a number of the more commonly used welding processes.

  14. Lessons Learned from the Application of Bulk Characterization to Individual Containers on the Brookhaven Graphite Research Reactor Decommissioning Project at Brookhaven National Laboratory - 12056

    SciTech Connect (OSTI)

    Kneitel, Terri [US DOE, Brookhaven Site Office (United States); Rocco, Diane [Brookhaven National Laboratory (United States)

    2012-07-01T23:59:59.000Z

    When conducting environmental cleanup or decommissioning projects, characterization of the material to be removed is often performed when the material is in-situ. The actual demolition or excavation and removal of the material can result in individual containers that vary significantly from the original bulk characterization profile. This variance, if not detected, can result in individual containers exceeding Department of Transportation regulations or waste disposal site acceptance criteria. Bulk waste characterization processes were performed to initially characterize the Brookhaven Graphite Research Reactor (BGRR) graphite pile and this information was utilized to characterize all of the containers of graphite. When the last waste container was generated containing graphite dust from the bottom of the pile, but no solid graphite blocks, the material contents were significantly different in composition from the bulk waste characterization. This error resulted in exceedance of the disposal site waste acceptance criteria. Brookhaven Science Associates initiated an in-depth investigation to identify the root causes of this failure and to develop appropriate corrective actions. The lessons learned at BNL have applicability to other cleanup and demolition projects which characterize their wastes in bulk or in-situ and then extend that characterization to individual containers. (authors)

  15. EPA Update: NESHAP Uranium Activities

    E-Print Network [OSTI]

    for underground uranium mining operations (Subpart B) EPA regulatory requirements for operating uranium mill for Underground Uranium Mining Operations (Subpart B) #12;5 EPA Regulatory Requirements for Underground Uranium uranium mines include: · Applies to 10,000 tons/yr ore production, or 100,000 tons/mine lifetime · Ambient

  16. Uranium hexafluoride public risk

    SciTech Connect (OSTI)

    Fisher, D.R.; Hui, T.E.; Yurconic, M.; Johnson, J.R.

    1994-08-01T23:59:59.000Z

    The limiting value for uranium toxicity in a human being should be based on the concentration of uranium (U) in the kidneys. The threshold for nephrotoxicity appears to lie very near 3 {mu}g U per gram kidney tissue. There does not appear to be strong scientific support for any other improved estimate, either higher or lower than this, of the threshold for uranium nephrotoxicity in a human being. The value 3 {mu}g U per gram kidney is the concentration that results from a single intake of about 30 mg soluble uranium by inhalation (assuming the metabolism of a standard person). The concentration of uranium continues to increase in the kidneys after long-term, continuous (or chronic) exposure. After chronic intakes of soluble uranium by workers at the rate of 10 mg U per week, the concentration of uranium in the kidneys approaches and may even exceed the nephrotoxic limit of 3 {mu}g U per gram kidney tissue. Precise values of the kidney concentration depend on the biokinetic model and model parameters assumed for such a calculation. Since it is possible for the concentration of uranium in the kidneys to exceed 3 {mu}g per gram tissue at an intake rate of 10 mg U per week over long periods of time, we believe that the kidneys are protected from injury when intakes of soluble uranium at the rate of 10 mg U per week do not continue for more than two consecutive weeks. For long-term, continuous occupational exposure to low-level, soluble uranium, we recommend a reduced weekly intake limit of 5 mg uranium to prevent nephrotoxicity in workers. Our analysis shows that the nephrotoxic limit of 3 {mu}g U per gram kidney tissues is not exceeded after long-term, continuous uranium intake at the intake rate of 5 mg soluble uranium per week.

  17. Fast reactors and nuclear nonproliferation

    SciTech Connect (OSTI)

    Avrorin, E.N. [Russian Federal Nuclear Center - Zababakhin Institute of Applied Physics, Snezhinsk (Russian Federation); Rachkov, V.I.; Chebeskov, A.N. [State Scientific Center of the Russian Federation - Institute for Physics and Power Engineering, Bondarenko Square, 1, Obninsk, Kaluga region, 249033 (Russian Federation)

    2013-07-01T23:59:59.000Z

    Problems are discussed with regard to nuclear fuel cycle resistance in fast reactors to nuclear proliferation risk due to the potential for use in military programs of the knowledge, technologies and materials gained from peaceful nuclear power applications. Advantages are addressed for fast reactors in the creation of a more reliable mode of nonproliferation in the closed nuclear fuel cycle in comparison with the existing fully open and partially closed fuel cycles of thermal reactors. Advantages and shortcomings are also discussed from the point of view of nonproliferation from the start with fast reactors using plutonium of thermal reactor spent fuel and enriched uranium fuel to the gradual transition using their own plutonium as fuel. (authors)

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

    DOE Patents [OSTI]

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

    1995-06-06T23:59:59.000Z

    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.

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

    DOE Patents [OSTI]

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

    1995-01-01T23:59:59.000Z

    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.

  20. Preparation of uranium compounds

    DOE Patents [OSTI]

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

    2013-02-19T23:59:59.000Z

    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.

  1. Environmental analysis and data report prepared for the environmental assessment of remedial action at the Lowman uranium mill tailings site near Lowman, Idaho. [Urnanium Mill Tailings Remedial Action (UMTRA) Project

    SciTech Connect (OSTI)

    Not Available

    1991-01-01T23:59:59.000Z

    This document contains information and data gathered in support of the preparation of the environmental assessment (EA) of the proposed remedial action at the Uranium Mill Tailings Remedial Action (UMTRA) Project site near Lowman, Idaho. The Lowman EA was prepared pursuant to the National Environmental Policy Act (NEPA), which requires Federal agencies to assess the effects of their actions on the environment. It examines the short-term and the long-term effects of the US Department of Energy's (DOE) proposed remedial action for the Lowman site as well as the no action alternative. The DOE will use the information and analyses presented in the EA to determine whether the proposed action would have a significant impact on the environment. If the impacts are determined to be significant, an environmental impact statement will be prepared. If the impacts are not judged to be significant, the DOE may issue a Finding of No Significant Impact and implement the proposed action. The information and data presented in this environmental analyses and data report are for background purposes only and are not required as part of the NEPA decision-making process.

  2. University Reactor Conversion Lessons Learned Workshop for Purdue University Reactor

    SciTech Connect (OSTI)

    Eric C. Woolstenhulme; Dana M. Hewit

    2008-09-01T23:59:59.000Z

    The Department of Energys Idaho National Laboratory, under its programmatic responsibility for managing the University Research Reactor Conversions, has completed the conversion of the reactor at Purdue University Reactor. With this work completed and in anticipation of other impending conversion projects, the INL convened and engaged the project participants in a structured discussion to capture the lessons learned. The lessons learned process has allowed us to capture gaps, opportunities, and good practices, drawing from the project teams experiences. These lessons will be used to raise the standard of excellence, effectiveness, and efficiency in all future conversion projects.

  3. MANHATTAN PROJECT NATIONAL HISTORICAL PARK | Department of Energy

    Energy Savers [EERE]

    PARK MANHATTAN PROJECT NATIONAL HISTORICAL PARK Shift change at the Y-12 facility in Oak Ridge, TN, where uranium-235 was separated from uranium-238. August 1945. Shift change...

  4. Uranium industry annual 1993

    SciTech Connect (OSTI)

    Not Available

    1994-09-01T23:59:59.000Z

    Uranium production in the United States has declined dramatically from a peak of 43.7 million pounds U{sub 3}O{sub 8} (16.8 thousand metric tons uranium (U)) in 1980 to 3.1 million pounds U{sub 3}O{sub 8} (1.2 thousand metric tons U) in 1993. This decline is attributed to the world uranium market experiencing oversupply and intense competition. Large inventories of uranium accumulated when optimistic forecasts for growth in nuclear power generation were not realized. The other factor which is affecting U.S. uranium production is that some other countries, notably Australia and Canada, possess higher quality uranium reserves that can be mined at lower costs than those of the United States. Realizing its competitive advantage, Canada was the world`s largest producer in 1993 with an output of 23.9 million pounds U{sub 3}O{sub 8} (9.2 thousand metric tons U). The U.S. uranium industry, responding to over a decade of declining market prices, has downsized and adopted less costly and more efficient production methods. The main result has been a suspension of production from conventional mines and mills. Since mid-1992, only nonconventional production facilities, chiefly in situ leach (ISL) mining and byproduct recovery, have operated in the United States. In contrast, nonconventional sources provided only 13 percent of the uranium produced in 1980. ISL mining has developed into the most cost efficient and environmentally acceptable method for producing uranium in the United States. The process, also known as solution mining, differs from conventional mining in that solutions are used to recover uranium from the ground without excavating the ore and generating associated solid waste. This article describes the current ISL Yang technology and its regulatory approval process, and provides an analysis of the factors favoring ISL mining over conventional methods in a declining uranium market.

  5. Studies of epithermal neutrons in uranium, heavy water lattices

    E-Print Network [OSTI]

    D'Ardenne, Walter Herbert

    1964-01-01T23:59:59.000Z

    Measurements related to reactor physics parameters were made in three heavy water lattices. The three lattices studied consisted of 0.250-inch-diameter, 1.03 w/o U2 3 5 uranium fuel rods arranged in triangular arrays and ...

  6. Candidate processes for diluting the {sup 235}U isotope in weapons-capable highly enriched uranium

    SciTech Connect (OSTI)

    Snider, J.D.

    1996-02-01T23:59:59.000Z

    The United States Department of Energy (DOE) is evaluating options for rendering its surplus inventories of highly enriched uranium (HEU) incapable of being used to produce nuclear weapons. Weapons-capable HEU was earlier produced by enriching uranium in the fissile {sup 235}U isotope from its natural occurring 0.71 percent isotopic concentration to at least 20 percent isotopic concentration. Now, by diluting its concentration of the fissile {sup 235}U isotope in a uranium blending process, the weapons capability of HEU can be eliminated in a manner that is reversible only through isotope enrichment, and therefore, highly resistant to proliferation. To the extent that can be economically and technically justified, the down-blended uranium product will be made suitable for use as commercial reactor fuel. Such down-blended uranium product can also be disposed of as waste if chemical or isotopic impurities preclude its use as reactor fuel.

  7. The Ambrosia Lake project archaeological investigations of three small sites associated with the southern Chacoan outlier of Kin Nizhoni, McKinley County, New Mexico. [Uranium Mill Tailings Remedial Action (UMTRA) Project

    SciTech Connect (OSTI)

    Cullington, B.J. (ed.); Hammack, L.C.; Baugh, T.G. (Complete Archaeological Service Associates, Cortez, CO (United States))

    1990-03-15T23:59:59.000Z

    During the fall of 1987, Complete Archaeological Service Associates conducted mitigative excavations at three sites (LA50363, LA50364, and LA50371) in McKinley County, New Mexico. These sites are adjacent to the Phillips/United Nuclear Inactive Uranium Mill and Tailings site at Ambrosia Lake, New Mexico. The primary deposition at each of these sites appears to be related to a Pueblo II or Bonito Phase occupation. Temporal placement is based primarily on the cross dating of ceramics and archaeomagnetic determinations when possible. No tree-ring or radiocarbon samples are available from these sites. These Ambrosia Lake sites indicate that this area was occupied primarily by Pueblo II people who may have had close social, economic, and ceremonial ties with the people living at the nuclear community of Lower Nizhoni about 3 km south-southeast. The later component at LA50364 indicates a Pueblo III occupation by people who may have had similar ties to the people of the Kin Nizhoni nuclear community. The Ambrosia Lake sites, then, provide important information on the structure of subnuclear communities within the southern Chaco periphery.

  8. Analytical electron microscopy characterization of uranium-contaminated soils from the Fernald Site, FY1993 report

    SciTech Connect (OSTI)

    Buck, E.C.; Cunnane, J.C.; Brown, N.R.; Dietz, N.L.

    1994-10-01T23:59:59.000Z

    A combination of optical microscopy, scanning electron microscopy with backscattered electron detection (SEM/BSE), and analytical electron microscopy (AEM) is being used to determine the nature of uranium in soils from the Fernald Environmental Management Project. The information gained from these studies is being used to develop and test remediation technologies. Investigations using SEM have shown that uranium is contained within particles that are typically 1 to 100 {mu}m in diameter. Further analysis with AEM has shown that these uranium-rich regions are made up of discrete uranium-bearing phases. The distribution of these uranium phases was found to be inhomogeneous at the microscopic level.

  9. REFERENCES CITED IN: Draft Technical Report: Considerations Related to Post-Closure Monitoring Of Uranium In-Situ

    E-Print Network [OSTI]

    , 2008, Cogema Mining, Inc, and Petrotek Engineering Corp. ML081060131. http://www.wise- uranium://www.wiley.com/WileyCDA/ Crow Butte 2000. Mine Unit 1 Restoration Report, Crow Butte Uranium Project. Submitted to NRC January Of Uranium In-Situ Leach/In-Situ Recovery (ISL/ISR) Sites. How to obtain the references: Most

  10. CHARACTERIZATION OF URANIUM, URANIUM OXIDE AND SILICON MULTILAYER THIN FILMS

    E-Print Network [OSTI]

    Hart, Gus

    CHARACTERIZATION OF URANIUM, URANIUM OXIDE AND SILICON MULTILAYER THIN FILMS by David T. Oliphant. Woolley Dean, College of Physical and Mathematical Sciences #12;ABSTRACT CHARACTERIZATION OF URANIUM, URANIUM OXIDE AND SILICON MULTILAYER THIN FILMS David T. Oliphant Department of Physics and Astronomy

  11. Reactor Physics

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

    Reactor Physics Reactor and nuclear physics is a key area of research at INL. Much of the research done in reactor physics can be separated into one of three categories:...

  12. Fast Reactor Spent Fuel Processing: Experience and Criticality Safety

    SciTech Connect (OSTI)

    Chad Pope

    2007-05-01T23:59:59.000Z

    This paper discusses operational and criticality safety experience associated with the Idaho National Laboratory Fuel Conditioning Facility which uses a pyrometallurgical process to treat spent fast reactor metallic fuel. The process is conducted in an inert atmosphere hot cell. The process starts with chopping metallic fuel elements into a basket. The basket is lowered into molten salt (LiCl-KCl) along with a steel mandrel. Active metal fission products, transuranic metals and sodium metal in the spent fuel undergo chemical oxidation and form chlorides. Voltage is applied between the basket, which serves as an anode, and the mandrel, which serves as a cathode, causing metallic uranium in the spent fuel to undergo electro-chemical oxidation thereby forming uranium chloride. Simultaneously at the cathode, uranium chloride undergoes electro-chemical reduction and deposits uranium metal onto the mandrel. The uranium metal and accompanying entrained salt are placed in a distillation furnace where the uranium melts forming an ingot and the entrained salt boils and subsequently condenses in a separate crucible. The uranium ingots are placed in long term storage. During the ten year operating history, over one hundred criticality safety evaluations were prepared. All criticality safety related limits and controls for the entire process are contained in a single document which required over thirty revisions to accommodate the process changes. Operational implementation of the limits and controls includes use of a near real-time computerized tracking system. The tracking system uses an Oracle database coupled with numerous software applications. The computerized tracking system includes direct fuel handler interaction with every movement of material. Improvements to this system during the ten year history include introduction of web based operator interaction, tracking of moderator materials and the development of a plethora database queries to assist in day to day operations as well as obtaining historical information. Over 12,000 driver fuel elements have been processed resulting in the production of 2500 kg of 20% enriched uranium. Also, over one thousand blanket fuel elements have been processed resulting in the production of 2400 kg of depleted uranium. These operations required over 35,000 fissile material transfers between zones and over 6000 transfers between containers. Throughout all of these movements, no mass limit violations occurred. Numerous lessons were learned over the ten year operating history. From a criticality safety perspective, the most important lesson learned was the involvement of a criticality safety practitioner in daily operations. A criticality safety engineer was assigned directly to facility operations, and was responsible for implementation of limits and controls including upkeep of the associated computerized tracking files. The criticality safety engineer was also responsible for conducting fuel handler training activities including serving on fuel handler qualification oral boards, and continually assessing operations from a criticality control perspective. The criticality safety engineer also attended bimonthly project planning meetings to identify upcoming process changes that would require criticality safety evaluation. Finally, the excellent criticality safety record was due in no small part to the continual support, involvement, trust, and confidence of project and operations mana

  13. Corrosion Minimization for Research Reactor Fuel

    SciTech Connect (OSTI)

    Eric Shaber; Gerard Hofman

    2005-06-01T23:59:59.000Z

    Existing university research reactors are being converted to use low-enriched uranium fue to eliminate the use of highly-enriched uranium. These conversions require increases in fuel loading that will result in the use of elements with more fuel plates, resulting in a net decrease in the water annulus between fuel plates. The proposed decrease in the water annulus raises questions about the requirements and stability of the surface hydroxide on the aluminum fuel cladding and the potential for runaway corrosion resulting in fuel over-temperature incidents. The Nuclear Regulatory Commission (NRC), as regulator for these university reactors, must ensure that proposed fuel modifications will not result in any increased risk or hazard to the reactor operators or the public. This document reviews the characteristics and behavior of aluminum hydroxides, analyzes the drivers for fuel plate corrosion, reviews relevant historical incidents, and provides recommendations on fuel design, surface treatment, and reactor operational practices to avoid corrosion issues.

  14. The Fast-spectrum Transmutation Experimental Facility FASTEF: Main design achievements (part 2: Reactor building design and plant layout) within the FP7-CDT collaborative project of the European Commission

    SciTech Connect (OSTI)

    De Bruyn, D.; Engelen, J. [Belgian Nuclear Research Centre SCK CEN, Boeretang 200, 2400 Mol (Belgium); Ortega, A.; Aguado, M. P. [Empresarios Agrupados A.I.E., Magallanes 3, 28015 Madrid (Spain)

    2012-07-01T23:59:59.000Z

    MYRRHA (Multi-purpose hybrid Research Reactor for High-tech Applications) is the flexible experimental accelerator-driven system (ADS) in development at SCK-CEN in replacement of its material testing reactor BR2. SCK-CEN in association with 17 European partners from industry, research centres and academia, responded to the FP7 (Seventh Framework Programme) call from the European Commission to establish a Central Design Team (CDT) for the design of a Fast Spectrum Transmutation Experimental Facility (FASTEF) able to demonstrate efficient transmutation and associated technology through a system working in subcritical and/or critical mode. The project has started on April 01, 2009 for a period of three years. In this paper, we present the latest concept of the reactor building and the plant layout. The FASTEF facility has evolved quite a lot since the intermediate reporting done at the ICAPP'10 and ICAPP'11 conferences 1,2. Many iterations have been performed to take into account the safety requirements. The present configuration enables an easy operation and maintenance of the facility, including the possibility to change large components of the reactor. In a companion paper 3, we present the latest configuration of the reactor core and primary system. (authors)

  15. Teaching About Nature's Nuclear Reactors

    E-Print Network [OSTI]

    Herndon, J M

    2005-01-01T23:59:59.000Z

    Naturally occurring nuclear reactors existed in uranium deposits on Earth long before Enrico Fermi built the first man-made nuclear reactor beneath Staggs Field in 1942. In the story of their discovery, there are important lessons to be learned about scientific inquiry and scientific discovery. Now, there is evidence to suggest that the Earth's magnetic field and Jupiter's atmospheric turbulence are driven by planetary-scale nuclear reactors. The subject of planetocentric nuclear fission reactors can be a jumping off point for stimulating classroom discussions about the nature and implications of planetary energy sources and about the geomagnetic field. But more importantly, the subject can help to bring into focus the importance of discussing, debating, and challenging current thinking in a variety of areas.

  16. Thermal analysis of uranium zirconium hydride fuel using a lead-bismuth gap at LWR operating temperatures

    E-Print Network [OSTI]

    Ensor, Brendan M. (Brendan Melvin)

    2012-01-01T23:59:59.000Z

    Next generation nuclear technology calls for more advanced fuels to maximize the effectiveness of new designs. A fuel currently being studied for use in advanced light water reactors (LWRs) is uranium zirconium hydride ...

  17. POTENTIAL BENCHMARKS FOR ACTINIDE PRODUCTION IN HANFORD REACTORS

    SciTech Connect (OSTI)

    PUIGH RJ; TOFFER H

    2011-10-19T23:59:59.000Z

    A significant experimental program was conducted in the early Hanford reactors to understand the reactor production of actinides. These experiments were conducted with sufficient rigor, in some cases, to provide useful information that can be utilized today in development of benchmark experiments that may be used for the validation of present computer codes for the production of these actinides in low enriched uranium fuel.

  18. Depleted uranium management alternatives

    SciTech Connect (OSTI)

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

    1994-08-01T23:59:59.000Z

    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.

  19. Proceedings of Workshop on Uranium Production Environmental Restoration: An exchange between the United States and Germany

    SciTech Connect (OSTI)

    Not Available

    1993-12-31T23:59:59.000Z

    Scientists, engineers, elected officials, and industry regulators from the United, States and Germany met in Albuquerque, New Mexico, August 16--20, 1993, in the first joint international workshop to discuss uranium tailings remediation. Entitled ``Workshop on Uranium Production Environmental Restoration: An Exchange between the US and Germany,`` the meeting was hosted by the US Department of Energy`s (DOE) Uranium Mill Tailings Remedial Action (UMTRA) Project. The goal of the workshop was to further understanding and communication on the uranium tailings cleanup projects in the US and Germany. Many communities around the world are faced with an environmental legacy -- enormous quantities of hazardous and low-level radioactive materials from the production of uranium used for energy and nuclear weapons. In 1978, the US Congress passed the Uranium Mill Tailings Radiation Control Act. Title I of the law established a program to assess the tailings at inactive uranium processing sites and provide a means for joint federal and state funding of the cleanup efforts at sites where all or substantially all of the uranium was produced for sale to a federal agency. The UMTRA Project is responsible for the cleanup of 24 sites in 10 states. Germany is facing nearly identical uranium cleanup problems and has established a cleanup project. At the workshop, participants had an opportunity to interact with a broad cross section of the environmental restoration and waste disposal community, discuss common concerns and problems, and develop a broader understanding of the issues. Abstracts are catalogued individually for the data base.

  20. The use of reduced-moderation light water reactors for transuranic isotope burning in thorium fuel

    E-Print Network [OSTI]

    Lindley, Benjamin A.

    2015-02-03T23:59:59.000Z

    Light water reactors (LWRs) are the worlds dominant nuclear reactor system. Uranium (U)-fuelled LWRs produce long-lived transuranic (TRU) isotopes. TRUs can be recycled in LWRs or fast reactors. The thermal neutron spectrum in LWRs is less suitable...

  1. Uranium series disequilibrium in the Bargmann property area of Karnes County, Texas

    SciTech Connect (OSTI)

    Davidson, J.R.

    1998-02-01T23:59:59.000Z

    Historical evidence is presented for natural uranium series radioactive disequilibrium in uranium bearing soils in the Bargmann property area of karnes County on the Gulf Coastal Plain of south Texas. The early history of uranium exploration in the area is recounted and records of disequilibrium before milling and mining operations began are given. The property contains an open pit uranium mine associated with a larger ore body. In 1995, the US Department of Energy (DOE) directed Oak Ridge National Laboratory (ORNL) to evaluate the Bargmann tract for the presence of uranium mill tailings (ORNL 1996). There was a possibility that mill tailings had washed onto or blown onto the property from the former tailings piles in quantities that would warrant remediation under the Uranium Mill Tailings Remediation Action Project. Activity ratios illustrating disequilibrium between {sup 226}Ra and {sup 238}U in background soils during 1986 are listed and discussed. Derivations of uranium mass-to-activity conversion factors are covered in detail.

  2. IPNS enriched uranium booster target

    SciTech Connect (OSTI)

    Schulke, A.W. Jr.

    1985-01-01T23:59:59.000Z

    Since startup in 1981, IPNS has operated on a fully depleted /sup 238/U target. With the booster as in the present system, high energy protons accelerated to 450 MeV by the Rapid Cycling Synchrotron are directed at the target and by mechanisms of spallation and fission of the uranium, produce fast neutrons. The neutrons from the target pass into adjacent moderator where they slow down to energies useful for spectroscopy. The target cooling systems and monitoring systems have operated very reliably and safely during this period. To provide higher neutron intensity, we have developed plans for an enriched uranium (booster) target. HETC-VIM calculations indicate that the target will produce approx.90 kW of heat, with a nominal x5 gain (k/sub eff/ = 0.80). The neutron beam intensity gain will be a factor of approx.3. Thermal-hydraulic and heat transport calculations indicate that approx.1/2 in. thick /sup 235/U discs are subject to about the same temperatures as the present /sup 238/U 1 in. thick discs. The coolant will be light demineralized water (H/sub 2/O) and the coolant flow rate must be doubled. The broadening of the fast neutron pulse width should not seriously affect the neutron scattering experiments. Delayed neutrons will appear at a level about 3% of the total (currently approx.0.5%). This may affect backgrounds in some experiments, so that we are assessing measures to control and correct for this (e.g., beam tube choppers). Safety analyses and neutronic calculations are nearing completion. Construction of the /sup 235/U discs at the ORNL Y-12 facility is scheduled to begin late 1985. The completion of the booster target and operation are scheduled for late 1986. No enriched uranium target assembly operating at the projected power level now exists in the world. This effort thus represents an important technological experiment as well as being a ''flux enhancer''.

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

    DOE Patents [OSTI]

    Kreuzmann, Alvin B. (Cincinnati, OH)

    1983-01-01T23:59:59.000Z

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

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

    DOE Patents [OSTI]

    Kreuzmann, A.B.

    1982-10-27T23:59:59.000Z

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

  5. Heavy Water Test Reactor Dome Removal

    SciTech Connect (OSTI)

    None

    2011-01-01T23:59:59.000Z

    A high speed look at the removal of the Heavy Water Test Reactor Dome Removal. A project sponsored by the Recovery Act on the Savannah River Site.

  6. Light Water Reactor Sustainability Program Contact Information

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

    Program Organization LWRS Program Management Richard Reister Federal Project Director Light Water Reactor Deployment Office of Nuclear Energy U.S. Department of Energy...

  7. Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production, Nuclear Energy Research Initiative Project 2001-001, Westinghouse Electric Co. Grant Number: DE-FG07-02SF22533, Final Report

    SciTech Connect (OSTI)

    Philip E. MacDonald

    2005-01-01T23:59:59.000Z

    The supercritical water-cooled reactor (SCWR) is one of the six reactor technologies selected for research and development under the Generation IV program. SCWRs are promising advanced nuclear systems because of their high thermal efficiency (i.e., about 45% versus about 33% efficiency for current Light Water Reactors [LWRs]) and considerable plant simplification. SCWRs are basically LWRs operating at higher pressure and temperatures with a direct once-through cycle. Operation above the critical pressure eliminates coolant boiling, so the coolant remains single-phase throughout the system. Thus, the need for a pressurizer, steam generators, steam separators, and dryers is eliminated. The main mission of the SCWR is generation of low-cost electricity. It is built upon two proven technologies: LWRs, which are the most commonly deployed power generating reactors in the world, and supercritical fossil-fired boilers, a large number of which are also in use around the world. The reference SCWR design for the U.S. program is a direct cycle system operating at 25.0 MPa, with core inlet and outlet temperatures of 280 and 500 C, respectively. The coolant density decreases from about 760 kg/m3 at the core inlet to about 90 kg/m3 at the core outlet. The inlet flow splits with about 10% of the inlet flow going down the space between the core barrel and the reactor pressure vessel (the downcomer) and about 90% of the inlet flow going to the plenum at the top of the rector pressure vessel, to then flow down through the core in special water rods to the inlet plenum. Here it mixes with the feedwater from the downcomer and flows upward to remove the heat in the fuel channels. This strategy is employed to provide good moderation at the top of the core. The coolant is heated to about 500 C and delivered to the turbine. The purpose of this NERI project was to assess the reference U.S. Generation IV SCWR design and explore alternatives to determine feasibility. The project was organized into three tasks: Task 1. Fuel-cycle Neutronic Analysis and Reactor Core Design Task 2. Fuel Cladding and Structural Material Corrosion and Stress Corrosion Cracking Task 3. Plant Engineering and Reactor Safety Analysis. moderator rods. materials.

  8. Standard specification for sintered gadolinium oxide-uranium dioxide pellets

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2008-01-01T23:59:59.000Z

    1.1 This specification is for finished sintered gadolinium oxide-uranium dioxide pellets for use in light-water reactors. It applies to gadolinium oxide-uranium dioxide pellets containing uranium of any 235U concentration and any concentration of gadolinium oxide. 1.2 This specification recognizes the presence of reprocessed uranium in the fuel cycle and consequently defines isotopic limits for gadolinium oxide-uranium dioxide pellets made from commercial grade UO2. Such commercial grade UO2 is defined so that, regarding fuel design and manufacture, the product is essentially equivalent to that made from unirradiated uranium. UO2 falling outside these limits cannot necessarily be regarded as equivalent and may thus need special provisions at the fuel fabrication plant or in the fuel design. 1.3 This specification does not include (1) provisions for preventing criticality accidents or (2) requirements for health and safety. Observance of this specification does not relieve the user of the obligation to be aw...

  9. Innovative Elution Processes for Recovering Uranium from Seawater

    SciTech Connect (OSTI)

    Wai, Chien; Tian, Guoxin; Janke, Christopher

    2014-05-29T23:59:59.000Z

    Utilizing amidoxime-based polymer sorbents for extraction of uranium from seawater has attracted considerable interest in recent years. Uranium collected in the sorbent is recovered typically by elution with an acid. One drawback of acid elution is deterioration of the sorbent which is a significant factor that limits the economic competitiveness of the amidoxime-based sorbent systems for sequestering uranium from seawater. Developing innovative elution processes to improve efficiency and to minimize loss of sorbent capacity become essential in order to make this technology economically feasible for large-scale industrial applications. This project has evaluated several elution processes including acid elution, carbonate elution, and supercritical fluid elution for recovering uranium from amidoxime-based polymer sorbents. The elution efficiency, durability and sorbent regeneration for repeated uranium adsorption- desorption cycles in simulated seawater have been studied. Spectroscopic techniques are used to evaluate chemical nature of the sorbent before and after elution. A sodium carbonate-hydrogen peroxide elution process for effective removal of uranium from amidoxime-based sorbent is developed. The cause of this sodium carbonate and hydrogen peroxide synergistic leaching of uranium from amidoxime-based sorbent is attributed to the formation of an extremely stable uranyl peroxo-carbonato complex. The efficiency of uranium elution by the carbonate-hydrogen peroxide method is comparable to that of the hydrochloric acid elution but damage to the sorbent material is much less for the former. The carbonate- hydrogen peroxide elution also does not need any elaborate step to regenerate the sorbent as those required for hydrochloric acid leaching. Several CO2-soluble ligands have been tested for extraction of uranium from the sorbent in supercritical fluid carbon dioxide. A mixture of hexafluoroacetylacetone and tri-n-butylphosphate shows the best result but uranium removal from the sorbent reaches only 80% after 10 hours of leaching. Some information regarding coordination of vanadium with amidoxime molecules and elution of vanadium from amidoxime- based sorbents is also given in the report.

  10. Reactor Cost Analysis Brian James

    E-Print Network [OSTI]

    Reactor Cost Analysis Brian James Directed Technologies, Inc. 6-7 November 2007 This presentation specification & optimization · Capital cost estimation · Projected hydrogen $/kg #12;Directed Technologies, Inc/WGS Membrane Reactor OTM/ Water-Splitting ANL With WGS #12;Directed Technologies, Inc. 6-7 November 2007 BILIWG

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

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

    Uranium Enrichment Decontamination and Decommissioning Fund's Fiscal Year 2008 and 2007 Financial Statement Audit Uranium Enrichment Decontamination and Decommissioning Fund's...

  12. Research Program of a Super Fast Reactor

    SciTech Connect (OSTI)

    Oka, Yoshiaki; Ishiwatari, Yuki; Liu, Jie; Terai, Takayuki; Nagasaki, Shinya; Muroya, Yusa; Abe, Hiroaki [Nuclear Professional School / Department of Nuclear Engineering and Management, The University of Tokyo, Tokaimura, Naka-gun, Ibaraki, 319-1188 (Japan); Mori, Hideo [Department of Mechanical Engineering, Kyushu University (Japan); Akiba, Masato; Akimoto, Hajime; Okumura, Keisuke; Akasaka, Naoaki [Japan Atomic Energy Agency (Japan); GOTO, Shoji [Tokyo Electric Power Company (Japan)

    2006-07-01T23:59:59.000Z

    Research program of a supercritical-pressure light water cooled fast reactor (Super Fast Reactor) is funded by MEXT (Ministry of Education, Culture, Sports, Science and Technology) in December 2005 as one of the research programs of Japanese NERI (Nuclear Energy Research Initiative). It consists of three programs. (1) development of Super Fast Reactor concept; (2) thermal-hydraulic experiments; (3) material developments. The purpose of the concept development is to pursue the advantage of high power density of fast reactor over thermal reactors to achieve economic competitiveness of fast reactor for its deployment without waiting for exhausting uranium resources. Design goal is not breeding, but maximizing reactor power by using plutonium from spent LWR fuel. MOX will be the fuel of the Super Fast Reactor. Thermal-hydraulic experiments will be conducted with HCFC22 (Hydro chlorofluorocarbons) heat transfer loop of Kyushu University and supercritical water loop at JAEA. Heat transfer data including effect of grid spacers will be taken. The critical flow and condensation of supercritical fluid will be studied. The materials research includes the development and testing of austenitic stainless steel cladding from the experience of PNC1520 for LMFBR. Material for thermal insulation will be tested. SCWR (Supercritical-Water Cooled Reactor) of GIF (Generation-4 International Forum) includes both thermal and fast reactors. The research of the Super Fast Reactor will enhance SCWR research and the data base. The research period will be until March 2010. (authors)

  13. Public participation in UMTRA Project program management

    SciTech Connect (OSTI)

    Majors, M.J.; Ulland, L.M. [Weston (Roy F.), Inc., Albuquerque, NM (United States)

    1993-12-31T23:59:59.000Z

    Innovative techniques for overcoming barriers to public participation on the US Department of Energy`s Uranium Mill Tailings Remedial Action (UMTRA) Project have led to improved communications with stakeholders at project sites and improved communications within the project. On the UMTRA Project, it`s been shown that an effective public participation program is an essential element to successful project implementation.

  14. Process for electrolytically preparing uranium metal

    DOE Patents [OSTI]

    Haas, Paul A. (Knoxville, TN)

    1989-01-01T23:59:59.000Z

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

  15. Controlling uranium reactivity March 18, 2008

    E-Print Network [OSTI]

    Meyer, Karsten

    for the last decade. Most of their work involves depleted uranium, a more common form of uraniumMarch 2008 Controlling uranium reactivity March 18, 2008 Uranium is an often misunderstood metal uranium research. In reality, uranium presents a wealth of possibilities for funda- mental chemistry. Many

  16. Influence of uranium hydride oxidation on uranium metal behaviour

    SciTech Connect (OSTI)

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

    2013-07-01T23:59:59.000Z

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

  17. Summary - Uranium233 Downblending and Disposition Project

    Office of Environmental Management (EM)

    issues id be included in plans andor de he TRA Team m identified the ts (CTEs) and t ess Level (TRL) ratory (TRL3) rocess - Wiped ing (TRL3) nt (TRL2) RA reports, please v...

  18. Summary - Uranium233 Downblending and Disposition Project

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of Energy Power.pdf11-161-LNGInternational EnergyCommittee onGASRainey STAR Center |Product EM wa in Buil to

  19. Raytheon explores thorium for next generation nuclear reactor

    SciTech Connect (OSTI)

    Crawford, M.

    1994-03-08T23:59:59.000Z

    Few new orders for nuclear power plants have been placed anywhere in the world in the last 20 years, but that is not discouraging Raytheon Engineers Constructors from making plans to explore new light water reactor technologies for commercial markets. The Lexington, Mass.-based company, which has extensive experience in nuclear power engineering and construction, has a vision for the light water reactor of the future - one that is based on the use of thorium-232, an element that decays over several steps to uranium-233. The use of thorium and a small amount of uranium that is 20 percent enriched is seen as providing operational, environmental, and safety advantages over reactors using the standard fuel mixture of uranium-238 and enriched uranium-235. According to Raytheon, the system could improve the economics of some reactors' operations by reducing fuel costs and lowering related waste volumes. At the same time, reactor safety could be improved by simpler control rod systems and the absence from reactor coolant of corrosive boric acid, which is used to slow neutrons in order to enhance reactions. Using thorium is also attractive because more of the fuel is burned up by the reactor, an estimated 12 percent as compared to about 4 percent for U-235. However, the technology's greatest attraction may well be its implications for nuclear proliferation. Growing plutonium inventories embedded in spent fuel rods from light water reactors have sparked concern worldwide. But according to Raytheon, using a thorium-based fuel core would alleviate this concern because it would produce only small quantities of plutonium. A thorium-based fuel system would produce 12 kilograms of plutonium over a decade versus 2,235 kilograms for an equivalent reactor operating with conventional uranium fuel.

  20. Uncertainty clouds uranium enrichment corporation's plans

    SciTech Connect (OSTI)

    Lane, E.

    1993-03-24T23:59:59.000Z

    An expected windfall to the US Treasury from the sale of the Energy Dept.'s commercial fuel enrichment facilities may evaporate in the next few weeks when the Clinton administration submits its fiscal 1994 budget proposal to Congress, according to congressional and administration officials. Under the Energy Policy Act of 1992, DOE is required to lease two uranium enrichment facilities, Portsmouth, Ohio, and Paducah, KY., to the government-owned US Enrichment Corp. (USEC) by July 1. Estimates by OMB and Treasury indicate a potential yearly payoff of $300 million from the government-owned company's sale of fuel for commercial reactors. Those two facilities use a process of gaseous diffusion to enrich uranium to about 3 percent for use as fuel in commercial power plants. DOE has contracts through at least 1996 to provide about 12 million separative work units (SWUs) yearly to US utilities and others world-wide. But under an agreement signed between the US and Russia last August, at least 10 metric tons, or 1.5 million SWUs, of low-enriched uranium (LEU) blended down from Russia warheads is expected to be delivered to the US starting in 1994. It could be sold at $50 to $60 per SWU, far below what DOE currently charges for its SWUs - $135 per SWU for 70 percent of the contract price and $90 per SWU for the remaining 30 percent.

  1. Uranium-titanium-niobium alloy

    DOE Patents [OSTI]

    Ludtka, Gail M. (Oak Ridge, TN); Ludtka, Gerard M. (Oak Ridge, TN)

    1990-01-01T23:59:59.000Z

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

  2. Annual status report on the Uranium Mill Tailings Remedial Action Program

    SciTech Connect (OSTI)

    Not Available

    1992-12-01T23:59:59.000Z

    This fourteenth annual status report for the Uranium Mill Tailings Remedial Action (UMTRA) Project Office summarizes activities of the Uranium Mill Tailings Remedial Action Surface (UMTRA-Surface) and Uranium Mill Tailings Remedial Action Groundwater (UMTRA-Groundwater) Projects undertaken during fiscal year (FY) 1992 by the US Department of Energy (DOE) and other agencies. Project goals for FY 1993 are also presented. An annual report of this type was a statutory requirement through January 1, 1986, pursuant to the Uranium Mill Tailings Radiation Control Act (UMTRCA) of 1978, Public Law (PL) 95-604. The DOE will continue to submit annual reports to DOE-Headquarters, the states, tribes, and local representatives through Project completion in order to inform the public of the yearly Project status. The purpose of the remedial action is to stabilize and control the tailings and other residual radioactive material (RRM) located on the inactive uranium processing sites in a safe and environmentally sound manner, and to minimize or eliminate potential health hazards. Commercial and residential properties near designated processing sites that are contaminated with material from the sites, herein referred to as ``vicinity properties (VP),`` are also eligible for remedial action. Included in the UMTRA Project are 24 inactive uranium processing sites and associated VPs located in 10 states, and the VPs associated with the Edgemont, South Dakota, uranium mill currently owned by the Tennessee Valley Authority (TVA) (Figure A.1, Appendix A).

  3. Irradiation behavior of miniature experimental uranium silicide fuel plates

    SciTech Connect (OSTI)

    Hofman, G.L.; Neimark, L.A.; Mattas, R.F.

    1983-01-01T23:59:59.000Z

    Uranium silicides, because of their relatively high uranium density, were selected as candidate dispersion fuels for the higher fuel densities required in the Reduced Enrichment Research and Test Reactor (RERTR) Program. Irradiation experience with this type of fuel, however, was limited to relatively modest fission densities in the bulk from, on the order of 7 x 10/sup 20/ cm/sup -3/, far short of the approximately 20 x 10/sup 20/ cm/sup -3/ goal established for the RERTR program. The purpose of the irradiation experiments on silicide fuels on the ORR, therefore, was to investigate the intrinsic irradiation behavior of uranium silicide as a dispersion fuel. Of particular interest was the interaction between the silicide particles and the aluminum matrix, the swelling behavior of the silicide particles, and the maximum volume fraction of silicide particles that could be contained in the aluminum matrix.

  4. Uranium immobilization and nuclear waste

    SciTech Connect (OSTI)

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

    1982-02-01T23:59:59.000Z

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

  5. The development of uranium foil farication technology utilizing twin roll method for Mo-99 irradiation target

    E-Print Network [OSTI]

    Kim, C K; Park, H D

    2002-01-01T23:59:59.000Z

    MDS Nordion in Canada, occupying about 75% of global supply of Mo-99 isotope, has provided the irradiation target of Mo-99 using the rod-type UAl sub x alloys with HEU(High Enrichment Uranium). ANL (Argonne National Laboratory) through co-operation with BATAN in Indonesia, leading RERTR (Reduced Enrichment for Research and Test Reactors) program substantially for nuclear non-proliferation, has designed and fabricated the annular cylinder of uranium targets, and successfully performed irradiation test, in order to develop the fabrication technology of fission Mo-99 using LEU(Low Enrichment Uranium). As the uranium foils could be fabricated in laboratory scale, not in commercialized scale by hot rolling method due to significant problems in foil quality, productivity and economic efficiency, attention has shifted to the development of new technology. Under these circumstances, the invention of uranium foil fabrication technology utilizing twin-roll casting method in KAERI is found to be able to fabricate LEU or...

  6. Modeling of the performance of weapons MOX fuel in light water reactors

    SciTech Connect (OSTI)

    Alvis, J.; Bellanger, P.; Medvedev, P.G.; Peddicord, K.L. [Texas A and M Univ., College Station, TX (United States). Nuclear Engineering Dept.; Gellene, G.I. [Texas Tech Univ., Lubbock, TX (United States). Dept. of Chemistry and Biochemistry

    1999-05-01T23:59:59.000Z

    Both the Russian Federation and the US are pursing mixed uranium-plutonium oxide (MOX) fuel in light water reactors (LWRs) for the disposition of excess plutonium from disassembled nuclear warheads. Fuel performance models are used which describe the behavior of MOX fuel during irradiation under typical power reactor conditions. The objective of this project is to perform the analysis of the thermal, mechanical, and chemical behavior of weapons MOX fuel pins under LWR conditions. If fuel performance analysis indicates potential questions, it then becomes imperative to assess the fuel pin design and the proposed operating strategies to reduce the probability of clad failure and the associated release of radioactive fission products into the primary coolant system. Applying the updated code to anticipated fuel and reactor designs, which would be used for weapons MOX fuel in the US, and analyzing the performance of the WWER-100 fuel for Russian weapons plutonium disposition are addressed in this report. The COMETHE code was found to do an excellent job in predicting fuel central temperatures. Also, despite minor predicted differences in thermo-mechanical behavior of MOX and UO{sub 2} fuels, the preliminary estimate indicated that, during normal reactor operations, these deviations remained within limits foreseen by fuel pin design.

  7. E-Print Network 3.0 - athene reactor Sample Search Results

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

    which country... will host a nuclear fusion reactor for the International Thermonuclear Experimental Reactor (ITER) project... . The candidate sites for the international...

  8. Spent fuel utilization in a compact traveling wave reactor

    SciTech Connect (OSTI)

    Hartanto, Donny; Kim, Yonghee [Korea Advanced Institute of Science and Technology 373-1 Kusong-dong, Yuseong-gu, Daejeon, 305-701 (Korea, Republic of)

    2012-06-06T23:59:59.000Z

    In recent years, several innovative designs of nuclear reactors are proposed. One of them is Traveling Wave Reactor (TWR). The unique characteristic of a TWR is the capability of breeding its own fuel in the reactor. The reactor is fueled by mostly depleted, natural uranium or spent nuclear fuel and a small amount of enriched uranium to initiate the fission process. Later on in the core, the reactor gradually converts the non-fissile material into the fissile in a process like a traveling wave. In this work, a TWR with spent nuclear fuel blanket was studied. Several parameters such as reactivity coefficients, delayed neutron fraction, prompt neutron generation lifetime, and fission power, were analyzed. The discharge burnup composition was also analyzed. The calculation is performed by a continuous energy Monte Carlo code McCARD.

  9. Corrosion-resistant uranium

    DOE Patents [OSTI]

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

    1981-10-21T23:59:59.000Z

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

  10. Corrosion-resistant uranium

    DOE Patents [OSTI]

    Hovis, Jr., Victor M. (Kingston, TN); Pullen, William C. (Knoxville, TN); Kollie, Thomas G. (Oak Ridge, TN); Bell, Richard T. (Knoxville, TN)

    1983-01-01T23:59:59.000Z

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

  11. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines About U.S. NaturalA. Michael SchaalNovember1. Foreign sales of uranium froma. Uranium

  12. Criticality Safety of Low-Enriched Uranium and High-Enriched Uranium Fuel Elements in Heavy Water Lattices

    SciTech Connect (OSTI)

    Pesic, Milan P

    2003-10-15T23:59:59.000Z

    The RB reactor was designed as a natural-uranium, heavy water, nonreflected critical assembly in the Vinca Institute of Nuclear Sciences, Belgrade, Yugoslavia, in 1958. From 1962 until 2002, numerous critical experiments were carried out with low-enriched uranium and high-enriched uranium fuel elements of tubular shape, known as the Russian TVR-S fuel assembly type, placed in various heavy water square lattices within the RB cylindrical aluminum tank. Some of these well-documented experiments were selected, described, evaluated, and accepted for inclusion in the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments', contributing to the preservation of a rather small number of heavy water benchmark critical experiments.

  13. Evaluation of in vitro dissolution rates of throum in uranium mill tailings

    SciTech Connect (OSTI)

    Reif, R.G. [Department of Energy, Albuquerque, NM (United States)

    1996-06-01T23:59:59.000Z

    Dissolution rates of thorium from the uranium mill tailings piles at two Department of Energy Uranium Mill Tailings Remedial Action Project (UMTRAP) sites have been evaluated. The thorium dissolution rates were evaluated in vitro using simulated lung fluid. The former uranium mills at the UMTRAP sites employed different chemical processes (acid leach and alkaline pressure leach) to extract the uranium from the ore, and the thorium dissolution rates at these sites were found to be markedly different. A site specific annual limit on intake (ALI) value for {sup 230}Th was calculated for the UMTRAP Site that was associated with a multiple component dissolution curve.

  14. Evaluation of in vitro dissolution rates of thorium in uranium mill tailings

    SciTech Connect (OSTI)

    Reif, R.H. [RUST Federal Services, Albuquerque, NM (United States)

    1994-11-01T23:59:59.000Z

    Dissolution rates of thorium from the uranium mill tailings piles at two Department of Energy Uranium Mill Tailings Remedial Action Project (UMTRAP) sites have been evaluated. The thorium dissolution rates were evaluated in vitro using simulated lung fluid. The former uranium mills at the UMTRAP sites employee different chemical processes (acid leach and alkaline pressure leach) to extract the uranium from the ore, and the thorium dissolution rates at these sites were found to be markedly different. A site specific annual limit on intake (ALI) value for {sup 230}Th was calculated for the UMTRAP site that was associated with a multiple component dissolution curve. 9 refs., 1 fig., 3 tabs.

  15. Geological and geochemical aspects of uranium deposits: a selected, annotated bibliography. [474 references

    SciTech Connect (OSTI)

    Thomas, J.M.; Garland, P.A.; White, M.B.; Daniel, E.W.

    1980-09-01T23:59:59.000Z

    This bibliography, a compilation of 474 references, is the fourth in a series compiled from the National Uranium Resource Evaluation (NURE) Bibliographic Data Base. This data base was created for the Grand Junction Office of the Department of Energy's National Uranium Resource Evaluation Project by the Ecological Sciences Information Center, Oak Ridge National Laboratory. The references in the bibliography are arranged by subject category: (1) geochemistry, (2) exploration, (3) mineralogy, (4) genesis of deposits, (5) geology of deposits, (6) uranium industry, (7) geology of potential uranium-bearing areas, and (8) reserves and resources. The references are indexed by author, geographic location, quadrangle name, geoformational feature, and keyword.

  16. The ALTO Project at IPN Orsay CNRS/IN2P3/Universit Paris Sud

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    . The technique originally proposed [3] is the use of energetic neutrons to induce fission of depleted uranium II and EURISOL projects). To get such beams, a large R&D effort is required. Uranium fission

  17. High loading uranium fuel plate

    DOE Patents [OSTI]

    Wiencek, Thomas C. (Bolingbrook, IL); Domagala, Robert F. (Indian Head Park, IL); Thresh, Henry R. (Palos Heights, IL)

    1990-01-01T23:59:59.000Z

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

  18. An autonomous long-term fast reactor system and the principal design limitations of the concept

    E-Print Network [OSTI]

    Tsvetkova, Galina Valeryevna

    2004-09-30T23:59:59.000Z

    Actinides MOX Mixed OXide MSR Molten-Salt Reactors NERI Nuclear Energy Research Initiative vii PWR Pressurized Water Reactor RGPu Reactor-Grade Plutonium SCNES Self-Consistent Nuclear Energy System STAR Secure Transportable Autonomous Reactor... of LWR?s, the drastic increase of Am and Cm inventories are observed after uranium fuel irradiation and the second recycling of MOX fuel.1 Therefore, partitioning and transmutation of the recovered MA?s could significantly reduce the long...

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

    SciTech Connect (OSTI)

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

    2008-02-01T23:59:59.000Z

    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 MissouriColumbia, and the MIT Reactor-II at the Massachusetts Institute of Technology. Characteristics for each of the HPRRs are given in Appendix A. The GTRI Convert Fuel Development program is currently engaged in the development of a novel nuclear fuel that will enable these conversions. The fuel design is based on a monolithic fuel meat (made from a uranium-molybdenum alloy) clad in Al-6061 that has shown excellent performance in irradiation testing. The unique aspects of the fuel design, however, necessitate the development and implementation of new fabrication techniques and, thus, establishment of the infrastructure to ensure adequate fuel fabrication capability. A conceptual fabrication process description and rough estimates of the total facility throughput are described in this document as a basis for establishing preconceptual fabrication facility designs.

  20. Prompt Neutron Lifetime for the NBSR Reactor

    SciTech Connect (OSTI)

    Hanson, A.L.; Diamond, D.

    2012-06-24T23:59:59.000Z

    In preparation for the proposed conversion of the National Institute of Standards and Technology (NIST) research reactor (NBSR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel, certain point kinetics parameters must be calculated. We report here values of the prompt neutron lifetime that have been calculated using three independent methods. All three sets of calculations demonstrate that the prompt neutron lifetime is shorter for the LEU fuel when compared to the HEU fuel and longer for the equilibrium end-of-cycle (EOC) condition when compared to the equilibrium startup (SU) condition for both the HEU and LEU fuels.

  1. Uranium from Seawater Program Review; Fuel Resources Uranium from Seawater Program DOE Office of Nuclear Energy

    SciTech Connect (OSTI)

    none,

    2013-07-01T23:59:59.000Z

    For nuclear energy to remain sustainable in the United States, economically viable sources of uranium beyond terrestrial ores must be developed. The goal of this program is to develop advanced adsorbents that can extract uranium from seawater at twice the capacity of the best adsorbent developed by researchers at the Japan Atomic Energy Agency (JAEA), 1.5 mg U/g adsorbent. A multidisciplinary team from Oak Ridge National Laboratory, Lawrence Berkeley National Laboratory, Pacific Northwest National Laboratory, and the University of Texas at Austin was assembled to address this challenging problem. Polymeric adsorbents, based on the radiation grafting of acrylonitrile and methacrylic acid onto high surface-area polyethylene fibers followed by conversion of the nitriles to amidoximes, have been developed. These poly(acrylamidoxime-co-methacrylic acid) fibers showed uranium adsorption capacities for the extraction of uranium from seawater that exceed 3 mg U/g adsorbent in testing at the Pacific Northwest National Laboratory Marine Sciences Laboratory. The essence of this novel technology lies in the unique high surface-area trunk material that considerably increases the grafting yield of functional groups without compromising its mechanical properties. This technology received an R&D100 Award in 2012. In addition, high surface area nanomaterial adsorbents are under development with the goal of increasing uranium adsorption capacity by taking advantage of the high surface areas and tunable porosity of carbon-based nanomaterials. Simultaneously, de novo structure-based computational design methods are being used to design more selective and stable ligands and the most promising candidates are being synthesized, tested and evaluated for incorporation onto a support matrix. Fundamental thermodynamic and kinetic studies are being carried out to improve the adsorption efficiency, the selectivity of uranium over other metals, and the stability of the adsorbents. Understanding the rate-limiting step of uranium uptake from seawater is also essential in designing an effective uranium recovery system. Finally, economic analyses have been used to guide these studies and highlight what parameters, such as capacity, recyclability, and stability, have the largest impact on the cost of extraction of uranium from seawater. Initially, the cost estimates by the JAEA for extraction of uranium from seawater with braided polymeric fibers functionalized with amidoxime ligands were evaluated and updated. The economic analyses were subsequently updated to reflect the results of this project while providing insight for cost reductions in the adsorbent development through cradle-to-grave case studies for the extraction process. This report highlights the progress made over the last three years on the design, synthesis, and testing of new materials to extract uranium for seawater. This report is organized into sections that highlight the major research activities in this project: (1) Chelate Design and Modeling, (2) Thermodynamics, Kinetics and Structure, (3) Advanced Polymeric Adsorbents by Radiation Induced Grafting, (4) Advanced Nanomaterial Adsorbents, (5) Adsorbent Screening and Modeling, (6) Marine Testing, and (7) Cost and Energy Assessment. At the end of each section, future research directions are briefly discussed to highlight the challenges that still remain to reduce the cost of extractions of uranium for seawater. Finally, contributions from the Nuclear Energy University Programs (NEUP), which complement this research program, are included at the end of this report.

  2. National Uranium Resource Evaluation. Volume 1. Summary of the geology and uranium potential of Precambrian conglomerates in southeastern Wyoming

    SciTech Connect (OSTI)

    Karlstrom, K.E.; Houston, R.S.; Flurkey, A.J.; Coolidge, C.M.; Kratochvil, A.L.; Sever, C.K.

    1981-02-01T23:59:59.000Z

    A series of uranium-, thorium-, and gold-bearing conglomerates in Late Archean and Early Proterozoic metasedimentary rocks have been discovered in southern Wyoming. The mineral deposits were found by applying the time and strata bound model for the origin of uranium-bearing quartz-pebble conglomerates to favorable rock types within a geologic terrane known from prior regional mapping. No mineral deposits have been discovered that are of current (1981) economic interest, but preliminary resource estimates indicate that over 3418 tons of uranium and over 1996 tons of thorium are present in the Medicine Bow Mountains and that over 440 tons of uranium and 6350 tons of thorium are present in Sierra Madre. Sampling has been inadequate to determine gold resources. High grade uranium deposits have not been detected by work to date but local beds of uranium-bearing conglomerate contain as much as 1380 ppM uranium over a thickness of 0.65 meters. This project has involved geologic mapping at scales from 1/6000 to 1/50,000 detailed sampling, and the evaluation of 48 diamond drill holes, but the area is too large to fully establish the economic potential with the present information. This first volume summarizes the geologic setting and geologic and geochemical characteristics of the uranium-bearing conglomerates. Volume 2 contains supporting geochemical data, lithologic logs from 48 drill holes in Precambrian rocks, and drill site geologic maps and cross-sections from most of the holes. Volume 3 is a geostatistical resource estimate of uranium and thorium in quartz-pebble conglomerates.

  3. Light-Water Breeder Reactor

    DOE Patents [OSTI]

    Beaudoin, B. R.; Cohen, J. D.; Jones, D. H.; Marier, Jr, L. J.; Raab, H. F.

    1972-06-20T23:59:59.000Z

    Described is a light-water-moderated and -cooled nuclear breeder reactor of the seed-blanket type characterized by core modules comprising loosely packed blanket zones enriched with fissile fuel and axial zoning in the seed and blanket regions within each core module. Reactivity control over lifetime is achieved by axial displacement of movable seed zones without the use of poison rods in the embodiment illustrated. The seed is further characterized by a hydrogen-to-uranium-233 atom ratio in the range 10 to 200 and a uranium-233-to-thorium-232 atom ratio ranging from 0.012 to 0.200. The seed occupies from 10 to 35 percent of the core volume in the form of one or more individual islands or annuli. (NSA 26: 55130)

  4. Special Analysis for the Disposal of the Consolidated Edison Uranium Solidification Project Waste Stream at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

    SciTech Connect (OSTI)

    NSTec Environmental Management

    2013-01-31T23:59:59.000Z

    The purpose of this Special Analysis (SA) is to determine if the Oak Ridge (OR) Consolidated Edison Uranium Solidification Project (CEUSP) uranium-233 (233U) waste stream (DRTK000000050, Revision 0) is acceptable for shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS) on the Nevada National Security Site (NNSS). The CEUSP 233U waste stream requires a special analysis because the concentrations of thorium-229 (229Th), 230Th, 232U, 233U, and 234U exceeded their NNSS Waste Acceptance Criteria action levels. The acceptability of the waste stream is evaluated by determining if performance assessment (PA) modeling provides a reasonable expectation that SLB disposal is protective of human health and the environment. The CEUSP 233U waste stream is a long-lived waste with unique radiological hazards. The SA evaluates the long-term acceptability of the CEUSP 233U waste stream for near-surface disposal as a two tier process. The first tier, which is the usual SA process, uses the approved probabilistic PA model to determine if there is a reasonable expectation that disposal of the CEUSP 233U waste stream can meet the performance objectives of U.S. Department of Energy Manual DOE M 435.1-1, Radioactive Waste Management, for a period of 1,000 years (y) after closure. The second tier addresses the acceptability of the OR CEUSP 233U waste stream for near-surface disposal by evaluating long-term site stability and security, by performing extended (i.e., 10,000 and 60,000 y) modeling analyses, and by evaluating the effect of containers and the depth of burial on performance. Tier I results indicate that there is a reasonable expectation of compliance with all performance objectives if the OR CEUSP 233U waste stream is disposed in the Area 5 RWMS SLB disposal units. The maximum mean and 95th percentile PA results are all less than the performance objective for 1,000 y. Monte Carlo uncertainty analysis indicates that there is a high likelihood of compliance with all performance objectives. Tier II results indicate that the long-term performance of the OR CEUSP 233U waste stream is protective of human health and the environment. The Area 5 RWMS is located in one of the least populated and most arid regions of the U.S. Site characterization data indicate that infiltration of precipitation below the plant root zone at 2.5 meters (8.2 feet) ceased 10,000 to 15,000 y ago. The site is not expected to have a groundwater pathway as long as the current arid climate persists. The national security mission of the NNSS and the location of the Area 5 RWMS within the Frenchman Flat Corrective Action Unit require that access controls and land use restrictions be maintained indefinitely. PA modeling results for 10,000 to 60,000 y also indicate that the OR CEUSP 233U waste stream is acceptable for near-surface disposal. The mean resident air pathway annual total effective dose (TED), the resident all-pathways annual TED, and the acute drilling TED are less than their performance objectives for 10,000 y after closure. The mean radon-222 (222Rn) flux density exceeds the performance objective at 4,200 y, but this is due to waste already disposed at the Area 5 RWMS and is only slightly affected by disposal of the CEUSP 233U. The peak resident all-pathways annual TED from CEUSP key radionuclides occurs at 48,000 y and is less than the 0.25 millisievert performance objective. Disposal of the OR CEUSP 233U waste stream in a typical SLB trench slightly increases PA results. Increasing the depth was found to eliminate any impacts of the OR CEUSP 233U waste stream. Containers could not be shown to have any significant impact on performance due to the long half-life of the waste stream and a lack of data for pitting corrosion rates of stainless steel in soil. The results of the SA indicate that all performance objectives can be met with disposal of the OR CEUSP 233U waste stream in the SLB units at the Area 5 RWMS. The long-term performance of the OR CEUSP 233U waste stream disposed in the near surface is protective of human health

  5. 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.

    1993-02-01T23:59:59.000Z

    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.

  6. 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.

    1993-02-01T23:59:59.000Z

    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.

  7. Uranium from seawater

    SciTech Connect (OSTI)

    Gregg, D.; Folkendt, M.

    1982-09-21T23:59:59.000Z

    A novel process for recovering uranium from seawater is proposed and some of the critical technical parameters are evaluated. The process, in summary, consists of two different options for contacting adsorbant pellets with seawater without pumping the seawater. It is expected that this will reduce the mass handling requirements, compared to pumped seawater systems, by a factor of approximately 10/sup 5/, which should also result in a large reduction in initial capital investment. Activated carbon, possibly in combination with a small amount of dissolved titanium hydroxide, is expected to be the preferred adsorbant material instead of the commonly assumed titanium hydroxide alone. The activated carbon, after exposure to seawater, can be stripped of uranium with an appropriate eluant (probably an acid) or can be burned for its heating value (possible in a power plant) leaving the uranium further enriched in its ash. The uranium, representing about 1% of the ash, is then a rich ore and would be recovered in a conventional manner. Experimental results have indicated that activated carbon, acting alone, is not adequately effective in adsorbing the uranium from seawater. We measured partition coefficients (concentration ratios) of approximately 10/sup 3/ in seawater instead of the reported values of 10/sup 5/. However, preliminary tests carried out in fresh water show considerable promise for an extraction system that uses a combination of dissolved titanium hydroxide (in minute amounts) which forms an insoluble compound with the uranyl ion, and the insoluble compound then being sorbed out on activated carbon. Such a system showed partition coefficients in excess of 10/sup 5/ in fresh water. However, the system was not tested in seawater.

  8. National uranium resource evaluation program: hydrogeochemical and stream sediment reconnaissance basic data for Sacramento quadrangle, California

    SciTech Connect (OSTI)

    Not Available

    1981-10-15T23:59:59.000Z

    Field and laboratory data are presented for 1890 sediment samples from the Sacramento Quadrangle, California. The samples were collected by Savannah River Laboratory; laboratory analysis and data reporting were performed by the Uranium Resource Evaluation Project at Oak Ridge, Tennessee.

  9. Proliferation resistance for fast reactors and related fuel cycles: issues and impacts

    SciTech Connect (OSTI)

    Pilat, Joseph F [Los Alamos National Laboratory

    2010-01-01T23:59:59.000Z

    The prospects for a dramatic growth in nuclear power may depend to a significant degree on the effectiveness of, and the resources devoted to, plans to develop and implement technologies and approaches that strengthen proliferation resistance and nuclear materials accountability. The challenges for fast reactors and related fuel cycles are especially critical. They are being explored in the Generation IV Tnternational Forum (GIF) and the Tnternational Atomic Energy Agency's (IAEA's) International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) initiative, as well as by many states that are looking to these systems for the efficient lise of uranium resources and long-term energy security. How do any proliferation risks they may pose compare to other reactors, both existing and under development, and their fuel cycles? Can they be designed with intrinsic (technological) features to make these systems more proliferation resistant? What roles can extrinsic (institutional) features play in proliferation resistance? What are the anticipated safeguards requirements, and will new technologies and approaches need to be developed? How can safeguards be facilitated by the design process? These and other questions require a rethinking of proliferation resistance and the prospects for new technologies and other intrinsic and extrinsic features being developed that are responsive to specific issues for fast reactors and related fuel cycles and to the broader threat environment in which these systems will have to operate. There are no technologies that can wholly eliminate the risk of proliferation by a determined state, but technology and design can playa role in reducing state threats and perhaps in eliminating non-state threats. There will be a significant role for extrinsic factors, especially the various measures - from safeguards and physical protection to export controls - embodied in the international nuclear nonproliferation regime. This paper will offer an assessment of the issues surrounding, and the prospects for, efforts to develop proliferation resistance for fast reactors and related fuel cycles in the context of a nuclear renaissance. The focus of the analysis is on fast reactors.

  10. Method of preparation of uranium nitride

    DOE Patents [OSTI]

    Kiplinger, Jaqueline Loetsch; Thomson, Robert Kenneth James

    2013-07-09T23:59:59.000Z

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

  11. URANIUM MILLING ACTIVITIES AT SEQUOYAH FUELS CORPORATION

    E-Print Network [OSTI]

    unknown authors

    Sequoyah Fuels Corporation (SFC) describes previous operations at its Gore, Oklahoma, uranium conversion facility as: (1) the recovery of uranium by concentration and purification processes; and (2) the conversion of concentrated and purified uranium ore into uranium hexafluoride (UF 6), or the reduction of depleted uranium tetrafluoride (UF 4) to UF 6. SFC contends that these

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

    SciTech Connect (OSTI)

    Not Available

    1993-07-01T23:59:59.000Z

    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.

  13. Evaluation of nuclear facility decommissioning projects. Three Mile Island Unit 2 reactor coolant system and systems decontamination. Summary status report. Volume 1

    SciTech Connect (OSTI)

    Doerge, D.H.; Miller, R.L.; Scotti, K.S.

    1986-05-01T23:59:59.000Z

    This document summarizes information relating to the decontamination and restoration of the Three Mile Island Unit 2 reactor coolant system and other plant systems. Data have been collected from activity reports, reactor containment entry records, and other sources and entered in a computerized data system which permits extraction/manipulation of specific data which can be used in planning for recovery from a loss of coolant event similar to that experienced by the Three Mile Island Unit 2 on March 28, 1979. This report contains a summary of radiation exposures, manpower, and time spent in radiation areas during the referenced period. Support activities conducted outside of radiation areas are not included. Computer reports included are: A chronological listing of all activities related to decomtamination and restoration of the reactor coolant system and other plant systems for the period of April 5, 1979, through December 19, 1984; a summary of labor and exposures by department for the same time period; and summary reports for each major task undertaken in connection with this specific work scope during the referenced time period.

  14. Method for fabricating uranium foils and uranium alloy foils

    DOE Patents [OSTI]

    Hofman, Gerard L. (Downers Grove, IL); Meyer, Mitchell K. (Idaho Falls, ID); Knighton, Gaven C. (Moore, ID); Clark, Curtis R. (Idaho Falls, ID)

    2006-09-05T23:59:59.000Z

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

  15. UMTRA Project value engineering plan

    SciTech Connect (OSTI)

    Not Available

    1990-06-01T23:59:59.000Z

    The objective of value engineering (VE) on the Uranium MILL Tailings Remedial Action (UMTRA) Project is to ensure that remedial action at the UMTRA Project sites is performed to meet the US Environmental Protection Agency (EPA) standards for inactive uranium mill tailings sites at the lowest cost, while maintaining a high quality of work. Through review of designs and consideration of reasonable, less expensive alternatives, VE can be an effective cost reduction tool and a means to improve the design. The UMTRA Project products are the design and construction of stabilized tailings embankments.

  16. Global Threat Reduction Initiative Fuel-Thermo-Physical Characterization Project Quality Assurance Plan

    SciTech Connect (OSTI)

    Pereira, Mario M.; Slonecker, Bruce D.

    2012-06-01T23:59:59.000Z

    The charter of the Fuel Thermo-Physical Characterization Project is to ready Pacific Northwest National Laboratory (PNNL) facilities and processes for the receipt of unirradiated and irradiated low enriched uranium (LEU) molybdenum (U-Mo) fuel element samples, and to perform analysis to support the Global Threat Reduction Initiative conversion program. PNNLs support for the program will include the establishment of post-irradiation examination processes, including thermo-physical properties, unique to the U.S. Department of Energy laboratories. These processes will ultimately support the submission of the base fuel qualification (BFQ) to the U.S. Nuclear Regulatory Commission (NRC) and revisions to High Performance Research Reactor Safety Analysis Reports to enable conversion from highly enriched uranium to LEU fuel. This quality assurance plan (QAP) provides the quality assurance requirements and processes that support the NRC BFQ. This QAP is designed to be used by project staff, and prescribes the required management control elements that are to be met and how they are implemented. Additional controls are captured in Fuel Thermo-Physical Characterization Project plans, existing procedures, and procedures to be developed that provide supplemental information on how work is conducted on the project.

  17. Design of a 25-kWe Surface Reactor System Based on SNAP Reactor Technologies

    SciTech Connect (OSTI)

    Dixon, David D. [Nuclear Systems Design Group, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Hiatt, Matthew T. [Nuclear Systems Design Group, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Department of Nuclear Engineering, Texas A and M University, College Station, TX 77843 (United States); Poston, David I.; Kapernick, Richard J. [Nuclear Systems Design Group, Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

    2006-01-20T23:59:59.000Z

    A Hastelloy-X clad, sodium-potassium (NaK-78) cooled, moderated spectrum reactor using uranium zirconium hydride (UZrH) fuel based on the SNAP program reactors is a promising design for use in surface power systems. This paper presents a 98 kWth reactor for a power system the uses multiple Stirling engines to produce 25 kWe-net for 5 years. The design utilizes a pin type geometry containing UZrHx fuel clad with Hastelloy-X and NaK-78 flowing around the pins as coolant. A compelling feature of this design is its use of 49.9% enriched U, allowing it to be classified as a category III-D attractiveness and reducing facility costs relative to highly-enriched space reactor concepts. Presented below are both the design and an analysis of this reactor's criticality under various safety and operations scenarios.

  18. Thermonuclear Reflect AB-Reactor

    E-Print Network [OSTI]

    Alexander Bolonkin

    2008-03-26T23:59:59.000Z

    The author offers a new kind of thermonuclear reflect reactor. The remarkable feature of this new reactor is a three net AB reflector, which confines the high temperature plasma. The plasma loses part of its energy when it contacts with the net but this loss can be compensated by an additional permanent plasma heating. When the plasma is rarefied (has a small density), the heat flow to the AB reflector is not large and the temperature in the triple reflector net is lower than 2000 - 3000 K. This offered AB-reactor has significantly less power then the currently contemplated power reactors with magnetic or inertial confinement (hundreds-thousands of kW, not millions of kW). But it is enough for many vehicles and ships and particularly valuable for tunnelers, subs and space apparatus, where air to burn chemical fuel is at a premium or simply not available. The author has made a number of innovations in this reactor, researched its theory, developed methods of computation, made a sample computation of typical project. The main point of preference for the offered reactor is its likely cheapness as a power source. Key words: Micro-thermonuclear reactor, Multi-reflex AB-thermonuclear reactor, Self-magnetic AB-thermonuclear reactor, aerospace thermonuclear engine.

  19. Uranium-plutonium-neptunium fuel cycle to produce isotopically denatured plutonium

    SciTech Connect (OSTI)

    Wydler, P.; Heer, W.; Stiller, P.; Wenger, H.U.

    1980-06-01T23:59:59.000Z

    In view of the considerable amount of /sup 237/ Np produced as a by-product in nuclear power reactors, possible utilization of this nuclide in the nuclear fuel cycle has been studied. In particular, the performance of a gas-cooled fast breeder reactor as a neptunium burner was assessed. A strategy was developed and mass flows were computed for a denatured plutonium LWR strategy using uranium, plutonium and neptunium recycling. 10 refs.

  20. Field-Portable Immunoassay Instruments and Reagents to Measure Chelators and Mobile Forms of Uranium

    SciTech Connect (OSTI)

    Blake, Diane A.

    2006-01-23T23:59:59.000Z

    Progress Report Date: 01/23/06 (report delayed due to Hurricane Katrina) Report of results to date: The goals of this 3-year project are to: (1) update and successfully deploy our present immunosensors at DOE sites; (2) devise immunosensor-based assays for Pb(II), Hg(II), chelators, and/or Cr(III) in surface and groundwater; and (3) develop new technologies in antibody engineering that will enhance this immunosensor program. Note: Work on this project was temporarily disrupted when Hurricane Katrina shut down the University on August 29, 2005. While most of the reagents stored in our refrigerators and freezers were destroyed, all of our hybridoma cell lines were saved because they had been stored in liquid nitrogen. We set up new tissue culture reactors with the hybridomas that synthesize the anti-uranium antibodies, and are purifying new monoclonal antibodies from these culture supernatants. Both the in-line and the field-portable sensor were rescued from our labs in New Orleans in early October, and we continued experiments with these sensors in the temporary laboratory we set up in Hammond, LA at Southeastern Louisiana University.

  1. The Next Generation Nuclear Plant (NGNP) Project

    SciTech Connect (OSTI)

    F. H. Southworth; P. E. MacDonald

    2003-11-01T23:59:59.000Z

    The Next Generation Nuclear Power (NGNP) Project will demonstrate emissions-free nuclearassisted electricity and hydrogen production by 2015. The NGNP reactor will be a helium-cooled, graphite moderated, thermal neutron spectrum reactor with a design goal outlet temperature of 1000 C or higher. The reactor thermal power and core configuration will be designed to assure passive decay heat removal without fuel damage during hypothetical accidents. The fuel cycle will be a once-through very high burnup low-enriched uranium fuel cycle. This paper provides a description of the project to build the NGNP at the Idaho National Engineering and Environmental Laboratory (INEEL). The NGNP Project includes an overall reactor design activity and four major supporting activities: materials selection and qualification, NRC licensing and regulatory support, fuel development and qualification, and the hydrogen production plant. Each of these activities is discussed in the paper. All the reactor design and construction activities will be managed under the DOEs project management system as outlined in DOE Order 413.3. The key elements of the overall project management system discussed in this paper include the client and project management organization relationship, critical decisions (CDs), acquisition strategy, and the project logic and timeline. The major activities associated with the materials program include development of a plan for managing the selection and qualification of all component materials required for the NGNP; identification of specific materials alternatives for each system component; evaluation of the needed testing, code work, and analysis required to qualify each identified material; preliminary selection of component materials; irradiation of needed sample materials; physical, mechanical, and chemical testing of unirradiated and irradiated materials; and documentation of final materials selections. The NGNP will be licensed by the NRC under 10 CFR 50 or 10 CFR 52, for the purpose of demonstrating the suitability of high-temperature gas-cooled reactors for commercial electric power and hydrogen production. Products that will support the licensing of the NGNP include the environmental impact statement, the preliminary safety analysis report, the NRC construction permit, the final safety analysis report, and the NRC operating license. The fuel development and qualification program consists of five elements: development of improved fuel manufacturing technologies, fuel and materials irradiations, safety testing and post-irradiation examinations, fuel performance modeling, and fission product transport and source term modeling. Two basic approaches will be explored for using the heat from the high-temperature helium coolant to produce hydrogen. The first technology of interest is the thermochemical splitting of water into hydrogen and oxygen. The most promising processes for thermochemical splitting of water are sulfur-based and include the sulfur-iodine, hybrid sulfur-electrolysis, and sulfur-bromine processes. The second technology of interest is thermally assisted electrolysis of water. The efficiency of this process can be substantially improved by heating the water to high-temperature steam before applying electrolysis.

  2. NRC review of Electric Power Research Institute`s advanced light water reactor utility requirements document. Passive plant designs, chapter 1, project number 669

    SciTech Connect (OSTI)

    Not Available

    1994-08-01T23:59:59.000Z

    The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the {open_quotes}Advanced Light Water Reactor [ALWR] Utility Requirements Document{close_quotes}, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume 1, {open_quotes}ALWR Policy and Summary of Top-Tier Requirements{close_quotes}, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, {open_quotes}NRC Review of Electric Power Research Institute`s Advanced Light Water Reactor Utility Requirements Document - Program Summary{close_quotes}, provided a discussion of the overall purpose and scope of the Requirements Document, the background of the staff`s review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review.

  3. NRC review of Electric Power Research Institute`s advanced light water reactor utility requirements document. Passive plant designs, chapters 2-13, project number 669

    SciTech Connect (OSTI)

    Not Available

    1994-08-01T23:59:59.000Z

    The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the {open_quotes}Advanced Light Water Reactor [ALWR] Utility Requirements Document{close_quotes}, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume I, {open_quotes}ALWR Policy and Summary of Top-Tier Requirements{close_quotes}, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, {open_quotes}NRC Review of Electric Power Research Institute`s Advanced Light Water Reactor Utility Requirements Document - Program Summary{close_quotes}, provided a discussion of the overall purpose and scope of the Requirements Document, the background of the staff`s review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review.

  4. Unexpected, Stable Form of Uranium Detected | EMSL

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

    Unexpected, Stable Form of Uranium Detected Unexpected, Stable Form of Uranium Detected Insights on underappreciated reaction could shed light on environmental cleanup options...

  5. Molten uranium dioxide structure and dynamics

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

    Skinner, L. B. [Argonne National Laboratory (ANL), Argonne, IL (United States); Stony Brook Univ., Stony Brook, NY (United States); Materials Development Inc., Arlington Heights, IL (United States); Parise, J. B. [Stony Brook Univ., Stony Brook, NY (United States); Benmore, C. J. [Argonne National Laboratory (ANL), Argonne, IL (United States); Weber, J. K.R. [Materials Development Inc., Arlington Heights, IL (United States); Williamson, M. A. [Argonne National Laboratory (ANL), Argonne, IL (United States); Tamalonis, A. [Materials Development Inc., Arlington Heights, IL (United States); Hebden, A. [Argonne National Laboratory (ANL), Argonne, IL (United States); Wiencek, T. [Argonne National Laboratory (ANL), Argonne, IL (United States); Alderman, O. L.G. [Materials Development Inc., Arlington Heights, IL (United States); Guthrie, M. [Carnegie Inst., Washington, DC (United States); Leibowitz, L. [Argonne National Laboratory (ANL), Argonne, IL (United States)

    2014-11-20T23:59:59.000Z

    Uranium dioxide (UO2) is the major nuclear fuel component of fission power reactors. A key concern during severe accidents is the melting and leakage of radioactive UO2 as it corrodes through its zirconium cladding and steel containment. Yet, the very high temperatures (>3140 kelvin) and chemical reactivity of molten UO2 have prevented structural studies. In this work, we combine laser heating, sample levitation, and synchrotron x-rays to obtain pair distribution function measurements of hot solid and molten UO2. The hot solid shows a substantial increase in oxygen disorder around the lambda transition (2670 K) but negligible U-O coordination change. On melting, the average U-O coordination drops from 8 to 6.7 0.5. Molecular dynamics models refined to this structure predict higher U-U mobility than 8-coordinated melts.

  6. Experimental method for reactor-noise measurements of effective beta. [LMFBR

    SciTech Connect (OSTI)

    Bennett, E.F.

    1981-09-01T23:59:59.000Z

    A variance-to-mean noise technique, modified to eliminate systematic errors from drifting of reactor power, has been used to infer integral values of effective beta for uranium and plutonium fueled fast reactor modk-ups. The measurement technique, including corrections for a finite detector-electrometer time response, is described together with preliminary beta measurement results.

  7. Comparison of REMIX vs. MOX fuel characteristics in multiple recycling in VVER reactor

    SciTech Connect (OSTI)

    Dekusar, V.M.; Kalashnikov, A.G.; Kapranova, E.N.; Korobitsyn, V.E.; Puzakov, A.Y. [State Scientific Centre of Russian Federation, Institute for Physics and Power Engineering, Obninsk (Russian Federation)

    2013-07-01T23:59:59.000Z

    Multiple recycling of regenerated uranium-plutonium fuel in thermal reactors of VVER-1000 type with high enriched uranium feeding (REMIX-fuel) gives a possibility to terminate the accumulation of spent nuclear fuels (SNF) and Pu and decrease the accumulation of irradiated uranium by an order of magnitude. Results of comparison of VVER-1000 nuclear fuel cycle characteristics vs different fuel types such as UOX, MOX and REMIX-fuel have been presented. REMIX fuel (Regenerated Mixture of U-, Pu oxides) is the mixture of plutonium and uranium extracted from SNF and refined from other actinides and fission products with the addition of enriched uranium to provide the power potential necessary. The savings in terms of uranium quantities and separation works in the nuclear energy system (NES) with reactors using REMIX-fuel compared to the NES with uranium-fuelled reactors are shown to be of about 30% and 8%, respectively. For the NES with thermal reactors partially loaded with MOX-fuel, the uranium and separation works saving of about 14% would be obtained. Production of neptunium and americium in reactors with REMIX-fuel in steady state increases by a factor 3, and production of curium - by 10 compared to the reactors with UOX-fuel. This increase of minor actinide buildup is owed to the multiple recycling of plutonium. It should be noted that in this case all fuel assemblies contain high-background plutonium, and their manufacturing involves an expensive technology. Besides, management of REMIX-fuel will require special protection measures even during the fresh fuel manufacturing phase. The above-said gives ground to state that the use of REMIX fuel would be questionable in economic aspect.

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

    DOE Patents [OSTI]

    Rothman, Alan B. (Willowbrook, IL); Graczyk, Donald G. (Lemont, IL); Essling, Alice M. (Elmhurst, IL); Horwitz, E. Philip (Naperville, IL)

    2001-01-01T23:59:59.000Z

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

  9. Annual status report on the Uranium Mill Tailings Remedial Action Program

    SciTech Connect (OSTI)

    Not Available

    1989-12-01T23:59:59.000Z

    This eleventh annual status report summarizes activities of the Uranium Mill Tailings Remedial Action (UMTRA) Project undertaken during Fiscal Year (FY) 1989 by the US Department of Energy (DOE) and other agencies. Project goals for FY 1990 are also presented. An annual report of this type was a statutory requirement through January 1, 1986, pursuant to the Uranium Mill Tailings Radiation Control Act (UMTRCA) of 1978, Public Law (PL) 95--604. The DOE will continue to submit an annual report through project completion in order to inform the public of yearly project status. Title I of the UMTRCA authorizes the DOE, in cooperation with affected states and Indian tribes within whose boundaries designated uranium processing sites are located, to provide a program of assessment and remedial action at such sites. The purpose of the remedial action is to stabilize and control the tailings and other residual radioactive materials located on the inactive uranium processing sites in a safe and environmentally sound manner and to minimize or eliminate potential radiation health hazards. Commercial and residential properties in the vicinity of designated processing sites that are contaminated with material from the sites, herein referred to as vicinity properties,'' are also eligible for remedial action. Included in the UMTRA Project are 24 inactive uranium processing sites and associated vicinity properties located in 10 states, and the vicinity properties associated with Edgemont, South Dakota, an inactive uranium mill currently owned by the Tennessee Valley Authority (TVA).

  10. Project Management Plan

    SciTech Connect (OSTI)

    Not Available

    1988-01-01T23:59:59.000Z

    The mission of the Uranium Mill Tailings Remedial Action (UMTRA) Project is explicitly stated and directed in the Uranium Mill Tailings Radiation Control Act of 1978, Public Law 95-604, 42 USC 7901 (hereinafter referred to as the Act''). Title I of the Act authorizes the Department of Energy (DOE) to undertake remedial actions at 24 designated inactive uranium processing sites and associated vicinity properties containing uranium mill tailings and other residual radioactive materials derived from the processing sites. The Act, amended in January 1983, by Public Law 97-415, also authorizes DOE to perform remedial actions at vicinity properties in Edgemont, South Dakota. Cleanup of the Edgemont processing site is the responsibility of the Tennessee Valley Authority. This document describes the plan, organization, system, and methodologies used to manage the design, construction, and other activities required to clean up the designated sites and associated vicinity properties in accordance with the Act. The plan describes the objectives of the UMTRA Project, defines participants' roles and responsibilities, outlines the technical approach for accomplishing the objectives, and describes the planning and managerial controls to be used in integrating and performing the Project mission. 21 figs., 21 tabs.

  11. Trace Fission Product Ratios for Nuclear Forensics Attribution of Weapons-Grade Plutonium from Fast Breeder Reactor Blankets

    E-Print Network [OSTI]

    Osborn, Jeremy

    2014-08-13T23:59:59.000Z

    ), whereas that used in an FBR blanket fuel is depleted uranium (0.25 atom percent 235U). The energy production in the FBR core is from the seed fuel subassemblies containing mixed oxides (MOX) of PuO2 and UO2. A plot of fast and thermal neutron energy... of the program involves a fleet of fast breeder reactors. The stage two fast breeder reactors, beginning with the PFBR, will be fueled with reactor-grade plutonium and depleted uranium from the reprocessed spent fuel of stage one reactors and will breed more...

  12. Reduced-Enrichment Research and Test Reactor Program: Environmental assessment

    SciTech Connect (OSTI)

    Not Available

    1980-05-01T23:59:59.000Z

    The principal program objective and principal part of the proposed action is to improve the proliferation resistance of nuclear fuels used in research and test reactors by providing the technical means (through technical development, design, and testing) for reducing the uranium enrichment requirements of these fuels to substantially less than the 90 to 93% enrichment currently used. Operator acceptance of the reduced-enrichment-uranium (REU) fuel alternative will require minimizing of reactor performance reduction, fuel cycle cost increases, the number of new safety and licensing issues raised, and reactor and facility modifications. The other part of the proposed action is to assure the capability for commercial production and supply of REU fuel for use both in the US and abroad. The RERTR Program scope is limited to generic design studies, technical support to reactor operating organizations in preparing for conversions to REU fuels, fuel development, fuel demonstrations, and technical support for commercialization of REU fuels. This environmental assessment addresses the environmental consequences of RERTR Program activities and of specific conversions of typical reactors (the Ford Nuclear Reactor and one or two other to-be-designated demonstrations) to REU-fuel cycles, including domestic and international shipments of enriched uranium pertinent to the conduct of RERTR Program activities.

  13. GLOBAL THREAT REDUCTION INITIATIVE REACTOR CONVERSION PROGRAM: STATUS AND CURRENT PLANS

    SciTech Connect (OSTI)

    Staples, Parrish A.; Leach, Wayne; Lacey, Jennifer M.

    2009-10-07T23:59:59.000Z

    The U.S. Department of Energys National Nuclear Security Administration (NNSA) Reactor Conversion Program supports the minimization, and to the extent possible, elimination of the use of high enriched uranium (HEU) in civilian nuclear applications by working to convert research and test reactors and radioisotope production processes to the use of low enriched uranium (LEU). The Reactor Conversion Program is a technical pillar of the NNSA Global Threat Reduction Initiative (GTRI) which is a key organization for implementing U.S. HEU minimization policy and works to reduce and protect vulnerable nuclear and radiological material domestically and abroad.

  14. Oxidation Protection of Uranium Nitride Fuel using Liquid Phase Sintering

    SciTech Connect (OSTI)

    Dr. Paul A. Lessing

    2012-03-01T23:59:59.000Z

    Two methods are proposed to increase the oxidation resistance of uranium nitride (UN) nuclear fuel. These paths are: (1) Addition of USi{sub x} (e.g. U3Si2) to UN nitride powder, followed by liquid phase sintering, and (2) 'alloying' UN nitride with various compounds (followed by densification via Spark Plasma Sintering or Liquid Phase Sintering) that will greatly increase oxidation resistance. The advantages (high thermal conductivity, very high melting point, and high density) of nitride fuel have long been recognized. The sodium cooled BR-10 reactor in Russia operated for 18 years on uranium nitride fuel (UN was used as the driver fuel for two core loads). However, the potential advantages (large power up-grade, increased cycle lengths, possible high burn-ups) as a Light Water Reactor (LWR) fuel are offset by uranium nitride's extremely low oxidation resistance (UN powders oxidize in air and UN pellets decompose in hot water). Innovative research is proposed to solve this problem and thereby provide an accident tolerant LWR fuel that would resist water leaks and high temperature steam oxidation/spalling during an accident. It is proposed that we investigate two methods to increase the oxidation resistance of UN: (1) Addition of USi{sub x} (e.g. U{sub 3}Si{sub 2}) to UN nitride powder, followed by liquid phase sintering, and (2) 'alloying' UN nitride with compounds (followed by densification via Spark Plasma Sintering) that will greatly increase oxidation resistance.

  15. 2013 Domestic Uranium Production Report

    E-Print Network [OSTI]

    Administration (EIA), the statistical and analytical agency within the U.S. Department of Energy. By law, EIA.S. Energy Information Administration | 2013 Domestic Uranium Production Report iii Preface The U.S. Energy://www.eia.doe.gov/glossary/. #12;U.S. Energy Information Administration | 2013 Domestic Uranium Production Report iv Contents

  16. Uranium Metal Reaction Behavior in Water, Sludge, and Grout Matrices

    SciTech Connect (OSTI)

    Delegard, Calvin H.; Schmidt, Andrew J.

    2008-09-25T23:59:59.000Z

    This report summarizes information and data on the reaction behavior of uranium metal in water, in water-saturated simulated and genuine K Basin sludge, and in grout matrices. This information and data are used to establish the technical basis for metallic uranium reaction behavior for the K Basin Sludge Treatment Project (STP). The specific objective of this report is to consolidate the various sources of information into a concise document to serve as a high-level reference and road map for customers, regulators, and interested parties outside the STP (e.g., external reviewers, other DOE sites) to clearly understand the current basis for the corrosion of uranium metal in water, sludge, and grout.

  17. Uranium Metal Reaction Behavior in Water, Sludge, and Grout Matrices

    SciTech Connect (OSTI)

    Delegard, Calvin H.; Schmidt, Andrew J.

    2009-05-27T23:59:59.000Z

    This report summarizes information and data on the reaction behavior of uranium metal in water, in water-saturated simulated and genuine K Basin sludge, and in grout matrices. This information and data are used to establish the technical basis for metallic uranium reaction behavior for the K Basin Sludge Treatment Project (STP). The specific objective of this report is to consolidate the various sources of information into a concise document to serve as a high-level reference and road map for customers, regulators, and interested parties outside the STP (e.g., external reviewers, other DOE sites) to clearly understand the current basis for the corrosion of uranium metal in water, sludge, and grout.

  18. Fernald vacuum transfer system for uranium materials repackaging

    SciTech Connect (OSTI)

    Kaushiva, Shirley; Weekley, Clint; Molecke, Martin; Polansky, Gary

    2002-02-24T23:59:59.000Z

    The Fernald Environmental Management Project (FEMP) is the site of a former Department of Energy (DOE) uranium processing plant. When production was halted, many materials were left in an intermediate state. Some of this product material included enriched uranium compounds that had to be repackaged for shipment of off-site storage. This paper provides an overview, technical description, and status of a new application of existing technology, a vacuum transfer system, to repackage the uranium bearing compounds for shipment. The vacuum transfer system provides a method of transferring compounds from their current storage configuration into packages that meet the Department of Transportation (DOT) shipping requirements for fissile materials. This is a necessary activity, supporting removal of nuclear materials prior to site decontamination and decommissioning, key to the Fernald site's closure process.

  19. Evaluation of integrated data sets: four examples. [Uranium deposits (exploration)

    SciTech Connect (OSTI)

    Bolivar, S.L.; Freeman, S.B.; Weaver, T.A.

    1982-01-01T23:59:59.000Z

    Several large data sets have been integrated and utilized for rapid evaluation on a reconnaissance scale for the Montrose 1/sup 0/ x 2/sup 0/ quadrangle, Colorado. The data sets include Landsat imagery, hydrogeochemical and stream sediment analyses, airborne geophysical data, known mineral occurrences, and a geologic map. All data sets were registered to a 179 x 119 rectangular grid and projected onto Universal Transverse Mercator coordinates. A grid resolution of 1 km was used. All possible combinations of three, for most data sets, were examined for general geologic correlations by utilizing a color microfilm output. In addition, gray-level pictures of statistical output, e.g., factor analysis, have been employed to aid evaluations. Examples for the data sets dysprosium-calcium, lead-copper-zinc, and equivalent uranium-uranium in water-uranium in sediment are described with respect to geologic applications, base-metal regimes, and geochemical associations.

  20. Effect of reduced enrichment on the fuel cycle for research reactors

    SciTech Connect (OSTI)

    Travelli, A.

    1982-01-01T23:59:59.000Z

    The new fuels developed by the RERTR Program and by other international programs for application in research reactors with reduced uranium enrichment (<20% EU) are discussed. It is shown that these fuels, combined with proper fuel-element design and fuel-management strategies, can provide at least the same core residence time as high-enrichment fuels in current use, and can frequently significantly extend it. The effect of enrichment reduction on other components of the research reactor fuel cycle, such as uranium and enrichment requirements, fuel fabrication, fuel shipment, and reprocessing are also briefly discussed with their economic implications. From a systematic comparison of HEU and LEU cores for the same reference research reactor, it is concluded that the new fuels have a potential for reducing the research reactor fuel cycle costs while reducing, at the same time, the uranium enrichment of the fuel.

  1. Domestic Uranium Production Report

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines AboutDecember 2005 (Thousand9, 2015Year109 AppendixCostsDistributedSep-1410. Uranium

  2. Domestic Uranium Production Report

    Annual Energy Outlook 2013 [U.S. Energy Information Administration (EIA)]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines AboutDecember 2005 (Thousand9, 2015Year109 AppendixCostsDistributedSep-1410. Uranium9.

  3. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines About U.S. NaturalA. Michael SchaalNovember1. Foreign sales of uranium from U.S.

  4. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines About U.S. NaturalA. Michael SchaalNovember1. Foreign sales of uranium from U.S.2.

  5. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines About U.S. NaturalA. Michael SchaalNovember1. Foreign sales of uranium from U.S.2.3.

  6. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines About U.S. NaturalA. Michael SchaalNovember1. Foreign sales of uranium from U.S.2.3.5.

  7. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines About U.S. NaturalA. Michael SchaalNovember1. Foreign sales of uranium from U.S.2.3.5.3.

  8. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines About U.S. NaturalA. Michael SchaalNovember1. Foreign sales of uranium from

  9. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines About U.S. NaturalA. Michael SchaalNovember1. Foreign sales of uranium froma. Uraniumb.

  10. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines About U.S. NaturalA. Michael SchaalNovember1. Foreign sales of uranium froma. Uraniumb.7.

  11. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines About U.S. NaturalA. Michael SchaalNovember1. Foreign sales of uranium froma.

  12. Uranium Marketing Annual Report -

    Gasoline and Diesel Fuel Update (EIA)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) " ,"ClickPipelines About U.S. NaturalA. Michael SchaalNovember1. Foreign sales of uranium froma.9.

  13. Fingerprinting Uranium | EMSL

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth7-1D: Vegetation ProposedUsing ZirconiaPolicyFeasibilityField Office FinalFinancingFingerprinting Uranium

  14. Uranium- and thorium-bearing pegmatites of the United States

    SciTech Connect (OSTI)

    Adams, J.W.; Arengi, J.T.; Parrish, I.S.

    1980-04-01T23:59:59.000Z

    This report is part of the National Uranium Resource Evaluation (NURE) Program designed to identify criteria favorable for the occurrence of the world's significant uranium deposits. This project deals specifically with uranium- and thorium-bearing pegmatites in the United States and, in particular, their distribution and origin. From an extensive literature survey and field examination of 44 pegmatite localities in the United States and Canada, the authors have compiled an index to about 300 uranium- and thorium-bearing pegmatites in the United States, maps giving location of these deposits, and an annotated bibliography to some of the most pertinent literature on the geology of pegmatites. Pegmatites form from late-state magma differentiates rich in volatile constituents with an attendant aqueous vapor phase. It is the presence of an aqueous phase which results in the development of the variable grain size which characterizes pegmatites. All pegmatites occur in areas of tectonic mobility involving crustal material usually along plate margins. Those pegmatites containing radioactive mineral species show, essentially, a similar distribution to those without radioactive minerals. Criteria such as tectonic setting, magma composition, host rock, and elemental indicators among others, all serve to help delineate areas more favorable for uranium- and thorium-bearing pegmatites. The most useful guide remains the radioactivity exhibited by uranium- and thorium-bearing pegmatites. Although pegmatites are frequently noted as favorable hosts for radioactive minerals, the general paucity and sporadic distribution of these minerals and inherent mining and milling difficulties negate the resource potential of pegmatites for uranium and thorium.

  15. Fuel and cladding nano-technologies based solutions for long life heat-pipe based reactors

    SciTech Connect (OSTI)

    Popa-Simil, L. [LAVM LLC, Los Alamos (United States)

    2012-07-01T23:59:59.000Z

    A novel nuclear reactor concept, unifying the fuel pipe with fuel tube functionality has been developed. The structure is a quasi-spherical modular reactor, designed for a very long life. The reactor module unifies the fuel tube with the heat pipe and a graphite beryllium reflector. It also uses a micro-hetero-structure that allows the fission products to be removed in the heat pipe flow and deposited in a getter area in the cold zone of the heat pipe, but outside the neutron flux. The reactor operates as a breed and burn reactor - it contains the fuel pipe with a variable enrichment, starting from the hot-end of the pipe, meant to assure the initial criticality, and reactor start-up followed by area with depleted uranium or thorium that get enriched during the consumption of the first part of the enriched uranium. (authors)

  16. E-Print Network 3.0 - auxiliary reactor area-1 Sample Search...

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

    plans eyed for fusion project site Summary: said. The six parties involved in the Thermonuclear Experimental Reactor (ITER) project have been... unable to agree where to locate...

  17. Safety requirements, facility user needs, and reactor concepts for a new Broad Application Test Reactor

    SciTech Connect (OSTI)

    Ryskamp, J.M. [ed.; Liebenthal, J.L.; Denison, A.B.; Fletcher, C.D.

    1992-07-01T23:59:59.000Z

    This report describes the EG&G Laboratory Directed Research and Development Program (LDRD) Broad Application Test Reactor (BATR) Project that was conducted in fiscal year 1991. The scope of this project was divided into three phases: a project process definition phase, a requirements development phase, and a preconceptual reactor design and evaluation phase. Multidisciplinary teams of experts conducted each phase. This report presents the need for a new test reactor, the project process definition, a set of current and projected regulatory compliance and safety requirements, a set of facility user needs for a broad range of projected testing missions, and descriptions of reactor concepts capable of meeting these requirements. This information can be applied to strategic planning to provide the Department of Energy with management options.

  18. Safety requirements, facility user needs, and reactor concepts for a new Broad Application Test Reactor

    SciTech Connect (OSTI)

    Ryskamp, J.M. (ed.); Liebenthal, J.L.; Denison, A.B.; Fletcher, C.D.

    1992-07-01T23:59:59.000Z

    This report describes the EG G Laboratory Directed Research and Development Program (LDRD) Broad Application Test Reactor (BATR) Project that was conducted in fiscal year 1991. The scope of this project was divided into three phases: a project process definition phase, a requirements development phase, and a preconceptual reactor design and evaluation phase. Multidisciplinary teams of experts conducted each phase. This report presents the need for a new test reactor, the project process definition, a set of current and projected regulatory compliance and safety requirements, a set of facility user needs for a broad range of projected testing missions, and descriptions of reactor concepts capable of meeting these requirements. This information can be applied to strategic planning to provide the Department of Energy with management options.

  19. Life cycle costs for the domestic reactor-based plutonium disposition option

    SciTech Connect (OSTI)

    Williams, K.A.

    1999-10-01T23:59:59.000Z

    Projected constant dollar life cycle cost (LCC) estimates are presented for the domestic reactor-based plutonium disposition program being managed by the US Department of Energy Office of Fissile Materials Disposition (DOE/MD). The scope of the LCC estimate includes: design, construction, licensing, operation, and deactivation of a mixed-oxide (MOX) fuel fabrication facility (FFF) that will be used to purify and convert weapons-derived plutonium oxides to MOX fuel pellets and fabricate MOX fuel bundles for use in commercial pressurized-water reactors (PWRs); fuel qualification activities and modification of facilities required for manufacture of lead assemblies that will be used to qualify and license this MOX fuel; and modification, licensing, and operation of commercial PWRs to allow irradiation of a partial core of MOX fuel in combination with low-enriched uranium fuel. The baseline cost elements used for this document are the same as those used for examination of the preferred sites described in the site-specific final environmental impact statement and in the DOE Record of Decision that will follow in late 1999. Cost data are separated by facilities, government accounting categories, contract phases, and expenditures anticipated by the various organizations who will participate in the program over a 20-year period. Total LCCs to DOE/MD are projected at approximately $1.4 billion for a 33-MT plutonium disposition mission.

  20. Procedures for preparation, printing, and distribution of UMTRA Project National Environmental Policy Act documents

    SciTech Connect (OSTI)

    Not Available

    1988-08-01T23:59:59.000Z

    This paper discusses the procedures for preparation, printing and distribution of Uranium Mill Tailings Remedial Action (UMTRA) Project National Environmental Policy Act documents. (KJD)

  1. Nuclear reactor engineering

    SciTech Connect (OSTI)

    Glasstone, S.; Sesonske, A.

    1981-01-01T23:59:59.000Z

    Chapters are presented concerning energy from nuclear fission; nuclear reactions and radiations; diffusion and slowing-down of neutrons; principles of reactor analysis; nuclear reactor kinetics and control; energy removal; non-fuel reactor materials; the reactor fuel system; radiation protection and environmental effects; nuclear reactor shielding; nuclear reactor safety; and power reactor systems.

  2. Solidus and liquidus temperatures in the uranium-plutonium-zirconium system

    SciTech Connect (OSTI)

    Leibowitz, L.; Veleckis, E.; Blomquist, R.A.; Pelton, A.D.

    1987-01-01T23:59:59.000Z

    Renewed interest in metallic fuel for nuclear reactors has prompted study of the solidus and liquidus for the uranium-plutonium-zirconium system. These temperatures are of importance in assessing the possibility of fuel melting during abnormal reactor conditions. Data obtained in previous work in this area were found to be inadequate for the needs of the current reactor development effort. A dual effort was undertaken to provide the needed data. These were (1) thermodynamic phase diagram analysis and calculation of the ternary solidus and liquidus surfaces and (2) experimental determination of solidus and liquidus temperatures for selected alloys. The methods used and results obtained are described.

  3. Advanced Test Reactor National Scientific User Facility Progress

    SciTech Connect (OSTI)

    Frances M. Marshall; Todd R. Allen; James I. Cole; Jeff B. Benson; Mary Catherine Thelen

    2012-10-01T23:59:59.000Z

    The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) is one of the worlds premier test reactors for studying the effects of intense neutron radiation on reactor materials and fuels. The ATR began operation in 1967, and has operated continuously since then, averaging approximately 250 operating days per year. The combination of high flux, large test volumes, and multiple experiment configuration options provide unique testing opportunities for nuclear fuels and material researchers. The ATR is a pressurized, light-water moderated and cooled, beryllium-reflected highly-enriched uranium fueled, reactor with a maximum operating power of 250 MWth. The ATR peak thermal flux can reach 1.0 x1015 n/cm2-sec, and the core configuration creates five main reactor power lobes (regions) that can be operated at different powers during the same operating cycle. In addition to these nine flux traps there are 68 irradiation positions in the reactor core reflector tank. The test positions range from 0.5 to 5.0 in diameter and are all 48 in length, the active length of the fuel. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material radiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. Goals of the ATR NSUF are to define the cutting edge of nuclear technology research in high temperature and radiation environments, contribute to improved industry performance of current and future light water reactors, and stimulate cooperative research between user groups conducting basic and applied research. The ATR NSUF has developed partnerships with other universities and national laboratories to enable ATR NSUF researchers to perform research at these other facilities, when the research objectives cannot be met using the INL facilities. The ATR NSUF program includes a robust education program enabling students to participate in their research at INL and the partner facilities, attend the ATR NSUF annual User Week, and compete for prizes at sponsored conferences. Development of additional research capabilities is also a key component of the ATR NSUF Program; researchers are encouraged to propose research projects leading to these enhanced capabilities. Some ATR irradiation experiment projects irradiate more specimens than are tested, resulting in irradiated materials available for post irradiation examination by other researchers. These extra specimens comprise the ATR NSUF Sample Library. This presentation will highlight the ATR NSUF Sample Library and the process open to researchers who want to access these materials and how to propose research projects using them. This presentation will provide the current status of all the ATR NSUF Program elements. Many of these were not envisioned in 2007, when DOE established the ATR NSUF.

  4. APPENDIX J Partition Coefficients For Uranium

    E-Print Network [OSTI]

    APPENDIX J Partition Coefficients For Uranium #12;Appendix J Partition Coefficients For Uranium J.1.0 Background The review of uranium Kd values obtained for a number of soils, crushed rock and their effects on uranium adsorption on soils are discussed below. The solution pH was also used as the basis

  5. Research reactors - an overview

    SciTech Connect (OSTI)

    West, C.D.

    1997-03-01T23:59:59.000Z

    A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs.

  6. FROM CONCEPT TO REALITY, IN-SITU DECOMMISSIONING OF THE P AND R REACTORS AT THE SAVANNAH RIVER SITE

    SciTech Connect (OSTI)

    Musall, J.; Blankenship, J.; Griffin, W.

    2012-01-09T23:59:59.000Z

    SRS recently completed an approximately three year effort to decommission two SRS reactors: P-Reactor (Building 105-P) and R-Reactor (Building 105-R). Completed in December 2011, the concurrent decommissionings marked the completion of two relatively complex and difficult facility disposition projects at the SRS. Buildings 105-P and 105-R began operating as production reactors in the early 1950s with the mission of producing weapons material (e.g., tritium and plutonium-239). The 'P' Reactor and was shutdown in 1991 while the 'R' Reactor and was shutdown in 1964. In the intervening period between shutdown and deactivation & decommissioning (D&D), Buildings 105-P and 105-R saw limited use (e.g., storage of excess heavy water and depleted uranium oxide). For Building 105-P, deactivation was initiated in April 2007 and was essentially complete by June 2010. For Building 105-R, deactivation was initiated in August 2008 and was essentially complete by September 2010. For both buildings, the primary objective of deactivation was to remove/mitigate hazards associated with the remaining hazardous materials, and thus prepare the buildings for in-situ decommissioning. Deactivation removed the following hazardous materials to the extent practical: combustibles/flammables, residual heavy water, acids, friable asbestos (as needed to protect workers performing deactivation and decommissioning), miscellaneous chemicals, lead/brass components, Freon(reg sign), oils, mercury/PCB containing components, mold and some radiologically-contaminated equipment. In addition to the removal of hazardous materials, deactivation included the removal of hazardous energy, exterior metallic components (representing an immediate fall hazard), and historical artifacts along with the evaporation of water from the two Disassembly Basins. Finally, so as to facilitate occupancy during the subsequent in-situ decommissioning, deactivation implemented repairs to the buildings and provided temporary power.

  7. Standard specification for sintered (Uranium-Plutonium) dioxide pellets

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2001-01-01T23:59:59.000Z

    1.1 This specification covers finished sintered and ground (uranium-plutonium) dioxide pellets for use in thermal reactors. It applies to uranium-plutonium dioxide pellets containing plutonium additions up to 15 % weight. This specification may not completely cover the requirements for pellets fabricated from weapons-derived plutonium. 1.2 This specification does not include (1) provisions for preventing criticality accidents or (2) requirements for health and safety. Observance of this specification does not relieve the user of the obligation to be aware of and conform to all applicable international, federal, state, and local regulations pertaining to possessing, processing, shipping, or using source or special nuclear material. Examples of U.S. government documents are Code of Federal Regulations Title 10, Part 50Domestic Licensing of Production and Utilization Facilities; Code of Federal Regulations Title 10, Part 71Packaging and Transportation of Radioactive Material; and Code of Federal Regulations Tit...

  8. Investigation of the low enrichment conversion of the Texas A and M Nuclear Science Center Reactor

    SciTech Connect (OSTI)

    Reuscher, J.A.

    1988-01-01T23:59:59.000Z

    The use of highly enriched uranium as a fuel research reactors is of concern due to the possibility of diversion for nuclear weapons applications. The Texas A M TRIGA reactor currently uses 70% enriched uranium in a FLIP (Fuel Life Improvement Program) fuel element manufactured by General Atomics. Thus fuel also contains 1.5 weight percent of erbium as a burnable poison to prolong useful core life. US university reactors that use highly enriched uranium will be required to covert to 20% or less enrichment to satisfy Nuclear Regulatory Commission requirements for the next core loading if the fuel is available. This investigation examined the feasibility of a material alternate to uranium-zirconium hydride for LEU conversion of a TRIGA reactor. This material is a beryllium oxide uranium dioxide based fuel. The theoretical aspects of core physics analyses were examined to assess the potential advantages of the alternative fuel. A basic model was developed for the existing core configuration since it is desired to use the present fuel element grid for the replacement core. The computing approach was calibrated to the present core and then applied to a core of BeO-UO{sub 2} fuel elements. Further calculations were performed for the General Atomics TRIGA low-enriched uranium zirconium hydride fuel.

  9. Light Water Reactor Sustainability

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

    3 Light Water Reactor Sustainability Program ACCOMPLISHMENTS REPORT 2013 Accomplishments Report | Light Water Reactor Sustainability 2 T he mission of the Light Water Reactor...

  10. Light Water Reactor Sustainability

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

    4 Light Water Reactor Sustainability ACCOMPLISHMENTS REPORT 2014 Accomplishments Report | Light Water Reactor Sustainability 2 T he mission of the Light Water Reactor...

  11. James P. Mosquera Director, Reactor Plant Components

    E-Print Network [OSTI]

    of the application of nuclear reactor power to capital ships of the U.S. Navy, and other assigned projects. Mr for steam generator technology (within the Nuclear Components Division); and power plant systems engineer working for the U.S. Naval Nuclear Propulsion Program (a.k.a. Naval Reactors). This program is a joint

  12. University of Virginia Reactor Facility Decommissioning Results

    SciTech Connect (OSTI)

    Ervin, P. F.; Lundberg, L. A.; Benneche, P. E.; Mulder, R. U.; Steva, D. P.

    2003-02-24T23:59:59.000Z

    The University of Virginia Reactor Facility started accelerated decommissioning in 2002. The facility consists of two licensed reactors, the CAVALIER and the UVAR. This paper will describe the progress in 2002, remaining efforts and the unique organizational structure of the project team.

  13. The End of Cheap Uranium

    E-Print Network [OSTI]

    Michael Dittmar

    2011-06-21T23:59:59.000Z

    Historic data from many countries demonstrate that on average no more than 50-70% of the uranium in a deposit could be mined. An analysis of more recent data from Canada and Australia leads to a mining model with an average deposit extraction lifetime of 10+- 2 years. This simple model provides an accurate description of the extractable amount of uranium for the recent mining operations. Using this model for all larger existing and planned uranium mines up to 2030, a global uranium mining peak of at most 58 +- 4 ktons around the year 2015 is obtained. Thereafter we predict that uranium mine production will decline to at most 54 +- 5 ktons by 2025 and, with the decline steepening, to at most 41 +- 5 ktons around 2030. This amount will not be sufficient to fuel the existing and planned nuclear power plants during the next 10-20 years. In fact, we find that it will be difficult to avoid supply shortages even under a slow 1%/year worldwide nuclear energy phase-out scenario up to 2025. We thus suggest that a worldwide nuclear energy phase-out is in order. If such a slow global phase-out is not voluntarily effected, the end of the present cheap uranium supply situation will be unavoidable. The result will be that some countries will simply be unable to afford sufficient uranium fuel at that point, which implies involuntary and perhaps chaotic nuclear phase-outs in those countries involving brownouts, blackouts, and worse.

  14. United States Domestic Research Reactor Infrastrucutre TRIGA Reactor Fuel Support

    SciTech Connect (OSTI)

    Douglas Morrell

    2011-03-01T23:59:59.000Z

    The United State Domestic Research Reactor Infrastructure Program at the Idaho National Laboratory manages and provides project management, technical, quality engineering, quality inspection and nuclear material support for the United States Department of Energy sponsored University Reactor Fuels Program. This program provides fresh, unirradiated nuclear fuel to Domestic University Research Reactor Facilities and is responsible for the return of the DOE-owned, irradiated nuclear fuel over the life of the program. This presentation will introduce the program management team, the universities supported by the program, the status of the program and focus on the return process of irradiated nuclear fuel for long term storage at DOE managed receipt facilities. It will include lessons learned from research reactor facilities that have successfully shipped spent fuel elements to DOE receipt facilities.

  15. Accelerating the Reduction of Excess Russian Highly Enriched Uranium

    SciTech Connect (OSTI)

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

    2004-02-18T23:59:59.000Z

    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.

  16. Uranium Neutron Coincidence Collar Model Utilizing Boron-10 Lined Tubes

    SciTech Connect (OSTI)

    Rogers, Jeremy L.; Ely, James H.; Kouzes, Richard T.; Lintereur, Azaree T.; Siciliano, Edward R.

    2012-09-18T23:59:59.000Z

    The Department of Energy Office of Nuclear Safeguards and Security (NA-241) is supporting the project Coincidence Counting With Boron-Based Alternative Neutron Detection Technology at Pacific Northwest National Laboratory (PNNL) for the development of a 3He proportional counter alternative neutron coincidence counter. The goal of this project is to design, build and demonstrate a system based upon 10B-lined proportional tubes in a configuration typical for 3He-based coincidence counter applications. This report, providing results for model development of Alternative Boron-Based Uranium Neutron Coincidence Collar (ABUNCL) designs, is a deliverable under Task 2 of the project.

  17. Safe Operating Procedure SAFETY PROTOCOL: URANIUM

    E-Print Network [OSTI]

    Farritor, Shane

    involve the use of natural or depleted uranium. Natural isotopes of uranium are U-238, U-235 and U-234 (see Table 1 for natural abundances). Depleted uranium contains less of the isotopes: U-235 and U-234. The specific activity of depleted uranium (5.0E-7 Ci/g) is less than that of natural uranium (7.1E-7 Ci

  18. DEPARTMENT OF ENERGY Excess Uranium Management: Effects of DOE...

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

    Excess Uranium Management: Effects of DOE Transfers of Excess Uranium on Domestic Uranium Mining, Conversion, and Enrichment Industries; Request for Information AGENCY: Office of...

  19. The U.S. Uranium Mill Tailings Radiation Control Act -- An environmental legacy of the Cold War

    SciTech Connect (OSTI)

    Watson, C.D.; Nelson, R.A. [Jacobs Engineering Group Inc., Albuquerque, NM (United States). Albuquerque Operations Office; Mann, P. [USDOE Albuquerque Operations Office, NM (United States)

    1993-12-31T23:59:59.000Z

    The US Department of Energy (DOE) has guided the Uranium Mill Tailings Remedial Action (UMTRA) Project through its first 10 years of successful remediation. The Uranium Mill Tailings Radiation Control Act (UMTRCA), passed in 1978, identified 24 uranium mill tailings sites in need of remediation to protect human health and the environment from the residual contamination resulting from the processing of uranium ore. The UMTRCA was promulgated in two titles: Title 1 and Title 2. This paper describes the regulatory structure, required documentation, and some of the technical approaches used to meet the Act`s requirements for managing and executing the $1.4 billion project under Title 1. Remedial actions undertaken by private industry under Title 2 of the Act are not addressed in this paper. Some of the lessons learned over the course of the project`s history are presented so that other countries conducting similar remedial action activities may benefit.

  20. Environmental health and safety plan for the Molten Salt Reactor Experiment Remediation Project at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    SciTech Connect (OSTI)

    Burman, S.N.; Tiner, P.F.; Gosslee, R.C.

    1998-01-01T23:59:59.000Z

    The Lockheed Martin Energy Systems, Inc. (Energy Systems) policy is to provide a safe and healthful workplace for all employees and subcontractors. The accomplishment of this policy requires that operations at the Molten Salt Reactor Experiment (MSRE) facility at the Department of Energy (DOE) Oak Ridge National Laboratory (ORNL) are guided by an overall plan and consistent proactive approach to environmental protection and safety and health (S and H) issues. The policy and procedures in this plan apply to all MSRE operations. The provisions of this plan are to be carried out whenever activities are initiated at the MSRE that could be a threat to human health or the environment. This plan implements a policy and establishes criteria for the development of procedures for day-to-day operations to prevent or minimize any adverse impact to the environment and personnel safety and health and to meet standards that define acceptable management of hazardous and radioactive materials and wastes. The plan is written to utilize past experience and the best management practices to minimize hazards to human health or the environment from events such as fires, explosions, falls, mechanical hazards, or any unplanned release of hazardous or radioactive materials to the air.

  1. Adsorbent materials development and testing for the extraction of uranium from seawater

    SciTech Connect (OSTI)

    Felker, L.K.; Dai, S.; Hay, B.P.; Janke, C.J.; Mayes, R.T.; Sun, X.; Tsouris, C. [Oak Ridge National Laboratory, Oak Ridge, Tennessee, 37831-6384 (United States)

    2013-07-01T23:59:59.000Z

    The extraction of uranium from seawater has been the focus of a research project for the U.S. Department of Energy to develop amidoxime functional group adsorbents using radiation-induced graphing on polymer-based fiber materials and subsequent chemical conversion of the radical sites to form the desired adsorbent material. Materials with promising uranium adsorption capacities were prepared through a series of parametric studies on radiation dose, time, temperature, graphing solutions, and properties of the base polymer materials. A laboratory screening protocol was developed to determine the uranium adsorption capacity to identify the most promising candidate materials for seawater testing. (authors)

  2. Catalytic reactor

    DOE Patents [OSTI]

    Aaron, Timothy Mark (East Amherst, NY); Shah, Minish Mahendra (East Amherst, NY); Jibb, Richard John (Amherst, NY)

    2009-03-10T23:59:59.000Z

    A catalytic reactor is provided with one or more reaction zones each formed of set(s) of reaction tubes containing a catalyst to promote chemical reaction within a feed stream. The reaction tubes are of helical configuration and are arranged in a substantially coaxial relationship to form a coil-like structure. Heat exchangers and steam generators can be formed by similar tube arrangements. In such manner, the reaction zone(s) and hence, the reactor is compact and the pressure drop through components is minimized. The resultant compact form has improved heat transfer characteristics and is far easier to thermally insulate than prior art compact reactor designs. Various chemical reactions are contemplated within such coil-like structures such that as steam methane reforming followed by water-gas shift. The coil-like structures can be housed within annular chambers of a cylindrical housing that also provide flow paths for various heat exchange fluids to heat and cool components.

  3. Bioconversion reactor

    DOE Patents [OSTI]

    McCarty, Perry L. (Stanford, CA); Bachmann, Andre (Palo Alto, CA)

    1992-01-01T23:59:59.000Z

    A bioconversion reactor for the anaerobic fermentation of organic material. The bioconversion reactor comprises a shell enclosing a predetermined volume, an inlet port through which a liquid stream containing organic materials enters the shell, and an outlet port through which the stream exits the shell. A series of vertical and spaced-apart baffles are positioned within the shell to force the stream to flow under and over them as it passes from the inlet to the outlet port. The baffles present a barrier to the microorganisms within the shell causing them to rise and fall within the reactor but to move horizontally at a very slow rate. Treatment detention times of one day or less are possible.

  4. Operating and life-cycle costs for uranium-contaminated soil treatment technologies

    SciTech Connect (OSTI)

    Douthat, D.M.; Armstrong, A.Q. [Oak Ridge National Lab., TN (United States). Health Sciences Research Div.; Stewart, R.N. [Univ. of Tennessee, Knoxville, TN (United States)

    1995-09-01T23:59:59.000Z

    The development of a nuclear industry in the US required mining, milling, and fabricating a large variety of uranium products. One of these products was purified uranium metal which was used in the Savannah River and Hanford Site reactors. Most of this feed material was produced at the US Department of Energy (DOE) facility formerly called the Feed Materials Production Center at Fernald, Ohio. During operation of this facility, soils became contaminated with uranium from a variety of sources. To avoid disposal of these soils in low-level radioactive waste burial sites, increasing emphasis has been placed on the remediating soils contaminated with uranium and other radionuclides. To address remediation and management of uranium-contaminated soils at sites owned by DOE, the DOE Office of Technology Development (OTD) evaluates and compares the versatility, efficiency, and economics of various technologies that may be combined into systems designed to characterize and remediate uranium-contaminated soils. Each technology must be able to (1) characterize the uranium in soil, (2) decontaminate or remove uranium from soil, (3) treat or dispose of resulting waste streams, (4) meet necessary state and federal regulations, and (5) meet performance assessment objectives. The role of the performance assessment objectives is to provide the information necessary to conduct evaluations of the technologies. These performance assessments provide the basis for selecting the optimum system for remediation of large areas contaminated with uranium. One of the performance assessment tasks is to address the economics of full-scale implementation of soil treatment technologies. The cost of treating contaminated soil is one of the criteria used in the decision-making process for selecting remedial alternatives.

  5. University Reactor Matching Grants Program

    SciTech Connect (OSTI)

    John Valentine; Farzad Rahnema; Said Abdel-Khalik

    2003-02-14T23:59:59.000Z

    During the 2002 Fiscal year, funds from the DOE matching grant program, along with matching funds from the industrial sponsors, have been used to support research in the area of thermal-hydraulics. Both experimental and numerical research projects have been performed. Experimental research focused on two areas: (1) Identification of the root cause mechanism for axial offset anomaly in pressurized water reactors under prototypical reactor conditions, and (2) Fluid dynamic aspects of thin liquid film protection schemes for inertial fusion reactor chambers. Numerical research focused on two areas: (1) Multi-fluid modeling of both two-phase and two-component flows for steam conditioning and mist cooling applications, and (2) Modeling of bounded Rayleigh-Taylor instability with interfacial mass transfer and fluid injection through a porous wall simulating the ''wetted wall'' protection scheme in inertial fusion reactor chambers. Details of activities in these areas are given.

  6. RERTR 2009 (Reduced Enrichment for Research and Test Reactors)

    SciTech Connect (OSTI)

    Totev, T.; Stevens, J.; Kim, Y. S.; Hofman, G.; Matos, J.; Hanan, N.; Garner, P.; Dionne, B.; Olson, A.; Feldman, E.; Dunn, F.; Nuclear Engineering Division; Atomic Research Center; Inst. of Nuclear Physics; LLNL; INL; Korea Atomic Energy Research Inst.; Comisi?n Nacional de Energ?a At?mica; Nuclear Reactor Lab.; Inst. of Atomic Energy-Poland; AECL-Canada; Hungarian Academy of Sciences KFKI Atomic Energy Research Inst.; Japan Atomic Energy Agency; Nuclear Power Inst. of China; Kyoto Univ. Research Reactor Inst.

    2010-03-01T23:59:59.000Z

    The U.S. Department of Energy/National Nuclear Security Administration's Office of Global Threat Reduction in cooperation with the China Atomic Energy Authority and International Atomic Energy Agency hosted the 'RERTR 2009 International Meeting on Reduced Enrichment for Research and Test Reactors.' The meeting was organized by Argonne National Laboratory, China Institute of Atomic Energy and Idaho National Laboratory and was held in Beijing, China from November 1-5, 2009. This was the 31st annual meeting in a series on the same general subject regarding the conversion of reactors within the Global Threat Reduction Initiative (GTRI). The Reduced Enrichment for Research and Test Reactors (RERTR) Program develops technology necessary to enable the conversion of civilian facilities using high enriched uranium (HEU) to low enriched uranium (LEU) fuels and targets.

  7. Reduced enrichment for research and test reactors: Proceedings

    SciTech Connect (OSTI)

    Not Available

    1988-05-01T23:59:59.000Z

    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.

  8. Low power reactor for remote applications

    SciTech Connect (OSTI)

    Meier, K.L.; Palmer, R.G.; Kirchner, W.L.

    1985-01-01T23:59:59.000Z

    A compact, low power reactor is being designed to provide electric power for remote, unattended applications. Because of the high fuel and maintenance costs for conventional power sources such as diesel generators, a reactor power supply appears especially attractive for remote and inaccessible locations. Operating at a thermal power level of 135 kWt, the power supply achieves a gross electrical output of 25 kWe from an organic Rankine cycle (ORC) engine. By intentional selection of design features stressing inherent safety, operation in an unattended mode is possible with minimal risk to the environment. Reliability is achieved through the use of components representing existing, proven technology. Low enrichment uranium particle fuel, in graphite core blocks, cooled by heat pipes coupled to an ORC converter insures long-term, virtually maintenance free, operation of this reactor for remote applications. 10 refs., 7 figs., 3 tabs.

  9. Baseline risk assessment of ground water contamination at the Uranium Mill Tailings Site near Green River, Utah

    SciTech Connect (OSTI)

    Not Available

    1994-09-01T23:59:59.000Z

    This document evaluates potential impacts to public health and the environment resulting from ground water contamination at the former uranium mill processing site. The tailings and other contaminated material at this site were placed in a disposal cell on the site in 1989 by the US DOE`s Uranium Mill Tailings Remedial Action (UMTRA) Project. Currently, UMTRA Project is evaluating ground water contamination in this risk assessment.

  10. Starfire - a commercial tokamak reactor

    SciTech Connect (OSTI)

    Baker, C.C.; Abdou, M.A.; DeFreece, D.A.; Trachsel, C.A.; Kokoszenski, J.; Graumann, D.

    1981-01-01T23:59:59.000Z

    The basic objective of the STARFIRE Project is to develop a design concept for a commercial tokamak fusion electric power plant based on the deuterium/tritium/lithium fuel cycle. The key technical objective is to develop the best embodiment of the tokamak as a power reactor consistent with credible engineering solutions to design problems. 10 refs.

  11. Uranium-233 purification and conversion to stabilized ceramic grade urania for LWBR fuel fabrication (LWBR Development Program)

    SciTech Connect (OSTI)

    Lloyd, R.

    1980-10-01T23:59:59.000Z

    High purity ceramic grade urania (/sup 233/UO/sub 2/) used in manufacturing the fuel for the Light Water Breeder Reactor (LWBR) core was made from uranium-233 that was obtained by irradiating thoria under special conditions to result in not more than 10 ppM of uranium-232 in the recovered uranium-233 product. A developmental study established the operating parameters of the conversion process for transforming the uranium-233 into urania powder with the appropriate chemical and physical attributes for use in fabricating the LWBR core fuel. This developmental study included the following: (a) design of an ion exchange purification process for removing the gamma-emitting alpha-decay daughters of uranium-232, to reduce the gamma-radiation field of the uranium-233 during LWBR fuel manufacture; (b) definition of the parameters for precipitating the uranium-233 as ammonium uranate (ADU) and for reducing the ADU with hydrogen to yield a urania conversion product of the proper particle size, surface area and sinterability for use in manufacturing the LWBR fuel; (c) establishment of parameters and design of equipment for stabilizing the urania conversion product to prevent it from undergoing excessive oxidation on exposure to the air during LWBR fuel manufacturing operations; and (d) development of a procedure and a facility to reprocess the unirradiated thoria-urania fuel scrap from the LWBR core manufacturing operations to recover the uranium-233 and convert it into high purity ceramic grade urania for LWBR core fabrication.

  12. NUCLEAR ISOTOPIC DILUTION OF HIGHLY ENRICHED URANIUM BY DRY BLENDING VIA THE RM-2 MILL TECHNOLOGY

    SciTech Connect (OSTI)

    Raj K. Rajamani; Sanjeeva Latchireddi; Vikas Devrani; Harappan Sethi; Roger Henry; Nate Chipman

    2003-08-01T23:59:59.000Z

    DOE has initiated numerous activities to focus on identifying material management strategies to disposition various excess fissile materials. In particular the INEEL has stored 1,700 Kg of offspec HEU at INTEC in CPP-651 vault facility. Currently, the proposed strategies for dispositioning are (a) aqueous dissolution and down blending to LEU via facilities at SRS followed by shipment of the liquid LEU to NFS for fabrication into LWR fuel for the TVA reactors and (b) dilution of the HEU to 0.9% for discard as a waste stream that would no longer have a criticality or proliferation risk without being processed through some type of enrichment system. Dispositioning this inventory as a waste stream via aqueous processing at SRS has been determined to be too costly. Thus, dry blending is the only proposed disposal process for the uranium oxide materials in the CPP-651 vault. Isotopic dilution of HEU to typically less than 20% by dry blending is the key to solving the dispositioning issue (i.e., proliferation) posed by HEU stored at INEEL. RM-2 mill is a technology developed and successfully tested for producing ultra-fine particles by dry grinding. Grinding action in RM-2 mill produces a two million-fold increase in the number of particles being blended in a centrifugal field. In a previous study, the concept of achieving complete and adequate blending and mixing (i.e., no methods were identified to easily separate and concentrate one titanium compound from the other) in remarkably short processing times was successfully tested with surrogate materials (titanium dioxide and titanium mono-oxide) with different particle sizes, hardness and densities. In the current project, the RM-2 milling technology was thoroughly tested with mixtures of natural uranium oxide (NU) and depleted uranium oxide (DU) stock to prove its performance. The effects of mill operating and design variables on the blending of NU/DU oxides were evaluated. First, NU and DU both made of the same oxide, UO{sub 3}, was used in the testing. Next, NU made up of UO{sub 3} and DU made up of UO{sub 2} was used in the test work. In every test, the blend achieved was characterized by spatial sampling of the ground product and analyzing for {sup 235}U concentration. The test work proved that these uranium oxide materials can be blended successfully. The spatial concentration was found to be uniform. Next, sintered thorium oxide pellets were used as surrogate for light water breeder reactor pellets (LWBR). To simulate LWBR pellet dispositioning, the thorium oxide pellets were first ground to a powder form and then the powder was blended with NU. In these tests also the concentration of {sup 235}U and {sup 232}Th in blended products fell within established limits proving the success of RM-2 milling technology. RM-2 milling technology is applicable to any dry radioactive waste, especially brittle solids that can be ground up and mixed with the non-radioactive stock.

  13. Uranium Pyrophoricity Phenomena and Prediction (FAI/00-39)

    SciTech Connect (OSTI)

    PLYS, M.G.

    2000-10-10T23:59:59.000Z

    The purpose of this report is to provide a topical reference on the phenomena and prediction of uranium pyrophoricity for the Hanford Spent Nuclear Fuel (SNF) Project with specific applications to SNF Project processes and situations. Spent metallic uranium nuclear fuel is currently stored underwater at the K basins in the Hanford 100 area, and planned processing steps include: (1) At the basins, cleaning and placing fuel elements and scrap into stainless steel multi-canister overpacks (MCOs) holding about 6 MT of fuel apiece; (2) At nearby cold vacuum drying (CVD) stations, draining, vacuum drying, and mechanically sealing the MCOs; (3) Shipping the MCOs to the Canister Storage Building (CSB) on the 200 Area plateau; and (4) Welding shut and placing the MCOs for interim (40 year) dry storage in closed CSB storage tubes cooled by natural air circulation through the surrounding vault. Damaged fuel elements have exposed and corroded fuel surfaces, which can exothermically react with water vapor and oxygen during normal process steps and in off-normal situations, A key process safety concern is the rate of reaction of damaged fuel and the potential for self-sustaining or runaway reactions, also known as uranium fires or fuel ignition. Uranium metal and one of its corrosion products, uranium hydride, are potentially pyrophoric materials. Dangers of pyrophoricity of uranium and its hydride have long been known in the U.S. Department of Energy (Atomic Energy Commission/DOE) complex and will be discussed more below; it is sufficient here to note that there are numerous documented instances of uranium fires during normal operations. The motivation for this work is to place the safety of the present process in proper perspective given past operational experience. Steps in development of such a perspective are: (1) Description of underlying physical causes for runaway reactions, (2) Modeling physical processes to explain runaway reactions, (3) Validation of the method against experimental data, (4) Application of the method to plausibly explain operational experience, and (5) Application of the method to present process steps to demonstrate process safety and margin. Essentially, the logic above is used to demonstrate that runaway reactions cannot occur during normal SNF Project process steps, and to illustrate the depth of the technical basis for such a conclusion. Some off-normal conditions are identified here that could potentially lead to runaway reactions. However, this document is not intended to provide an exhaustive analysis of such cases. In summary, this report provides a ''toolkit'' of models and approaches for analysis of pyrophoricity safety issues at Hanford, and the technical basis for the recommended approaches. A summary of recommended methods appears in Section 9.0.

  14. Fabrication and Characterization of Uranium-based High Temperature Reactor

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmospheric Optical Depth7-1D: Vegetation ProposedUsing ZirconiaPolicy andExsolutionFES6FY 2011 OIG(SC) 2FY98 ToFuel | ornl.gov

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

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-SeriesFlickr Flickr Editor'sshortGeothermal Heat PumpsChad Simmons Gerdenis Kodis

  16. Laser induced phosphorescence uranium analysis

    DOE Patents [OSTI]

    Bushaw, Bruce A. (Kennewick, WA)

    1986-01-01T23:59:59.000Z

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

  17. Laser induced phosphorescence uranium analysis

    DOE Patents [OSTI]

    Bushaw, B.A.

    1983-06-10T23:59:59.000Z

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

  18. Neutronic reactor

    DOE Patents [OSTI]

    Wende, Charles W. J. (Augusta, GA); Babcock, Dale F. (Wilmington, DE); Menegus, Robert L. (Wilmington, DE)

    1983-01-01T23:59:59.000Z

    A nuclear reactor includes an active portion with fissionable fuel and neutron moderating material surrounded by neutron reflecting material. A control element in the active portion includes a group of movable rods constructed of neutron-absorbing material. Each rod is movable with respect to the other rods to vary the absorption of neutrons and effect control over neutron flux.

  19. Implications of Fast Reactor Transuranic Conversion Ratio

    SciTech Connect (OSTI)

    Steven J. Piet; Edward A. Hoffman; Samuel E. Bays

    2010-11-01T23:59:59.000Z

    Theoretically, the transuranic conversion ratio (CR), i.e. the transuranic production divided by transuranic destruction, in a fast reactor can range from near zero to about 1.9, which is the average neutron yield from Pu239 minus 1. In practice, the possible range will be somewhat less. We have studied the implications of transuranic conversion ratio of 0.0 to 1.7 using the fresh and discharge fuel compositions calculated elsewhere. The corresponding fissile breeding ratio ranges from 0.2 to 1.6. The cases below CR=1 (burners) do not have blankets; the cases above CR=1 (breeders) have breeding blankets. The burnup was allowed to float while holding the maximum fluence to the cladding constant. We graph the fuel burnup and composition change. As a function of transuranic conversion ratio, we calculate and graph the heat, gamma, and neutron emission of fresh fuel; whether the material is attractive for direct weapon use using published criteria; the uranium utilization and rate of consumption of natural uranium; and the long-term radiotoxicity after fuel discharge. For context, other cases and analyses are included, primarily once-through light water reactor (LWR) uranium oxide fuel at 51 MWth-day/kg-iHM burnup (UOX-51). For CR<1, the heat, gamma, and neutron emission increase as material is recycled. The uranium utilization is at or below 1%, just as it is in thermal reactors as both types of reactors require continuing fissile support. For CR>1, heat, gamma, and neutron emission decrease with recycling. The uranium utilization exceeds 1%, especially as all the transuranic elements are recycled. exceeds 1%, especially as all the transuranic elements are recycled. At the system equilibrium, heat and gamma vary by somewhat over an order of magnitude as a function of CR. Isotopes that dominate heat and gamma emission are scattered throughout the actinide chain, so the modest impact of CR is unsurprising. Neutron emitters are preferentially found among the higher actinides, so the neutron emission varies much stronger with CR, about three orders of magnitude.

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

    SciTech Connect (OSTI)

    Dyer, R.H. [USDOE Oak Ridge Operations Office, TN (United States); Kovac, F.M. [Oak Ridge National Lab., TN (United States); Pryor, W.A. [PAI Corp., Oak Ridge, TN (United States)

    1993-10-01T23:59:59.000Z

    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. Conversion feasibility studies for the Grenoble high flux reactor

    SciTech Connect (OSTI)

    Mo, S.C.; Matos, J.E.

    1989-01-01T23:59:59.000Z

    Feasibility studies for conversion of the High Flux Reactor (RHF) at Grenoble France have been performed at the Argonne National Laboratory in cooperation with the Institut Laue-Langevin (ILL). The uranium densities required for conversion of the RHF to reduced enrichment fuels were computed to be 7.9 g/cm{sup 3} with 20% enrichment, 4.8 g/cm{sup 3} with 29% enrichment, and 2.8 g/cm{sup 3} with 45% enrichment. Thermal flux reductions at the peak in the heavy water reflector were computed to be 3% with 45% enriched fuel and 7% with 20% enriched fuel. In each case, the reactor's 44 day cycle length was preserved and no changes were made in the fuel element geometry. If the cladding thickness could be reduced from 0.38 mm to 0.30 mm, the required uranium density with 20% enrichment would be about 6.0 g/cm{sup 3} and the thermal flux reduction at the peak in the heavy water reflector would be about 7%. Significantly higher uranium densities are required in the RHF than in heavy water reactors with more conventional designs because the neutron spectrum is much harder in the RHF. Reduced enrichment fuels with the uranium densities required for use in the RHF are either not available or are not licensable at the present time. 6 refs., 6 figs., 3 tabs.

  2. Guidelines for preparing and reviewing applications for the licensing of non-power reactors: Standard review plan and acceptance criteria. NUREG - 1537, Part 2

    SciTech Connect (OSTI)

    NONE

    1996-02-01T23:59:59.000Z

    NUREG - 1537, Part 2 gives guidance on the conduct of licensing action reviews to NRC staff who review non-power reactor licensing applications. These licensing actions include construction permits and initial operating licenses, license renewals, amendments, conversions from highly enriched uranium to low-enriched uranium, decommissioning, and license termination.

  3. Guidelines for preparing and reviewing applications for the licensing of non-power reactors: Format and Content. NUREG-1537, Part 1

    SciTech Connect (OSTI)

    NONE

    1996-02-01T23:59:59.000Z

    NUREG - 1537, Part 1 gives guidance to non-power reactor licensees and applicants on the format and content of applications to the Nuclear Regulatory Commission for licensing actions. These licensing actions include construction permits and initial operating licenses, license renewals, amendments, conversions from highly enriched uranium to low-enriched uranium, decommissioning, and license termination.

  4. Standard Test Method for Determination of Uranium, Oxygen to Uranium (O/U), and Oxygen to Metal (O/M) in Sintered Uranium Dioxide and Gadolinia-Uranium Dioxide Pellets by Atmospheric Equilibration

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2007-01-01T23:59:59.000Z

    Standard Test Method for Determination of Uranium, Oxygen to Uranium (O/U), and Oxygen to Metal (O/M) in Sintered Uranium Dioxide and Gadolinia-Uranium Dioxide Pellets by Atmospheric Equilibration

  5. Determining Reactor Flux from Xenon-136 and Cesium-135 in Spent Fuel

    E-Print Network [OSTI]

    A. C. Hayes; Gerard Jungman

    2012-05-30T23:59:59.000Z

    The ability to infer the reactor flux from spent fuel or seized fissile material would enhance the tools of nuclear forensics and nuclear nonproliferation significantly. We show that reactor flux can be inferred from the ratios of xenon-136 to xenon-134 and cesium-135 to cesium-137. If the average flux of a reactor is known, the flux inferred from measurements of spent fuel could help determine whether that spent fuel was loaded as a blanket or close to the mid-plane of the reactor. The cesium ratio also provides information on reactor shutdowns during the irradiation of fuel, which could prove valuable for identifying the reactor in question through comparisons with satellite reactor heat monitoring data. We derive analytic expressions for these correlations and compare them to experimental data and to detailed reactor burn simulations. The enrichment of the original uranium fuel affects the correlations by up to 3 percent, but only at high flux.

  6. BDDR, a new CEA technological and operating reactor database

    SciTech Connect (OSTI)

    Soldevilla, M.; Salmons, S.; Espinosa, B. [CEA-Saclay, CEA/DEN/DANS/DM2S/SERMA, 91191 Gif-sur-Yvette (France); Clanet, M.; Boudin, X. [CEA-Bruyeres-le-Chatel, 91297 Arpajon (France)

    2013-07-01T23:59:59.000Z

    The new application BDDR (Reactor database) has been developed at CEA in order to manage nuclear reactors technological and operating data. This application is a knowledge management tool which meets several internal needs: -) to facilitate scenario studies for any set of reactors, e.g. non-proliferation assessments; -) to make core physics studies easier, whatever the reactor design (PWR-Pressurized Water Reactor-, BWR-Boiling Water Reactor-, MAGNOX- Magnesium Oxide reactor-, CANDU - CANada Deuterium Uranium-, FBR - Fast Breeder Reactor -, etc.); -) to preserve the technological data of all reactors (past and present, power generating or experimental, naval propulsion,...) in a unique repository. Within the application database are enclosed location data and operating history data as well as a tree-like structure containing numerous technological data. These data address all kinds of reactors features and components. A few neutronics data are also included (neutrons fluxes). The BDDR application is based on open-source technologies and thin client/server architecture. The software architecture has been made flexible enough to allow for any change. (authors)

  7. Decommissioning the UHTREX Reactor Facility at Los Alamos, New Mexico

    SciTech Connect (OSTI)

    Salazar, M.; Elder, J.

    1992-08-01T23:59:59.000Z

    The Ultra-High Temperature Reactor Experiment (UHTREX) facility was constructed in the late 1960s to advance high-temperature and gas-cooled reactor technology. The 3-MW reactor was graphite moderated and helium cooled and used 93% enriched uranium as its fuel. The reactor was run for approximately one year and was shut down in February 1970. The decommissioning of the facility involved removing the reactor and its associated components. This document details planning for the decommissioning operations which included characterizing the facility, estimating the costs of decommissioning, preparing environmental documentation, establishing a system to track costs and work progress, and preplanning to correct health and safety concerns in the facility. Work to decommission the facility began in 1988 and was completed in September 1990 at a cost of $2.9 million. The facility was released to Department of Energy for other uses in its Los Alamos program.

  8. Weld monitor and failure detector for nuclear reactor system

    DOE Patents [OSTI]

    Sutton, Jr., Harry G. (Mt. Lebanon, PA)

    1987-01-01T23:59:59.000Z

    Critical but inaccessible welds in a nuclear reactor system are monitored throughout the life of the reactor by providing small aperture means projecting completely through the reactor vessel wall and also through the weld or welds to be monitored. The aperture means is normally sealed from the atmosphere within the reactor. Any incipient failure or cracking of the weld will cause the environment contained within the reactor to pass into the aperture means and thence to the outer surface of the reactor vessel where its presence is readily detected.

  9. Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)

    SciTech Connect (OSTI)

    J. K. Wright; R. N. Wright

    2010-07-01T23:59:59.000Z

    The U.S. Department of Energy (DOE) has selected the High-Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production, with an outlet gas temperature in the range of 750C, and a design service life of 60 years. The reactor design will be a graphite-moderated, helium-cooled, prismatic, or pebble bed reactor and use low-enriched uranium, Tri-Isotopic (TRISO)-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. This technology development plan details the additional research and development (R&D) required to design and license the NGNP RPV, assuming that A 508/A 533 is the material of construction. The majority of additional information that is required is related to long-term aging behavior at NGNP vessel temperatures, which are somewhat above those commonly encountered in the existing database from LWR experience. Additional data are also required for the anticipated NGNP environment. An assessment of required R&D for a Grade 91 vessel has been retained from the first revision of the R&D plan in Appendix B in somewhat less detail. Considerably more development is required for this steel compared to A 508/A 533 including additional irradiation testing for expected NGNP operating temperatures, high-temperature mechanical properties, and extensive studies of long-term microstructural stability.

  10. Evolutionary/advanced light water reactor data report

    SciTech Connect (OSTI)

    NONE

    1996-02-09T23:59:59.000Z

    The US DOE Office of Fissile Material Disposition is examining options for placing fissile materials that were produced for fabrication of weapons, and now are deemed to be surplus, into a condition that is substantially irreversible and makes its use in weapons inherently more difficult. The principal fissile materials subject to this disposition activity are plutonium and uranium containing substantial fractions of plutonium-239 uranium-235. The data in this report, prepared as technical input to the fissile material disposition Programmatic Environmental Impact Statement (PEIS) deal only with the disposition of plutonium that contains well over 80% plutonium-239. In fact, the data were developed on the basis of weapon-grade plutonium which contains, typically, 93.6% plutonium-239 and 5.9% plutonium-240 as the principal isotopes. One of the options for disposition of weapon-grade plutonium being considered is the power reactor alternative. Plutonium would be fabricated into mixed oxide (MOX) fuel and fissioned (``burned``) in a reactor to produce electric power. The MOX fuel will contain dioxides of uranium and plutonium with less than 7% weapon-grade plutonium and uranium that has about 0.2% uranium-235. The disposition mission could, for example, be carried out in existing power reactors, of which there are over 100 in the United States. Alternatively, new LWRs could be constructed especially for disposition of plutonium. These would be of the latest US design(s) incorporating numerous design simplifications and safety enhancements. These ``evolutionary`` or ``advanced`` designs would offer not only technological advances, but also flexibility in siting and the option of either government or private (e.g., utility) ownership. The new reactor designs can accommodate somewhat higher plutonium throughputs. This data report deals solely with the ``evolutionary`` LWR alternative.

  11. Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications

    E-Print Network [OSTI]

    Helmreich, Grant

    2012-02-14T23:59:59.000Z

    The sintering behavior of uranium and uranium-zirconium alloys in the alpha phase were characterized in this research. Metal uranium powder was produced from pieces of depleted uranium metal acquired from the Y-12 plant via hydriding...

  12. Baseline risk assessment of ground water contamination at the uranium mill tailings site Salt Lake City, Utah

    SciTech Connect (OSTI)

    Not Available

    1994-09-01T23:59:59.000Z

    This baseline risk assessment of groundwater contamination at the uranium mill tailings site near Salt Lake City, Utah, evaluates potential public health or environmental impacts resulting from ground water contamination at the former uranium ore processing site. The tailings and other contaminated material at this site were placed in a disposal cell located at Clive, Utah, in 1987 by the US Department of Energy`s Uranium Mill Tailings Remedial Action (UMTRA) Project. The second phase of the UMTRA Project is to evaluate residual ground water contamination at the former uranium processing site, known as the Vitro processing site. This risk assessment is the first site-specific document under the Ground Water Project. It will help determine the appropriate remedial action for contaminated ground water at the site.

  13. Lessons learned from decommissioning projects at Los Alamos National Laboratory

    SciTech Connect (OSTI)

    Salazar, M.

    1995-09-01T23:59:59.000Z

    This paper describes lessons learned over the last 20 years from 12 decommissioning projects at Los Alamos National Laboratory. These lessons relate both to overall program management and to management of specific projects during the planning and operations phases. The issues include waste management; the National Environmental Policy Act (NEPA); the Resource Conservation and Recovery Act (RCRA); the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA); contracting; public involvement; client/customer interface; and funding. Key elements of our approach are to be proactive; follow the observation method; perform field activities concurrently; develop strategies to keep reportable incidents from delaying work; seek and use programs, methods, etc., in existence to shorten learning curves; network to help develop solutions; and avoid overstudying and overcharacterizing. This approach results in preliminary plans that require very little revision before implementation, reasonable costs and schedules, early acquisition of permits and NEPA documents, preliminary characterization reports, and contracting documents. Our track record is good -- the last four projects (uranium and plutonium-processing facility and three research reactors) have been on budget and on schedule.

  14. UMTRA Project document control system manual

    SciTech Connect (OSTI)

    Not Available

    1994-09-01T23:59:59.000Z

    This manual defines the Project Document Control System (PDCS) operated by the US DOE Uranium Mill tailings Remedial Action (UMTRA) project Office. The purpose of the PDCS is to provide an active and continuing program for acquiring, controlling, retaining, retrieving, retiring and disposing of all UMTRA Project documents. The PDCS also provides guidance and coordination in transferring documents by various UMTRA Projection document control centers to a central location.

  15. Updating reactor control: mini-computers

    SciTech Connect (OSTI)

    Crawford, K.C.; Sandquist, G.M. [University of Utah, Salt Lake City, UT (United States)

    1984-07-01T23:59:59.000Z

    An aging reactor control console and a limited operating budget have impeded many research projects in the TRIGA reactor facility at the University of Utah. The, University's present console is Circa 1959 vintage and repairs to the console are frequently required which present many electronic problems to a staff with little electronic training. As an alternative to a single function control console we are developing a TRIGA control system based upon a mini-computer. The system hardware has been specified and the hardware is currently being acquired. The software will be programmed by the staff to customize the system to the reactor's physical systems and technical specifications. The software will be designed to monitor and control all reactor functions, control a pneumatic sample transfer system, acquire and analyze neutron activation data, provide reactor facility security surveillance, provide reactor documentation including online logging of physical parameters, and record regularly scheduled reactor calibrations and laboratory accounting procedures. The problem of hardware rewiring and changing technical specifications and changing safety system characteristics can be easily handled in the software. Our TRIGA reactor also functions as a major educational resource using available reactor based software. The computer control system can be employed to provide on-line training in reactor physics and kinetics. (author)

  16. Updated Conceptual Model for the 300 Area Uranium Groundwater Plume

    SciTech Connect (OSTI)

    Zachara, John M.; Freshley, Mark D.; Last, George V.; Peterson, Robert E.; Bjornstad, Bruce N.

    2012-11-01T23:59:59.000Z

    The 300 Area uranium groundwater plume in the 300-FF-5 Operable Unit is residual from past discharge of nuclear fuel fabrication wastes to a number of liquid (and solid) disposal sites. The source zones in the disposal sites were remediated by excavation and backfilled to grade, but sorbed uranium remains in deeper, unexcavated vadose zone sediments. In spite of source term removal, the groundwater plume has shown remarkable persistence, with concentrations exceeding the drinking water standard over an area of approximately 1 km2. The plume resides within a coupled vadose zone, groundwater, river zone system of immense complexity and scale. Interactions between geologic structure, the hydrologic system driven by the Columbia River, groundwater-river exchange points, and the geochemistry of uranium contribute to persistence of the plume. The U.S. Department of Energy (DOE) recently completed a Remedial Investigation/Feasibility Study (RI/FS) to document characterization of the 300 Area uranium plume and plan for beginning to implement proposed remedial actions. As part of the RI/FS document, a conceptual model was developed that integrates knowledge of the hydrogeologic and geochemical properties of the 300 Area and controlling processes to yield an understanding of how the system behaves and the variables that control it. Recent results from the Hanford Integrated Field Research Challenge site and the Subsurface Biogeochemistry Scientific Focus Area Project funded by the DOE Office of Science were used to update the conceptual model and provide an assessment of key factors controlling plume persistence.

  17. Evidence of uranium biomineralization in sandstone-hosted roll-front uranium deposits, northwestern China

    E-Print Network [OSTI]

    Fayek, Mostafa

    Evidence of uranium biomineralization in sandstone-hosted roll-front uranium deposits, northwestern Available online 25 January 2005 Abstract We show evidence that the primary uranium minerals, uraninite-front uranium deposits, Xinjiang, northwestern China were biogenically precipitated and psuedomorphically

  18. The Ames Project (1942-1946)

    ScienceCinema (OSTI)

    None

    2013-06-05T23:59:59.000Z

    The Ames Laboratory was officially founded on May 17, 1947, following development of a process to purify uranium metal for the historic Manhattan Project. From 1942 to 1946, Ames Lab scientists produced over two-million pounds of uranium metal. A U.S. Department of Energy national research laboratory, the Ames Laboratory creates materials and energy solutions. Iowa State University operates Ames Laboratory under contract with the DOE.

  19. Inherently safe in situ uranium recovery

    DOE Patents [OSTI]

    Krumhansl, James L; Brady, Patrick V

    2014-04-29T23:59:59.000Z

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

  20. Uranium Acquisition | Y-12 National Security Complex

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

    of Interest (EOI) to acquire up to 6,800 metric tons of Uranium (MTU) of high purity depleted uranium metal (DU) and related material and services. This request for EOI does...

  1. The End of Cheap Uranium

    E-Print Network [OSTI]

    Dittmar, Michael

    2011-01-01T23:59:59.000Z

    Historic data from many countries demonstrate that on average no more than 50-70% of the uranium in a deposit could be mined. An analysis of more recent data from Canada and Australia leads to a mining model with an average deposit extraction lifetime of 10+- 2 years. This simple model provides an accurate description of the extractable amount of uranium for the recent mining operations. Using this model for all larger existing and planned uranium mines up to 2030, a global uranium mining peak of at most 58 +- 4 ktons around the year 2015 is obtained. Thereafter we predict that uranium mine production will decline to at most 54 +- 5 ktons by 2025 and, with the decline steepening, to at most 41 +- 5 ktons around 2030. This amount will not be sufficient to fuel the existing and planned nuclear power plants during the next 10-20 years. In fact, we find that it will be difficult to avoid supply shortages even under a slow 1%/year worldwide nuclear energy phase-out scenario up to 2025. We thus suggest that a world...

  2. Uranium Mill Tailings Remedial Action fiscal year 1992 roadmap

    SciTech Connect (OSTI)

    Not Available

    1993-02-01T23:59:59.000Z

    The Uranium Mill Tailings Remedial Action (UMTRA) Project is funded and managed as two separate projects: Surface remediation (UMTRA-S) and Groundwater compliance (UMTRA-G). Surface remediation is a Major System Acquisition and has been completed at 10 sites, 7 sites are under construction, and 7 sites are in the planning stage. The planning stages of the UMTRA-G Project, a major project, began in April 1991. A programmatic environmental impact statement (PEIS) has been started. Site characterization work and baseline risk assessment will begin FY 1993. Thus, the UMTRA-S Project is a mature and ongoing program with the roles of various organizations well defined, while the UMTRA-G Project is still being formulated and the interfaces between the DOE, states and tribes, and the EPA are being established. The Office of Environmental Restoration and Waste Management (EM) directed that all projects under its authority develop roadmaps for their activities. The UMTRA Project Roadmap was developed by the UMTRA Project Office with input from the TAC, RAC, the GJPO, and assistance from SAIC. A single roadmap has been prepared for both the UMTRA-S and UMTRA-G Projects. This was deemed appropriate due to the close relationship between the projects and to the fact that the same Government and contractor personnel are preparing the roadmaps. Roadmap development is a planning process that focuses on issue identification, root-cause analysis, and issues resolution. The methodology is divided into three phases: assessment, analysis, and issues resolution.

  3. Environmental Management Brookhaven Graphite Research Reactor

    E-Print Network [OSTI]

    Homes, Christopher C.

    -out report · Transition to long-term surveillance and maintenance · Office of Environmental ManagementEnvironmental Management Brookhaven Graphite Research Reactor (BGRR) Project Completion John Sattler Federal Project Director Office of Environmental Management U.S. Department of Energy BNL

  4. Role of small lead-cooled fast reactors for international deployment in worldwide sustainable nuclear energy supply.

    SciTech Connect (OSTI)

    Sienicki, J. J.; Wade, D. C.; Moisseytsev, A.; Nuclear Engineering Division

    2008-01-01T23:59:59.000Z

    Most recently, the global nuclear energy partnership (GNEP) has identified, as one of its key objectives, the development and demonstration of concepts for small and medium-sized reactors (SMRs) that can be globally deployed while assuring a high level of proliferation resistance. Lead-cooled systems offer several key advantages in meeting these goals. The small lead-cooled fast reactor concept known as the small secure transportable autonomous reactor (SSTAR) has been under ongoing development as part of the US advanced nuclear energy systems programs. Meeting future worldwide projected energy demands during this century (e.g., 1000 to 2000 GWe by 2050) in a sustainable manner while maintaining CO2 emissions at or below today's level will require massive deployments of nuclear reactors in non-fuel cycle states as well as fuel cycle states. The projected energy demands of non-fuel cycle states will not be met solely through the deployment of Light Water Reactors (LWRs) in those states without using up the world's resources of fissile material (e.g., known plus speculative virgin uranium resources = 15 million tonnes). The present U.S. policy is focused upon domestic deployment of large-scale LWRs and sodium-cooled fast spectrum Advanced Burner Reactors (ABRs) working in a symbiotic relationship that burns existing fissile material while destroying the actinides which are generated. Other major nuclear nations are carrying out the development and deployment of SFR breeders as witness the planning for SFR breeder deployments in France, Japan, China, India, and Russia. Small (less that 300 MWe) and medium (300 to 700 MWe) size reactors are better suited to the growing economies and infrastructures of many non-fuel cycle states and developing nations. For those deployments, fast reactor converters which are fissile self-sufficient by creating as much fissile material as they consume are preferred to breeders that create more fissile material than they consume. Thus, there is a need for small and medium size fast reactors in non-fuel cycle states operating in a converter mode as well as large sodium-cooled fast breeders in fuel cycle states. Desired attributes for exportable small fast reactors include: proliferation resistance features such as restricted access to fuel; long core life further restricting access by reducing or eliminating the need for refueling; restricted potential to be misused in a breeding mode; fuel form that is unattractive in the safeguards sense; and a conversion ratio of unity to self-generate as much fissile material as is consumed. Desired attributes for exportable small reactor deployments in developing nations and remote sites also include: a small power level to match the smaller demand of towns or sites that are off-grid or on immature local grids; low enough cost to be economically competitive with alternative energy sources available to developing nation customers (e.g. diesel generators in remote locations); readily transported and assembled from transportable modules; simple to operate and highly reliable reducing plant operating staff requirements; as well as high reliability and passive safety reducing the number of accident initiators and need for safety systems as well as reducing the size of the exclusion and emergency planning zones. The Lead-Cooled Fast Reactor (LFR) has the desired attributes. An example of a small exportable LFR concept is the 20 MWe (45 MWt) Small Secure Transportable Autonomous Reactor (SSTAR) incorporating proliferation resistance, fissile selfsufficiency, autonomous load following, a high degree of passive safety, and supercritical carbon dioxide Brayton cycle energy conversion for high plant efficiency and improved economic competitiveness.

  5. Radiation measurements of uranium ingots from the electrometallurgical treatment of spent fuel.

    SciTech Connect (OSTI)

    Westphal, B. R.; Liaw, J. R.; Krsul, J. R.; Maddison, D. W.; Jensen, B. A.

    2003-03-24T23:59:59.000Z

    Radiation measurements and gamma spectroscopy analyses were made on numerous uranium ingots produced during the treatment of Experimental Breeder Reactor-II (EBR-II) spent nuclear fuel. The objective of these measurements was to provide background data for shielding concerns and potential process optimization. The uranium ingots resulted from the processing of both driver and blanket fuel by the electrometallurgical treatment process. The observed variation in the measurements was traced to the levels of certain fission product residues that remained in the uranium ingots produced during spent fuel treatment. A minor process change to hold the material at an elevated temperature for a specified length of time was found to significantly reduce concentrations of high-activity fission products and, thus the radiation field.

  6. Gamma-spectrometric determination of 232U in uranium-bearing materials

    E-Print Network [OSTI]

    Jozsef Zsigrai; Cong Tam Nguyen; Andriy Berlizov

    2015-01-19T23:59:59.000Z

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

  7. Gamma-spectrometric determination of 232U in uranium-bearing materials

    E-Print Network [OSTI]

    Zsigrai, Jozsef; Berlizov, Andriy

    2015-01-01T23:59:59.000Z

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

  8. Neutron-Rich Isotope Production Using a Uranium Carbide Carbon Nanotubes SPES Target Prototype

    SciTech Connect (OSTI)

    Corradetti, Stefano [ORNL; Biasetto, Lisa [INFN, Laboratori Nazionali di Legnaro, Italy; Manzolaro, Mattia [INFN, Laboratori Nazionali di Legnaro, Italy; Scarpa, Daniele [ORNL; Carturan, S. [INFN, Laboratori Nazionali di Legnaro, Italy; Andrighetto, Alberto [INFN, Laboratori Nazionali di Legnaro, Italy; Prete, Gianfranco [ORNL; Vasquez, Jose L [ORNL; Zanonato, P. [Dipartimento di Scienze Chimiche, Padova, Italy; Colombo, P. [Dipartimento di Ingegneria Meccanica, Padova, Italy; Jost, Carola [University of Tennessee, Knoxville (UTK); Stracener, Daniel W [ORNL

    2013-01-01T23:59:59.000Z

    The SPES (Selective Production of Exotic Species) project, under development at the Istituto Nazionale di Fisica Nucleare - Laboratori Nazionali di Legnaro (INFN-LNL), is a new-generation Isotope Separation On-Line (ISOL) facility for the production of radioactive ion beams by means of the proton-induced fission of uranium. In the framework of the research on the SPES target, seven uranium carbide discs, obtained by reacting uranium oxide with graphite and carbon nanotubes, were irradiated with protons at the Holifield Radioactive Ion Beam Facility (HRIBF) of Oak Ridge National Laboratory (ORNL). In the following, the yields of several fission products obtained during the experiment are presented and discussed. The experimental results are then compared to those obtained using a standard uranium carbide target. The reported data highlights the capability of the new type of SPES target to produce and release isotopes of interest for the nuclear physics community.

  9. High strength uranium-tungsten alloys

    DOE Patents [OSTI]

    Dunn, Paul S. (Santa Fe, NM); Sheinberg, Haskell (Los Alamos, NM); Hogan, Billy M. (Los Alamos, NM); Lewis, Homer D. (Bayfield, CO); Dickinson, James M. (Los Alamos, NM)

    1991-01-01T23:59:59.000Z

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

  10. High strength uranium-tungsten alloy process

    DOE Patents [OSTI]

    Dunn, Paul S. (Santa Fe, NM); Sheinberg, Haskell (Los Alamos, NM); Hogan, Billy M. (Los Alamos, NM); Lewis, Homer D. (Bayfield, CO); Dickinson, James M. (Los Alamos, NM)

    1990-01-01T23:59:59.000Z

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

  11. Clean Air Act Requirements: Uranium Mill Tailings

    E-Print Network [OSTI]

    EPA'S Clean Air Act Requirements: Uranium Mill Tailings Radon Emissions Rulemaking Reid J. Rosnick requirements for operating uranium mill tailings (Subpart W) Status update on Subpart W activities Outreach/Communications #12;3 EPA Regulatory Requirements for Operating Uranium Mill Tailings (Clean Air Act) · 40 CFR 61

  12. Remediation and Recovery of Uranium from Contaminated

    E-Print Network [OSTI]

    Lovley, Derek

    Remediation and Recovery of Uranium from Contaminated Subsurface Environments with Electrodes K E L that Geobacter species can effectively remove uranium from contaminated groundwater by reducing soluble U was stably precipitated until reoxidized in the presence of oxygen. When an electrode was placed in uranium

  13. Uranium Watch REGULATORY CONFUSION: FEDERALAND STATE

    E-Print Network [OSTI]

    Uranium Watch Report REGULATORY CONFUSION: FEDERALAND STATE ENFORCEMENT OF 40 C.F.R. PART 61 SUBPART W INTRODUCTION 1. This Uranium Watch Report, Regulatory Confusion: Federal and State Enforcement at the White Mesa Uranium Mill, San Juan County, Utah. 2. The DAQ, a Division of the Utah Department

  14. D Riso-R-429 Automated Uranium

    E-Print Network [OSTI]

    routinely used analytical techniques for uranium determina- tions in geological samples, fissionCM i D Riso-R-429 Automated Uranium Analysis by Delayed-Neutron Counting H. Kunzendorf, L. Løvborg AUTOMATED URANIUM ANALYSIS BY DELAYED-NEUTRON COUNTING H. Kunzendorf, L. Løvborg and E.M. Christiansen

  15. Y-12 Uranium Exposure Study

    SciTech Connect (OSTI)

    Eckerman, K.F.; Kerr, G.D.

    1999-08-05T23:59:59.000Z

    Following the recent restart of operations at the Y-12 Plant, the Radiological Control Organization (RCO) observed that the enriched uranium exposures appeared to involve insoluble rather than soluble uranium that presumably characterized most earlier Y-12 operations. These observations necessitated changes in the bioassay program, particularly the need for routine fecal sampling. In addition, it was not reasonable to interpret the bioassay data using metabolic parameter values established during earlier Y-12 operations. Thus, the recent urinary and fecal bioassay data were interpreted using the default guidance in Publication 54 of the International Commission on Radiological Protection (ICRP); that is, inhalation of Class Y uranium with an activity median aerodynamic diameter (AMAD) of 1 {micro}m. Faced with apparently new workplace conditions, these actions were appropriate and ensured a cautionary approach to worker protection. As additional bioassay data were accumulated, it became apparent that the data were not consistent with Publication 54. Therefore, this study was undertaken to examine the situation.

  16. Impact of conversion to mixed-oxide fuels on reactor structural components

    SciTech Connect (OSTI)

    Yahr, G.T.

    1997-04-01T23:59:59.000Z

    The use of mixed-oxide (MOX) fuel to replace conventional uranium fuel in commercial light-water power reactors will result in an increase in the neutron flux. The impact of the higher flux on the structural integrity of reactor structural components must be evaluated. This report briefly reviews the effects of radiation on the mechanical properties of metals. Aging degradation studies and reactor operating experience provide a basis for determining the areas where conversion to MOX fuels has the potential to impact the structural integrity of reactor components.

  17. Comparison of actinide production in traveling wave and pressurized water reactors

    SciTech Connect (OSTI)

    Osborne, A.G.; Smith, T.A.; Deinert, M.R. [Department of Mechanical Engineering, University of Texas at Austin, Austin, TX (United States)

    2013-07-01T23:59:59.000Z

    The geopolitical problems associated with civilian nuclear energy production arise in part from the accumulation of transuranics in spent nuclear fuel. A traveling wave reactor is a type of breed-burn reactor that could, if feasible, reduce the overall production of transuranics. In one possible configuration, a cylinder of natural or depleted uranium would be subjected to a fast neutron flux at one end. The neutrons would transmute the uranium, producing plutonium and higher actinides. Under the right conditions, the reactor could become critical, at which point a self-stabilizing fission wave would form and propagate down the length of the reactor cylinder. The neutrons from the fission wave would burn the fissile nuclides and transmute uranium ahead of the wave to produce additional fuel. Fission waves in uranium are driven largely by the production and fission of {sup 239}Pu. Simulations have shown that the fuel burnup can reach values greater than 400 MWd/kgIHM, before fission products poison the reaction. In this work we compare the production of plutonium and minor actinides produced in a fission wave to that of a UOX fueled light water reactor, both on an energy normalized basis. The nuclide concentrations in the spent traveling wave reactor fuel are computed using a one-group diffusion model and are verified using Monte Carlo simulations. In the case of the pressurized water reactor, a multi-group collision probability model is used to generate the nuclide quantities. We find that the traveling wave reactor produces about 0.187 g/MWd/kgIHM of transuranics compared to 0.413 g/MWd/kgIHM for a pressurized water reactor running fuel enriched to 4.95 % and burned to 50 MWd/kgIHM. (authors)

  18. Electron Backscatter Diffraction (EBSD) Characterization of Uranium and Uranium Alloys

    SciTech Connect (OSTI)

    McCabe, Rodney J. [Los Alamos National Laboratory; Kelly, Ann Marie [Los Alamos National Laboratory; Clarke, Amy J. [Los Alamos National Laboratory; Field, Robert D. [Los Alamos National Laboratory; Wenk, H. R. [University of California, Berkeley

    2012-07-25T23:59:59.000Z

    Electron backscatter diffraction (EBSD) was used to examine the microstructures of unalloyed uranium, U-6Nb, U-10Mo, and U-0.75Ti. For unalloyed uranium, we used EBSD to examine the effects of various processes on microstructures including casting, rolling and forming, recrystallization, welding, and quasi-static and shock deformation. For U-6Nb we used EBSD to examine the microstructural evolution during shape memory loading. EBSD was used to study chemical homogenization in U-10Mo, and for U-0.75Ti, we used EBSD to study the microstructure and texture evolution during thermal cycling and deformation. The studied uranium alloys have significant microstructural and chemical differences and each of these alloys presents unique preparation challenges. Each of the alloys is prepared by a sequence of mechanical grinding and polishing followed by electropolishing with subtle differences between the alloys. U-6Nb and U-0.75Ti both have martensitic microstructures and both require special care in order to avoid mechanical polishing artifacts. Unalloyed uranium has a tendency to rapidly oxidize when exposed to air and a two-step electropolish is employed, the first step to remove the damaged surface layer resulting from the mechanical preparation and the second step to passivate the surface. All of the alloying additions provide a level of surface passivation and different one and two step electropolishes are employed to create good EBSD surfaces. Because of its low symmetry crystal structure, uranium exhibits complex deformation behavior including operation of multiple deformation twinning modes. EBSD was used to observe and quantify twinning contributions to deformation and to examine the fracture behavior. Figure 1 shows a cross section of two mating fracture surfaces in cast uranium showing the propensity of deformation twinning and intergranular fracture largely between dissimilarly oriented grains. Deformation of U-6Nb in the shape memory regime occurs by the motion of twin boundaries formed during the martensitic transformation. Deformation actually results in a coarsening of the microstructure making EBSD more practical following a limited amount of strain. Figure 2 shows the microstructure resulting from 6% compression. Casting of U-10Mo results in considerable chemical segregation as is apparent in Figure 2a. The segregation subsists through rolling and heat treatment processes as shown in Figure 2b. EBSD was used to study the effects of homogenization time and temperature on chemical heterogeneity. It was found that times and temperatures that result in a chemically homogeneous microstructure also result in a significant increase in grain size. U-0.75Ti forms an acicular martinsite as shown in Figure 4. This microstructure prevails through cycling into the higher temperature solid uranium phases.

  19. Process for alloying uranium and niobium

    DOE Patents [OSTI]

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

    1991-01-01T23:59:59.000Z

    Alloys such as U-6Nb are prepared by forming a stacked sandwich array of uraniun sheets and niobium powder disposed in layers between the sheets, heating the array in a vacuum induction melting furnace to a temperature such as to melt the uranium, holding the resulting mixture at a temperature above the melting point of uranium until the niobium dissolves in the uranium, and casting the uranium-niobium solution. Compositional uniformity in the alloy product is enabled by use of the sandwich structure of uranium sheets and niobium powder.

  20. Hybrid adsorptive membrane reactor

    DOE Patents [OSTI]

    Tsotsis, Theodore T. (Huntington Beach, CA); Sahimi, Muhammad (Altadena, CA); Fayyaz-Najafi, Babak (Richmond, CA); Harale, Aadesh (Los Angeles, CA); Park, Byoung-Gi (Yeosu, KR); Liu, Paul K. T. (Lafayette Hill, PA)

    2011-03-01T23:59:59.000Z

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.