National Library of Energy BETA

Sample records for uranium enrichment plant

  1. Aseismic design criteria for uranium enrichment plants

    SciTech Connect (OSTI)

    Beavers, J.E.

    1980-01-01

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

  2. The uranium cylinder assay system for enrichment plant safeguards

    SciTech Connect (OSTI)

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

    2010-01-01

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

  3. Uranium enrichment

    SciTech Connect (OSTI)

    Not Available

    1991-04-01

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

  4. Uranium enrichment

    SciTech Connect (OSTI)

    Not Available

    1991-08-01

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

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

    SciTech Connect (OSTI)

    Swindle, D.W.

    1990-03-01

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

  6. Measurements of uranium holdup in an operating gaseous diffusion enrichment plant

    SciTech Connect (OSTI)

    Augustson, R.H.; Walton, R.B.; Harris, R.; Harbarger, W.; Hicks, J.; Timmons, G.; Shissler, D.; Tayloe, R.; Jones, S.; Fields, L.

    1983-01-01

    Holdup of nuclear material in process equipment is one of the major sources of uncertainty in materials balances, particularly for high-throughput facilities with large equipment and extensive piping, such as gaseous diffusion uranium-enrichment plants. Locating and measuring the holdup while the plant is operating is a challenging problem because of background from the process material and the neighboring equipment. This paper reports NDA measurements performed at the Goodyear Atomic Gaseous Diffusion Plant, Portsmouth, Ohio, on enrichment equipment at the higher enrichment and (>10% /sup 235/U isotopic abundance) of the cascade. Both neutron and gamma-ray measurements were made to locate anomalously large deposits in converters and compressors and, within the limitations of the techniques, to quantify the amount of the deposit.

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

    SciTech Connect (OSTI)

    Singhal, M.K.; Kincaid, J.H.; Hammond, C.R.; Stockdale, B.I.; Walls, J.C. [Oak Ridge National Lab., TN (United States). Technical Programs and Services; Brock, W.R.; Denton, D.R. [Lockheed Martin Energy Systems, Inc., Oak Ridge, TN (United States)

    1995-12-31

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

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

    SciTech Connect (OSTI)

    JOE,J.

    2007-07-08

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

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

    SciTech Connect (OSTI)

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

    2010-08-08

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

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

    SciTech Connect (OSTI)

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

    2014-01-10

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

  11. Highly Enriched Uranium Materials Facility

    National Nuclear Security Administration (NNSA)

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

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

    SciTech Connect (OSTI)

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

    2010-04-01

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

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

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

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

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

    SciTech Connect (OSTI)

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

    1980-02-05

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

  15. U. S. forms uranium enrichment corporation

    SciTech Connect (OSTI)

    Seltzer, R.

    1993-07-12

    After almost 40 years of operation, the federal government is withdrawing from the uranium enrichment business. On July 1, the Department of Energy turned over to a new government-owned entity--the US Enrichment Corp. (USEC)--both the DOE enrichment plants at Paducah, Ky., and Portsmouth, Ohio, and domestic and international marketing of enriched uranium from them. Pushed by the inability of DOE's enrichment operations to meet foreign competition, Congress established USEC under the National Energy Policy Act of 1992, envisioning the new corporation as the first step to full privatization. With gross revenues of $1.5 billion in fiscal 1992, USEC would rank 275th on the Fortune 500 list of top US companies. USEC will lease from DOE the Paducah and Portsmouth facilities, built in the early 1950s, which use the gaseous diffusion process for uranium enrichment. USEC's stock is held by the US Treasury, to which it will pay annual dividends. Martin Marietta Energy Systems, which has operated Paducah since 1984 and Portsmouth since 1986 for DOE, will continue to operate both plants for USEC. Closing one of the two facilities will be studied, especially in light of a 40% world surplus of capacity over demand. USEC also will consider other nuclear-fuel-related ventures. USEC will produce only low-enriched uranium, not weapons-grade material. Indeed, USEC will implement a contract now being completed under which the US will purchase weapons-grade uranium from dismantled Russian nuclear weapons and convert it into low-enriched uranium for power reactor fuel.

  16. THE THEORY OF URANIUM ENRICHMENT BY THE GAS CENTRIFUGE

    E-Print Network [OSTI]

    Olander, Donald R.

    2013-01-01

    Soubbaramayer, (1979) in "Uranium Enrichment", S. Villani,and Davies, E. (1973) "Uranium Enrichment by Gas Centrifuge"Nuclear Energy THE THEORY OF URANIUM ENRICHMENT BY THE GAS

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

    SciTech Connect (OSTI)

    NONE

    1996-08-01

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

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

    SciTech Connect (OSTI)

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

    2009-07-04

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

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

    SciTech Connect (OSTI)

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

    2008-01-01

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

  20. THE THEORY OF URANIUM ENRICHMENT BY THE GAS CENTRIFUGE

    E-Print Network [OSTI]

    Olander, Donald R.

    2013-01-01

    E. (1973) "Uranium Enrichment by Gas Centrifuge" Mills andTHEORY OF URANIUM ENRICHMENT BY THE GAS CENTRIFUGE Donald R.THEORY OF URANIUM ENRICHMENT BY THE GAS CENTRIFUGE by Donald

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

    Energy Savers [EERE]

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

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

    Office of Environmental Management (EM)

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

  3. Safety aspects of gas centrifuge enrichment plants

    SciTech Connect (OSTI)

    Hansen, A.H.

    1987-01-01

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

  4. Measurements of Low-Enriched Uranium Holdup.

    SciTech Connect (OSTI)

    Belian, A. P. (Anthony P.); Reilly, T. D. (T. Douglas); Russo, P. A. (Phyllis A.); Tobin, S. J. (Stephen J.)

    2005-01-01

    A recent effort determined uranium holdup at a large fuel fabrication facility abroad where low enriched ({approx} 3%) uranium (LEU) oxide feeds the pellet manufacturing process. Measurements taken with both high- and low-resolution gamma-ray spectrometry systems include extensive data for the ventilation and vacuum systems. Equipment dimensions and the corresponding holdup deposit masses are large for LEU. Because deposits are infinitely thick to the 186 keV gamma ray in many locations in an LEU environment, measurements of both the 186 and 1001 keV gamma-rays were required, and self-attenuation was significant at 1001 keV in many cases. These wide-dynamic-range measruements used short count times, portable scintillator detectors, and portable MCAs. Because equipment is elevated above floor levels, most measurements were made with detectors mounted on extended telescoping poles. One of the main goals of this effort was to demonstrate and validate methods for measurement and quantitative analysis of LEU holdup using low-resolution detectors and the Generalized Geometry Holdup (GGH) techniques. The current GGH approach is applied elsewhere for holdup measurements of plutonium and high-enriched uranium. The recent experience is directly applicable to holdup measruements at LEU facilities such as the Paducah and Portmouth gaseous diffusion enrichment plants and elsewhere, including LEU sites where D and D is active. This report discusses the measurement methodology, calibration of the measurement equipment, measurement control, analysis of the data, and the global and local assay results including random and systematic uncertainties. It includes field-validation exercises (multiple calibrated systems that perform measruements on the same extended equipment) as well as quantitative validation results obtained on reference materials assembled to emulate the deposits in an extended vacuum line that was also measured by these techniques. The paper examines the differences in assay results between the low-resolution system using the GGH method and the high-resolution system utilizing the commercially available ISOCS analysis method.

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

    SciTech Connect (OSTI)

    Kuzminski, Jozef [Los Alamos National Laboratory; Nesuhoff, J [NBL; Cratto, P [NBL; Pfennigwerth, G [Y12 NATIONAL SEC. COMPLEX; Mikhailenko, A [ULBA METALLURGICAL PLANT; Maliutina, I [ULBA METALLURGICAL PLANT; Nations, J [GREGG PROTECTION SERVICES

    2009-01-01

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

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

    National Nuclear Security Administration (NNSA)

    NNSA seeks to recover the economic value of the material by using the resulting LEU as nuclear reactor fuel. U.S.-Russian Highly Enriched Uranium Purchase Agreement NNSA's HEU...

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

    SciTech Connect (OSTI)

    Brim, Cornelia P.

    2013-04-01

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

  8. Uranium enrichment: investment options for the long term

    SciTech Connect (OSTI)

    Not Available

    1983-01-01

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

  9. Application of the HGSYSTEM/UF{sub 6} model to simulate atmospheric dispersion of UF{sub 6} releases from uranium enrichment plants

    SciTech Connect (OSTI)

    Goode, W.D. Jr.; Bloom, S.G.; Keith, K.D. Jr.

    1995-03-01

    Uranium hexafluoride is a dense, reactive gas used in Gaseous Diffusion Plants (GDPs) to make uranium enriched in the {sup 235}U isotope. Large quantities of UF{sub 6} exist at the GDPs in the form of in-process gas and as a solid in storage cylinders; smaller amounts exist as hot liquid during transfer operations. If liquid UF{sub 6} is released to the environment, it immediately flashes to a solid and a dense gas that reacts rapidly with water vapor in the air to form solid particles of uranyl fluoride and hydrogen fluoride gas. Preliminary analyses were done on various accidental release scenarios to determine which scenarios must be considered in the safety analyses for the GDPS. These scenarios included gas releases due to failure of process equipment and liquid/gas releases resulting from a breach of transfer piping from a cylinder. A major goal of the calculations was to estimate the response time for mitigating actions in order to limit potential off-site consequences of these postulated releases. The HGSYSTEM/UF{sub 6} code was used to assess the consequences of these release scenarios. Inputs were developed from release calculations which included two-phase, choked flow followed by expansion to atmospheric pressure. Adjustments were made to account for variable release rates and multiple release points. Superpositioning of outputs and adjustments for exposure time were required to evaluate consequences based on health effects due to exposures to uranium and HF at a specific location.

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

    SciTech Connect (OSTI)

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

    2008-01-01

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

  11. Uranium enrichment management review: summary of analysis

    SciTech Connect (OSTI)

    Not Available

    1981-01-01

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

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

    SciTech Connect (OSTI)

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

    2012-06-06

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

  13. Conversion and Blending Facility Highly enriched uranium to low enriched uranium as uranium hexafluoride. Revision 1

    SciTech Connect (OSTI)

    NONE

    1995-07-05

    This report describes the Conversion and Blending Facility (CBF) which will have two missions: (1) convert surplus HEU materials to pure HEU UF{sub 6} and a (2) blend the pure HEU UF{sub 6} with diluent UF{sub 6} to produce LWR grade LEU-UF{sub 6}. The primary emphasis of this blending be to destroy the weapons capability of large, surplus stockpiles of HEU. The blended LEU product can only be made weapons capable again by the uranium enrichment process. The chemical and isotopic concentrations of the blended LEU product will be held within the specifications required for LWR fuel. The blended LEU product will be offered to the United States Enrichment Corporation (USEC) to be sold as feed material to the commercial nuclear industry.

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

    SciTech Connect (OSTI)

    Robert Bean; Casey Durst

    2009-10-01

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

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

    SciTech Connect (OSTI)

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

    2008-06-13

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

  16. Supply of enriched uranium for research reactors

    SciTech Connect (OSTI)

    Mueller, H. [NUKEM GmbH, Alzenau (Germany)

    1997-08-01

    Since the RERTR-meeting In Newport/USA in 1990 the author delivered a series of papers in connection with the fuel cycle for research reactors dealing with its front-end. In these papers the author underlined the need for unified specifications for enriched uranium metal suitable for the production of fuel elements and made proposals with regard to the re-use of in Europe reprocessed highly enriched uranium. With regard to the fuel cycle of research reactors the research reactor community was since 1989 more concentrating on the problems of its back-end since the USA stopped the acceptance of spent research reactor fuel on December 31, 1988. Now, since it is apparent that these back-end problem have been solved by AEA`s ability to reprocess and the preparedness of the USA to again accept physically spent research reactor fuel the author is focusing with this paper again on the front-end of the fuel cycle on the question whether there is at all a safe supply of low and high enriched uranium for research reactors in the future.

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

    SciTech Connect (OSTI)

    Sewell, P.G.

    1991-11-01

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

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

    SciTech Connect (OSTI)

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

    2003-02-27

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

  19. Overview of enrichment plant safeguards

    SciTech Connect (OSTI)

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

    1982-01-01

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

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

    Office of Environmental Management (EM)

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

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

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

    440 Highly Enriched Uranium Materials Facility (HEUMF) Major Design Changes Late Lessons Learned Report Apr 2010.pdf More Documents & Publications EIS-0387: Draft Site-Wide...

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

    Office of Environmental Management (EM)

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

  3. US developments in technology for uranium enrichment

    SciTech Connect (OSTI)

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

    1982-01-01

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

  4. Uranium enrichment: heading for a cliff

    SciTech Connect (OSTI)

    Norman, C.

    1987-05-22

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

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

    SciTech Connect (OSTI)

    Myers, Astasia [Stanford University, Stanford, CA 94305, USA and MonAme Scientific Research Center, Ulaanbaatar (Mongolia)

    2011-06-28

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

  6. An Optically Stimulated Luminescence Uranium Enrichment Monitor

    SciTech Connect (OSTI)

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

    2010-08-11

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

  7. Gas Centrifuge Enrichment Plant Safeguards System Modeling

    SciTech Connect (OSTI)

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

    2006-06-05

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

  8. Standard specification for uranium hexafluoride enriched to less than 5 % 235U

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This specification covers nuclear grade uranium hexafluoride (UF6) that either has been processed through an enrichment plant, or has been produced by the blending of Highly Enriched Uranium with other uranium to obtain uranium of any 235U concentration below 5 % and that is intended for fuel fabrication. The objectives of this specification are twofold: (1) To define the impurity and uranium isotope limits for Enriched Commercial Grade UF6 so that, with respect to fuel design and manufacture, it is essentially equivalent to enriched uranium made from natural UF6; and (2) To define limits for Enriched Reprocessed UF6 to be expected if Reprocessed UF6 is to be enriched without dilution with Commercial Natural UF6. For such UF6, special provisions, not defined herein, may be needed to ensure fuel performance and to protect the work force, process equipment, and the environment. 1.2 This specification is intended to provide the nuclear industry with a standard for enriched UF6 that is to be used in the pro...

  9. Basic characterization of highly enriched uranium by gamma spectrometry

    E-Print Network [OSTI]

    Cong Tam Nguyen; Jozsef Zsigrai

    2005-08-25

    Gamma-spectrometric methods suitable for the characterization of highly enriched uranium samples encountered in illicit trafficking of nuclear materials are presented. In particular, procedures for determining the 234U, 235U, 238U, 232U and 236U contents and the age of highly enriched uranium are described. Consequently, the total uranium content and isotopic composition can be calculated. For determining the 238U and 232U contents a low background chamber was used. In addition, age dating of uranium was also performed using low-background spectrometry.

  10. Basic characterization of highly enriched uranium by gamma spectrometry

    E-Print Network [OSTI]

    Nguyen, C T

    2006-01-01

    Gamma-spectrometric methods suitable for the characterization of highly enriched uranium samples encountered in illicit trafficking of nuclear materials are presented. In particular, procedures for determining the 234U, 235U, 238U, 232U and 236U contents and the age of highly enriched uranium are described. Consequently, the total uranium content and isotopic composition can be calculated. For determining the 238U and 232U contents a low background chamber was used. In addition, age dating of uranium was also performed using low-background spectrometry.

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

    Office of Environmental Management (EM)

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

  12. 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)

    2014-01-01

    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 ...

  13. Description of the Portsmouth Gas Centrifuge Enrichment Plant

    SciTech Connect (OSTI)

    Arthur, W.B.

    1980-12-16

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

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

    SciTech Connect (OSTI)

    Not Available

    1990-06-01

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

  15. Criticality safety evaluation for Portsmouth X-345 High-Enriched-Uranium storage area

    SciTech Connect (OSTI)

    Koponen, B.L.

    1993-09-20

    This report evaluates nuclear criticality safety for the High-Enriched Uranium storage area of the X-345 building of the Portsmouth Gaseous Diffusion Plant. The effects of loss of moderation or mass control are examined for storage units in or out of the storage receptacles. Recommendations are made for decreasing criticality hazards under some conditions of storage or handling considered to be hazardous.

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

    SciTech Connect (OSTI)

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

    1996-12-31

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

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

    SciTech Connect (OSTI)

    Not Available

    1982-01-01

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

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

    SciTech Connect (OSTI)

    Curtis, Michael M.

    2008-03-31

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

  19. 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...

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

    Broader source: Energy.gov [DOE]

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

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

    SciTech Connect (OSTI)

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

    2008-06-05

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

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

    E-Print Network [OSTI]

    Newton, Thomas Henderson

    2006-01-01

    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 ...

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

    Broader source: Energy.gov [DOE]

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

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

    Office of Energy Efficiency and Renewable Energy (EERE)

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

  5. Measurement of uranium enrichment by gamma spectroscopy: result of an experimental design

    E-Print Network [OSTI]

    PAPER Measurement of uranium enrichment by gamma spectroscopy: result of an experimental design Gamma spectroscopy is commonly used in nuclear safeguards to measure uranium enrichment. An experimental design has been carried out for the measurement of uranium enrichment using this technique with different

  6. Issues and recommendations related to replacement of CFC-114 at the uranium enrichment gaseous diffusion plant. Task title: Chlorofluorocarbon (CFC) Program Review, Final report, August 1, 1991--October 1, 1992

    SciTech Connect (OSTI)

    Anderson, B.L.; Banaghan, E.

    1993-03-31

    The operating uranium enrichment gaseous diffusion plants (GDPs) in Portsmouth, Ohio and Paducah, Kentucky, which are operated for the United States Department for Energy by Martin Marietta Energy Systems (MMES), currently use a chlorofluorocarbon (CFC-114) as the primary process stream coolant. Due to recent legislation embodied in the Clean Air Act, the production of this and other related chlorofluorocarbons (CFCS) are to be phased out with no production occurring after 1995. Since the plants lose approximately 500,000 pounds per year of this process stream coolant through various leaks, the GDPs are faced with the challenge of identifying a replacement coolant that will allow continued operation of the plants. MMES formed the CFC Task Team to identify and solve the various problems associated with identifying and implementing a replacement coolant. This report includes a review of the work performed by the CFC Task Team, and recommendations that were formulated based on this review and upon original work. The topics covered include; identifying a replacement coolant, coolant leak detection and repair efforts, coolant safety concerns, coolant level sensors, regulatory issues, and an analytical decision analysis.

  7. Simulation of transportation of low enriched uranium solutions

    SciTech Connect (OSTI)

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

    1996-08-01

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

  8. 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

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

    SciTech Connect (OSTI)

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

    1995-12-31

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

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

    SciTech Connect (OSTI)

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

    2012-06-07

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

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

    SciTech Connect (OSTI)

    Rothe, R.E.

    1996-04-01

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

  12. Chapter 20 - Uranium Enrichment Decontamination & Decommissioning Fund

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

    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 Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustmentsShirley Ann JacksonDepartment|Marketing, LLCEfficiencyCOP 21:Department ofUranium Enrichment

  13. Natural uranium/conversion services/enrichment services

    SciTech Connect (OSTI)

    1993-12-31

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

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

    SciTech Connect (OSTI)

    Not Available

    1991-04-01

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

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

    SciTech Connect (OSTI)

    Anheier, Norman C.; Bushaw, Bruce A.

    2010-04-15

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

  16. Uranium mineralization in fluorine-enriched volcanic rocks

    SciTech Connect (OSTI)

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

    1980-09-01

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

  17. Engineering analysis of low enriched uranium fuel using improved zirconium hydride cross sections 

    E-Print Network [OSTI]

    Candalino, Robert Wilcox

    2006-10-30

    for the change out of the existing high enriched uranium fuel to this high-burnup, low enriched uranium fuel design. The codes MCNP and Monteburns were utilized for the neutronic analysis while the code PARET was used to determine fuel and cladding temperatures...

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

    SciTech Connect (OSTI)

    Anheier, Norman C.; Bushaw, Bruce A.

    2010-08-11

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

  19. 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)

    2011-01-01

    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 ...

  20. 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

    2008-01-01

    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 monolithic UMo fuel. The design of an optimum LEU core for the ...

  1. 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

    2012-01-01

    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 ...

  2. Validation of the Monte Carlo Criticality Program KENO V. a for highly-enriched uranium systems

    SciTech Connect (OSTI)

    Knight, J.R.

    1984-11-01

    A series of calculations based on critical experiments have been performed using the KENO V.a Monte Carlo Criticality Program for the purpose of validating KENO V.a for use in evaluating Y-12 Plant criticality problems. The experiments were reflected and unreflected systems of single units and arrays containing highly enriched uranium metal or uranium compounds. Various geometrical shapes were used in the experiments. The SCALE control module CSAS25 with the 27-group ENDF/B-4 cross-section library was used to perform the calculations. Some of the experiments were also calculated using the 16-group Hansen-Roach Library. Results are presented in a series of tables and discussed. Results show that the criteria established for the safe application of the KENO IV program may also be used for KENO V.a results.

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

    SciTech Connect (OSTI)

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

    2010-04-05

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

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

    SciTech Connect (OSTI)

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

    2012-01-01

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

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

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

    Administration (NNSA) will remove up to 200 metric tons (MT) of Highly Enriched Uranium (HEU), in the coming decades, from further use as fissile material in U.S. nuclear...

  6. 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.

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

    SciTech Connect (OSTI)

    Not Available

    1987-08-01

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

  8. Uranium Purchases Report

    Reports and Publications (EIA)

    1996-01-01

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

  9. Validation of KENO V.a for highly enriched uranium systems with hydrogen and/or carbon moderation

    SciTech Connect (OSTI)

    Elliott, E.P.; Vornehm, R.G. [Oak Ridge Y-12 Plant, TN (United States); Dodds, H.L. Jr. [Univ. of Tennessee, Knoxville, TN (United States). Nuclear Engineering Dept.

    1993-06-04

    This paper describes the validation in accordance with ANSI/ANS-8.1-1983(R1988) of KENO V.a using the 27-group ENDF/B-IV cross-section library for systems containing highly-enriched uranium, carbon, and hydrogen and for systems containing highly-enriched uranium and carbon with high carbon to uranium (C/U) atomic ratios. The validation has been performed for two separate computational platforms: an IBM 3090 mainframe and an HP 9000 Model 730 workstation, both using the Oak Ridge Y-12 Plant Nuclear Criticality Safety Software (NCSS) code package. Critical experiments performed at the Oak Ridge Critical Experiments Facility, in support of the Rover reactor program, and at the Pajarito site at Los Alamos National Laboratory were identified as having the constituents desired for this validation as well as sufficient experimental detail to allow accurate construction of KENO V.a calculational models. Calculated values of k{sub eff} for the Rover experiments, which contain uranium, carbon, and hydrogen, are between 1.0012 {+-} 0.0026 and 1.0245 {+-} 0.0023. Calculation of the Los Alamos experiments, which contain uranium and carbon at high C/U ratios, yields values of k{sub eff} between 0.9746 {+-} 0.0028 and 0.9983 {+-} 0.0027. Safety criteria can be established using this data for both types of systems.

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

    SciTech Connect (OSTI)

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

    2014-02-01

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

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

    SciTech Connect (OSTI)

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

    2010-09-13

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

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

    DOE Patents [OSTI]

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

    1995-05-02

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

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

    DOE Patents [OSTI]

    McLean, II, William (Oakland, CA); Miller, Philip E. (Livermore, CA); Horton, James A. (Livermore, CA)

    1995-01-01

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

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

    SciTech Connect (OSTI)

    Freeman, Corey R; Geist, William H

    2010-01-01

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

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

    SciTech Connect (OSTI)

    Primm, Trent; Guida, Tracey

    2010-02-01

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

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

    SciTech Connect (OSTI)

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

    2012-07-01

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

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

    SciTech Connect (OSTI)

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

    2012-07-17

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

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

    SciTech Connect (OSTI)

    Cantrell, J.

    2012-05-23

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

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

    Broader source: Energy.gov [DOE]

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

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

    Broader source: Energy.gov [DOE]

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

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

    Broader source: Energy.gov [DOE]

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

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

    Broader source: Energy.gov [DOE]

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

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

    Broader source: Energy.gov [DOE]

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

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

    Broader source: Energy.gov [DOE]

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

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

    Broader source: Energy.gov [DOE]

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

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

    Broader source: Energy.gov [DOE]

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

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

    SciTech Connect (OSTI)

    Rothe, R.E.

    1996-09-30

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

  8. Soil to plant transfer of 238 Th on a uranium

    E-Print Network [OSTI]

    Hu, Qinhong "Max"

    Soil to plant transfer of 238 U, 226 Ra and 232 Th on a uranium mining-impacted soil from species grown in soils from southeastern China contaminated with uranium mine tailings were analyzed. Keywords: Uranium; Thorium; Radium; Tailings-contaminated soil; Soileplant transfer 1. Introduction

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

    SciTech Connect (OSTI)

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

    2010-01-01

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

  10. Uranium enrichment decontamination and decommissioning fund, 1995 report

    SciTech Connect (OSTI)

    NONE

    1996-11-01

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

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

    SciTech Connect (OSTI)

    Not Available

    1990-06-01

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

  12. Initial report on characterization of excess highly enriched uranium

    SciTech Connect (OSTI)

    NONE

    1996-07-01

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

  13. Genetic engineering of syringyl-enriched lignin in plants

    DOE Patents [OSTI]

    Chiang, Vincent Lee; Li, Laigeng

    2004-11-02

    The present invention relates to a novel DNA sequence, which encodes a previously unidentified lignin biosynthetic pathway enzyme, sinapyl alcohol dehydrogenase (SAD) that regulates the biosynthesis of syringyl lignin in plants. Also provided are methods for incorporating this novel SAD gene sequence or substantially similar sequences into a plant genome for genetic engineering of syringyl-enriched lignin in plants.

  14. Radiological health aspects of commercial uranium conversion, enrichment, and fuel fabrication

    SciTech Connect (OSTI)

    Stoetzel, G.A.; Hoenes, G.R.; Cummings, F.M.; McCormack, W.D.

    1982-11-01

    Detailed information concerning occupational exposures, health physics practices, and regulatory procedures at commercial conversion, enrichment and fuel fabrication facilities is given. Sites visits were the primary source of information, which is divided into four sections. The first section discusses health physics practices that are common to the conversion, enrichment, and fuel fabrication phases of the commercial uranium industry. The next three sections review process descriptions, radiological health practices, and regulatory procedures for the three phases. Nonradiological exposures are considered only as they influence the interpretation of the health effects of radiological exposures. The review of regulatory procedures indicates the types of exposure evaluation records being kept on uranium workers and the responsibility for maintaining the records.

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

    SciTech Connect (OSTI)

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

    2010-10-01

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

  16. Current status and future plan of uranium enrichment technology

    SciTech Connect (OSTI)

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

    1994-12-31

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

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

    SciTech Connect (OSTI)

    Ilas, Germina [ORNL; Primm, Trent [ORNL

    2009-11-01

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

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

    Broader source: Energy.gov [DOE]

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

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

    SciTech Connect (OSTI)

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

    2010-03-01

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

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

    SciTech Connect (OSTI)

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

    2010-02-01

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

  1. THE THEORY OF URANIUM ENRICHMENT BY THE GAS CENTRIFUGE

    SciTech Connect (OSTI)

    Olander, Donald R.

    1981-03-01

    Onsager's analysis of the hydrodynamics of fluid circulation in the boundary layer on the rotor wall of a gas centrifuge is reviewed. The description of the flow in the boundary layers on the top and bottom end caps due to Carrier and Maslen is summarized. The method developed by Wood and Morton of coupling the flow models in the rotor wall and end cap boundary layers to complete the hydrodynamic analysis of the centrifuge is presented. Mechanical and thermal methods of driving the internal gas circulation are described. The isotope enrichment which results from the superposition of the elementary separation effect due to the centrifugal field in the gas and its internal circulation is analyzed by the Onsager-Cohen theory. The performance function representing the optimized separative power of a centrifuge as a function of throughput and cut is calculated for several simplified internal flow models. The use of asymmetric ideal cascades to exploit the distinctive features of centrifuge performance functions is illustrated.

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

    SciTech Connect (OSTI)

    FINFROCK SH

    2009-12-10

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

  3. 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

    2011-02-01

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

  4. Centrifuge enrichment plants. (Latest citations from the NTIS bibliographic database). Published Search

    SciTech Connect (OSTI)

    Not Available

    1994-03-01

    The bibliography contains citations concerning the design, control, monitoring, and safety of centrifuge enrichment plants. Power supplies, enrichment plant safeguards, facility design, cascade heater test loops to monitor the enrichment process, inspection strategies, and the socioeconomic effects of centrifuge enrichment plants are examined. Radioactive waste disposal problems are considered. (Contains a minimum of 172 citations and includes a subject term index and title list.)

  5. Centrifuge enrichment plants. (Latest citations from the NTIS bibliographic database). Published Search

    SciTech Connect (OSTI)

    1993-09-01

    The bibliography contains citations concerning the design, control, monitoring, and safety of centrifuge enrichment plants. Power supplies, enrichment plant safeguards, facility design, cascade heater test loops to monitor the enrichment process, inspection strategies, and the socioeconomic effects of centrifuge enrichment plants are examined. Radioactive waste disposal problems are considered. (Contains a minimum of 171 citations and includes a subject term index and title list.)

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

    DOE Patents [OSTI]

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

    1995-05-30

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

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

    DOE Patents [OSTI]

    Horton, James A. (Livermore, CA); Hayden, Jr., Howard W. (Oakridge, TN)

    1995-01-01

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

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

    SciTech Connect (OSTI)

    Marwick, P.

    1994-12-15

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

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

    SciTech Connect (OSTI)

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

    2009-03-01

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

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

    SciTech Connect (OSTI)

    Marshall, Margaret A.

    2014-11-04

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

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

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

    Marshall, Margaret A.

    2014-11-04

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

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

    SciTech Connect (OSTI)

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

    2010-07-01

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

  13. Partial Safety Analysis for a Reduced Uranium Enrichment Core for the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Primm, Trent [ORNL; Gehin, Jess C [ORNL

    2009-04-01

    A computational model of the reactor core of the High Flux Isotope Rector (HFIR) was developed in order to analyze non-destructive accidents caused by transients during reactor operation. The reactor model was built for the latest version of the nuclear analysis software package called Program for the Analysis of Reactor Transients (PARET). Analyses performed with the model constructed were compared with previous data obtained with other tools in order to benchmark the code. Finally, the model was used to analyze the behavior of the reactor under transients using a different nuclear fuel with lower enrichment of uranium (LEU) than the fuel currently used, which has a high enrichment of uranium (HEU). The study shows that the presence of fertile isotopes in LEU fuel, which increases the neutron resonance absorption, reduces the impact of transients on the fuel and enhances the negative reactivity feedback, thus, within the limitations of this study, making LEU fuel appear to be a safe alternative fuel for the reactor core.

  14. Conversion and standardization of university reactor fuels using low-enrichment uranium - options and costs

    SciTech Connect (OSTI)

    Harris, D.R.; Matos, J.E.; Young, H.H.

    1985-01-01

    The highly-enriched uranium (HEU) fuel used in twenty United States university reactors can be viewed as contributing to the risk of theft or diversion of weapons-useable material. The US Nuclear Regulatory Commission has issued a policy statement expressing its concern and has published a proposed rule on limiting the use of HEU in NRC-licensed non-power reactors. The fuel options, functional impacts, licensing, and scheduling of conversion and standardization of these reactor fuels to use of low-enrichment uranium (LEU) have been assessed. The university reactors span a wide range in form and function, from medium-power intense neutron sources where HEU fuel may be required, to low-power training and research facilities where HEU fuel is unnecessary. Conversion provides an opportunity to standardize university reactor fuels and improve reactor utilization in some cases. The entire program is estimated to cost about $10 million and to last about five years. Planning for conversion and standardization is facilitated by the US Department of Energy. 20 refs., 1 tab.

  15. Design of an Unattended Environmental Aerosol Sampling and Analysis System for Gaseous Centrifuge Enrichment Plants

    SciTech Connect (OSTI)

    Anheier, Norman C.; Munley, John T.; Alexander, M. L.

    2011-07-19

    The resources of the IAEA continue to be challenged by the rapid, worldwide expansion of nuclear energy production. Gaseous centrifuge enrichment plants (GCEPs) represent an especially formidable dilemma to the application of safeguard measures, as the size and enrichment capacity of GCEPs continue to escalate. During the early part of the 1990's, the IAEA began to lay the foundation to strengthen and make cost-effective its future safeguard regime. Measures under Part II of 'Programme 93+2' specifically sanctioned access to nuclear fuel production facilities and environmental sampling by IAEA inspectors. Today, the Additional Protocol grants inspection and environmental sample collection authority to IAEA inspectors at GCEPs during announced and low frequency unannounced (LFUA) inspections. During inspections, IAEA inspectors collect environmental swipe samples that are then shipped offsite to an analytical laboratory for enrichment assay. This approach has proven to be an effective deterrence to GCEP misuse, but this method has never achieved the timeliness of detection goals set forth by IAEA. Furthermore it is questionable whether the IAEA will have the resources to even maintain pace with the expansive production capacity of the modern GCEP, let alone improve the timeliness in reaching current safeguards conclusions. New safeguards propositions, outside of familiar mainstream safeguard measures, may therefore be required that counteract the changing landscape of nuclear energy fuel production. A new concept is proposed that offers rapid, cost effective GCEP misuse detection, without increasing LFUA inspection access or introducing intrusive access demands on GCEP operations. Our approach is based on continuous onsite aerosol collection and laser enrichment analysis. This approach mitigates many of the constraints imposed by the LFUA protocol, reduces the demand for onsite sample collection and offsite analysis, and overcomes current limitations associated with the in-facility misuse detection devices. Onsite environmental sample collection offers the ability to collect fleeting uranium hexafluoride emissions before they are lost to the ventilation system or before they disperse throughout the facility, to become deposited onto surfaces that are contaminated with background and historical production material. Onsite aerosol sample collection, combined with enrichment analysis, provides the unique ability to quickly detect stepwise enrichment level changes within the facility, leading to a significant strengthening of facility misuse deterence. We report in this paper our study of several GCEP environmental sample release scenarios and simulation results of a newly designed aerosol collection and particle capture system that is fully integrated with the Laser Ablation, Absorbance Ratio Spectrometry (LAARS) uranium particle enrichment analysis instrument that was developed at the Pacific Northwest National Laboratory.

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

    SciTech Connect (OSTI)

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

    2009-01-01

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

  17. An Inspector's Assessment of the New Model Safeguards Approach for Enrichment Plants

    SciTech Connect (OSTI)

    Curtis, Michael M.

    2007-07-31

    This conference paper assesses the changes that are being made to the Model Safeguards Approach for Gas Centrifuge Enrichment Plants.

  18. LISSAT Analysis of a Generic Centrifuge Enrichment Plant

    SciTech Connect (OSTI)

    Lambert, H; Elayat, H A; O?Connell, W J; Szytel, L; Dreicer, M

    2007-05-31

    The U.S. Department of Energy (DOE) is interested in developing tools and methods for use in designing and evaluating safeguards systems for current and future plants in the nuclear power fuel cycle. The DOE is engaging several DOE National Laboratories in efforts applied to safeguards for chemical conversion plants and gaseous centrifuge enrichment plants. As part of the development, Lawrence Livermore National Laboratory has developed an integrated safeguards system analysis tool (LISSAT). This tool provides modeling and analysis of facility and safeguards operations, generation of diversion paths, and evaluation of safeguards system effectiveness. The constituent elements of diversion scenarios, including material extraction and concealment measures, are structured using directed graphs (digraphs) and fault trees. Statistical analysis evaluates the effectiveness of measurement verification plans and randomly timed inspections. Time domain simulations analyze significant scenarios, especially those involving alternate time ordering of events or issues of timeliness. Such simulations can provide additional information to the fault tree analysis and can help identify the range of normal operations and, by extension, identify additional plant operational signatures of diversions. LISSAT analyses can be used to compare the diversion-detection probabilities for individual safeguards technologies and to inform overall strategy implementations for present and future plants. Additionally, LISSAT can be the basis for a rigorous cost-effectiveness analysis of safeguards and design options. This paper will describe the results of a LISSAT analysis of a generic centrifuge enrichment plant. The paper will describe the diversion scenarios analyzed and the effectiveness of various safeguards systems alternatives.

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

    SciTech Connect (OSTI)

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

    2012-07-01

    Protection and control of nuclear fuels is paramount for nuclear security and safeguards; therefore, it is important to develop fast and robust controlling mechanisms to ensure the safety of nuclear fuels. Through both passive- and active-interrogation methods we can use fast-neutron detection to perform real-time measurements of fission neutrons for process monitoring. Active interrogation allows us to use different ranges of incident neutron energy to probe for different isotopes of uranium. With fast-neutron detectors, such as organic liquid scintillation detectors, we can detect the induced-fission neutrons and photons and work towards quantifying a sample’s mass and enrichment. Using MCNPX-PoliMi, a system was designed to measure induced-fission neutrons from U-235 and U-238. Measurements were then performed in the summer of 2010 at the Joint Research Centre in Ispra, Italy. Fissions were induced with an associated particle D-T generator and an isotopic Am-Li source. The fission neutrons, as well as neutrons from (n, 2n) and (n, 3n) reactions, were measured with five 5” by 5” EJ-309 organic liquid scintillators. The D-T neutron generator was available as part of a measurement campaign in place by Padova University. The measurement and data-acquisition systems were developed at the University of Michigan utilizing a CAEN V1720 digitizer and pulse-shape discrimination algorithms to differentiate neutron and photon detections. Low-enriched uranium samples of varying mass and enrichment were interrogated. Acquired time-of-flight curves and cross-correlation curves are currently analyzed to draw relationships between detected neutrons and sample mass and enrichment. In the full paper, the promise of active-interrogation measurements and fast-neutron detection will be assessed through the example of this proof-of-concept measurement campaign. Additionally, MCNPX-PoliMi simulation results will be compared to the measured data to validate the MCNPX-PoliMi code when used for active-interrogation simulations.

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

    SciTech Connect (OSTI)

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

    2015-08-01

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

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

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

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

    2015-08-01

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

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

    SciTech Connect (OSTI)

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

    2010-01-01

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

  3. Centrifuge enrichment plants. (Latest citations from the NTIS data base). Published Search

    SciTech Connect (OSTI)

    Not Available

    1992-09-01

    The bibliography contains citations concerning the design, control, monitoring, and safety of centrifuge enrichment plants. Power supplies, enrichment plant safeguards, facility design, cascade heater test loops to monitor the enrichment process, inspection strategies, and the socioeconomic effects of centrifuge enrichment plants are examined. Radioactive waste disposal problems are briefly considered. (Contains a minimum of 169 citations and includes a subject term index and title list.)

  4. Characterization of past and present solid waste streams from the Plutonium-Uranium Extraction Plant

    SciTech Connect (OSTI)

    Pottmeyer, J.A.; Weyns, M.I.; Lorenzo, D.S.; Vejvoda, E.J. [Los Alamos Technical Associates, Inc., NM (US); Duncan, D.R. [Westinghouse Hanford Co., Richland, WA (US)

    1993-04-01

    During the next two decades the transuranic wastes, now stored in the burial trenches and storage facilities at the Hanford Site, are to be retrieved, processed at the Waste Receiving and Processing Facility, and shipped to the Waste Isolation Pilot Plant near Carlsbad, New Mexico for final disposal. Over 7% of the transuranic waste to be retrieved for shipment to the Waste Isolation Pilot Plant has been generated at the Plutonium-Uranium Extraction (PUREX) Plant. The purpose of this report is to characterize the radioactive solid wastes generated by PUREX using process knowledge, existing records, and oral history interviews. The PUREX Plant is currently operated by the Westinghouse Hanford Company for the US Department of Energy and is now in standby status while being prepared for permanent shutdown. The PUREX Plant is a collection of facilities that has been used primarily to separate plutonium for nuclear weapons from spent fuel that had been irradiated in the Hanford Site`s defense reactors. Originally designed to reprocess aluminum-clad uranium fuel, the plant was modified to reprocess zirconium alloy clad fuel elements from the Hanford Site`s N Reactor. PUREX has provided plutonium for research reactor development, safety programs, and defense. In addition, the PUREX was used to recover slightly enriched uranium for recycling into fuel for use in reactors that generate electricity and plutonium. Section 2.0 provides further details of the PUREX`s physical plant and its operations. The PUREX Plant functions that generate solid waste are as follows: processing operations, laboratory analyses and supporting activities. The types and estimated quantities of waste resulting from these activities are discussed in detail.

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

    SciTech Connect (OSTI)

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

    1991-01-01

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

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

    SciTech Connect (OSTI)

    Lewis, K.D.

    1994-09-01

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

  7. A Monte Carlo Analysis of Gas Centrifuge Enrichment Plant Process Load Cell Data

    SciTech Connect (OSTI)

    Garner, James R; Whitaker, J Michael

    2013-01-01

    As uranium enrichment plants increase in number, capacity, and types of separative technology deployed (e.g., gas centrifuge, laser, etc.), more automated safeguards measures are needed to enable the IAEA to maintain safeguards effectiveness in a fiscally constrained environment. Monitoring load cell data can significantly increase the IAEA s ability to efficiently achieve the fundamental safeguards objective of confirming operations as declared (i.e., no undeclared activities), but care must be taken to fully protect the operator s proprietary and classified information related to operations. Staff at ORNL, LANL, JRC/ISPRA, and University of Glasgow are investigating monitoring the process load cells at feed and withdrawal (F/W) stations to improve international safeguards at enrichment plants. A key question that must be resolved is what is the necessary frequency of recording data from the process F/W stations? Several studies have analyzed data collected at a fixed frequency. This paper contributes to load cell process monitoring research by presenting an analysis of Monte Carlo simulations to determine the expected errors caused by low frequency sampling and its impact on material balance calculations.

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

    SciTech Connect (OSTI)

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

    2008-09-01

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

  9. Onsite Gaseous Centrifuge Enrichment Plant UF6 Cylinder Destructive Analysis

    SciTech Connect (OSTI)

    Anheier, Norman C.; Cannon, Bret D.; Qiao, Hong; Carter, Jennifer C.; McNamara, Bruce K.; O'Hara, Matthew J.; Phillips, Jon R.; Curtis, Michael M.

    2012-07-17

    The IAEA safeguards approach for gaseous centrifuge enrichment plants (GCEPs) includes measurements of gross, partial, and bias defects in a statistical sampling plan. These safeguard methods consist principally of mass and enrichment nondestructive assay (NDA) verification. Destructive assay (DA) samples are collected from a limited number of cylinders for high precision offsite mass spectrometer analysis. DA is typically used to quantify bias defects in the GCEP material balance. Under current safeguards measures, the operator collects a DA sample from a sample tap following homogenization. The sample is collected in a small UF6 sample bottle, then sealed and shipped under IAEA chain of custody to an offsite analytical laboratory. Current practice is expensive and resource intensive. We propose a new and novel approach for performing onsite gaseous UF6 DA analysis that provides rapid and accurate assessment of enrichment bias defects. DA samples are collected using a custom sampling device attached to a conventional sample tap. A few micrograms of gaseous UF6 is chemically adsorbed onto a sampling coupon in a matter of minutes. The collected DA sample is then analyzed onsite using Laser Ablation Absorption Ratio Spectrometry-Destructive Assay (LAARS-DA). DA results are determined in a matter of minutes at sufficient accuracy to support reliable bias defect conclusions, while greatly reducing DA sample volume, analysis time, and cost.

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

    SciTech Connect (OSTI)

    Rudisill, T; Fernando Fondeur, F

    2009-01-15

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

  11. Seven years of carbon dioxide enrichment, nitrogen fertilization and plant diversity influence arbuscular

    E-Print Network [OSTI]

    Thomas, David D.

    Seven years of carbon dioxide enrichment, nitrogen fertilization and plant diversity influence by examining the joint effects of carbon dioxide (CO2) enrichment, nitrogen (N) fertilization and plant. Increasing atmospheric carbon dioxide (CO2) influences plant water relations and often pref- erentially

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

    SciTech Connect (OSTI)

    SCHWINKENDORF, K.N.

    2006-05-12

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

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

    SciTech Connect (OSTI)

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

    2010-09-30

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

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

    SciTech Connect (OSTI)

    Ilas, Germina [ORNL; Primm, Trent [ORNL

    2011-05-01

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

  15. Approach to IAEA material-balance verification with intermittent inspection at the Portsmouth Gas Centrifuge Enrichment Plant

    SciTech Connect (OSTI)

    Gordon, D.M.; Sanborn, J.B.

    1984-05-18

    This paper describes a potential approach by which the International Atomic Energy Agency (IAEA) might verify the nuclear-material balance at the Portsmouth Gas Centrifuge Enrichment Plant (GCEP) for the circumstance in which the IAEA inspections occur on an intermittent basis. The verification approach is a variation of the standard IAEA attributes/variables measurement-verification method. This alternative approach is useful and applicable at the Portsmouth GCEP, which will ship all its product and tails UF/sub 6/ to United States facilities not eligible for IAEA safeguards. The paper reviews some of the relevant results of the Hexapartite Safeguards Project (HSP), describes the standard IAEA material-balance-verification approach for bulk-handling facilities, and provides the procedures to be followed in handling and processing UF/sub 6/ cylinders at the Portsmouth GCEP. The paper then discusses the assumptions made in the approach, and derives a formula for the probability with which the IAEA could detect the diversion of a significant quantity of uranium (75 kg of U-235 in depleted, normal, and low-enriched uranium) if this method were applied. The paper also provides numerical examples of IAEA detection probability should the operator divert uranium from the feed, product, or tails streams for the Portsmouth GCEP with a capacity of 1100 tonnes of separative work per year.

  16. Two Methods for Converting a Heavy-Water Research Reactor to Use Low-Enriched-Uranium Fuel to Improve Proliferation Resistance After Startup

    E-Print Network [OSTI]

    Kemp, R. Scott

    This article demonstrates the feasibility of converting a heavy-water research reactor from natural to low-enriched uranium in order to slow the production of weapon-usable plutonium, even if the core cannot be physically ...

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

    SciTech Connect (OSTI)

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

    2004-07-01

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

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

    SciTech Connect (OSTI)

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

    2012-03-01

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

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

    SciTech Connect (OSTI)

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

    2014-10-30

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

  20. Centrifuge enrichment plants. January 1970-October 1988 (Citations from the NTIS data base). Report for January 1970-October 1988

    SciTech Connect (OSTI)

    Not Available

    1988-11-01

    This bibliography contains citations concerning the design, control, monitoring, and safety of centrifuge enrichment plants. Power supplies, enrichment plant safeguards, facility design, cascade heater test loops to monitor the enrichment process, inspection strategies, and the socio-economic effects of centrifuge enrichment plants are examined. Radioactive waste disposal problems are briefly considered. (Contains 151 citations fully indexed and including a title list.)

  1. Systems approach used in the Gas Centrifuge Enrichment Plant

    SciTech Connect (OSTI)

    Rooks, W.A. Jr.

    1982-01-01

    A requirement exists for effective and efficient transfer of technical knowledge from the design engineering team to the production work force. Performance-Based Training (PBT) is a systematic approach to the design, development, and implementation of technical training. This approach has been successfully used by the US Armed Forces, industry, and other organizations. The advantages of the PBT approach are: cost-effectiveness (lowest life-cycle training cost), learning effectiveness, reduced implementation time, and ease of administration. The PBT process comprises five distinctive and rigorous phases: Analysis of Job Performance, Design of Instructional Strategy, Development of Training Materials and Instructional Media, Validation of Materials and Media, and Implementation of the Instructional Program. Examples from the Gas Centrifuge Enrichment Plant (GCEP) are used to illustrate the application of PBT.

  2. 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

    2003-08-01

    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.

  3. Preconceptual design studies and cost data of depleted uranium hexafluoride conversion plants

    SciTech Connect (OSTI)

    Jones, E

    1999-07-26

    One of the more important legacies left with the Department of Energy (DOE) after the privatization of the United States Enrichment Corporation is the large inventory of depleted uranium hexafluoride (DUF6). The DOE Office of Nuclear Energy, Science and Technology (NE) is responsible for the long-term management of some 700,000 metric tons of DUF6 stored at the sites of the two gaseous diffusion plants located at Paducah, Kentucky and Portsmouth, Ohio, and at the East Tennessee Technology Park in Oak Ridge, Tennessee. The DUF6 management program resides in NE's Office of Depleted Uranium Hexafluoride Management. The current DUF6 program has largely focused on the ongoing maintenance of the cylinders containing DUF6. However, the long-term management and eventual disposition of DUF6 is the subject of a Programmatic Environmental Impact Statement (PEIS) and Public Law 105-204. The first step for future use or disposition is to convert the material, which requires construction and long-term operation of one or more conversion plants. To help inform the DUF6 program's planning activities, it was necessary to perform design and cost studies of likely DUF6 conversion plants at the preconceptual level, beyond the PEIS considerations but not as detailed as required for conceptual designs of actual plants. This report contains the final results from such a preconceptual design study project. In this fast track, three month effort, Lawrence Livermore National Laboratory and Bechtel National Incorporated developed and evaluated seven different preconceptual design cases for a single plant. The preconceptual design, schedules, costs, and issues associated with specific DUF6 conversion approaches, operating periods, and ownership options were evaluated based on criteria established by DOE. The single-plant conversion options studied were similar to the dry-conversion process alternatives from the PEIS. For each of the seven cases considered, this report contains information on the conversion process, preconceptual plant description, rough capital and operating costs, and preliminary project schedule.

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

    SciTech Connect (OSTI)

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

    2007-10-22

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

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

    SciTech Connect (OSTI)

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

    2011-05-12

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

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

    SciTech Connect (OSTI)

    Christopher Landers; Igor Bolshinsky; Ken Allen; Stanley Moses

    2010-07-01

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

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

    SciTech Connect (OSTI)

    Michael A. Pope

    2012-07-01

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

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

    SciTech Connect (OSTI)

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

    1991-09-01

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

  9. Laser and gas centrifuge enrichment

    SciTech Connect (OSTI)

    Heinonen, Olli

    2014-05-09

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

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

    SciTech Connect (OSTI)

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

    1991-09-01

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

  11. A Laser-Based Method for On-Site Analysis of UF6 at Enrichment Plants

    SciTech Connect (OSTI)

    Anheier, Norman C.; Cannon, Bret D.; Martinez, Alonzo; Barrett, Christopher A.; Taubman, Matthew S.; Anderson, Kevin K.; Smith, Leon E.

    2014-11-23

    The International Atomic Energy Agency’s (IAEA’s) long-term research and development plan calls for more cost-effective and efficient safeguard methods to detect and deter misuse of gaseous centrifuge enrichment plants (GCEPs). The IAEA’s current safeguards approaches at GCEPs are based on a combination of routine and random inspections that include environmental sampling and destructive assay (DA) sample collection from UF6 in-process material and selected cylinders. Samples are then shipped offsite for subsequent laboratory analysis. In this paper, a new DA sample collection and onsite analysis approach that could help to meet challenges in transportation and chain of custody for UF6 DA samples is introduced. This approach uses a handheld sampler concept and a Laser Ablation, Laser Absorbance Spectrometry (LAARS) analysis instrument, both currently under development at the Pacific Northwest National Laboratory. A LAARS analysis instrument could be temporarily or permanently deployed in the IAEA control room of the facility, in the IAEA data acquisition cabinet, for example. The handheld PNNL DA sampler design collects and stabilizes a much smaller DA sample mass compared to current sampling methods. The significantly lower uranium mass reduces the sample radioactivity and the stabilization approach diminishes the risk of uranium and hydrogen fluoride release. These attributes enable safe sample handling needed during onsite LAARS assay and may help ease shipping challenges for samples to be processed at the IAEA’s offsite laboratory. The LAARS and DA sampler implementation concepts will be described and preliminary technical viability results presented.

  12. New Measures to Safeguard Gas Centrifuge Enrichment Plants

    SciTech Connect (OSTI)

    Whitaker, Jr., James; Garner, James R; Whitaker, Michael; Lockwood, Dunbar; Gilligan, Kimberly V; Younkin, James R; Hooper, David A; Henkel, James J; Krichinsky, Alan M

    2011-01-01

    As Gas Centrifuge Enrichment Plants (GCEPs) increase in separative work unit (SWU) capacity, the current International Atomic Energy Agency (IAEA) model safeguards approach needs to be strengthened. New measures to increase the effectiveness of the safeguards approach are being investigated that will be mutually beneficial to the facility operators and the IAEA. One of the key concepts being studied for application at future GCEPs is embracing joint use equipment for process monitoring of load cells at feed and withdrawal (F/W) stations. A mock F/W system was built at Oak Ridge National Laboratory (ORNL) to generate and collect F/W data from an analogous system. The ORNL system has been used to collect data representing several realistic normal process and off-normal (including diversion) scenarios. Emphasis is placed on the novelty of the analysis of data from the sensors as well as the ability to build information out of raw data, which facilitates a more effective and efficient verification process. This paper will provide a progress report on recent accomplishments and next steps.

  13. Next Generation Safeguards Initiative: Analysis of Probability of Detection of Plausible Diversion Scenarios at Gas Centrifuge Enrichment Plants Using Advanced Safeguards

    SciTech Connect (OSTI)

    Hase, Kevin R.; Hawkins Erpenbeck, Heather; Boyer, Brian D.

    2012-07-10

    Over the last decade, efforts by the safeguards community, including inspectorates, governments, operators and owners of centrifuge facilities, have given rise to new possibilities for safeguards approaches in enrichment plants. Many of these efforts have involved development of new instrumentation to measure uranium mass and uranium-235 enrichment and inspection schemes using unannounced and random site inspections. We have chosen select diversion scenarios and put together a reasonable system of safeguards equipment and safeguards approaches and analyzed the effectiveness and efficiency of the proposed safeguards approach by predicting the probability of detection of diversion in the chosen safeguards approaches. We analyzed the effect of redundancy in instrumentation, cross verification of operator instrumentation by inspector instrumentation, and the effects of failures or anomalous readings on verification data. Armed with these esults we were able to quantify the technical cost benefit of the addition of certain instrument suites and show the promise of these new systems.

  14. Selection of potential IAEA inspection strategies involving cascade access at the Portsmouth Gas Centrifuge Enrichment Plant (GCEP)

    SciTech Connect (OSTI)

    Not Available

    1981-04-13

    This report has been prepared as a US contribution to Team 4 of the Hexapartite Safeguards Project. It provides to the Team 4 participants one example of an approach, which has been used in the United States, to developing a range of safeguards strategies involving differing degrees of access to cascade areas of centrifuge enrichment plants. Its purpose is to facilitate the work of other Hexapartite participants in completing Task II of Team 4's terms of reference. The scope of this report is limited to identifying safeguards approaches for the Portsmouth Gas Centrifuge Enrichment Plant (GCEP) which involve differing degrees of access to the cascade area. This report provides a method for selecting cascade access inspection strategies at GCEP which appear promising for more detailed evaluation. It is quite important to note, however, that the effectiveness and practicability of these strategies have not been established at the present. In addition, some strategies have been included on the basis of very preliminary calculations and considerations which have not been validated. Thus, some of these strategies may ultimately be rejected because they prove to be impracticable. Considerations of cost and the possible transfer of information and technology related to the production of enriched uranium will also be pertinent in considering the degrees and frequency of access to the cascade areas of centrifuge enrichment plants. This report describes the process for combining technical measures, implementation approaches and objectives to arrive at the total number of theoretically possible combinations. It then describes how these combinations may be reduced in a series of steps to a number that is more manageable for detailed evaluation. The process is shown schematically.

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

    SciTech Connect (OSTI)

    TOFFER, H.

    2006-07-18

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

  16. A "Proof-of-Concept" Demonstration of RF-Based Technologies for UF6 Cylinder Tracking at Centrifuge Enrichment Plant

    SciTech Connect (OSTI)

    Pickett, Chris A; Younkin, James R; Kovacic, Donald N; Dixon, E. T.; Martinez, B.

    2007-01-01

    This effort describes how radio-frequency (RF) technology can be integrated into a uranium enrichment facility's nuclear materials accounting and control program to enhance uranium hexafluoride (UF6) cylinder tracking and thus provide benefits to both domestic and international safeguards. Approved industry-standard cylinders are used to handle and store UF6 feed, product, tails, and samples at uranium enrichment plants. In the international arena, the International Atomic Energy Agency (IAEA) relies on time-consuming manual cylinder inventory and tracking techniques to verify operator declarations and to detect potential diversion of UF6. Development of a reliable, automated, and tamper-resistant process for tracking and monitoring UF6 cylinders would greatly reduce the risk of false or misreported cylinder tare weights, diversion of nuclear material, concealment of excess production, utilization of undeclared cylinders, and misrepresentation of the cylinders contents. This paper will describe a "proof-of concept" system that was designed show the feasibility of using RF based technologies to track individual UF6 cylinders throughout their entire life cycle, and thus ensure both increased domestic accountability of materials and a more effective and efficient method for application of IAEA international safeguards at the site level. The proposed system incorporates RF-based identification devices, which provide a mechanism for a reliable, automated, and tamper-resistant tracking network. We explore how securely attached RF tags can be integrated with other safeguards technologies to better detect diversion of cylinders. The tracking system could also provide a foundation for integration of other types of safeguards that would further enhance detection of undeclared activities.

  17. CO2 enrichment accelerates successional development of an understory plant community

    E-Print Network [OSTI]

    Post, Wilfred M.

    CO2 enrichment accelerates successional development of an understory plant community Contact Research · Over 11 years of CO2 enrichment in the ORNL FACE experiment, the understory community changed dramatically. · Aboveground biomass was on average 25% greater in elevated [CO2] than in ambient [CO2] plots

  18. Uranium hexafluoride packaging tiedown systems overview at Portsmouth Gaseous Diffusion Plant, Piketon, Ohio

    SciTech Connect (OSTI)

    Becker, D.L.; Lindquist, M.R.

    1993-03-01

    The Portsmouth Gaseous Diffusion Plant (PORTS) in Piketon, Ohio operated by Martin Marietta Energy Systems, Inc., through the US Department of Energy-Oak Ridge Operations Office (DOE-ORO) for the US Department of Energy Headquarters, Office of Nuclear Energy. The PORTS conducts those operations that are necessary for the production, packaging, and shipment of enriched uranium hexafluoride (UF[sub 6]). Uranium hexafluoride enriched greater than 1.0 wt percent [sup 235]U shall be packaged in accordance with the US Department of Transportation (DOT) regulations of Title 49 CFR Parts 173 (Reference 1) and 178 (Reference 2), or in US Nuclear Regulatory Commission (NRC) or US Department of Energy (DOE) certified package designs. Concerns have been expressed regarding the various tiedown methods and condition of the trailers being used by some shippers/carriers for international transport of the UF[sub 6] cylinders/overpacks (Reference 3). Because of the concerns about international shipments, the US Department of Energy-Headquarters (DOE-HQ) Office of Nuclear Energy, through DOE-HQ Transportation Management Division, requested Westinghouse Hanford Company (Westinghouse Hanford) to review UF[sub 6] packaging tiedown and shipping practices used by PORTS, and where possible and appropriate, provide recommendations for enhancing these practices. Consequently, a tram of two individuals from Westinghouse Hanford visited PORTS on March 5 and 6, 1990, for the purpose of conducting this review. The paper provides a brief discussion of the review activities and a summary of the resulting findings and recommendations. A detailed reporting of the review is documented in Reference 4.

  19. Uranium hexafluoride packaging tiedown systems overview at Portsmouth Gaseous Diffusion Plant, Piketon, Ohio

    SciTech Connect (OSTI)

    Becker, D.L.; Lindquist, M.R.

    1993-03-01

    The Portsmouth Gaseous Diffusion Plant (PORTS) in Piketon, Ohio operated by Martin Marietta Energy Systems, Inc., through the US Department of Energy-Oak Ridge Operations Office (DOE-ORO) for the US Department of Energy Headquarters, Office of Nuclear Energy. The PORTS conducts those operations that are necessary for the production, packaging, and shipment of enriched uranium hexafluoride (UF{sub 6}). Uranium hexafluoride enriched greater than 1.0 wt percent {sup 235}U shall be packaged in accordance with the US Department of Transportation (DOT) regulations of Title 49 CFR Parts 173 (Reference 1) and 178 (Reference 2), or in US Nuclear Regulatory Commission (NRC) or US Department of Energy (DOE) certified package designs. Concerns have been expressed regarding the various tiedown methods and condition of the trailers being used by some shippers/carriers for international transport of the UF{sub 6} cylinders/overpacks (Reference 3). Because of the concerns about international shipments, the US Department of Energy-Headquarters (DOE-HQ) Office of Nuclear Energy, through DOE-HQ Transportation Management Division, requested Westinghouse Hanford Company (Westinghouse Hanford) to review UF{sub 6} packaging tiedown and shipping practices used by PORTS, and where possible and appropriate, provide recommendations for enhancing these practices. Consequently, a tram of two individuals from Westinghouse Hanford visited PORTS on March 5 and 6, 1990, for the purpose of conducting this review. The paper provides a brief discussion of the review activities and a summary of the resulting findings and recommendations. A detailed reporting of the review is documented in Reference 4.

  20. Uranium hexafluoride packaging tiedown systems overview at Portsmouth Gaseous Diffusion Plant, Piketon, Ohio. Revision 1

    SciTech Connect (OSTI)

    Becker, D.L.; Green, D.J.; Lindquist, M.R.

    1993-07-01

    The Portsmouth Gaseous Diffusion Plant (PORTS) in Piketon, Ohio, is operated by Martin Marietta Energy Systems, Inc., through the US Department of Energy-Oak Ridge Operations Office (DOE-ORO) for the US Department of Energy-Headquarters, Office of Nuclear Energy. The PORTS conducts those operations that are necessary for the production, packaging, and shipment of uranium hexafluoride (UF{sub 6}). Uranium hexafluoride enriched uranium than 1.0 wt percent {sup 235}U shall be packaged in accordance with the US Department of Transportation (DOT) regulations of Title 49 CFR Parts 173 (Reference 1) and 178 (Reference 2), or in US Nuclear Regulatory Commission (NRC) or US Department of Energy (DOE) certified package designs. Concerns have been expressed regarding the various tiedown methods and condition of the trailers being used by some shippers/carriers for international transport of the UF{sub 6} cylinders/overpacks. Because of the concerns about international shipments, the US Department of Energy-Headquarters (DOE-HQ) Office of Nuclear Energy, through DOE-HQ Transportation Management Division, requested Westinghouse Hanford Company (Westinghouse Hanford) to review UF{sub 6} packaging tiedown and shipping practices used by PORTS, and where possible and appropriate, provide recommendations for enhancing these practices. Consequently, a team of two individuals from Westinghouse Hanford visited PORTS on March 5 and 6, 1990, for the purpose of conducting this review. The paper provides a brief discussion of the review activities and a summary of the resulting findings and recommendations. A detailed reporting of the is documented in Reference 4.

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

    SciTech Connect (OSTI)

    Not Available

    1985-01-01

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

  2. Conversion and enrichment in the Soviet Union

    SciTech Connect (OSTI)

    1991-04-01

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

  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.; Novosel’tsev, 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)

    2014-12-15

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

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

    SciTech Connect (OSTI)

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

    2011-10-01

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

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

    SciTech Connect (OSTI)

    1995-12-01

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

  6. 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)

    2006-02-01

    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.

  7. Non-Destructive Analysis Calibration Standards for Gaseous Diffusion Plant (GDP) Decommissioning

    Broader source: Energy.gov [DOE]

    The decommissioning of Gaseous Diffusion Plant facilities requires accurate, non-destructive assay (NDA) of residual enriched uranium in facility components for safeguards and nuclear criticality...

  8. Solubility Classification of Airborne Uranium Products from LWR-Fuel Plants

    SciTech Connect (OSTI)

    kalkwarf, D. R.

    1980-08-01

    Airborne dust samples were obtained from various locations within plants manufacturing fuel elements for light-water reactors, and the dissolution rates of uranium from these samples into simulated lung fluid at 37°C were measured. These measurements were used to classify the solubilities of the samples in terms of the lung clearance model proposed by the International Commission on Radiological Protection. Similar evaluations were performed for samples of pure uranium compounds expected as components in plant dust. The variation in solubility classifications of dust encountered along the fuel production lines is described and correlated with the process chemistry and the solubility classifications of the pure uranium compounds.

  9. The strategy on rehabilitation of the former uranium facilities at the 'Pridneprovsky chemical plant' in Ukraine

    SciTech Connect (OSTI)

    Voitsekhovich, O.; Lavrova, T. [Ukrainian Hydrometeorological Institute, Kiev (Ukraine); Skalskiy, A.S. [Institute of Geological Sciences of Ac.of Sc., Kiev (Ukraine); Ryazantsev, V.F. [State Nuclear Regulatory Committee of Ukraine, 9/11 Arsenalna str., Kyiv-11, 01011 (Ukraine)

    2007-07-01

    This paper describes current status of the former Uranium Facilities at the Pridneprovsky Chemical Plant in Ukraine, which are currently under development of action plan for its territory rehabilitation. The monitoring data carried out during recent several years show its impact to the Environment and gives a basis for justification of the number of measures aiming to reduce radiological and ecological risks of the Uranium tailings situated at the territory of PChP. The monitoring data and strategy for its remediation are considered in the presentation. Uranium mining has been intensively conducted in Ukraine since the end of the 40-s. Most of the uranium deposits have been explored in the Dnieper river basin, while some smaller deposits can be found within the basins of the Southern Bug and Severskiy Donets rivers. There also several large Uranium Milling facilities were in operation since the end of the 40-s till 1991, when due to disintegration of the former Soviet Union system the own uranium production has been significantly declined. The Milling Plant and Uranium extraction Facilities in ZhevtiVody is still in operation with UkrAtomprom Industrial Consortium. Therefore rehabilitation programme for all Uranium facilities in this site are in duty of the East Mining Combine and the Consortium. The most difficult case is to provide rehabilitation Action Plan for Uranium tailings and number of other facilities situated in Dnieprodzerzhinsk town and which were in operation by the former State Industrial Enterprise Pridneprovskiy Chemical Plant (PChP). In past PChP was one of the largest Uranium Milling facilities of the Former Soviet Union and has been in operation since 1948 till 1991. During Soviet time the Uranium extraction at this legacy site has been carried out using the ore raw products delivered also from Central Asia, Germany and Checz Republic. After extraction the uranium residue has been putting to the nearest landscape depressions at the vicinity of the Milling facilities. This plant is being in the sanitation stage since 1991 with the 9 Uranium tailings dumps at its territory, containing about 42 million tonnes of Uranium Residues. There were no engineering barriers created at most of the tailings. After fulfilment of the tailing dumps capacity their surfaces usually were covering by the local soils, debris and other industrial wastes. (authors)

  10. Report on the Effect the Low Enriched Uranium Delivered Under the Highly

    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 DeliciousMathematicsEnergyInterestedReplacement-2-A Wholesale Power Rate ScheduleSHERMAN STREET,and S.isEnriched

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

    SciTech Connect (OSTI)

    Not Available

    1980-01-01

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

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

    DOE Patents [OSTI]

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

    1985-02-04

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

  13. Refurbishment of uranium hexafluoride cylinder storage yards C-745-K, L, M, N, and P and construction of a new uranium hexafluoride cylinder storage yard (C-745-T) at the Paducah Gaseous Diffusion Plant, Paducah, Kentucky

    SciTech Connect (OSTI)

    1996-07-01

    The Paducah Gaseous Diffusion Plant (PGDP) is a uranium enrichment facility owned by the US Department of Energy (DOE). A residual of the uranium enrichment process is depleted uranium hexafluoride (UF6). Depleted UF6, a solid at ambient temperature, is stored in 32,200 steel cylinders that hold a maximum of 14 tons each. Storage conditions are suboptimal and have resulted in accelerated corrosion of cylinders, increasing the potential for a release of hazardous substances. Consequently, the DOE is proposing refurbishment of certain existing yards and construction of a new storage yard. This environmental assessment (EA) evaluates the impacts of the proposed action and no action and considers alternate sites for the proposed new storage yard. The proposed action includes (1) renovating five existing cylinder yards; (2) constructing a new UF6 storage yard; handling and onsite transport of cylinders among existing yards to accommodate construction; and (4) after refurbishment and construction, restacking of cylinders to meet spacing and inspection requirements. Based on the results of the analysis reported in the EA, DOE has determined that the proposed action is not a major Federal action that would significantly affect the quality of the human environment within the context of the National Environmental Policy Act of 1969. Therefore, DOE is issuing a Finding of No Significant Impact. Additionally, it is reported in this EA that the loss of less than one acre of wetlands at the proposed project site would not be a significant adverse impact.

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

    SciTech Connect (OSTI)

    Michael Tyacke; Dr. Igor Bolshinsky; Jeff Chamberlin

    2008-07-01

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

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

    Broader source: Energy.gov [DOE]

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

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

    SciTech Connect (OSTI)

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

    2008-12-23

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

  17. Position paper Oak Ridge Y-12 Plant storage of uranium in plastics

    SciTech Connect (OSTI)

    Duerksen, W.K.

    1995-07-01

    As a result of the end of the Cold War, the United States nuclear weapon stockpile is being reduced from approximately 20,000 warheads to fewer than 10,000 by the end of the century. The Oak Ridge Y-12 Plant is the Department of Energy (DOE) site charged with the responsibility of providing safe, secure storage for the uranium recovered from these weapons. In addition to weapons material, Y-12 has traditionally processed and stored uranium from nonweapon programs and presumably will continue to do so. The purpose of this document is to evaluate the suitability of plastics for use in the containment of uranium.

  18. Department of Energy`s Uranium Enrichment Decontamination and Decommissioning Fund fiscal year 1997 financial statement audit

    SciTech Connect (OSTI)

    NONE

    1998-08-21

    This report presents the results of the independent certified public accountants` audit of the Department of Energy`s Uranium Enrichment Decontamination and Decommissioning Fund (D&D Fund) financial statements as of September 30, 1997. The auditors have expressed an unqualified opinion on the 1997 statement of financial position and the related statements of operations and changes in net position and cash flows. The 1997 financial statement audit was made under provisions of the Inspector General Act (5 U.S.C. App.) as amended, the Government Management Reform Act (31 U.S.C. 3515), and Office of Management and Budget implementing guidance. The auditor`s work was conducted in accordance with generally accepted government auditing standards. To fulfill our audit responsibilities, we contracted with the independent public accounting firm of KPMG Peat Marwick LLP (KPMG) to conduct the audit for us, subject to our review. The auditors` report on the D&D Fund`s internal control structure disclosed no reportable conditions. The auditors` report on compliance with laws and regulations disclosed one instance of noncompliance. This instance of noncompliance relates to the shortfall in Government appropriations. Since this instance was addressed in a previous audit, no further recommendation is made at this time. During the course of the audit, KPMG also identified other matters that, although not material to the financial statements, nevertheless, warrant management`s attention. These items are fully discussed in a separate letter to management.

  19. Integrated Information Technology Framework for Analysis of Data from Enrichment Plants to Support the Safeguards Mission

    SciTech Connect (OSTI)

    Marr, Clifton T.; Thurman, David A.; Jorgensen, Bruce V.

    2008-07-15

    ABSTRACT Many examples of software architectures exist that support process monitoring and analysis applications which could be applied to enrichment plants in a fashion that supports the Safeguards Mission. Pacific Northwest National Laboratory (PNNL) has developed mature solutions that will provide the framework to support online statistical analysis of enrichment plans and the entire nuclear fuel cycle. Most recently, PNNL has developed a refined architecture and supporting tools that address many of the common problems analysis and modeling environments experience: pipelining, handling large data volumes, and real-time performance. We propose the architecture and tools may be successfully used in furthering the goals of nuclear material control and accountability as both an aid to processing plant owners and as comprehensive monitoring for oversight teams.

  20. Uranium industry annual 1996

    SciTech Connect (OSTI)

    1997-04-01

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

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

    SciTech Connect (OSTI)

    NONE

    1997-05-01

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

  2. India's Worsening Uranium Shortage

    SciTech Connect (OSTI)

    Curtis, Michael M.

    2007-01-15

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

  3. Advanced uranium enrichment technologies

    SciTech Connect (OSTI)

    Merriman, R.

    1983-03-10

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

  4. highly enriched uranium

    National Nuclear Security Administration (NNSA)

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

  5. highly enriched uranium

    National Nuclear Security Administration (NNSA)

    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 Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of NaturalDukeWakefield Municipal GasAdministration Medal01 Sandia4) August 20123/%2A en46Afed feed families|fleet6/%2A

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

    SciTech Connect (OSTI)

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

    1986-08-01

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

  7. The ultimate disposition of depleted uranium

    SciTech Connect (OSTI)

    Lemons, T.R. [Uranium Enrichment Organization, Oak Ridge, TN (United States)

    1991-12-31

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

  8. Status of Uranium Atomic Vapor Laser Isotope Separation Program

    SciTech Connect (OSTI)

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

    1993-06-01

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

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

    SciTech Connect (OSTI)

    Parish, T.A.

    1995-03-02

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

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

    SciTech Connect (OSTI)

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

    2014-11-01

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

  11. Approach to IAEA material-balance verification at the Portsmouth Gas Centrifuge Enrichment Plant

    SciTech Connect (OSTI)

    Gordon, D.M.; Sanborn, J.B.; Younkin, J.M.; DeVito, V.J.

    1983-01-01

    This paper describes a potential approach by which the International Atomic Energy Agency (IAEA) might verify the nuclear-material balance at the Portsmouth Gas Centrifuge Enrichment Plant (GCEP). The strategy makes use of the attributes and variables measurement verification approach, whereby the IAEA would perform independent measurements on a randomly selected subset of the items comprising the U-235 flows and inventories at the plant. In addition, the MUF-D statistic is used as the test statistic for the detection of diversion. The paper includes descriptions of the potential verification activities, as well as calculations of: (1) attributes and variables sample sizes for the various strata, (2) standard deviations of the relevant test statistics, and (3) the detection sensitivity which the IAEA might achieve by this verification strategy at GCEP.

  12. Production plant separator system conceptual design

    SciTech Connect (OSTI)

    Ng, E.; Kan, T.

    1994-12-31

    A full conceptual design has been completed for a Uranium Atomic Vapor Laser Isotope Separation (U-AVLIS) production plant capable of producing {approximately}1700 metric tons of enriched uranium per year (MTU/y). This plant is the first step in the deployment of AVLIS enrichment technology, which will provide inexpensive, dependable, and environmentally safe uranium enrichment services to utility customers. Previous issues of the ISAM Semiannual Report describe other major systems in the plant, namely the laser, feed and product systems. This article describes the design of the separator system. The separator system is a a key component in the plant. After the feed conversion system converts uranium trioxide (UO{sub 3}) to a uranium-iron alloy, the alloy enters the separator system. In the separator, and intense electron beam vaporizes uranium metal in a vacuum chamber. In the laser system, fixed-frequency copper-vapor lasers pump tunable dye lasers. These precisely tuned dye lasers then selectively excite and ionize uranium-235 atoms in the vapor stream, leaving the uranium-238 atoms untouched. The photo-ions of uranium-235 are then drawn to an electrically biased collector, producing the enriched product stream. The remaining vapor flows through, producing the depleted tails stream. Both product and tails streams are continuously removed from the separator pod as flowing liquid uranium metal. Withdrawal containers are used to collect separately the enriched and depleted uranium. The enriched product will be converted by fuel fabricators to uranium dioxide (UO{sub 2}) and used to fabricate reactor fuel assemblies for utility customers.

  13. Challenges using a {sup 252}Cf shuffler instrument in a plant environment to measure mixtures of uranium and plutonium transuranic waste

    SciTech Connect (OSTI)

    Hurd, J.R.

    1999-08-29

    An active-passive {sup 252}Cf shuffler instrument, installed and certified several years ago at Los Alamos National Laboratory's plutonium facility, has now been calibrated for different matrices to measure Waste Isolation Pilot Plant (WIPP)-destined transuranic (TRU) waste. Little or no data currently exist for these types of measurements in plant environments where sudden large changes in the neutron background radiation can significantly distort the results. Measurements and analyses of twenty-two 55-gallon drums, consisting of mixtures of varying quantities of uranium and plutonium in mostly noncombustible matrices, have been recently completed at the plutonium facility. The calibration and measurement techniques, including the method used to separate out the plutonium component, will be presented and discussed. Calculations used to adjust for differences in uranium enrichment from that of the calibration standards will be shown. Methods used to determine various sources of both random and systematic error will be indicated. Particular attention will be directed to those problems identified as arising from the plant environment. The results of studies to quantify the aforementioned distortion effects in the data will be presented. Various solution scenarios will be outlined, along with those adopted here.

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

    SciTech Connect (OSTI)

    Rasmussen B.

    2010-01-01

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

  15. Gamma/neutron time-correlation for special nuclear material characterization %3CU%2B2013%3E active stimulation of highly enriched uranium.

    SciTech Connect (OSTI)

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

    2013-09-01

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

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

    SciTech Connect (OSTI)

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

    1986-06-01

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

  17. Radiometric Determination of Uranium in Natural Waters after Enrichment and Separation by Cation-Exchange and Liquid-Liquid Extraction

    E-Print Network [OSTI]

    I. Pashalidis; H. Tsertos

    2003-04-28

    The alpha-radiometric determination of uranium after its pre-concentration from natural water samples using the cation-exchange resin Chelex-100, its selective extraction by tributylphosphate and electrodeposition on stainless steel discs is reported. The validity of the separation procedure and the chemical recoveries were checked by addition of uranium standard solution as well as by tracing with U-232. The average uranium yield was determined to be (97 +- 2) % for the cation-exchange, (95 +- 2) % for the liquid-liquid extraction, and more than 99% for the electrodeposition. Employing high-resolution alpha-spectroscopy, the measured activity of the U-238 and U-234 radioisotopes was found to be of similar magnitude; i.e. ~7 mBq/L and ~35 mBq/L for ground- and seawater samples, respectively. The energy resolution (FWHM) of the alpha-peaks was 22 keV, while the Minimum Detectable Activity (MDA) was estimated to be 1 mBq/L (at the 95% confidence limit).

  18. EIS-0089: PUREX Plant and Uranium Oxide Plant Facilities, Hanford Site, Richland, Washington

    Broader source: Energy.gov [DOE]

    The U.S. Department of Energy developed this statement to evaluate the environmental impacts of resumption of operations of the PUREX/Uranium Oxide facilities at the Hanford Site to produce plutonium and other special nuclear materials for national defense needs.

  19. Spectroscopic evidence of uranium immobilization in acidic wetlands by natural organic matter and plant roots

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

    Li, Dien; Kaplan, Daniel I.; Chang, Hyun-Shik; Seaman, John C.; Jaffé, Peter R.; Koster van Groos, Paul; Scheckel, Kirk G.; Segre, Carlo U.; Chen, Ning; Jiang, De-Tong; et al

    2015-03-03

    Biogeochemistry of uranium in wetlands plays important roles in U immobilization in storage ponds of U mining and processing facilities but has not been well understood. The objective of this work was to study molecular mechanisms responsible for high U retention by Savannah River Site (SRS) wetland sediments under varying redox and acidic (pH = 2.6–5.8) conditions using U L?-edge X-ray absorption spectroscopy. Uranium in the SRS wetland sediments existed primarily as U(VI) bonded as a bidentate to carboxylic sites (U–C bond distance at ~2.88 Å), rather than phenolic or other sites of natural organic matter (NOM). In microcosms simulatingmore »the SRS wetland processes, U immobilization on roots was two orders of magnitude higher than on the adjacent brown or more distant white sands in which U was U(VI). Uranium on the roots were both U(IV) and U(VI), which were bonded as a bidentate to carbon, but the U(VI) may also form a U phosphate mineral. After 140 days of air exposure, all U(IV) was re-oxidized to U(VI) but remained as a bidentate bonding to carbon. This study demonstrated NOM and plant roots can highly immobilize U(VI) in the SRS acidic sediments, which has significant implication for the long-term stewardship of U-contaminated wetlands.« less

  20. Results of Active Test of Uranium-Plutonium Co-denitration Facility at Rokkasho Reprocessing Plant

    SciTech Connect (OSTI)

    Numao, Teruhiko; Nakayashiki, Hiroshi; Arai, Nobuyuki; Miura, Susumu; Takahashi, Yoshiharu [Denitration Section, Plant Operation Dept., Reprocessing Plant, Reprocessing Business Division, Japan Nuclear Fuel Limited Rokkasho-mura, Kamikita-gun, Aomori-ken (Japan); Nakamura, Hironobu; Tanaka, Izumi [Technical Support Dept., Reprocessing Plant, Reprocessing Business Division, Japan Nuclear Fuel Limited Rokkasho-mura, Kamikita-gun, Aomori-ken (Japan)

    2007-07-01

    In the U-Pu co-denitration facility at Rokkasho Reprocessing Plant (RRP), Active Test which composes of 5 steps was performed by using uranium-plutonium nitrate solution that was extracted from spent fuels. During Active Test, two kinds of tests were performed in parallel. One was denitration performance test in denitration ovens, and expected results were successfully obtained. The other was validation and calibration of non-destructive assay (NDA) systems, and expected performances were obtained and their effectiveness as material accountancy and safeguards system was validated. (authors)

  1. BIO-MONITORING FOR URANIUM USING STREAM-SIDE TERRESTRIAL PLANTS AND MACROPHYTES

    SciTech Connect (OSTI)

    Caldwell, E.; Duff, M.; Hicks, T.; Coughlin, D.; Hicks, R.; Dixon, E.

    2012-01-12

    This study evaluated the abilities of various plant species to act as bio-monitors for environmental uranium (U) contamination. Vegetation and soil samples were collected from a U processing facility. The water-way fed from facility storm and processing effluents was the focal sample site as it represented a primary U transport mechanism. Soils and sediments from areas exposed to contamination possessed U concentrations that averaged 630 mg U kg{sup -1}. Aquatic mosses proved to be exceptional accumulators of U with dry weight (dw) concentrations measuring as high as 12500 mg U kg{sup -1} (approximately 1% of the dw mass was attributable to U). The macrophytes (Phragmites communis, Scripus fontinalis and Sagittaria latifolia) were also effective accumulators of U. In general, plant roots possessed higher concentrations of U than associated upper portions of plants. For terrestrial plants, the roots of Impatiens capensis had the highest observed levels of U accumulation (1030 mg kg{sup -1}), followed by the roots of Cyperus esculentus and Solidago speciosa. The concentration ratio (CR) characterized dry weight (dw) vegetative U levels relative to that in associated dw soil. The plant species that accumulated U at levels in excess of that found in the soil were: P. communis root (CR, 17.4), I. capensis root (CR, 3.1) and S. fontinalis whole plant (CR, 1.4). Seven of the highest ten CR values were found in the roots. Correlations with concentrations of other metals with U were performed, which revealed that U concentrations in the plant were strongly correlated with nickel (Ni) concentrations (correlation: 0.992; r-squared: 0.984). Uranium in plant tissue was also strongly correlated with strontium (Sr) (correlation: 0.948; r-squared: 0.899). Strontium is chemically and physically similar to calcium (Ca) and magnesium (Mg), which were also positively-correlated with U. The correlation with U and these plant nutrient minerals, including iron (Fe), suggests that active uptake mechanisms may influence plant U accumulation.

  2. Uranium Industry Annual, 1992

    SciTech Connect (OSTI)

    Not Available

    1993-10-28

    The Uranium Industry Annual provides current statistical data on the US uranium industry for the Congress, Federal and State agencies, the uranium and electric utility industries, and the public. The feature article, ``Decommissioning of US Conventional Uranium Production Centers,`` is included. Data on uranium raw materials activities including exploration activities and expenditures, resources and reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities including domestic uranium purchases, commitments by utilities, procurement arrangements, uranium imports under purchase contracts and exports, deliveries to enrichment suppliers, inventories, secondary market activities, utility market requirements, and uranium for sale by domestic suppliers are presented in Chapter 2.

  3. OXYGEN DIFFUSION IN HYPOSTOICHIOMETRIC URANIUM DIOXIDE

    E-Print Network [OSTI]

    Kim, Kee Chul

    2010-01-01

    IN HYPOSTOICHIOMETRIC URANIUM DIOXIDE Kee Chul Kim Ph.D.727-366; Figure 1. Oxygen-uranium phase-equilibrium _ystem [18]. uranium dioxide powders and 18 0 enriched carbon

  4. Literature review: Phytoaccumulation of chromium, uranium, and plutonium in plant systems

    SciTech Connect (OSTI)

    Hossner, L.R.; Loeppert, R.H.; Newton, R.J.; Szaniszlo, P.J.

    1998-05-01

    Phytoremediation is an integrated multidisciplinary approach to the cleanup of contaminated soils, which combines the disciplines of plant physiology, soil chemistry, and soil microbiology. Metal hyperaccumulator plants are attracting increasing attention because of their potential application in decontamination of metal-polluted soils. Traditional engineering technologies may be too expensive for the remediation of most sites. Removal of metals from these soils using accumulator plants is the goal of phytoremediation. The emphasis of this review has been placed on chromium (Cr), plutonium (Pu), and uranium (U). With the exception of Cr, these metals and their decay products exhibit two problems, specifically, radiation dose hazards and their chemical toxicity. The radiation hazard introduces the need for special precautions in reclamation beyond that associated with non-radioactive metals. The uptake of beneficial metals by plants occurs predominantly by way of channels, pores, and transporters in the root plasma membrane. Plants characteristically exhibit a remarkable capacity to absorb what they need and exclude what they don`t need. But most vascular plants absorb toxic and heavy metals through their roots to some extent, though to varying degrees, from negligible to substantial. Sometimes absorption occurs because of the chemical similarity between beneficial and toxic metals. Some plants utilize exclusion mechanisms, where there is a reduced uptake by the roots or a restricted transport of the metal from root to shoot. At the other extreme, hyperaccumulator plants absorb and concentrate metals in both roots and shoots. Some plant species endemic to metalliferous soils accumulate metals in percent concentrations in the leaf dry matter.

  5. Nitrogen control of 13C enrichment in heterotrophic organs relative to leaves in a landscape-building desert plant species

    SciTech Connect (OSTI)

    Zhang, Jinxin [Chinese Academy of Forestry; Gu, Lianhong [ORNL

    2014-01-01

    A longstanding puzzle in isotopic studies of C3 plant species is that heterotrophic plant organs (e.g., stems, roots, seeds, and fruits) tend to be enriched in 13C compared to the autotrophic organ (leaves) that provides them with photosynthate. Our inability to explain this puzzle suggests key deficiencies in understanding post-photosynthetic metabolic processes. It also limits the effectiveness of applications of stable carbon isotope analyses in a variety of scientific disciplines ranging from plant physiology to global carbon cycle studies. To gain insight into this puzzle, we excavated whole plant architectures of Nitraria tangutorum Bobrov, a C3 species that has an exceptional capability of fixing sands and building sand dunes, in two deserts in northwestern China. We systematically and simultaneously measured carbon isotopic ratios and nitrogen and phosphorous concentrations of different parts of the excavated plants. We also determined the seasonal variations in leaf carbon isotopic ratios on nearby intact plants of N. tangutorum. We found that higher nitrogen concentrations in heterotrophic organs were significantly correlated with increased heterotrophic 13C enrichment compared to leaves. However, phosphorous concentrations had no effect on the enrichment. In addition, new leaves had carbon isotopic ratios similar to roots but were progressively depleted in 13C as they matured. We concluded that a nitrogen-mediated process, probably the refixation of respiratory CO2 by phosphoenolpyruvate (PEP) carboxylase, was responsible for the differences in 13C enrichment among different heterotrophic organs while processes within leaves or during phloem loading may contribute to the overall autotrophic heterotrophic difference in carbon isotopic compositions.

  6. History of Uranium-233(sup233U)Processing at the Rocky Flats Plant. In support of the RFETS Acceptable Knowledge Program

    SciTech Connect (OSTI)

    Moment, R.L.; Gibbs, F.E.; Freiboth, C.J.

    1999-04-01

    This report documents the processing of Uranium-233 at the Rocky Flats Plant (Rocky Flats Environmental Technology Site). The information may be used to meet Waste Isolation Pilot Plant (WIPP) Waste Acceptance Criteria (WAC)and for determining potential Uranium-233 content in applicable residue waste streams.

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

    SciTech Connect (OSTI)

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

    1991-07-01

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

  8. Enriching stable isotopes: Alternative use for Urenco technology

    SciTech Connect (OSTI)

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

    1996-12-31

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

  9. In-Born Radio Frequency Identification Devices for Safeguards Use at Gas-Centrifuge Enrichment Plants

    SciTech Connect (OSTI)

    Ward,R.; Rosenthal,M.

    2009-07-12

    Global expansion of nuclear power has made the need for improved safeguards measures at Gas Centrifuge Enrichment Plants (GCEPs) imperative. One technology under consideration for safeguards applications is Radio Frequency Identification Devices (RFIDs). RFIDs have the potential to increase IAEA inspector"s efficiency and effectiveness either by reducing the number of inspection visits necessary or by reducing inspection effort at those visits. This study assesses the use of RFIDs as an integral component of the "Option 4" safeguards approach developed by Bruce Moran, U.S. Nuclear Regulatory Commission (NRC), for a model GCEP [1]. A previous analysis of RFIDs was conducted by Jae Jo, Brookhaven National Laboratory (BNL), which evaluated the effectiveness of an RFID tag applied by the facility operator [2]. This paper presents a similar evaluation carried out in the framework of Jo’s paper, but it is predicated on the assumption that the RFID tag is applied by the manufacturer at the birth of the cylinder, rather than by the operator. Relevant diversion scenarios are examined to determine if RFIDs increase the effectiveness and/ or efficiency of safeguards in these scenarios. Conclusions on the benefits offered to inspectors by using in-born RFID tagging are presented.

  10. Application of systems engineering techniques (reliability, availability, maintainability, and dollars) to the Gas Centrifuge Enrichment Plant

    SciTech Connect (OSTI)

    Boylan, J.G.; DeLozier, R.C.

    1982-01-01

    The systems engineering function for the Gas Centrifuge Enrichment Plant (GCEP) covers system requirements definition, analyses, verification, technical reviews, and other system efforts necessary to assure good balance of performance, safety, cost, and scheduling. The systems engineering function will support the design, installation, start-up, and operational phases of GCEP. The principal objectives of the systems engineering function are to: assure that the system requirements of the GCEP process are adequately specified and documented and that due consideration and emphasis are given to all aspects of the project; provide system analyses of the designs as they progress to assure that system requirements are met and that GCEP interfaces are compatible; assist in the definition of programs for the necessary and sufficient verification of GCEP systems; and integrate reliability, maintainability, logistics, safety, producibility, and other related specialties into a total system effort. This paper addresses the GCEP reliability, availability, maintainability, and dollars (RAM dollars) analyses which are the primary systems engineering tools for the development and implementation of trade-off studies. These studies are basic to reaching cost-effective project decisions. The steps necessary to achieve optimum cost-effective design are shown.

  11. Rotor dynamic analysis of GCEP (Gas Centrifuge Enrichment Plant) Tails Withdrawal Test Facility AC-12 compressor

    SciTech Connect (OSTI)

    Spencer, J.W.

    1982-01-22

    The reliable operation of the centrifugal compressors utilized in the gaseous diffusion process is of great importance due to the critical function of these machines in product and tails withdrawal, cascade purge and evacuation processes, the purge cascade and product booster applications. The same compressors will be used in equally important applications within the Gas Centrifuge Enrichment Plant (GCEP). In response to concern over the excessive vibration exhibited by the AC-12 compressor in the No. 3 position of the GCEP Tails Withdrawal Test Facility, a rotor-bearing dynamic analysis was performed on the compressor. This analysis included the acquisition and reduction of compressor vibration data, characterization and modeling of the rotorbearing system, a computer dynamic study, and recommendations for machine modification. The compressor dynamic analysis was performed for rotor speeds of 9000 rpm and 7200 to 7800 rpm, which includes all possible opreating speeds of the compressor in the GCEP Test Facility. While the analysis was performed on this particular AC-12 compressor, the results should be pertinent to other AC-12 applications as well. Similar diagnostic and analytical techniques can be used to evaluate operation of other types of centrifugal compressors.

  12. The Problem with Continuity of Knowledge in Enrichment Plant Process Monitoring

    SciTech Connect (OSTI)

    Curtis, Michael M.

    2009-08-01

    It has been three years since the new Gas Centrifuge Enrichment Plant (GCEP) Model Safeguards Approach was approved for implementation by the International Atomic Energy Agency’s Department of Safeguards. Among its recommendations are safeguard measures that place greater emphasis on instrumentation in the process area (Cooley 2007). Irrespective of the compelling technologies, an often overlooked impediment to the application of such instrumentation is maintenance of continuity of knowledge on material that has been identified as abnormal. Any instrument purporting to identify problems in the process area should include some means of containing or monitoring that material until International Atomic Energy Agency (IAEA) inspectors can arrive to confirm the discrepancy. If no containment or surveillance is employed in the interim, and no discrepancy or anomaly is subsequently uncovered in storage cylinders, it is unclear what follow-up action inspectors can take. Some CoK measures have been proposed, but they usually involve an array of cameras or host-applied seals—options that may require a backup system of their own.

  13. Y-12 Uranium Exposure Study

    SciTech Connect (OSTI)

    Eckerman, K.F.; Kerr, G.D.

    1999-08-05

    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.

  14. Material protection, control and accounting cooperation at the Urals Electrochemical Integrated Plant (UEIP), Novouralsk, Russia

    SciTech Connect (OSTI)

    McAllister, S., LLNL

    1998-07-15

    The Urals Electrochemical Integrated Plant is one of the Russian Ministry of Atomic Energy`s nuclear material production sites participating in the US Department of Energy`s Material Protection, Control and Accounting (MPC&A) Program. The Urals Electrochemical Integrated Plant is Russia`s largest uranium enrichment facility and blends tons of high-enriched uranium into low enriched uranium each year as part of the US high-enriched uranium purchase. The Electrochemical Integrated Plant and six participating national laboratories are cooperating to implement a series of enhancements to the nuclear material protection, control, and accountability systems at the site This paper outlines the overall objectives of the MPC&A program at Urals Electrochemical Integrated Plant and the work completed as of the date of the presentation.

  15. Criticality safety concerns of uranium deposits in cascade equipment

    SciTech Connect (OSTI)

    Plaster, M.J. [Lockheed Martin Utility Services, Inc., Piketon, OH (United States)

    1996-12-31

    The Paducah and Portsmouth Gaseous Diffusion Plants enrich uranium in the {sup 235}U isotope by diffusing gaseous uranium hexafluoride (UF{sub 6}) through a porous barrier. The UF{sub 6} gaseous diffusion cascade utilized several thousand {open_quotes}stages{close_quotes} of barrier to produce highly enriched uranium (HEU). Historically, Portsmouth has enriched the Paducah Gaseous Diffusion Plant`s product (typically 1.8 wt% {sup 235}U) as well as natural enrichment feed stock up to 97 wt%. Due to the chemical reactivity of UF{sub 6}, particularly with water, the formation of solid uranium deposits occur at a gaseous diffusion plant. Much of the equipment operates below atmospheric pressure, and deposits are formed when atmospheric air enters the cascade. Deposits may also be formed from UF{sub 6} reactions with oil, UF{sub 6} reactions with the metallic surfaces of equipment, and desublimation of UF{sub 6}. The major deposits form as a result of moist air in leakage due to failure of compressor casing flanges, blow-off plates, seals, expansion joint convolutions, and instrument lines. This report describes criticality concerns and deposit disposition.

  16. Technology, Safety and Costs of Decommissioning a Reference Uranium Hexafluoride Conversion Plant

    SciTech Connect (OSTI)

    Elder, H. K.

    1981-10-01

    Safety and cost information is developed for the conceptual decommissioning of a commercial uranium hexafluoride conversion (UF{sub 6}) plant. Two basic decommissioning alternatives are studied to obtain comparisons between cost and safety impacts: DECON, and passive SAFSTOR. A third alternative, DECON of the plant and equipment with stabilization and long-term care of lagoon wastes. is also examined. DECON includes the immediate removal (following plant shutdown) of all radioactivity in excess of unrestricted release levels, with subsequent release of the site for public use. Passive SAFSTOR requires decontamination, preparation, maintenance, and surveillance for a period of time after shutdown, followed by deferred decontamination and unrestricted release. DECON with stabilization and long-term care of lagoon wastes (process wastes generated at the reference plant and stored onsite during plant operation} is also considered as a decommissioning method, although its acceptability has not yet been determined by the NRC. The decommissioning methods assumed for use in each decommissioning alternative are based on state-of-the-art technology. The elapsed time following plant shutdown required to perform the decommissioning work in each alternative is estimated to be: for DECON, 8 months; for passive SAFSTOR, 3 months to prepare the plant for safe storage and 8 months to accomplish deferred decontamination. Planning and preparation for decommissioning prior to plant shutdown is estimated to require about 6 months for either DECON or passive SAFSTOR. Planning and preparation prior to starting deferred decontamination is estimated to require an additional 6 months. OECON with lagoon waste stabilization is estimated to take 6 months for planning and about 8 months to perform the decommissioning work. Decommissioning cost, in 1981 dollars, is estimated to be $5.91 million for OECON. For passive SAFSTOR, preparing the facility for safe storage is estimated to cost $0.88 million, the annual maintenance and surveillance cost is estimated to be about $0.095 million, and deferred decontamination is estimated to cost about $6.50 million. Therefore, passive SAFSTOR for 10 years is estimated to cost $8.33 million in nondiscounted 1981 dollars. DECON with lagoon waste stabilization is estimated to cost about $4.59 million, with an annual cost of $0.011 million for long-term care. All of these estimates include a 25% contingency. Waste management costs for DECON, including the net cost of disposal of the solvent extraction lagoon wastes by shipping those wastes to a uranium mill for recovery of residual uranium, comprise about 38% of the total decommissioning cost. Disposal of lagoon waste at a commercial low-level waste burial ground is estimated to add $10.01 million to decommissioning costs. Safety analyses indicate that radiological and nonradiological safety impacts from decommissioning activities should be small. The 50-year committed dose equivalent to members of the public from airborne releases during normal decommissioning activities is estimated to 'Je about 4.0 man-rem. Radiation doses to the public from accidents are found to be very low for all phases of decommissioning. Occupational radiation doses from normal decommissioning operations (excluding transport operations) are estimated to be about 79 man-rem for DECON and about 80 man-rem for passive SAFSTOR with 10 years of safe storage. Doses from DECON with lagoon waste stabilization are about the same as for DECON except there is less dose resulting from transportation of radioactive waste. The number of fatalities and serious lost-time injuries not related to radiation is found to be very small for all decommissioning alternatives. Comparison of the cost estimates shows that DECON with lagoon waste stabilization is the least expensive method. However, this alternative does not allow unrestricted release of the site. The cumulative cost of maintenance and surveillance and the higher cost of deferred decontamination makes passive SAFSTOR more expensive than DECON. Seve

  17. Approach to IAEA verification of the nuclear-material balance at the Portsmouth Gas Centrifuge Enrichment Plant (GCEP)

    SciTech Connect (OSTI)

    Gordon, D.M.; Sanborn, J.B.; Younkin, J.M.; DeVito, V.J.

    1982-01-01

    This paper describes a potential approach by which the International Atomic Energy Agency (IAEA) might verify the nuclear-material balance at the Portsmouth Gas Centrifuge Enrichment Plant (GCEP), should that plant be placed under IAEA safeguards. The strategy makes use of the attributes and variables measurement verification approach, whereby the IAEA would perform independent measurements on a randomly selected subset of the items comprising the U-235 flows and inventories at the plant. In addition, the MUF-D statistic is used as the test statistics for the detection of diversion. The paper includes descriptions of the potential verification activities, as well as calculations of (a) attributes and variables sample sizes for the various strata, (b) standard deviations of the relevant test statistics, and (c) the sensitivity for detection of diversion which the IAEA might achieve by this verification strategy at GCEP.

  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

    2012-09-01

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

  19. US-Russian collaboration for enhancing nuclear materials protection, control, and accounting at the Elektrostal uranium fuel-fabrication plant

    SciTech Connect (OSTI)

    Smith, H. [Los Alamos National Lab., NM (United States); Allentuck, J. [Brookhaven National Lab., Upton, NY (United States); Barham, M. [Oak Ridge National Lab., TN (United States); Bishop, M. [Sandia National Labs., Albuquerque, NM (United States); Wentz, D. [Lawrence Livermore National Lab., CA (United States); Steele, B.; Bricker, K. [Pacific Northwest National Lab., Richland, WA (United States); Cherry, R. [USDOE, Washington, DC (United States); Snegosky, T. [Dept. of Defense, Washington, DC (United States). Defense Nuclear Agency

    1996-09-01

    In September 1993, an implementing agreement was signed that authorized collaborative projects to enhance Russian national materials control and accounting, physical protection, and regulatory activities, with US assistance funded by the Nunn-Lugar Act. At the first US-Russian technical working group meeting in Moscow in February 1994, it was decided to identify a model facility where materials protection, control, and accounting (MPC and A) and regulatory projects could be carried out using proven technologies and approaches. The low-enriched uranium (LEU or RBMK and VVER) fuel-fabrication process at Elektrostal was selected, and collaborative work began in June 1994. Based on many factors, including initial successes at Elektrostal, the Russians expanded the cooperation by proposing five additional sites for MPC and A development: the Elektrostal medium-enriched uranium (MEU or BN) fuel-fabrication process and additional facilities at Podolsk, Dmitrovgrad, Obninsk, and Mayak. Since that time, multilaboratory teams have been formed to develop and implement MPC and A upgrades at the additional sites, and much new work is underway. This paper summarizes the current status of MPC and A enhancement projects in the LEU fuel-fabrication process and discusses the status of work that addresses similar enhancements in the MEU (BN) fuel processes at Elektrostal, under the recently expanded US-Russian MPC and A cooperation.

  20. Application of Condition-Based Monitoring Techniques for Remote Monitoring of a Simulated Gas Centrifuge Enrichment Plant

    SciTech Connect (OSTI)

    Hooper, David A; Henkel, James J; Whitaker, Michael

    2012-01-01

    This paper presents research into the adaptation of monitoring techniques from maintainability and reliability (M&R) engineering for remote unattended monitoring of gas centrifuge enrichment plants (GCEPs) for international safeguards. Two categories of techniques are discussed: the sequential probability ratio test (SPRT) for diagnostic monitoring, and sequential Monte Carlo (SMC or, more commonly, particle filtering ) for prognostic monitoring. Development and testing of the application of condition-based monitoring (CBM) techniques was performed on the Oak Ridge Mock Feed and Withdrawal (F&W) facility as a proof of principle. CBM techniques have been extensively developed for M&R assessment of physical processes, such as manufacturing and power plants. These techniques are normally used to locate and diagnose the effects of mechanical degradation of equipment to aid in planning of maintenance and repair cycles. In a safeguards environment, however, the goal is not to identify mechanical deterioration, but to detect and diagnose (and potentially predict) attempts to circumvent normal, declared facility operations, such as through protracted diversion of enriched material. The CBM techniques are first explained from the traditional perspective of maintenance and reliability engineering. The adaptation of CBM techniques to inspector monitoring is then discussed, focusing on the unique challenges of decision-based effects rather than equipment degradation effects. These techniques are then applied to the Oak Ridge Mock F&W facility a water-based physical simulation of a material feed and withdrawal process used at enrichment plants that is used to develop and test online monitoring techniques for fully information-driven safeguards of GCEPs. Advantages and limitations of the CBM approach to online monitoring are discussed, as well as the potential challenges of adapting CBM concepts to safeguards applications.

  1. ORNL/TM-2009/110 Profile of World Uranium

    E-Print Network [OSTI]

    Pennycook, Steve

    ORNL/TM-2009/110 Profile of World Uranium Enrichment Programs--2009 April 2009 Prepared by M. D PROFILE OF WORLD URANIUM ENRICHMENT PROGRAMS--2009 M. D. Laughter Date Published: April 2009 This work

  2. Species richness in a successional grassland: effects of nitrogen enrichment and plant litter

    E-Print Network [OSTI]

    Foster, Bryan L.; Gross, Katherine L.

    1998-12-01

    ) that was experimentally introduced to the study site as seed. Additions of nitrogen fertilizer to experimental plots over two growing seasons increased plant production as indicated by increases in both living plant biomass and litter biomass. Increased productivity...

  3. DOE uranium enrichment program. Hearing before the Subcommittee on Energy Conservation and Power of the Committee on Energy and Commerce, House of Representatives, Ninety-Ninth Congress, Second Session, February 19, 1986

    SciTech Connect (OSTI)

    Not Available

    1986-01-01

    Six witnesses representing the Congressional Budget Office, DOE, the National Taxpayers Union, the Edison Electric Institute, and the General Accounting Office testified on the implications of the uranium enrichment program to the national deficit. The program's $1.6 billion annual budget makes it DOE's largest program, but its benefits are aimed more at the nuclear utility customers at the expense of taxpayers. The administration's proposal to write off unrecovered costs will further hurt taxpayers, and is counter to its philosophy of allowing market forces to operate. The hearing addressed DOE's proposals for improving the economics of the program. Additional material for the record follows the testimony.

  4. Highly Enriched Uranium Materials Facility

    National Nuclear Security Administration (NNSA)

    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 Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of NaturalDukeWakefield Municipal GasAdministration Medal01 Sandia4) AugustA. -71- Particulate: Columns 59 and R e s0/%2A

  5. Nuclear criticality safety controls for uranium deposits during D and D at the Oak Ridge Gaseous Diffusion Plant

    SciTech Connect (OSTI)

    Haire, M.J.; Jordan, W.C. [Oak Ridge National Lab., TN (United States); Jollay, L.J. III; Dahl, T.L. [Oak Ridge K-25 Site, TN (United States)

    1997-02-01

    The US Department of Energy (DOE) Deputy Assistant Secretary of Energy for Environmental Management has issued a challenge to complete DOE environmental cleanup within a decade. The response for Oak Ridge facilities is in accordance with the DOE ten-year plan which calls for completion of > 95% of environmental management work by the year 2006. This will result in a 99% risk reduction and in a significant savings in base line costs in waste management (legacy waste); remedial action (groundwater, soil, etc.); and decontamination and decommissioning (D and D). It is assumed that there will be long-term institutional control of cascade equipment, i.e., there will be no walk away from sites, and that there will be firm radioactivity release limits by 1999 for recycle metals. An integral part of these plants is the removal of uranium deposits which pose nuclear criticality safety concerns in the shut down of the Oak Ridge Gaseous Diffusion Plant. DOE has initiated the Nuclear Criticality Stabilization Program to improve nuclear criticality safety by removing the larger uranium deposits from unfavorable geometry equipment. Nondestructive assay (NDA) measurements have identified the location of these deposits. The objective of the K-25 Site Nuclear Criticality Stabilization Program is to remove and place uranium deposits into safe geometry storage containers to meet the double contingency principle. Each step of the removal process results in safer conditions where multiple controls are present. Upon completion of the Program, nuclear criticality risks will be greatly reduced.

  6. statistical physics canonical ensemble Uranium Centrifuges

    E-Print Network [OSTI]

    statistical physics canonical ensemble Uranium Centrifuges The easiest type of nuclear weapon of the physics behind crude uranium enrichment methods. 2 The centrifuge concept is a very generic way of trying the uranium, we remove gas from the ends of the centrifuge, where the heavier uranium atoms are more

  7. Generic report on health effects for the US Gaseous Diffusion Plants. Sect. 8, Pt. 1

    SciTech Connect (OSTI)

    Just, R.A.; Emler, V.S.

    1984-06-01

    Toxic substances present in uranium enrichment plants include uranium hexafluoride (UF/sub 6/), hydrogen fluoride (HF), uranyl fluoride (UO/sub 2/F/sub 2/), chlorine (Cl/sub 2/), chlorine trifluoride (ClF/sub 3/), fluorine (F/sub 2/), uranium tetrafluoride (UF/sub 4/), and technetium (Tc). The current knowledge of the expected health effects of acute exposures to these substances is described. 10 references, 2 figures, 6 tables. (ACR)

  8. Analysis of HEU samples from the ULBA Metallurgical Plant

    SciTech Connect (OSTI)

    Gift, E.H.

    1995-05-01

    In early March 1994, eight highly enriched uranium (HEU) samples were collected from materials stored at the Ulba Metallurgical Plant in Oskamen (Ust Kamenogorsk), Kazakhstan. While at the plant site, portions of four samples were dissolved and analyzed by mass spectrograph at the Ulba analytical laboratory by Ulba analysts. Three of these mass spectrograph solutions and the eight HEU samples were subsequently delivered to the Y-12 Plant for complete chemical and isotopic analyses. Chemical forms of the eight samples were uranium metal chips, U0{sub 2} powder, uranium/beryllium oxide powder, and uranium/beryllium alloy rods. All were declared by the Ulba plant to have a uranium assay of {approximately}90 wt % {sup 235}U. The uranium/beryllium powder and alloy samples were also declared to range from about 8 to 28 wt % uranium. The chemical and uranium isotopic analyses done at the Y-12 Plant confirm the Ulba plant declarations. All samples appear to have been enriched using some reprocessed uranium, probably from recovery of uranium from plutonium production reactors. As a result, all samples contain some {sup 236}U and {sup 232}U and have small but measurable quantities of plutonium. This plutonium could be the result of either contamination carried over from the enrichment process or cross-contamination from weapons material. It is not the result of direct reactor exposure. Neither the {sup 232}U nor the plutonium concentrations are sufficiently high to provide a significant industrial health hazard. Both are well within established or proposed acceptance criteria for storage at Y-12. The trace metal analyses showed that, with the exception of beryllium, there are no trace metals in any of these HEU samples that pose a significant health hazard.

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

    SciTech Connect (OSTI)

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

    2012-06-01

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

  10. An assessment of plant biointrusion at the Uranium Mill Tailings Remedial Action Project rock-covered disposal cells

    SciTech Connect (OSTI)

    Not Available

    1990-10-01

    This study is one of a number of special studies that have been conducted regarding various aspects of the Uranium Mill Tailings Remedial Action (UMTRA) Project. This special study was proposed following routine surveillance and maintenance surveys and observations reported in a special study of vegetative covers (DOE, 1988), in which plants were observed growing up through the rock erosion layer at recently completed disposal cells. Some of the plants observed were deep-rooted woody species, and questions concerning root intrusion into disposal cells and the need to control plant growth were raised. The special study discussed in this report was designed to address some of the ramifications of plant growth on disposal cells that have rock covers. The NRC has chosen rock covers over vegetative covers in the arid western United States because licenses cannot substantiate that the vegetative covers will be significantly greater than 30 percent and preferably 70 percent,'' which is the amount of vegetation required to reduce flow to a point of stability.'' The potential impacts of vegetation growing in rock covers are not addressed by the NRC (1990). The objectives, then, of this study were to determine the species of plants growing on two rock-covered disposal cells, study the rooting pattern of plants on these cells, and identify possible impacts of plant root penetration on these and other UMTRA Project rock-covered cells.

  11. Nuclear non-proliferation regime effectiveness : an integrated methodology for analyzing highly enriched uranium production scenarios at gas centrifuge enrichment plants

    E-Print Network [OSTI]

    Kwak, Taeshin (Taeshin S.)

    2010-01-01

    The dramatic change in the international security environment after the collapse of the bipolar system has had a negative impact on the effectiveness of the existing nuclear non-proliferation regime. Furthermore, the success ...

  12. Enrichment Assay Methods for a UF6 Cylinder Verification Station

    SciTech Connect (OSTI)

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

    2010-11-30

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

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

    SciTech Connect (OSTI)

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

    2010-04-01

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

  14. Recent ORNL experience in site performance prediction: the Gas Centrifuge Enrichment Plant and the Oak Ridge Central Waste Disposal Facility

    SciTech Connect (OSTI)

    Pin, F.G.

    1985-01-01

    The suitability of the Portsmouth Gas Centrifuge Enrichment Plant Landfill and the Oak Ridge, Tennessee, Central Waste Disposal Facility for disposal of low-level radioactive waste was evaluated using pathways analyses. For these evaluations, a conservative approach was selected; that is, conservatism was built into the analyses when assumptions concerning future events had to be made or when uncertainties concerning site or waste characteristics existed. Data from comprehensive laboratory and field investigations were used in developing the conceptual and numerical models that served as the basis for the numerical simulations of the long-term transport of contamination to man. However, the analyses relied on conservative scenarios to describe the generation and migration of contamination and the potential human exposure to the waste. Maximum potential doses to man were calculated and compared to the appropriate standards. Even under this conservative framework, the sites were found to provide adequate buffer to persons outside the DOE reservations and conclusions concerning site capacity and site acceptability were drawn. Our experience through these studies has shown that in reaching conclusions in such studies, some consideration must be given to the uncertainties and conservatisms involved in the analyses. Analytical methods to quantitatively assess the probability of future events to occur and to quantitatively determine the sensitivity of the results to data uncertainty may prove useful in relaxing some of the conservatism built into the analyses. The applicability of such methods to pathways analyses is briefly discussed.

  15. 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; Shibahara, Yuji; Morimoto, Yasuyuki; Tokuyasu, Takashi; Takahashi, Nobuo; Tanaka, Yoshio; Sugitsue, Noritake

    2012-07-01

    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)

  16. Design considerations for instrumentation to monitor the enrichment of gaseous UF{sub 6}

    SciTech Connect (OSTI)

    Close, D.A.

    1999-07-01

    The measurement of the enrichment of gaseous UF{sub 6} presents unique measurement problems. The well-known uranium enrichment meter is not applicable. For solid samples of uranium, including metal, and oxide and fluoride compounds, the infinite thickness is {approximately}1 cm. Gaseous UF{sub 6}, at a pressure of tens of Torr, has an infinite thickness on the order of 350 m. This is a physically and operationally unrealistic situation for an operating facility. Pipe dimensions and composition also strongly influence the applicable measurement technique. Fundamentally, the definition of enrichment is the ratio of {sup 235}U to total uranium. The amount of {sup 235}U is determined by measuring the intensity of the 185.7-keV gamma ray from the decay of {sup 235}U. There are two methods that have been implemented to determine the amount of total uranium in the gas: X-ray fluorescence (XRF) and gamma-ray transmission. The technique used to measure the amount of total uranium is dependent on the pressure of the gas in the header pipe. The transmission measurement is applicable for higher pressures, generally pressures {lt}40 Torr. The XRF measurement can be used for pressures greater than a few Torr. An XRF measurement at pressures lower than a few Torr becomes very difficult. Two other constraints strongly influence the implementation of the measurement technique--pipe diameter and material composition. These two techniques have been implemented. The XRF technique is an approved measurement by the International Atomic Energy Agency (IAEA) for inspections at gaseous centrifuge facilities. The XRF technique has also been implemented at the Portsmouth Gaseous Diffusion Plant for the IAEA verification experiment during the period December 1997 to October 1998 to verify the downblending of US highly enriched uranium (HEU) to low-enriched uranium (LEU). The transmission technique was originally developed to verify the downblending of Russian HEU to LEU. This instrument was demonstrated at the Paducah Gaseous Diffusion Plant from April 1998 to July 1998 and installed at the Urals Electrochemical Integrated Plant, Novouralsk, Russia, during January 1999.

  17. Effect of Cd-Enriched Sewage Sludge on Plant Growth, Nutrients and Heavy Metals Concentrations in the Soil–Plant System

    E-Print Network [OSTI]

    Rusan, Munir Mohammad; Athamneh, Bayan Mahmoud

    2009-01-01

    utilization of sewage sludge. A twenty-year study atoriginal and Cd-enriched sewage sludge (SS) and diammoniumtoxic elements in sewage sludge as affected by soil organic

  18. Depleted Uranium Hexafluoride (DUF6) Fully Operational at the...

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

    Jack Zimmerman, DUF6 at the PortsmouthPaducah Project Office. DUF6 is depleted uranium hexafluoride, a byproduct of uranium enrichment that has taken place at U.S. gaseous...

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

    SciTech Connect (OSTI)

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

    2012-07-31

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

  20. Uranium and other heavy metals in the plant-animal-human food chain near abandoned mining sites and structures in an American Indian community in northwestern New Mexico

    E-Print Network [OSTI]

    Samuel-Nakamura, Christine

    2013-01-01

    environment of uranium deposits of the Grants region, Newan area with large uranium deposits. The area was initiallyepigenetic sandstone uranium ore deposits, the predominant

  1. Uranium and other heavy metals in the plant-animal-human food chain near abandoned mining sites and structures in an American Indian community in northwestern New Mexico

    E-Print Network [OSTI]

    Samuel-Nakamura, Christine

    2013-01-01

    of risk maps to minimize uranium exposures in the NavajoThe Navajo people and uranium mining. Albuquerque, NM:toxicity of natural uranium: A review. Reviews on

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

    SciTech Connect (OSTI)

    Not Available

    1991-09-01

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

  3. Material protection control and accounting program activities at the electrochemical plant

    SciTech Connect (OSTI)

    McAllister, S.

    1997-11-14

    The Electrochemical Plant (ECP) is the one of the Russian Federation`s four uranium enrichment plants and one of three sites in Russia blending high enriched uranium (HEU) into commercial grade low enriched uranium. ECP is located approximately 200 km east of Krasnoyarsk in the closed city of Zelenogorsk (formerly Krasnoyarsk- 45). DOE`s MPC&A program first met with ECP in September of 1996. The six national laboratories participating in DOE`s Material Protection Control and Accounting program are cooperating with ECP to enhance the capabilities of the physical protection, access control, and nuclear material control and accounting systems. The MPC&A work at ECP is expected to be completed during fiscal year 2001.

  4. Material protection control and accounting program activities at the Urals electrochemical integrated plant

    SciTech Connect (OSTI)

    McAllister, S.

    1997-11-14

    The Urals Electrochemical Integrated Plant (UEIP) is the Russian Federation`s largest uranium enrichment plant and one of three sites in Russia blending high enriched uranium (HEU) into commercial grade low enriched uranium. UEIP is located approximately 70 km north of Yekaterinburg in the closed city of Novouralsk (formerly Sverdlovsk- 44). DOE`s MPC&A program first met with UEIP in June of 1996, however because of some contractual issues the work did not start until September of 1997. The six national laboratories participating in DOE`s Material Protection Control and Accounting program are cooperating with UEIP to enhance the capabilities of the physical protection, access control, and nuclear material control and accounting systems. The MPC&A work at UEIP is expected to be completed during fiscal year 2001.

  5. Defining the needs for non-destructive assay of UF6 feed, product, and tails at gas centrifuge enrichment plants and possible next steps

    SciTech Connect (OSTI)

    Boyer, Brian D; Swinhoe, Martyn T; Moran, Bruce W; Lebrun, Alain

    2009-01-01

    Current safeguards approaches used by the IAEA at gas centrifuge enrichment plants (GCEPs) need enhancement in order to detect undeclared LEU production with adequate detection probability using non destructive assay (NDA) techniques. At present inspectors use attended systems, systems needing the presence of an inspector for operation, during inspections to verify the mass and {sup 235}U enrichment of UF{sub 6} bulk material used in the process of enrichment at GCEPS. The inspectors also take destructive assay (DA) samples for analysis off-site which provide accurate, on the order of 0.1 % to 0.5% uncertainty, data on the enrichment of the UF{sub 6} feed, tails, and product. However, DA sample taking is a much more labor intensive and resource intensive exercise for the operator and inspector. Furthermore, the operator must ship the samples off-site to the IAEA laboratory which delays the timeliness of the results and contains the possibility of the loss of the continuity of knowledge of the samples during the storage and transit of the material. Use of the IAEA's inspection sampling algorithm shows that while total sample size is fixed by the total population of potential samples and its intrinsic qualities, the split of the samples into NDA or DA samples is determined by the uncertainties in the NDA measurements. Therefore, the larger the uncertainties in the NDA methods, more of the sample taken must be DA samples. Since the DA sampling is arduous and costly, improvements in NDA methods would reduce the number of DA samples needed. Furthermore, if methods of on-site analysis of the samples could be developed that have uncertainties in the 1-2% range, a lot of the problems inherent in DA sampling could be removed. The use of an unattended system that could give an overview of the entire process giving complementary data on the enrichment process as well as accurate measures of enrichment and weights of the UF{sub 6} feed, tails, and product would be a major step in enhancing the ability of NDA beyond present attended systems. The possibility of monitoring the feed, tails, and product header pipes in such a way as to gain safeguards relevant flow and enrichment information without compromising the intellectual property of the operator including proprietary equipment and operational parameters would be a huge step forward. This paper contains an analysis of possible improvements in unattended and attended NDA systems including such process monitoring and possible on-site analysis of DA samples that could reduce the uncertainty of the inspector measurements reducing the difference between the operator's and inspector's measurements providing more effective and efficient IAEA GeEPs safeguards.

  6. Effect of Cd-Enriched Sewage Sludge on Plant Growth, Nutrients and Heavy Metals Concentrations in the Soil–Plant System

    E-Print Network [OSTI]

    Rusan, Munir Mohammad; Athamneh, Bayan Mahmoud

    2009-01-01

    from waste-activated sludge, Soil Sci. Soc. Am. J, 60:505-utilization of sewage sludge. A twenty-year study atCd-enriched sewage sludge (SS) and diammonium phosphate (

  7. FEMO, A FLOW AND ENRICHMENT MONITOR FOR VERIFYING COMPLIANCE WITH INTERNATIONAL SAFEGUARDS REQUIREMENTS AT A GAS CENTRIFUGE ENRICHMENT FACILITY

    SciTech Connect (OSTI)

    Gunning, John E; Laughter, Mark D; March-Leuba, Jose A

    2008-01-01

    A number of countries have received construction licenses or are contemplating the construction of large-capacity gas centrifuge enrichment plants (GCEPs). The capability to independently verify nuclear material flows is a key component of international safeguards approaches, and the IAEA does not currently have an approved method to continuously monitor the mass flow of 235U in uranium hexafluoride (UF6) gas streams. Oak Ridge National Laboratory is investigating the development of a flow and enrichment monitor, or FEMO, based on an existing blend-down monitoring system (BDMS). The BDMS was designed to continuously monitor both 235U mass flow and enrichment of UF6 streams at the low pressures similar to those which exists at GCEPs. BDMSs have been installed at three sites-the first unit has operated successfully in an unattended environment for approximately 10 years. To be acceptable to GCEP operators, it is essential that the instrument be installed and maintained without interrupting operations. A means to continuously verify flow as is proposed by FEMO will likely be needed to monitor safeguards at large-capacity plants. This will enable the safeguards effectiveness that currently exists at smaller plants to be maintained at the larger facilities and also has the potential to reduce labor costs associated with inspections at current and future plants. This paper describes the FEMO design requirements, operating capabilities, and development work required before field demonstration.

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

    SciTech Connect (OSTI)

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

    2009-10-22

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

  9. Criticality safety aspects of K-25 Building uranium deposit removal

    SciTech Connect (OSTI)

    Haire, M.J.; Jordan, W.C. [Oak Ridge National Lab., TN (United States); Ingram, J.C. III; Stinnet, E.C. Jr. [Oak Ridge K-25 Site, TN (United States)

    1995-12-31

    The K-25 Building of the Oak Ridge Gaseous Diffusion Plant (now the K-25 Site) went into operation during World War II as the first large scale production plant to separate {sup 235}U from uranium by the gaseous diffusion process. It operated successfully until 1964, when it was placed in a stand-by mode. The Department of Energy has initiated a decontamination and decommissioning program. The primary objective of the Deposit Removal (DR) Project is to improve the nuclear criticality safety of the K-25 Building by removing enriched uranium deposits from unfavorable-geometry process equipment to below minimum critical mass. The method utilized to accomplish this are detailed in this report.

  10. Validation of MCNP4a for highly enriched uranium using the Battelle process safety and risk management IBM RS/6000 workstation

    SciTech Connect (OSTI)

    Negron, S.B.; Lee, B.L. Jr.; Tayloe, R.W. Jr.

    1996-01-01

    This document has been prepared to allow use of the Radiation Shielding and Information Center (RSIC) release of MCNP4a, which has been installed on the Battelle Process Safety and Risk Management (PSRM) IBM RS/6000 workstation, for production calculations for the Portsmouth Gaseous Diffusion Plant (PORTS). This hardware/software configuration is under the configuration control plan listed in Reference 1. The first portion of this document outlines basic information with regard to validation of MCNP4a using the supplied cross sections and the additional MCNPDAT cross sections. A basic discussion of MCNP is provided, along with discussions of the validation database in general. A description of the statistical analysis then follows. The results of this validation indicate that the software and data libraries examined may be used with confidence for criticality calculations at the Portsmouth Gaseous Diffusion Plant (PORTS). When the validation results are treated as a single group, there is a 95% confidence that 99.9% of future calculations of similar critical systems will have a calculated k{sub eff} > 0.95. Based on this result, the Battelle PSRM Nuclear Safety Group has adopted the calculational acceptance criteria that a calculated k{sub eff} + 2{sigma}, {le} 0.95 is safely subcritical. The conclusion of this document is that MCNP4a and all associated cross section libraries installed on the PSRM IBM RS/6000 are acceptable for use in performing production criticality safety calculations for the Portsmouth Gaseous Diffusion Plant.

  11. USE OF MAILBOX APPROACH, VIDEO SURVEILLANCE, AND SHORT-NOTICE RANDOM INSPECTIONS TO ENHANCE DETECTION OF UNDECLARED LEU PRODUCTION AT GAS CENTRIFUGE ENRICHMENT PLANTS.

    SciTech Connect (OSTI)

    BOYER, B.D.; GORDON, D.M.; JO, J.

    2006-07-16

    Current safeguards approaches used by the IAEA at gas centrifuge enrichment plants (GCEPs) need enhancement in order to detect undeclared LEU production with adequate detection probability. ''Mailbox'' declarations have been used in the last two decades to verify receipts, production, and shipments at some bulk-handling facilities (e.g., fuel-fabrication plants). The operator declares the status of his plant to the IAEA on a daily basis using a secure ''Mailbox'' system such as a secure tamper-resistant computer. The operator agrees to hold receipts and shipments for a specified period of time, along with a specified number of annual inspections, to enable inspector access to a statistically large enough population of UF{sub 6} cylinders and fuel assemblies to achieve the desired detection probability. The inspectors can access the ''Mailbox'' during randomly timed inspections and then verify the operator's declarations for that day. Previously, this type of inspection regime was considered mainly for verifying the material balance at fuel-fabrication, enrichment, and conversion plants. Brookhaven National Laboratory has expanded the ''Mailbox'' concept with short-notice random inspections (SNRIs), coupled with enhanced video surveillance, to include declaration and verification of UF{sub 6} cylinder operational data to detect activities associated with undeclared LEU production at GCEPs. Since the ''Mailbox'' declarations would also include data relevant to material-balance verification, these randomized inspections would replace the scheduled monthly interim inspections for material-balance purposes; in addition, the inspectors could simultaneously perform the required number of Limited-Frequency Unannounced Access (LFUA) inspections used for HEU detection. This approach would provide improved detection capabilities for a wider range of diversion activities with not much more inspection effort than at present.

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

    SciTech Connect (OSTI)

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

    2012-05-02

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

  13. 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; Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Permana, Sidik; Aziz, Ferhat; Sekimoto, Hiroshi

    2014-02-12

    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.

  14. From the Lab to the real world : sources of error in UF {sub 6} gas enrichment monitoring

    SciTech Connect (OSTI)

    Lombardi, Marcie L.

    2012-03-01

    Safeguarding uranium enrichment facilities is a serious concern for the International Atomic Energy Agency (IAEA). Safeguards methods have changed over the years, most recently switching to an improved safeguards model that calls for new technologies to help keep up with the increasing size and complexity of today’s gas centrifuge enrichment plants (GCEPs). One of the primary goals of the IAEA is to detect the production of uranium at levels greater than those an enrichment facility may have declared. In order to accomplish this goal, new enrichment monitors need to be as accurate as possible. This dissertation will look at the Advanced Enrichment Monitor (AEM), a new enrichment monitor designed at Los Alamos National Laboratory. Specifically explored are various factors that could potentially contribute to errors in a final enrichment determination delivered by the AEM. There are many factors that can cause errors in the determination of uranium hexafluoride (UF{sub 6}) gas enrichment, especially during the period when the enrichment is being measured in an operating GCEP. To measure enrichment using the AEM, a passive 186-keV (kiloelectronvolt) measurement is used to determine the {sup 235}U content in the gas, and a transmission measurement or a gas pressure reading is used to determine the total uranium content. A transmission spectrum is generated using an x-ray tube and a “notch” filter. In this dissertation, changes that could occur in the detection efficiency and the transmission errors that could result from variations in pipe-wall thickness will be explored. Additional factors that could contribute to errors in enrichment measurement will also be examined, including changes in the gas pressure, ambient and UF{sub 6} temperature, instrumental errors, and the effects of uranium deposits on the inside of the pipe walls will be considered. The sensitivity of the enrichment calculation to these various parameters will then be evaluated. Previously, UF{sub 6} gas enrichment monitors have required empty pipe measurements to accurately determine the pipe attenuation (the pipe attenuation is typically much larger than the attenuation in the gas). This dissertation reports on a method for determining the thickness of a pipe in a GCEP when obtaining an empty pipe measurement may not be feasible. This dissertation studies each of the components that may add to the final error in the enrichment measurement, and the factors that were taken into account to mitigate these issues are also detailed and tested. The use of an x-ray generator as a transmission source and the attending stability issues are addressed. Both analytical calculations and experimental measurements have been used. For completeness, some real-world analysis results from the URENCO Capenhurst enrichment plant have been included, where the final enrichment error has remained well below 1% for approximately two months.

  15. Fabrication and Characterization of Uranium-Molybdenum-Zirconium Alloys 

    E-Print Network [OSTI]

    Woolum, Connor

    2014-12-12

    As part of a global effort to convert reactors that require highly enriched uranium to instead operate with low enriched uranium, monolithic fuel plates consisting of a U-Mo fuel meat with a zirconium foil barrier layer and clad in aluminum...

  16. Modeling of UF{sub 6} enrichment with gas centrifuges for nuclear safeguards activities

    SciTech Connect (OSTI)

    Mercurio, G.; Peerani, P.; Richir, P.; Janssens, W.; Eklund, G.

    2012-09-26

    The physical modeling of uranium isotopes ({sup 235}U, {sup 238}U) separation process by centrifugation of is a key aspect for predicting the nuclear fuel enrichment plant performances under surveillance by the Nuclear Safeguards Authorities. In this paper are illustrated some aspects of the modeling of fast centrifuges for UF{sub 6} gas enrichment and of a typical cascade enrichment plant with the Theoretical Centrifuge and Cascade Simulator (TCCS). The background theory for reproducing the flow field characteristics of a centrifuge is derived from the work of Cohen where the separation parameters are calculated using the solution of a differential enrichment equation. In our case we chose to solve the hydrodynamic equations for the motion of a compressible fluid in a centrifugal field using the Berman - Olander vertical velocity radial distribution and the solution was obtained using the Matlab software tool. The importance of a correct estimation of the centrifuge separation parameters at different flow regimes, lies in the possibility to estimate in a reliable way the U enrichment plant performances, once the separation external parameters are set (feed flow rate and feed, product and tails assays). Using the separation parameters of a single centrifuge allow to determine the performances of an entire cascade and, for this purpose; the software Simulink was used. The outputs of the calculation are the concentrations (assays) and the flow rates of the enriched (product) and depleted (tails) gas mixture. These models represent a valid additional tool, in order to verify the compliance of the U enrichment plant operator declarations with the 'on site' inspectors' measurements.

  17. Uranium and other heavy metals in the plant-animal-human food chain near abandoned mining sites and structures in an American Indian community in northwestern New Mexico

    E-Print Network [OSTI]

    Samuel-Nakamura, Christine

    2013-01-01

    The Navajo people and uranium mining. Albuquerque, NM:2002). The history of uranium mining and the Navajo people.Key, C.R. (2000). Uranium mining and lung cancer among

  18. SciTech Connect: "enriched uranium"

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

    Renewable Energy Laboratory (NREL), Golden, CO (United States) Naval Petroleum and Oil Shale Reserves (United States) Navarro Navarro Nevada Environmental Services Nevada Field...

  19. SciTech Connect: enriched uranium

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

    Renewable Energy Laboratory (NREL), Golden, CO (United States) Naval Petroleum and Oil Shale Reserves (United States) Navarro Navarro Nevada Environmental Services Nevada Field...

  20. highly enriched uranium | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

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

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

    National Nuclear Security Administration (NNSA)

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

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

    National Nuclear Security Administration (NNSA)

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

  3. Uranium Mining, Conversion, and Enrichment Industries

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

    would cause. The analysis evaluates six factors for each industry: changes to prices; changes in production levels at existing facilities; changes to employment in the...

  4. Highly Enriched Uranium Disposition | National Nuclear Security

    National Nuclear Security Administration (NNSA)

    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 Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of NaturalDukeWakefield Municipal GasAdministration Medal01 Sandia4) AugustA. -71- Particulate: Columns 59 and R e s

  5. highly enriched uranium | National Nuclear Security Administration

    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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantityBonneville Power AdministrationRobust,Field-effectWorkingLosThe 26th Annual ConferenceFallHemmertHighlights6, 1999highly

  6. Uranium Mining, Conversion, and Enrichment Industries

    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 QA:QA J-E-1 SECTION J APPENDIX E LIST OF APPLICABLEStatutoryin theNuclear EnergyPotomac RiverUpperEnvironmentali

  7. Enrichment and broad representation of plant biomass-degrading enzymes in the specialized hyphal swellings of Leucoagaricus gongylophorus, the fungal symbiont of leaf-cutter ants

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

    Aylward, Frank O.; Khadempour, Lily; Tremmel, Daniel M.; McDonald, Bradon R.; Nicora, Carrie D.; Wu, Si; Moore, Ronald J.; Orton, Daniel J.; Monroe, Matthew E.; Piehowski, Paul D.; et al

    2015-08-28

    Leaf-cutter ants are prolific and conspicuous constituents of Neotropical ecosystems that derive energy from specialized fungus gardens they cultivate using prodigious amounts of foliar biomass. The basidiomycetous cultivar of the ants, Leucoagaricus gongylophorus, produces specialized hyphal swellings called gongylidia that serve as the primary food source of ant colonies. Gongylidia also contain plant biomass-degrading enzymes that become concentrated in ant digestive tracts and are deposited within fecal droplets onto fresh foliar material as ants incorporate it into the fungus garden. Although the enzymes concentrated by L. gongylophorus within gongylidia are thought to be critical to the initial degradation of plantmore »biomass, only a few enzymes present in these hyphal swellings have been identified. Here we use proteomic methods to identify proteins present in the gongylidia of three Atta cephalotes colonies. Our results demonstrate that a diverse but consistent set of enzymes is present in gongylidia, including numerous plant biomass-degrading enzymes likely involved in the degradation of polysaccharides, plant toxins, and proteins. Overall, gongylidia contained over three quarters of all biomass-degrading enzymes identified in the L. gongylophorus genome, demonstrating that the majority of the enzymes produced by this fungus for biomass breakdown are ingested by the ants. We also identify a set of 40 of these enzymes enriched in gongylidia compared to whole fungus garden samples, suggesting that certain enzymes may be particularly important in the initial degradation of foliar material. Our work sheds light on the complex interplay between leaf-cutter ants and their fungal symbiont that allows for the host insects to occupy an herbivorous niche by indirectly deriving energy from plant biomass.« less

  8. Optical manufacturing requirements for an AVLIS plant

    SciTech Connect (OSTI)

    Primdahl, K.; Chow, R.; Taylor, J.R.

    1997-07-14

    A uranium enrichment plant utilizing Atomic Vapor Laser Isotope Separation (AVLIS) technology is currently being planned. Deployment of the Plant will require tens of thousands of commercial and custom optical components and subsystems. The Plant optical system will be expected to perform at a high level of optical efficiency and reliability in a high-average-power-laser production environment. During construction, demand for this large number of optics must be coordinated with the manufacturing capacity of the optical industry. The general requirements and approach to ensure supply of optical components is described. Dynamic planning and a closely coupled relationship with the optics industry will be required to control cost, schedule, and quality.

  9. DOE Announces Transfer of Depleted Uranium to Advance the U.S...

    Energy Savers [EERE]

    Transfer of Depleted Uranium to Advance the U.S. National Security Interests, Extend Operations at Paducah Gaseous Diffusion Plant DOE Announces Transfer of Depleted Uranium to...

  10. Beneficial Uses of Depleted Uranium

    SciTech Connect (OSTI)

    Brown, C. [U.S. Department of Energy, Germantown, MD (United States); Croff, A.G.; Haire, M. J. [Oak Ridge National Lab., TN (United States)

    1997-08-01

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

  11. Validation of criticality safety calculational methods for U-AVLIS plant project

    SciTech Connect (OSTI)

    Lewis, K.D.

    1993-07-14

    The objectives of the Uranium Atomic Vapor Laser isotope Separation (U-AVLIS) are to develop, demonstrate, and deploy a laser-based process to enrich natural uranium in the U-235 isotope to levels useful as fuel in commercial light-water power reactors. Current U-AVLIS production plant criteria call for uranium product enriched in {sup 235}U up to 5 wt%. Development of the U-AVLIS technology is in an advanced stage, and demonstration of the integrated enrichment process is currently in progress using plant-scale equipment in the Uranium Demonstration System (UDS) at Lawrence Livermore National Laboratory. In this paper several existing experimental data which are applicable to the critical systems of importance to the safe design of the U-AVLIS plant are identified. These were used to benchmark a configuration-controlled, work station based version of one state-of-the-art computer code employed by the U-AVLIS program in UDS equipment design, and in U-AVLIS plant conceptual design NCS analyses.

  12. A Mock UF6 Feed and Withdrawal System for Testing Safeguards Monitoring Systems and Strategies Intended for Nuclear Fuel Enrichment and Processing Plants

    SciTech Connect (OSTI)

    Krichinsky, Alan M; Bates, Bruce E; Chesser, Joel B; Koo, Sinsze; Whitaker, J Michael

    2009-12-01

    This report describes an engineering-scale, mock UF6 feed and withdrawal (F&W) system, its operation, and its intended uses. This system has been assembled to provide a test bed for evaluating and demonstrating new methodologies that can be used in remote, unattended, continuous monitoring of nuclear material process operations. These measures are being investigated to provide independent inspectors improved assurance that operations are being conducted within declared parameters, and to increase the overall effectiveness of safeguarding nuclear material. Testing applicable technologies on a mock F&W system, which uses water as a surrogate for UF6, enables thorough and cost-effective investigation of hardware, software, and operational strategies before their direct installation in an industrial nuclear material processing environment. Electronic scales used for continuous load-cell monitoring also are described as part of the basic mock F&W system description. Continuous monitoring components on the mock F&W system are linked to a data aggregation computer by a local network, which also is depicted. Data collection and storage systems are described only briefly in this report. The mock UF{sub 6} F&W system is economical to operate. It uses a simple process involving only a surge tank between feed tanks and product and withdrawal (or waste) tanks. The system uses water as the transfer fluid, thereby avoiding the use of hazardous UF{sub 6}. The system is not tethered to an operating industrial process involving nuclear materials, thereby allowing scenarios (e.g., material diversion) that cannot be conducted otherwise. These features facilitate conducting experiments that yield meaningful results with a minimum of expenditure and quick turnaround time. Technologies demonstrated on the engineering-scale system lead to field trials (described briefly in this report) for determining implementation issues and performance of the monitoring technologies under plant operating conditions. The ultimate use of technologies tested on the engineering-scale test bed is to work with safeguards agencies to install them in operating plants (e.g., enrichment and fuel processing plants), thereby promoting new safeguards measures with minimal impact to operating plants. In addition, this system is useful in identifying features for new plants that can be incorporated as part of 'safeguards by design,' in which load cells and other monitoring technologies are specified to provide outputs for automated monitoring and inspector evaluation.

  13. Policy Paper 37: Energy and Security in Northeast Asia: Proposals for Nuclear Cooperation

    E-Print Network [OSTI]

    Kaneko, Kumao; Suzuki, Atsuyuki; Choi, Jor-Shan; Fei, Edward

    1998-01-01

    reactors (LWRs) using low-enriched uranium (LEU) as fuelenrichment plant for low-enriched uranium, and a subcritical

  14. Domestic utility attitudes toward foreign uranium supply

    SciTech Connect (OSTI)

    Not Available

    1981-06-01

    The current embargo on the enrichment of foreign-origin uranium for use in domestic utilization facilities is scheduled to be removed in 1984. The pending removal of this embargo, complicated by a depressed worldwide market for uranium, has prompted consideration of a new or extended embargo within the US Government. As part of its on-going data collection activities, Nuclear Resources International (NRI) has surveyed 50 domestic utility/utility holding companies (representing 60 lead operator-utilities) on their foreign uranium purchase strategies and intentions. The most recent survey was conducted in early May 1981. A number of qualitative observations were made during the course of the survey. The major observations are: domestic utility views toward foreign uranium purchase are dynamic; all but three utilities had some considered foreign purchase strategy; some utilities have problems with buying foreign uranium from particular countries; an inducement is often required by some utilities to buy foreign uranium; opinions varied among utilities concerning the viability of the domestic uranium industry; and many utilities could have foreign uranium fed through their domestic uranium contracts (indirect purchases). The above observations are expanded in the final section of the report. However, it should be noted that two of the observations are particularly important and should be seriously considered in formulation of foreign uranium import restrictions. These important observations are the dynamic nature of the subject matter and the potentially large and imbalanced effect the indirect purchases could have on utility foreign uranium procurement.

  15. Application of pathways analyses for site performance prediction for the Gas Centrifuge Enrichment Plant and Oak Ridge Central Waste Disposal Facility

    SciTech Connect (OSTI)

    Pin, F.G.; Oblow, E.M.

    1984-01-01

    The suitability of the Gas Centrifuge Enrichment Plant and the Oak Ridge Central Waste Disposal Facility for shallow-land burial of low-level radioactive waste is evaluated using pathways analyses. The analyses rely on conservative scenarios to describe the generation and migration of contamination and the potential human exposure to the waste. Conceptual and numerical models are developed using data from comprehensive laboratory and field investigations and are used to simulate the long-term transport of contamination to man. Conservatism is built into the analyses when assumptions concerning future events have to be made or when uncertainties concerning site or waste characteristics exist. Maximum potential doses to man are calculated and compared to the appropriate standards. The sites are found to provide adequate buffer to persons outside the DOE reservations. Conclusions concerning site capacity and site acceptability are drawn. In reaching these conclusions, some consideration is given to the uncertainties and conservatisms involved in the analyses. Analytical methods to quantitatively assess the probability of future events to occur and the sensitivity of the results to data uncertainty may prove useful in relaxing some of the conservatism built into the analyses. The applicability of such methods to pathways analyses is briefly discussed. 18 refs., 9 figs.

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

    SciTech Connect (OSTI)

    Anheier, Norman C.; Bushaw, Bruce A.

    2010-01-01

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

  17. AVLIS Production Plant Project Management Plan

    SciTech Connect (OSTI)

    Not Available

    1984-11-15

    The AVLIS Production Plant is designated as a Major System Acquisition (in accordance with DOE Order 4240.IC) to deploy Atomic Vapor Laser Isotope Separation (AVLIS) technology at the Oak Ridge, Tennessee site, in support of the US Uranium Enrichment Program. The AVLIS Production Plant Project will deploy AVLIS technology by performing the design, construction, and startup of a production plant that will meet capacity production requirements of the Uranium Enrichment Program. The AVLIS Production Plant Project Management Plan has been developed to outline plans, baselines, and control systems to be employed in managing the AVLIS Production Plant Project and to define the roles and responsibilities of project participants. Participants will develop and maintain detailed procedures for implementing the management and control systems in agreement with this plan. This baseline document defines the system that measures work performed and costs incurred. This plan was developed by the AVLIS Production Plant Project staff of Martin Marietta Energy Systems, Inc. and Lawrence Livermore National Laboratory in accordance with applicable DOE directives, orders and notices. 38 figures, 19 tables.

  18. Prospects for the recovery of uranium from seawater

    E-Print Network [OSTI]

    Best, F. R.

    1980-01-01

    A computer program entitled URPE (Uranium Recovery Performance and Economics) has been developed to simulate the engineering performance and provide an economic analysis O of a plant recovering uranium from seawater. The ...

  19. Uranium hexafluoride handling. Proceedings

    SciTech Connect (OSTI)

    Not Available

    1991-12-31

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

  20. Privatization of the gaseous diffusion plants and impacts on nuclear criticality safety administration

    SciTech Connect (OSTI)

    D`Aquila, D.M.; Holliday, R.T. [Lockheed Martin Utility Services, Inc., Piketon, OH (United States); Dean, J.C. [Lockheed Martin Utility Services, Inc., Paducah, KY (United States)

    1996-12-31

    The Energy Policy Act of 1992 created the United States Enrichment Corporation (USEC) on July 1, 1993. The USEC is a government-owned business that leases those Gaseous Diffusion Plant (GDP) facilities at the Portsmouth, Ohio, and Paducah, Kentucky, sites from the U.S. Department of Energy (DOE) that are required for enriching uranium. Lockheed Martin Utility Services is the operating contractor for the USEC-leased facilities. The DOE has retained use of, and regulation over, some facilities and areas at the Portsmouth and Paducah sites for managing legacy wastes and environmental restoration activities. The USEC is regulated by the DOE, but is currently changing to regulation under the U.S. Nuclear Regulatory Commission (NRC). The USEC is also preparing for privatization of the uranium enrichment enterprise. These changes have significantly affected the nuclear criticality safety (NCS) programs at the sites.

  1. Uranium industry annual 1997

    SciTech Connect (OSTI)

    NONE

    1998-04-01

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

  2. ORNL/TM-2008/195 Model of a Generic Natural Uranium

    E-Print Network [OSTI]

    Pennycook, Steve

    ORNL/TM-2008/195 Model of a Generic Natural Uranium Conversion Plant--Suggested Measures OF A GENERIC NATURAL URANIUM CONVERSION PLANT-- SUGGESTED MEASURES TO STRENGTHEN INTERNATIONAL SAFEGUARD From ........................................................................................................................ 1 2. TECHNICAL PROCESS FOR NATURAL URANIUM CONVERSION PLANTS ...................... 2 2.1 Generic

  3. Safety Evaluation Report for the Claiborne Enrichment Center, Homer, Louisiana (Docket No. 70-3070)

    SciTech Connect (OSTI)

    Not Available

    1994-01-01

    This report documents the US Nuclear Regulatory Commission (NRC) staff review and safety evaluation of the Louisiana Energy Services, L.P. (LES, the applicant) application for a license to possess and use byproduct, source, and special nuclear material and to enrich natural uranium to a maximum of 5 percent U-235 by the gas centrifuge process. The plant, to be known as the Claiborne Enrichment Center (CEC), would be constructed near the town of Homer in Claiborne Parish, Louisiana. At full production in a given year, the plant will receive approximately 4,700 tonnes of feed UF{sub 6} and produce 870 tonnes of low-enriched UF{sub 6}, and 3,830 tonnes of depleted UF{sub 6} tails. Facility construction, operation, and decommissioning are expected to last 5, 30, and 7 years, respectively. The objective of the review is to evaluate the potential adverse impacts of operation of the facility on worker and public health and safety under both normal operating and accident conditions. The review also considers the management organization, administrative programs, and financial qualifications provided to assure safe design and operation of the facility. The NRC staff concludes that the applicant`s descriptions, specifications, and analyses provide an adequate basis for safety review of facility operations and that construction and operation of the facility does not pose an undue risk to public health and safety.

  4. The End of Cheap Uranium

    E-Print Network [OSTI]

    Michael Dittmar

    2011-06-21

    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.

  5. Degradation problems with the solvent extraction organic at Roessing uranium

    SciTech Connect (OSTI)

    Munyungano, Brodrick; Feather, Angus; Virnig, Michael

    2008-07-01

    Roessing Uranium Ltd recovers uranium from a low-grade ore in Namibia. Uranium is recovered and purified from an ion-exchange eluate in a solvent-extraction plant. The solvent-extraction plant uses Alamine 336 as the extractant for uranium, with isodecanol used as a phase modifier in Sasol SSX 210, an aliphatic hydrocarbon diluent. Since the plant started in the mid 1970's, there have been a few episodes where the tertiary amine has been quickly and severely degraded when the plant was operated outside certain operating parameters. The Rossing experience is discussed in more detail in this paper. (authors)

  6. Next Generation Nuclear Plant Materials Research and Development Program Plan

    SciTech Connect (OSTI)

    G. O. Hayner; E.L. Shaber

    2004-09-01

    The U.S Department of Energy (DOE) has selected the Very High Temperature Reactor (VHTR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production without greenhouse gas emissions. The reactor design will be a graphite moderated, helium-cooled, prismatic or pebble-bed, thermal neutron spectrum reactor that will produce electricity and hydrogen in a state-of-the-art thermodynamically efficient manner. The NGNP will use very high burn-up, low-enriched uranium, TRISO-coated fuel and have a projected plant design service life of 60 years.

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

    SciTech Connect (OSTI)

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

    2013-01-01

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

  8. Use of the UNCLE Facility to Assess Integrated Online Monitoring Systems for Detection of Diversions at Uranium Conversion Facilities

    SciTech Connect (OSTI)

    Dewji, Shaheen A; Chapman, Jeffrey Allen; Lee, Denise L; Rauch, Eric; Hertel, Nolan

    2011-01-01

    Historically, the approach to safeguarding nuclear material in the front end of the fuel cycle was implemented only at the stage when UF6 was declared as feedstock for enrichment plants. Recent International Atomic Energy Agency (IAEA) circulars and policy papers have sought to implement safeguards when any purified aqueous uranium solution or uranium oxides suitable for isotopic enrichment or fuel fabrication exist. Oak Ridge National Laboratory has developed the Uranyl Nitrate Calibration Loop Equipment (UNCLE) facility to simulate the full-scale operating conditions for a purified uranium-bearing aqueous stream exiting the solvent extraction process conducted in a natural uranium conversion plant (NUCP) operating at 6000 MTU/year. Monitoring instruments, including the 3He passive neutron detector developed at Los Alamos National Laboratory and the Endress+Hauser Promass 83F Coriolis meter, have been tested at UNCLE and field tested at Springfields. The field trials demonstrated the need to perform full-scale equipment testing under controlled conditions prior to field deployment of operations and safeguards monitoring at additional plants. Currently, UNCLE is testing neutron-based monitoring for detection of noncompliant activities; however, gamma-ray source term monitoring is currently being explored complementary to the neutron detector in order to detect undeclared activities in a more timely manner. The preliminary results of gamma-ray source term modeling and monitoring at UNCLE are being analyzed as part of a comprehensive source term and detector benchmarking effort. Based on neutron source term detection capabilities, alternative gamma-based detection and monitoring methods will be proposed to more effectively monitor NUCP operations in verifying or detecting deviations from declared conversion activities.

  9. Uranium 2009 resources, production and demand

    E-Print Network [OSTI]

    Organisation for Economic Cooperation and Development. Paris

    2010-01-01

    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...

  10. No Slide Title

    National Nuclear Security Administration (NNSA)

    (if not losstheftdiversion of significant srd) - Enrichment plants, plutoniumuranium fuel fabricators and down blending plants (>10% enriched uranium) report all...

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

    SciTech Connect (OSTI)

    John D. Bess

    2010-06-01

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

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

    SciTech Connect (OSTI)

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

    2013-07-01

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

  13. Development of Integrated Online Monitoring Systems for Detection of Diversion at Natural Uranium Conversion Facilities

    SciTech Connect (OSTI)

    Dewji, Shaheen A; Lee, Denise L; Croft, Stephen; McElroy, Robert Dennis; Hertel, Nolan; Chapman, Jeffrey Allen; Cleveland, Steven L

    2013-01-01

    Recent work at Oak Ridge National Laboratory (ORNL) has focused on some source term modeling of uranyl nitrate (UN) as part of a comprehensive validation effort employing gamma-ray detector instrumentation for the detection of diversion from declared conversion activities. Conversion, the process by which natural uranium ore (yellowcake) is purified and converted through a series of chemical processes into uranium hexafluoride gas (UF6), has historically been excluded from the nuclear safeguards requirements of the 235U-based nuclear fuel cycle. The undeclared diversion of this product material could potentially provide feedstock for a clandestine weapons program for state or non-state entities. Given the changing global political environment and the increased availability of dual-use nuclear technology, the International Atomic Energy Agency has evolved its policies to emphasize safeguarding this potential feedstock material in response to dynamic and evolving potential diversion pathways. To meet the demand for instrumentation testing at conversion facilities, ORNL developed the Uranyl Nitrate Calibration Loop Equipment (UNCLE) facility to simulate the full-scale operating conditions of a purified uranium-bearing aqueous stream exiting the solvent extraction process in a natural uranium conversion plant. This work investigates gamma-ray signatures of UN circulating in the UNCLE facility and evaluates detector instrumentation sensitivity to UN for safeguards applications. These detector validation activities include assessing detector responses to the UN gamma-ray signatures for spectrometers based on sodium iodide, lanthanum bromide, and germanium detectors. The results of measurements under static and dynamic operating conditions at concentrations ranging from 10-90g U/L of naturally enriched UN will be presented. A range of gamma-ray lines was examined and self-attenuation factors were calculated, in addition to attenuation for transmission measurement of density, concentration and enrichment. A detailed uncertainty analysis will be presented providing insights into instrumentation limitations to spoofing.

  14. Uranium accountancy in Atomic Vapor Laser Isotope Separation

    SciTech Connect (OSTI)

    Carver, R.D.

    1986-01-01

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

  15. 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.

    1994-05-01

    The Department of Energy (DOE), Office of Environmental Restoration and Waste Management, has chartered a study to evaluate alternative management strategies for depleted uranium (DU) currently stored throughout the DOE complex. Historically, DU has been maintained as a strategic resource because of uses for DU metal and potential uses for further enrichment or for uranium oxide as breeder reactor blanket fuel. This study has focused on evaluating the disposal options for DU if it were considered a waste. This report is in no way declaring these DU reserves a ``waste,`` but is intended to provide baseline data for comparison with other management options for use of DU. To PICS considered in this report include: Retrievable disposal; permanent disposal; health hazards; radiation toxicity and chemical toxicity.

  16. Microsoft PowerPoint - 2A_Wednesday 5-22 830 NMMSS_2013_Presentation...

    National Nuclear Security Administration (NNSA)

    Exports Natural uranium to conversion and enrichment plants Enriched uranium to fuel fabrication plants Return of spent fuel from foreign research reactors 6 Primary...

  17. URANIUM IN ALKALINE ROCKS

    E-Print Network [OSTI]

    Murphy, M.

    2011-01-01

    1962. "Diatremes and Uranium Deposits in the Hopi Buttes,H. , 1970. "Low-Grade Uranium Deposits in Agpaitic NephelineL. Torkild, 1974B. "The Uranium Deposit at Kvanefjeld, The

  18. URANIUM IN ALKALINE ROCKS

    E-Print Network [OSTI]

    Murphy, M.

    2011-01-01

    1977. "Geology of Brazil's Uranium and Thorium Occurrences,"A tantalo-niobate of uranium, near pyrochlore. Isometric,niobate and tantalate of uranium, with ferrous iron and rare

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

    SciTech Connect (OSTI)

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

    2010-04-01

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

  20. 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

    2002-01-01

    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...

  1. The End of Cheap Uranium

    E-Print Network [OSTI]

    Dittmar, Michael

    2011-01-01

    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. The Blend Down Monitoring System Demonstration at the Paducah Gaseous Diffusion Plant

    SciTech Connect (OSTI)

    Benton, J.; Close, D.; Johnson, W., Jr.; Kerr, P.; March-Leuba, J.; Mastal, E.; Moss, C.; Powell, D.; Sumner, J.; Uckan, T.; Vines, R.; Wright, P.D.

    1999-07-25

    Agreements between the governments of the US and the Russian Federation for the US purchase of low enriched uranium (LEU) derived from highly enriched uranium (HEU) from dismantled Russian nuclear weapons calls for the establishment of transparency measures to provide confidence that nuclear nonproliferation goals are being met. To meet these transparency goals, the agreements call for the installation of nonintrusive US instruments to monitor the down blending of HEU to LEU. The Blend Down Monitoring System (BDMS) has been jointly developed by the Los Alamos National Laboratory (LANL) and the Oak Ridge National Laboratory (ORNL) to continuously monitor {sup 235}U enrichments and mass flow rates at Russian blending facilities. Prior to its installation in Russian facilities, the BDMS was installed and operated in a UF{sub 6} flow loop in the Paducah Gaseous Diffusion Plant simulating flow and enrichment conditions expected in a typical down-blending facility. A Russian delegation to the US witnessed the equipment demonstration in June, 1998. To conduct the demonstration in the Paducah Gaseous Diffusion Plant (PGDP), the BDMS was required to meet stringent Nuclear Regulatory Commission licensing, safety and operational requirements. The Paducah demonstration was an important milestone in achieving the operational certification for the BDMS use in Russian facilities.

  3. The geochemistry of uranium in the Orca Basin 

    E-Print Network [OSTI]

    Weber, Frederick Fewell

    1979-01-01

    in each sample was also measur. ed to gain insight concerning the origin and nature of Urea Basin deposits. For comparison, cores from the brine- filled Suakin and Atlantis II Deeps, both in the Red Sea, were also analyzed. Ores Basin sediments show... Deep where no uranium enrichment was also observed. The Atlantis II Deep, however, contains sediments significantly enriched in uranium. This basin differs from the other two in that its brin. e temperature is close to 40'C warmer. than average Red...

  4. Uranium atomic vapor laser isotope separation (AVL1S)

    SciTech Connect (OSTI)

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

    1992-12-01

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

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

    SciTech Connect (OSTI)

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

    1997-07-01

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

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

    SciTech Connect (OSTI)

    Ade, Brian J; Gauld, Ian C

    2011-10-01

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

  7. Characterization of Thermal Properties of Depleted Uranium Metal Microspheres 

    E-Print Network [OSTI]

    Humrickhouse, Carissa Joy

    2012-07-16

    llment of the requirements for the degree of MASTER OF SCIENCE Approved by: Chair of Committee, Sean M. McDeavitt Committee Members, Kenneth L. Peddicord Lin Shao Head of Department, Yassin A. Hassan May 2012 Major Subject: Nuclear Engineering iii.../m-K) Density (units: g/cm3) CHTA Crucible Heater Test Assembly DU Depleted uranium EU Enriched uranium LFA Laser (or light) ash analysis LFA 447 Light ash analyzer, model 447, by Netzsch Instruments LWR Light water reactor ODU Oxidized depleted uranium...

  8. 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)

    2012-02-15

    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.

  9. U.S. transparency monitoring of HEU oxide conversion and blending to LEU hexafluoride at three Russian blending plants

    SciTech Connect (OSTI)

    Leich, D., LLNL

    1998-07-27

    The down-blending of Russian highly enriched uranium (HEU) takes place at three Russian gaseous centrifuge enrichment plants. The fluorination of HEU oxide and down-blending of HEU hexafluoride began in 1994, and shipments of low enriched uranium (LEU) hexafluoride product to the United States Enrichment Corporation (USEC) began in 1995 US transparency monitoring under the HEU Purchase Agreement began in 1996 and includes a permanent monitoring presence US transparency monitoring at these facilities is intended to provide confidence that HEU is received and down-blended to LEU for shipment to USEC The monitoring begins with observation of the receipt of HEU oxide shipments, including confirmation of enrichment using US nondestructive assay equipment The feeding of HEU oxide to the fluorination process and the withdrawal of HEU hexafluoride are monitored Monitoring is also conducted where the blending takes place and where shipping cylinders are filled with LEU product. A series of process and material accountancy documents are provided to US monitors.

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

    Office of Energy Efficiency and Renewable Energy (EERE)

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

  11. Towards a Systematic Threat Modeling Approach for Cyber-physical Systems Goncalo Martins1

    E-Print Network [OSTI]

    Koutsoukos, Xenofon D.

    that targeted nuclear centrifuges at the Iranian uranium enrichment plant [1]. In this attack, a worm

  12. Taxonomy for Description of Cross-Domain Attacks on CPS

    E-Print Network [OSTI]

    Koutsoukos, Xenofon D.

    . This attack has reportedly damaged over 1000 centrifuges at an Iranian uranium enrichment plant [1]. Multiple

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

    SciTech Connect (OSTI)

    Anheier, Norman C.; Bushaw, Bruce A.

    2009-07-01

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

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

    SciTech Connect (OSTI)

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

    2007-07-01

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

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

    Office of Environmental Management (EM)

    Enforcement Office of Enforcement U.S. Environmental Protection Agency Michael F. Wood, Director Compliance Division Office of Compliance Monitoring Office of Pesticides and...

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

    National Nuclear Security Administration (NNSA)

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

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

    National Nuclear Security Administration (NNSA)

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

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

    National Nuclear Security Administration (NNSA)

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

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

    National Nuclear Security Administration (NNSA)

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

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

    National Nuclear Security Administration (NNSA)

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

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

    National Nuclear Security Administration (NNSA)

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

  2. THE THEORY OF URANIUM ENRICHMENT BY THE GAS CENTRIFUGE

    E-Print Network [OSTI]

    Olander, Donald R.

    2013-01-01

    cascades. Fig. 1 MAGNETIC BEARING AND DAMPING ASSEMBLY '··bearing assembly on top, where contact between moving and stationary parts is avoided by magnetic

  3. THE THEORY OF URANIUM ENRICHMENT BY THE GAS CENTRIFUGE

    E-Print Network [OSTI]

    Olander, Donald R.

    2013-01-01

    Technical Basis of the Gas Centrifuge", Adv. in Nucl. Sci.D.R. , (1978) "The Gas Centrifuge", Scientific American,Fluid Dynamics of a Gas Centrifuge", J. Fluid Mech. , 101,

  4. Highly Enriched Uranium Materials Facility, Major Design Changes

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

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

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

    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 Delicious Rank EERE:Financing Tool FitsProjectDataSecretaryDepartment7 Annual ReportSouthwestEnergy2707-01and

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

    National Nuclear Security Administration (NNSA)

    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 Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of NaturalDukeWakefield Municipal Gas &SCE-SessionsSouthReport for the t-) S/,,5 'a C | National NuclearLibrary

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

    National Nuclear Security Administration (NNSA)

    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 Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of NaturalDukeWakefield Municipal Gas &SCE-SessionsSouthReport forRetirement PlanSupplementalEmergencyDetection Sensors

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

    National Nuclear Security Administration (NNSA)

    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 Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of NaturalDukeWakefield Municipal GasAdministration Medal of HonorPoster Session |SecurityNSDDfor

  9. Highly Enriched Uranium Materials Facility | National Nuclear Security

    National Nuclear Security Administration (NNSA)

    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 Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of NaturalDukeWakefield Municipal GasAdministration Medal01 Sandia NationalSecurityNuclearH-canyon | NationalNRC

  10. Toxic Substances Control Act Uranium Enrichment Federal Facility Compliance

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

    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 Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustmentsShirleyEnergyThe U.S.Laclede GasEfficiency| Department ofConstruction &

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

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

    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 DeliciousMathematics And Statistics » USAJobs Search USAJobs Search TheChlamydomonasMaterial from Building 30192008

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

    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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity ofkandz-cm11 Outreach Home Room NewsInformation Current HABFESOpportunities Nuclear PhysicsGE GlobalGetting Started at

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

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

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  14. Highly Enriched Uranium Transparency Program | National Nuclear Security

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

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

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

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

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

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

    Energy Savers [EERE]

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

  18. 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

  19. Safeguards for Uranium Extraction (UREX) +1a Process 

    E-Print Network [OSTI]

    Feener, Jessica S.

    2011-08-08

    As nuclear energy grows in the United States and around the world, the expansion of the nuclear fuel cycle is inevitable. All currently deployed commercial reprocessing plants are based on the Plutonium - Uranium Extraction ...

  20. Naval Nuclear Propulsion Plants | National Nuclear Security Administra...

    National Nuclear Security Administration (NNSA)

    Naval Nuclear Propulsion Plants In naval nuclear propulsion plants, fissioning of uranium atoms in the reactor core produces heat. Because the fission process also produces...

  1. Assessing the environmental availability of uranium in soils and sediments

    SciTech Connect (OSTI)

    Amonette, J.E.; Holdren, G.R. Jr.; Krupa, K.M.; Lindenmeier, C.W. [Pacific Northwest Lab., Richland, WA (United States)

    1994-06-01

    Soils and sediments contaminated with uranium pose certain environmental and ecological risks. At low to moderate levels of contamination, the magnitude of these risks depends not only on the absolute concentrations of uranium in the material but also on the availability of the uranium to drinking water supplies, plants, or higher organisms. Rational approaches for regulating the clean-up of sites contaminated with uranium, therefore, should consider the value of assessing the environmental availability of uranium at the site before making decisions regarding remediation. The purpose of this work is to review existing approaches and procedures to determine their potential applicability for assessing the environmental availability of uranium in bulk soils or sediments. In addition to making the recommendations regarding methodology, the authors have tabulated data from the literature on the aqueous complexes of uranium and major uranium minerals, examined the possibility of predicting environmental availability of uranium based on thermodynamic solubility data, and compiled a representative list of analytical laboratories capable of performing environmental analyses of uranium in soils and sediments.

  2. Corrosion Evaluation of RERTR Uranium Molybdenum Fuel

    SciTech Connect (OSTI)

    A K Wertsching

    2012-09-01

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

  3. Method for converting uranium oxides to uranium metal

    DOE Patents [OSTI]

    Duerksen, Walter K. (Norris, TN)

    1988-01-01

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

  4. Environmental restoration and waste management site specific plan for Oak Ridge Operation Office Paducah Gaseous Diffusion Plant

    SciTech Connect (OSTI)

    Not Available

    1990-07-18

    The Paducah Gaseous Diffusion Plant (PGDP) occupies 748 security-fenced acres located on a 3,400-acre tract in McCracken County, Kentucky, which was previously part of the Kentucky Ordnance Works. The principle objective on-site process at PGDP is the separation of uranium isotopes through gaseous diffusion. The process produces enriched uranium, which is used for nuclear fuel in commercial power plants and for military purposes. This document provides an overview of the major environmental and waste management concerns at PGDP, requirements for implementation, organization/management, corrective activities, environmental restoration, waste management options, compliance with National Environmental Policy Act (NEPA), reporting and data management, quality assurance and federal, state and local interactions. 12 refs., 6 figs., 5 tabs.

  5. The Mailbox Computer System for the IAEA verification experiment on HEU downlending at the Portsmouth Gaseous Diffusion Plant

    SciTech Connect (OSTI)

    Aronson, A.L.; Gordon, D.M.

    2000-07-31

    IN APRIL 1996, THE UNITED STATES (US) ADDED THE PORTSMOUTH GASEOUS DIFFUSION PLANT TO THE LIST OF FACILITIES ELIGIBLE FOR THE APPLICATION OF INTERNATIONAL ATOMIC ENERGY AGENCY (IAEA) SAFEGUARDS. AT THAT TIME, THE US PROPOSED THAT THE IAEA CARRY OUT A ''VERIFICATION EXPERIMENT'' AT THE PLANT WITH RESPECT TO DOOWNBLENDING OF ABOUT 13 METRIC TONS OF HIGHLY ENRICHED URANIUM (HEU) IN THE FORM OF URANIUM HEXAFLUROIDE (UF6). DURING THE PERIOD DECEMBER 1997 THROUGH JULY 1998, THE IAEA CARRIED OUT THE REQUESTED VERIFICATION EXPERIMENT. THE VERIFICATION APPROACH USED FOR THIS EXPERIMENT INCLUDED, AMONG OTHER MEASURES, THE ENTRY OF PROCESS-OPERATIONAL DATA BY THE FACILITY OPERATOR ON A NEAR-REAL-TIME BASIS INTO A ''MAILBOX'' COMPUTER LOCATED WITHIN A TAMPER-INDICATING ENCLOSURE SEALED BY THE IAEA.

  6. Development of Novel Sorbents for Uranium Extraction from Seawater

    SciTech Connect (OSTI)

    Lin, Wenbin; Taylor-Pashow, Kathryn

    2014-01-08

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

  7. Depleted uranium: A DOE management guide

    SciTech Connect (OSTI)

    1995-10-01

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

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

    SciTech Connect (OSTI)

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

    1981-04-01

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

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

    SciTech Connect (OSTI)

    Vasil’ev, I. V. Ivanov, A. S.; Churin, V. A.

    2014-12-15

    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 1500°C. 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.

  10. Uranium hexafluoride liquid thermal expansion, elusive eutectic with hydrogen fluoride, and very first production using chlorine trifluoride

    SciTech Connect (OSTI)

    Rutledge, G.P. [Central Environmental, Inc., Anchorage, AK (United States)

    1991-12-31

    Three unusual incidents and case histories involving uranium hexafluoride in the enrichment facilities of the USA in the late 1940`s and early 1950`s are presented. The history of the measurements of the thermal expansion of liquids containing fluorine atoms within the molecule is reviewed with special emphasis upon uranium hexafluoride. A comparison is made between fluorinated esters, fluorocarbons, and uranium hexafluoride. The quantitative relationship between the thermal expansion coefficient, a, of liquids and the critical temperature, T{sub c} is presented. Uranium hexafluoride has an a that is very high in a temperature range that is used by laboratory and production workers - much higher than any other liquid measured. This physical property of UF{sub 6} has resulted in accidents involving filling the UF{sub 6} containers too full and then heating with a resulting rupture of the container. Such an incident at a uranium gaseous diffusion plant is presented. Production workers seldom {open_quotes}see{close_quotes} uranium hexafluoride. The movement of UF{sub 6} from one container to another is usually trailed by weight, not sight. Even laboratory scientists seldom {open_quotes}see{close_quotes} solid or liquid UF{sub 6} and this can be a problem at times. This inability to {open_quotes}see{close_quotes} the UF{sub 6}-HF mixtures in the 61.2{degrees}C to 101{degrees}C temperature range caused a delay in the understanding of the phase diagram of UF{sub 6}-HF which has a liquid - liquid immiscible region that made the eutectic composition somewhat elusive. Transparent fluorothene tubes solved the problem both for the UF{sub 6}-HF phase diagram as well as the UF{sub 6}-HF-CIF{sub 3} phase diagram with a miscibility gap starting at 53{degrees}C. The historical background leading to the first use of CIF{sub 3} to produce UF{sub 6} in both the laboratory and plant at K-25 is presented.

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

    DOE Patents [OSTI]

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

    1995-06-06

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

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

    DOE Patents [OSTI]

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

    1995-01-01

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

  13. Preparation of uranium compounds

    DOE Patents [OSTI]

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

    2013-02-19

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

  14. Commissioning Measurements and Experience Obtained from the Installation of a Fissile Mass Flow monitor in the URAL Electrochemical Integrated Plant (UEIP) in Novouralsk

    SciTech Connect (OSTI)

    March-Leuba, J.; Mastal, E.; Powell, D.; Sumner, J.; Uckan, T.; Vines, V.

    1999-07-25

    The Blend Down Monitoring System (BDMS) equipment sent earlier to the Ural Electrochemical Integrated Plant (UEIP) at Novouralsk, Russia, was installed and implemented successfully on February 2, 1999. The BDMS installation supports the highly enriched uranium (HEU) Transparency Implementation Program for material subject to monitoring under the HEU purchase agreement between the United States of America (USA) and the Russian Federation (RF). The BDMS consists of the Oak Ridge National Laboratory (ORNL) Fissile (uranium-235) Mass Flow Monitor (FMFM) and the Los Alamos National Laboratory (LANL) Enrichment Monitor (EM). Two BDMS?s for monitoring the Main and Reserve HEU blending process lines were installed at UEIP. Independent operation of the FMFM Main and FMFM Reserve was successfully demonstrated for monitoring the fissile mass flow as well as the traceability of HEU to the product low enriched uranium. The FMFM systems failed when both systems were activated during the calibration phase due to a synchronization problem between the systems. This operational failure was caused by the presence of strong electromagnetic interference (EMI) in the blend point. The source-modulator shutter motion of the two FMFM systems was not being properly synchronized because of EMI producing a spurious signal on the synchronization cable connecting the two FMFM cabinets. The signature of this failure was successfully reproduced at ORNL after the visit. This unexpected problem was eliminated by a hardware modification and software improvements during a recent visit (June 9-11, 1999) to UEIP, and both systems are now operating as expected.

  15. U.S.Uranium Reserves

    Gasoline and Diesel Fuel Update (EIA)

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

  16. The Enriched Xenon Observatory

    SciTech Connect (OSTI)

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

    2009-12-17

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

  17. Development of an enrichment monitor for the Portsmouth GCEP

    SciTech Connect (OSTI)

    Strittmatter, R.B.; Stovall, L.A.; Sprinkle, J.K. Jr.

    1983-01-01

    We have developed a gas-phase UF/sub 6/ enrichment monitor for use by the International Atomic Energy Agency at the Portsmouth Gas Centrifuge Enrichment Plant. The enrichment monitoring system provides a method for effective nuclear materials accountability verification while reducing the effort for both the facility operator and the inspector. The experience with an inplant prototype monitor, the facility and operational constraints, and the constraints related to international safeguards inspection are described in terms of the impact on the monitor design.

  18. Uranium dioxide electrolysis

    DOE Patents [OSTI]

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

    2009-12-29

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

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

    SciTech Connect (OSTI)

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

    1998-12-31

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

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

    SciTech Connect (OSTI)

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

    1991-11-01

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

  1. 2013 Domestic Uranium Production Report

    E-Print Network [OSTI]

    2013 Domestic Uranium Production Report May 2014 Independent Statistics & Analysis www.eia.gov U Administration | 2013 Domestic Uranium Production Report ii Contacts This report was prepared by the staff of the Renewables and Uranium Statistics Team, Office of Electricity, Renewables, and Uranium Statistics. Questions

  2. Development of NF3 Deposit Removal Technology for the Portsmouth Gaseous Diffusion Plant

    SciTech Connect (OSTI)

    Scheele, R.D.; McNamara, B.K.; Rapko, B.M.; Edwards, M.K.; Kozelisky, A.E.; Daniel, R.C. [Battelle Pacific Northwest Division, PO Box 999, Battelle Blvd, Richland, Washington 99352 (United States); McSweeney, T.I.; Maharas, S.J.; Weaver, P.J.; Iwamasa, K.J. [Battelle Columbus Operations, 505 King Avenue, Columbus, Ohio 43201 (United States); Kefgen, R.B. [WASTREN, Inc., 1864 Shyville Road, Piketon, Ohio 45661 (United States)

    2006-07-01

    This paper summarizes the Battelle, Stoller, and WASTREN (BSW) team's efforts, to date, in support of the United States Department of Energy's plans to remove uranium and technetium deposits before decommissioning the Portsmouth Gaseous Diffusion Plant. The BSW team investigated nitrogen trifluoride (NF{sub 3}) as a safer yet effective alternative gaseous treatment to the chlorine trifluoride (ClF{sub 3})-elemental fluorine (F{sub 2}) treatment currently used to remove uranium and technetium deposits from the uranium enrichment cascade. Both ClF{sub 3} and F{sub 2} are highly reactive, toxic, and hazardous gases, while NF{sub 3}, although toxic [1], is no more harmful than moth balls [2]. BSW's laboratory thermo-analytical and laboratory-scale prototype studies with NF{sub 3} established that thermal NF{sub 3} can effectively remove likely and potential uranium (UO{sub 2}F{sub 2} and UF{sub 4}) and technetium deposits (a surrogate deposit material, TcO{sub 2}, and pertechnetates) by conversion to volatile compounds. Our engineering evaluations suggest that NF{sub 3}'s effectiveness could be enhanced by combining with a lesser concentration of ClF{sub 3}. BSW's and other's studies indicate compatibility with Portsmouth materials of construction (aluminum, copper, and nickel). (authors)

  3. BEHAVIOR OF METALLIC INCLUSIONS IN URANIUM DIOXIDE

    E-Print Network [OSTI]

    Yang, Rosa L.

    2013-01-01

    Metallic Inclusions in Uranium Dioxide", LBL-11117 (1980).in Hypostoichiornetric Uranium Dioxide 11 , LBL-11095 (OF METALLIC INCLUSIONS IN URANIUM DIOXIDE Rosa L. Yang and

  4. AVLIS production plant project schedule and milestones

    SciTech Connect (OSTI)

    Not Available

    1984-11-15

    An AVLIS Production Plant Deployment Schedule for the engineering, procurement, and construction for both the Initial Increment of Production and the fully Activated Plant, has been developed by the project team consisting of Lawrence Livermore National Laboratory, Martin Marietta Energy Systems, Inc. with architect-engineer support from Bechtel National, Inc., Stone and Webster Engineering Corporation, and Westinghouse Corporation. The initial deployment phase consists of six separators modules and the three laser power amplifier modules consistent with the FY84 reference design with a name plate capacity of 5 million separative work units/yr followed by a full plant activation to approximately 13 million separative work units/yr. The AVLIS Production Plant project team's strategy for deployment schedule analysis focused on three schedule options: engineering limited schedule; authorization limited schedule; and funding limited project schedule. The three deployment schedule options developed by AVLIS project team have been classified in ranges such as an optimistic, rapid/moderate, or moderate/pessimistic based on the probability of meeting the individual schedule option's major milestones or program objectives of enriching uranium by the AVLIS process in an effective cost and schedule manner. 47 figures, 7 tables.

  5. Thermal breeder fuel enrichment zoning

    DOE Patents [OSTI]

    Capossela, Harry J. (Schenectady, NY); Dwyer, Joseph R. (Albany, NY); Luce, Robert G. (Schenectady, NY); McCoy, Daniel F. (Latham, NY); Merriman, Floyd C. (Rotterdam, NY)

    1992-01-01

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

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

    SciTech Connect (OSTI)

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

    2010-12-01

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

  7. Ultraslow Wave Nuclear Burning of Uranium-Plutonium Fissile Medium on Epithermal Neutrons

    E-Print Network [OSTI]

    V. D. Rusov; V. A. Tarasov; M. V. Eingorn; S. A. Chernezhenko; A. A. Kakaev; V. M. Vashchenko; M. E. Beglaryan

    2014-09-29

    For a fissile medium, originally consisting of uranium-238, the investigation of fulfillment of the wave burning criterion in a wide range of neutron energies is conducted for the first time, and a possibility of wave nuclear burning not only in the region of fast neutrons, but also for cold, epithermal and resonance ones is discovered for the first time. For the first time the results of the investigation of the Feoktistov criterion fulfillment for a fissile medium, originally consisting of uranium-238 dioxide with enrichments 4.38%, 2.00%, 1.00%, 0.71% and 0.50% with respect to uranium-235, in the region of neutron energies 0.015-10.0eV are presented. These results indicate a possibility of ultraslow wave neutron-nuclear burning mode realization in the uranium-plutonium media, originally (before the wave initiation by external neutron source) having enrichments with respect to uranium-235, corresponding to the subcritical state, in the regions of cold, thermal, epithermal and resonance neutrons. In order to validate the conclusions, based on the slow wave neutron-nuclear burning criterion fulfillment depending on the neutron energy, the numerical modeling of ultraslow wave neutron-nuclear burning of a natural uranium in the epithermal region of neutron energies (0.1-7.0eV) was conducted for the first time. The presented simulated results indicate the realization of the ultraslow wave neutron-nuclear burning of the natural uranium for the epithermal neutrons.

  8. Measurement and Analysis of Fission Rates in a Spherical Mockup of Uranium and Polyethylene

    E-Print Network [OSTI]

    Tong-Hua, Zhu; Xin-Xin, Lu; Rong, Liu; Zi-Jie, Han; Li, Jiang; Mei, Wang

    2013-01-01

    Measurements of the reaction rate distribution were carried out using two kinds of Plate Micro Fission Chamber(PMFC). The first is a depleted uranium chamber and the second an enriched uranium chamber. The material in the depleted uranium chamber is strictly the same as the material in the uranium assembly. With the equation solution to conduct the isotope contribution correction, the fission rate of 238U and 235U were obtained from the fission rate of depleted uranium and enriched uranium. And then, the fission count of 238U and 235U in an individual uranium shell was obtained. In this work, MCNP5 and continuous energy cross sections ENDF/BV.0 were used for the analysis of fission rate distribution and fission count. The calculated results were compared with the experimental ones. The calculation of fission rate of DU and EU were found to agree with the measured ones within 10% except at the positions in polyethylene region and the two positions near the outer surface. Beacause the fission chamber was not co...

  9. Laser Isotope Enrichment for Medical and Industrial Applications

    SciTech Connect (OSTI)

    Leonard Bond

    2006-07-01

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

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

    SciTech Connect (OSTI)

    NONE

    1994-09-01

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

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

    DOE Patents [OSTI]

    Kreuzmann, Alvin B. (Cincinnati, OH)

    1983-01-01

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

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

    DOE Patents [OSTI]

    Kreuzmann, A.B.

    1982-10-27

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

  13. Decommissioning of U.S. uranium production facilities

    SciTech Connect (OSTI)

    Not Available

    1995-02-01

    From 1980 to 1993, the domestic production of uranium declined from almost 44 million pounds U{sub 3}O{sub 8} to about 3 million pounds. This retrenchment of the U.S. uranium industry resulted in the permanent closing of many uranium-producing facilities. Current low uranium prices, excess world supply, and low expectations for future uranium demand indicate that it is unlikely existing plants will be reopened. Because of this situation, these facilities eventually will have to be decommissioned. The Uranium Mill Tailings and Radiation Control Act of 1978 (UMTRCA) vests the U.S. Environmental Protection Agency (EPA) with overall responsibility for establishing environmental standards for decommissioning of uranium production facilities. UMTRCA also gave the U.S. Nuclear Regulatory Commission (NRC) the responsibility for licensing and regulating uranium production and related activities, including decommissioning. Because there are many issues associated with decommissioning-environmental, political, and financial-this report will concentrate on the answers to three questions: (1) What is required? (2) How is the process implemented? (3) What are the costs? Regulatory control is exercised principally through the NRC licensing process. Before receiving a license to construct and operate an uranium producing facility, the applicant is required to present a decommissioning plan to the NRC. Once the plan is approved, the licensee must post a surety to guarantee that funds will be available to execute the plan and reclaim the site. This report by the Energy Information Administration (EIA) represents the most comprehensive study on this topic by analyzing data on 33 (out of 43) uranium production facilities located in Colorado, Nebraska, New Mexico, South Dakota, Texas, Utah, and Washington.

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

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

    Glossary Home > Nuclear > U.S. Uranium Reserves Estimates U.S. Uranium Reserves Estimates Data for: 2008 Report Released: July 2010 Next Release Date: 2012 Summary The U.S. Energy...

  15. Uranium (VI) solubility in carbonate-free ERDA-6 brine

    SciTech Connect (OSTI)

    Lucchini, Jean-francois; Khaing, Hnin; Reed, Donald T

    2010-01-01

    When present, uranium is usually an element of importance in a nuclear waste repository. In the Waste Isolation Pilot Plant (WIPP), uranium is the most prevalent actinide component by mass, with about 647 metric tons to be placed in the repository. Therefore, the chemistry of uranium, and especially its solubility in the WIPP conditions, needs to be well determined. Long-term experiments were performed to measure the solubility of uranium (VI) in carbonate-free ERDA-6 brine, a simulated WIPP brine, at pC{sub H+} values between 8 and 12.5. These data, obtained from the over-saturation approach, were the first repository-relevant data for the VI actinide oxidation state. The solubility trends observed pointed towards low uranium solubility in WIPP brines and a lack of amphotericity. At the expected pC{sub H+} in the WIPP ({approx} 9.5), measured uranium solubility approached 10{sup -7} M. The objective of these experiments was to establish a baseline solubility to further investigate the effects of carbonate complexation on uranium solubility in WIPP brines.

  16. Process for electrolytically preparing uranium metal

    DOE Patents [OSTI]

    Haas, Paul A. (Knoxville, TN)

    1989-01-01

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

  17. 1.0 MAJOR STUDIES SUPPORTING THIS SCOPING RISK The most important period of past U.S. uranium production spanned from approximately 1948

    E-Print Network [OSTI]

    plants (U.S. DOE/EIA 2003a, 2003b, 2006). Uranium exploration, mining, and ore processing left a legacy of abandoned uranium mines. The major studies supporting this scoping analysis include EPA's 1983 Report to Congress on the Potential Health and Environmental Hazards of Uranium Mine Wastes (U.S. EPA 1983a, b, c

  18. Uranium-titanium-niobium alloy

    DOE Patents [OSTI]

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

    1990-01-01

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

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

    DOE Patents [OSTI]

    Travelli, Armando (Hinsdale, IL)

    1988-01-01

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

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

    DOE Patents [OSTI]

    Travelli, A.

    1985-10-25

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

  1. Corrosion-resistant uranium

    DOE Patents [OSTI]

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

    1981-10-21

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

  2. Uranium Transport Modeling

    SciTech Connect (OSTI)

    Bostick, William D.

    2008-01-15

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

  3. 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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustments (Billion Cubic Feet) Wyoming963 1.969 1.979Coal Consumers inYear JanSalesa. Uranium3.9.3.3. Uranium

  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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustments (Billion Cubic Feet) Wyoming963 1.969 1.979Coal Consumers inYear JanSalesa. Uranium3.9.3.3. Uranium5.

  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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustments (Billion Cubic Feet) Wyoming963 1.969 1.979Coal Consumers inYear JanSalesa. Uranium3.9.3.3.b. Uranium

  6. Paducah Gaseous Diffusion Plant Annual Site Environmental Report summary for 1993

    SciTech Connect (OSTI)

    Not Available

    1994-11-01

    This report contains summaries of the environmental programs at Paducah Gaseous Diffusion Plant, environmental monitoring and the results, and the impact of operations on the environment and the public for 1993. The environmental monitoring program at Paducah includes effluent monitoring and environmental surveillance. Effluent monitoring is measurement of releases as they occur. Contaminants are released through either airborne emissions or liquids discharged from the plant. These releases occur as part of normal site operations, such as cooling water discharged from the uranium enrichment cascade operations or airborne releases from ventilation systems. In the event of system failure, this monitoring provides timely warning so that corrective action can be taken before releases reach an unsafe level. Environmental surveillance tracks the dispersion of materials into the environment after they have been released. This involves the collection of samples from various media, such as water, soil, vegetation, and food crops, and the analysis of these samples for certain radionuclides, chemicals, and metals.

  7. Portsmouth Gaseous Diffusion Plant Annual Site Environmental Report summary for 1993

    SciTech Connect (OSTI)

    Not Available

    1994-11-01

    This report contains summaries of the environmental programs at Paducah Gaseous Diffusion Plant, environmental monitoring and the results, and the impact of operations on the environment and the public for 1993. The environmental monitoring program at Paducah includes effluent monitoring and environmental surveillance. Effluent monitoring is measurement of releases as they occur. Contaminants are released through either airborne emissions or liquids discharged from the plant. These releases occur as part of normal site operations, such as cooling water discharged from the uranium enrichment cascade operations or airborne releases from ventilation systems. In the event of system failure, this monitoring provides timely warning so that corrective action can be taken before releases reach an unsafe level. Environmental surveillance tracks the dispersion of materials into the environment after they have been released. This involves the collection of samples from various media, such as water, soil, vegetation, and food crops, and the analysis of these samples for certain radionuclides, chemicals, and metals.

  8. Characterization activities of the Waste Calcine Facility at the Idaho Chemical Processing Plant

    SciTech Connect (OSTI)

    Feldt, E.G.; Bilson, B.

    1994-12-31

    The Idaho Chemical Processing Plant (ICPP) was established in 1949 at the Idaho National Engineering Laboratory. Its mission was to reprocess nuclear fuel for the recovery of enriched uranium for defense purposes. The ICPP is a large complex encompassing 10 process buildings, 3 fuel storage facilities, 181 support facilities, and 1800 workers. The facilities being deactivated range from contaminated structures that do not meet current code requirements (seismic and electrical) to structures that have had extensive upgrades performed during the 1980s and represent multiple opportunities for reuse due to their seismic qualifications and code compliance status. The facilities declared to be excess and being deactivated at the ICPP include the fuel dissolution cell, the CPP-601/602 complex, the CPP-627 custom dissolution lab, the rare gas plant, the Rover facility, the waste calcine facility, and several small ancillary buildings.

  9. Report of Survey of Oak Ridge Isotope Enrichment (Calutron) Facility Building 9204-3

    Broader source: Energy.gov [DOE]

    The purpose of this document is to report the results of a survey conducted at the Isotope Enrichment Facility (IEF, Calutron, Building 9204-3) on the Y-12 Plant property at the Oak Ridge Site. The...

  10. 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)

    1990-01-01

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

  11. Plutonium Uranium Extraction Plant (PUREX) - Hanford Site

    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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity ofkandz-cm11 Outreach Home Room NewsInformationJesseworkSURVEYI/OPerformancePi Day Pi Day Pi DayPlasmaandAbout Us Projects

  12. Fissile Mass Flow Monitor Implementation for Transparency in HEU Blenddown at the URAL Electrochemical Integrated Plant (UEIP) in Novouralsk

    SciTech Connect (OSTI)

    March-Leuba, J.; Mastal, E.; Powell, D.; Sumner, J.; Uckan, T.; Vines, B.

    1999-07-25

    The Oak Ridge National Laboratory (ORNL) Fissile Mass Flow Monitor (FMFM) was deployed at the Ural Electrochemical Integrated Plant (UEIP) highly enriched uranium (HEU) blending facility in January and February 1999 at Novouralsk in Russia for the DOE HEU Transparency Program. The FMFM provides unattended monitoring of the fissile mass flow of the uranium hexafluoride (UF{sub 6}) gas in the process lines of HEU, the low enriched uranium (LEU) blend stock, and the product LEU (P-LEU) of the blending tee non-intrusively. To do this, uranium-235 (U-235) fissions are induced in the UF{sub 6} by a thermalized and modulated californium-252 (Cf-252) neutron source placed on each process line. A set of detectors, located downstream of source, measure delayed gamma rays emitted by the resulting fission fragments. The observed delay in the time correlated measurement between the source and the detector signal provides the velocity of UF{sub 6} and its amplitude is related to the U- 235 content in UF{sub 6}. An on-line computer controls the source modulator, processes the collected detector data, and displays the results. The UEIP Main and the Reserved process lines were implemented with minor modifications. The FMFM monitors the HEU blending operation by measuring UF{sub 6} flows in the process blending lines, and the traceability of the HEU flow from the blend point to the P-LEU. The detail operational characteristics of the FMFM software (FM2) and the measurement methodology used are presented.

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

    SciTech Connect (OSTI)

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

    2013-02-01

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

  14. Standard test methods for chemical, mass spectrometric, spectrochemical, nuclear, and radiochemical analysis of uranium hexafluoride

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2011-01-01

    1.1 These test methods cover procedures for subsampling and for chemical, mass spectrometric, spectrochemical, nuclear, and radiochemical analysis of uranium hexafluoride UF6. Most of these test methods are in routine use to determine conformance to UF6 specifications in the Enrichment and Conversion Facilities. 1.2 The analytical procedures in this document appear in the following order: Note 1—Subcommittee C26.05 will confer with C26.02 concerning the renumbered section in Test Methods C761 to determine how concerns with renumbering these sections, as analytical methods are replaced with stand-alone analytical methods, are best addressed in subsequent publications. Sections Subsampling of Uranium Hexafluoride 7 - 10 Gravimetric Determination of Uranium 11 - 19 Titrimetric Determination of Uranium 20 Preparation of High-Purity U3O 8 21 Isotopic Analysis 22 Isotopic Analysis by Double-Standard Mass-Spectrometer Method 23 - 29 Determination of Hydrocarbons, Chlorocarbons, and Partially Substitut...

  15. The removal of uranium from acidic media using ion exchange and/or extraction chromatography

    SciTech Connect (OSTI)

    FitzPatrick, J.R.; Schake, B.S.; Murphy, J.; Holmes, K; West, M.H.

    1996-06-01

    The separation and purification of uranium from either nitric acid or hydrochloric acid media can be accomplished by using either solvent extraction or ion-exchange. Over the past two years at Los Alamos, emerging programs are focused on recapturing the expertise required to do limited, small-quantity processing of enriched uranium. During this period of time, we have been investigating ion-addition, waste stream polishing is associated with this effort in order to achieve more complete removal of uranium prior to recycle of the acid. Extraction chromatography has been demonstrated to further polish the uranium from both nitric and hydrochloric acid media thus allowing for a more complete recovery of the actinide material and creation of less waste during the processing steps.

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

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2008-01-01

    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...

  17. Study on Shielding Requirements for Radioactive Waste Transportation in a Mo-99 Production Plant - 13382

    SciTech Connect (OSTI)

    Melo Rego, Maria Eugenia de; Kazumi Sakata, Solange; Vicente, Roberto; Hiromoto, Goro [Nuclear and Energy Research Institute, IPEN-CNEN/SP (Brazil)] [Nuclear and Energy Research Institute, IPEN-CNEN/SP (Brazil)

    2013-07-01

    Brazil is currently planning to produce {sup 99}Mo from fission of low enriched uranium (LEU) targets. The planned end of irradiation activity of {sup 99}Mo is about 185 TBq (5 kCi) per week to meet the present domestic demand of {sup 99m}Tc generators. The radioactive wastes from the production plant will be transferred to a waste treatment facility at the same site. The total activity of the actinides, fission and activation products present in the wastes can be predicted based on the yields of fission and activation data for the irradiation conditions, such as composition and mass of uranium targets, irradiation time, neutron flux, production schedule, etc., which were in principle already established by the project management. The transportation of the wastes from the production plant to the treatment facility will be done by means of special shielded packages. An assessment of the shielding required for the packages has been done and the results are presented here, aiming at contributing to the design of the waste management facility for the {sup 99}Mo production plant. (authors)

  18. Controlling uranium reactivity March 18, 2008

    E-Print Network [OSTI]

    Meyer, Karsten

    March 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 research groups have been involved in utilizing the large size and unique reactivity of the uranium atom

  19. Method of preparation of uranium nitride

    DOE Patents [OSTI]

    Kiplinger, Jaqueline Loetsch; Thomson, Robert Kenneth James

    2013-07-09

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

  20. Enrichment of light hydrocarbon mixture

    DOE Patents [OSTI]

    Yang, Dali (Los Alamos, NM); Devlin, David (Santa Fe, NM); Barbero, Robert S. (Santa Cruz, NM); Carrera, Martin E. (Naperville, IL); Colling, Craig W. (Warrenville, IL)

    2011-11-29

    Light hydrocarbon enrichment is accomplished using a vertically oriented distillation column having a plurality of vertically oriented, nonselective micro/mesoporous hollow fibers. Vapor having, for example, both propylene and propane is sent upward through the distillation column in between the hollow fibers. Vapor exits neat the top of the column and is condensed to form a liquid phase that is directed back downward through the lumen of the hollow fibers. As vapor continues to ascend and liquid continues to countercurrently descend, the liquid at the bottom of the column becomes enriched in a higher boiling point, light hydrocarbon (propane, for example) and the vapor at the top becomes enriched in a lower boiling point light hydrocarbon (propylene, for example). The hollow fiber becomes wetted with liquid during the process.

  1. Enrichment of light hydrocarbon mixture

    DOE Patents [OSTI]

    Yang; Dali (Los Alamos, NM); Devlin, David (Santa Fe, NM); Barbero, Robert S. (Santa Cruz, NM); Carrera, Martin E. (Naperville, IL); Colling, Craig W. (Warrenville, IL)

    2010-08-10

    Light hydrocarbon enrichment is accomplished using a vertically oriented distillation column having a plurality of vertically oriented, nonselective micro/mesoporous hollow fibers. Vapor having, for example, both propylene and propane is sent upward through the distillation column in between the hollow fibers. Vapor exits neat the top of the column and is condensed to form a liquid phase that is directed back downward through the lumen of the hollow fibers. As vapor continues to ascend and liquid continues to countercurrently descend, the liquid at the bottom of the column becomes enriched in a higher boiling point, light hydrocarbon (propane, for example) and the vapor at the top becomes enriched in a lower boiling point light hydrocarbon (propylene, for example). The hollow fiber becomes wetted with liquid during the process.

  2. A more accurate and penetrating method to measure the enrichment and mass of UF6 storage containers using passive neutron self-interrogation

    SciTech Connect (OSTI)

    Menlove, Howard O; Swinhoe, Martyn T; Miller, Karen A

    2010-01-01

    This paper describes an unattended mode neutron measurement that can provide the enrichment of the uranium in UF{sub 6} cylinders. The new passive neutron measurement provides better penetration into the uranium mass than prior gamma-ray enrichment measurement methods. The Passive Neutron Enrichment Monitor (PNEM) provides a new measurement technique that uses passive neutron totals and coincidence counting together with neutron self-interrogation to measure the enrichment in the cylinders. The measurement uses the neutron rates from two detector pods. One of the pods has a bare polyethylene surface next to the cylinder and the other polyethylene surface is covered with Cd to prevent thermal neutrons from returning to the cylinder. The primary neutron source from the enriched UF{sub 6} is the alpha-particle decay from the {sub 234}U that interacts with the fluorine to produce random neutrons. The singles neutron counting rate is dominated by the {sub 234}U neutrons with a minor contribution from the induced fissions in the {sub 235}U. However, the doubles counting rate comes primarily from the induced fissions (i.e., multiplication) in the {sub 235}U in enriched uranium. The PNEM concept makes use of the passive neutrons that are initially produced from the {sub 234}U reactions that track the {sub 235}U enrichment during the enrichment process. The induced fission reactions from the thermal-neutron albedo are all from the {sub 235}U and provide a measurement of the {sub 235}U. The Cd ratio has the desirable feature that all of the thermal-neutron-induced fissions in {sub 235}U are independent of the original neutron source. Thus, the ratio is independent of the uranium age, purity, and prior reactor history.

  3. 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)

    2006-09-05

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

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

    SciTech Connect (OSTI)

    Not Available

    1993-07-01

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

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

    SciTech Connect (OSTI)

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

    2013-07-01

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

  6. AVLIS enrichment of medical isotopes

    SciTech Connect (OSTI)

    Haynam, C.A.; Scheibner, K.F.; Stern, R.C.; Worden, E.F.

    1996-12-31

    Under the Sponsorship of the United states Enrichment Corporation (USEC), we are currently investigating the large scale separation of several isotopes of medical interest using atomic vapor isotope separation (AVLIS). This work includes analysis and experiments in the enrichment of thallium 203 as a precursor to the production of thallium 201 used in cardiac imaging following heart attacks, on the stripping of strontium 84 from natural strontium as precursor to the production of strontium 89, and on the stripping of lead 210 from lead used in integrated circuits to reduce the number of alpha particle induced logic errors.

  7. Uranium in prehistoric Indian pottery 

    E-Print Network [OSTI]

    Filberth, Ernest William

    1976-01-01

    URANIUM IN PREHISTORIC INDIAN POTTERY A Thesis by ERNEST WILLIAM FILBERTH Submitted to the Graduate College of Texas A&M University in partial fulfillment of the requirement for the degree of MASTER OF SCIENCE December 1976 Major Subject...: Chemistry URANIUM IN PREHISTORIC INDIAN POTTERY A Thesis by ERNEST WILLIAM FILBERTH Approved as to style and content by: (Chairman of Committee) (Head of Department) (Member) (Membe (Member) (Member) December 1976 ABSTRACT Uranium in Prehistoric...

  8. Expansion capacity of an SX unit in uranium process pilot tests

    SciTech Connect (OSTI)

    Courtaud, B.; Auger, F.; Morel, P.

    2008-07-01

    The rising price of uranium has led uranium producers to increase their plant capacity. The new project proposed to increase capacity is based on processing low-grade uranium by heap leaching. It is necessary to modify the plant, particularly the solvent extraction unit, to handle the increased flow. The goal of our study is to determine the minimal changes necessary to process the whole flow. Several stages have been carried out (i) thermodynamic modelling of the solvent extraction process to determine the capacities of the SX plant and the impact of the modification and (ii) pilot tests at the plant of the different configurations proposed by modelling. This paper presents results of the pilot tests performed at the plant. (authors)

  9. Ultraslow Wave Nuclear Burning of Uranium-Plutonium Fissile Medium on Epithermal Neutrons

    E-Print Network [OSTI]

    Rusov, V D; Eingorn, M V; Chernezhenko, S A; Kakaev, A A

    2014-01-01

    For a fissile medium, originally consisting of uranium-238, the investigation of fulfillment of the wave burning criterion in a wide range of neutron energies is conducted for the first time, and a possibility of wave nuclear burning not only in the region of fast neutrons, but also for cold, epithermal and resonance ones is discovered for the first time. For the first time the results of the investigation of the Feoktistov criterion fulfillment for a fissile medium, originally consisting of uranium-238 dioxide with enrichments 4.38%, 2.00%, 1.00%, 0.71% and 0.50% with respect to uranium-235, in the region of neutron energies 0.015-10.0eV are presented. These results indicate a possibility of ultraslow wave neutron-nuclear burning mode realization in the uranium-plutonium media, originally (before the wave initiation by external neutron source) having enrichments with respect to uranium-235, corresponding to the subcritical state, in the regions of cold, thermal, epithermal and resonance neutrons. In order to...

  10. Engineering assessment of inactive uranium mill tailings

    SciTech Connect (OSTI)

    Not Available

    1981-07-01

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

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

    Office of Scientific and Technical Information (OSTI)

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

  12. Method and apparatus for measuring enrichment of UF6

    DOE Patents [OSTI]

    Hill, Thomas Roy (Santa Fe, NM); Ianakiev, Kiril Dimitrov (Los Alamos, NM)

    2011-06-07

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

  13. Uranium in the Oatman Creek granite of Central Texas and its economic potential 

    E-Print Network [OSTI]

    Conrad, Curtis Paul

    1982-01-01

    50 Fission track pattern corresponding to 5C showing uranium concentration 1n cracks which have been infilled with iron ox1de material. Note the extreme enrichment of uran1um in the large 1ron oxide mineral. 50 Photom1crograph 1n plane polarized...buted throughout the plagioclase grain along fractures cutt1ng perpendicular to the cleavage traces. 56 7C Fission track pattern corresponding to 7A showing uranium enr1chment 1n the fractures. 56 BA Photomicrograph in plane polar1zed light of a...

  14. 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)

    2001-01-01

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

  15. Spain`s uranium industry

    SciTech Connect (OSTI)

    Ferguson, M.P.

    1992-05-01

    Spain currently operates nine nuclear reactors totalling over 7,100 MWe of capacity, contributing about one-third of all electricity generated in Spain. Four reactors at advanced stages of construction remain mothballed as the result of a government-imposed moratorium, and a fire at Vandellos 1 in 1989 led to its premature closure and to a revival of anti-nuclear sentiment in the country. In the new national energy plan, which was sent to the Spanish Parliament on July 25, 1991, Spain opted to continue the nuclear moratorium that began in 1984 and rely upon conservation measures, additional natural gas imports, and electricity imports to meet expected demand. Under the new plan, nuclear power`s share of Spain`s total installed electrical generating capacity will fall from about 17 percent in 1990, to approximately 14 percent by the end of the century, as only the current nuclear facilities will continue to operate and no new nuclear plants will be built. Spain`s integration into the European Community also is affecting the country`s energy plans, prompting consolidation within the Spanish electricity sector in order to be more competitive in Europe. To supply the existing reactors, the government is supporting a major expansion of the country`s domestic uranium industry.

  16. Method for monitoring stack gases for uranium activity

    DOE Patents [OSTI]

    Beverly, Claude R. (Paducah, KY); Ernstberger, Harold G. (Paducah, KY)

    1988-01-01

    A method for monitoring the stack gases of a purge cascade of a gaseous diffusion plant for uranium activity. A sample stream is taken from the stack gases and contacted with a volume of moisture-laden air for converting trace levels of uranium hexafluoride, if any, in the stack gases into particulate uranyl fluoride. A continuous strip of filter paper from a supply roll is passed through this sampling stream to intercept and gather any uranyl fluoride in the sampling stream. This filter paper is then passed by an alpha scintillation counting device where any radioactivity on the filter paper is sensed so as to provide a continuous monitoring of the gas stream for activity indicative of the uranium content in the stack gases.

  17. Method for monitoring stack gases for uranium activity

    DOE Patents [OSTI]

    Beverly, C.R.; Ernstberger, E.G.

    1985-07-03

    A method for monitoring the stack gases of a purge cascade of gaseous diffusion plant for uranium activity. A sample stream is taken from the stack gases and contacted with a volume of moisture-laden air for converting trace levels of uranium hexafluoride, if any, in the stack gases into particulate uranyl fluoride. A continuous strip of filter paper from a supply roll is passed through this sampling stream to intercept and gather any uranyl fluoride in the sampling stream. This filter paper is then passed by an alpha scintillation counting device where any radioactivity on the filter paper is sensed so as to provide a continuous monitoring of the gas stream for activity indicative of the uranium content in the stack gases. 1 fig.

  18. Pulsed CO laser for isotope separation of uranium

    SciTech Connect (OSTI)

    Baranov, Igor Y.; Koptev, Andrey V. [Rocket-Space Technics Department, Baltic State Technical University, 1, 1st Krasnoarmeyskaya st.,St. Petersburg, 190005 (Russian Federation)

    2012-07-30

    This article proposes a technical solution for using a CO laser facility for the industrial separation of uranium used in the production of fuel for nuclear power plants, employing a method of laser isotope separation of uranium with condensation repression in a free jet. The laser operation with nanosecond pulse irradiation can provide an acceptable efficiency in the separating unit and a high efficiency of the laser with the wavelength of 5.3 {mu}m. In the present work we also introduce a calculation model and define the parameters of a mode-locked CO laser with a RF discharge in the supersonic stream. The average pulsed CO laser power of 3 kW is sufficient for efficient industrial isotope separation of uranium in one stage.

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

    SciTech Connect (OSTI)

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

    2011-01-01

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

  20. 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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustments (Billion Cubic Feet) Wyoming963 1.969 1.979Coal Consumers inYear JanSalesa. Uranium purchased by

  1. 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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustments (Billion Cubic Feet) Wyoming963 1.969 1.979Coal Consumers inYear JanSalesa. Uranium purchased byb.

  2. 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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustments (Billion Cubic Feet) Wyoming963 1.969 1.979Coal Consumers inYear JanSalesa. Uranium purchased byb.S2.

  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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustments (Billion Cubic Feet) Wyoming963 1.969 1.979Coal Consumers inYear JanSalesa. Uranium purchased

  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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustments (Billion Cubic Feet) Wyoming963 1.969 1.979Coal Consumers inYear JanSalesa. Uranium purchasedb.

  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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustments (Billion Cubic Feet) Wyoming963 1.969 1.979Coal Consumers inYear JanSalesa. Uranium purchasedb.4.

  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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustments (Billion Cubic Feet) Wyoming963 1.969 1.979Coal Consumers inYear JanSalesa. Uranium purchasedb.4..

  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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustments (Billion Cubic Feet) Wyoming963 1.969 1.979Coal Consumers inYear JanSalesa. Uranium purchasedb.4..0.

  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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustments (Billion Cubic Feet) Wyoming963 1.969 1.979Coal Consumers inYear JanSalesa. Uranium

  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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustments (Billion Cubic Feet) Wyoming963 1.969 1.979Coal Consumers inYear JanSalesa. Uranium3. Deliveries of

  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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustments (Billion Cubic Feet) Wyoming963 1.969 1.979Coal Consumers inYear JanSalesa. Uranium3. Deliveries of4.

  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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustments (Billion Cubic Feet) Wyoming963 1.969 1.979Coal Consumers inYear JanSalesa. Uranium3. Deliveries

  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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustments (Billion Cubic Feet) Wyoming963 1.969 1.979Coal Consumers inYear JanSalesa. Uranium3. Deliveries6.

  13. 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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustments (Billion Cubic Feet) Wyoming963 1.969 1.979Coal Consumers inYear JanSalesa. Uranium3. Deliveries6.7.

  14. 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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustments (Billion Cubic Feet) Wyoming963 1.969 1.979Coal Consumers inYear JanSalesa. Uranium3.

  15. 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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustments (Billion Cubic Feet) Wyoming963 1.969 1.979Coal Consumers inYear JanSalesa. Uranium3.9. Foreign

  16. 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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustments (Billion Cubic Feet) Wyoming963 1.969 1.979Coal Consumers inYear JanSalesa. Uranium3.9. Foreign.

  17. 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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustments (Billion Cubic Feet) Wyoming963 1.969 1.979Coal Consumers inYear JanSalesa. Uranium3.9. Foreign.0.

  18. 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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustments (Billion Cubic Feet) Wyoming963 1.969 1.979Coal Consumers inYear JanSalesa. Uranium3.9. Foreign.0.1.

  19. 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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustments (Billion Cubic Feet) Wyoming963 1.969 1.979Coal Consumers inYear JanSalesa. Uranium3.9.

  20. 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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustments (Billion Cubic Feet) Wyoming963 1.969 1.979Coal Consumers inYear JanSalesa. Uranium3.9.3. Inventories

  1. 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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustments (Billion Cubic Feet) Wyoming963 1.969 1.979Coal Consumers inYear JanSalesa. Uranium3.9.3.

  2. 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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustments (Billion Cubic Feet) Wyoming963 1.969 1.979Coal Consumers inYear JanSalesa. Uranium3.9.3.3.

  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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustments (Billion Cubic Feet) Wyoming963 1.969 1.979Coal Consumers inYear JanSalesa. Uranium3.9.3.3.b.

  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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustments (Billion Cubic Feet) Wyoming963 1.969 1.979Coal Consumers inYear JanSalesa. Uranium3.9.3.3.b.8.

  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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustments (Billion Cubic Feet) Wyoming963 1.969 1.979Coal Consumers inYear JanSalesa. Uranium3.9.3.3.b.8.9.

  6. Domestic Uranium Production Report

    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: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantityBonneville Power Administration would like submit theCovalentLaboratory |Sector Full reportTown2008 Final May1. U.S. uranium

  7. Student Science Enrichment Training Program

    SciTech Connect (OSTI)

    Sandhu, S.S.

    1990-12-31

    Funds are requested for the science enrichment training program (emphasis on chemistry and computer science), which will be held at Claflin College during the 1990 and 1991 summers, concomitant with summer school. The thirty participants will include high school students and some college freshmen; the students will come from rural South Carolina schools with limited science and computer facilities. Focus will be on high ability minority students.

  8. Impacts of uranium-utilization improvements on light water reactor radionuclide releases

    SciTech Connect (OSTI)

    Aaberg, R.L.

    1981-08-01

    This report discusses potential changes to radionuclide releases as a result of uranium-saving plant modifications and altered operating practices. Only releases to the environment from routine operation are considered; releases resulting from abnormal conditions outside the technical specifications covering plant operation are not considered.

  9. 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

  10. SHEEP MOUNTAIN URANIUM PROJECT CROOKS GAP, WYOMING

    E-Print Network [OSTI]

    SHEEP MOUNTAIN URANIUM PROJECT CROOKS GAP, WYOMING US EPA Project Meeting April 7 2011April 7, 2011/Titan Uranium, VP Development · Deborah LebowAal/EPA Region 8 Air Program Introduction to Titan Uranium USA;PROJECT OVERVIEW ·Site Location·Site Location ·Fremont , Wyoming ·Existing Uranium Mine Permit 381C

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

    DOE Patents [OSTI]

    Wiencek, Thomas C. (Orland Park, IL); Matos, James E. (Oak Park, IL); Hofman, Gerard L. (Downers Grove, IL)

    2000-12-12

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

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

    DOE Patents [OSTI]

    Wiencek, Thomas C. (Orland Park, IL); Matos, James E. (Oak Park, IL); Hofman, Gerard L. (Downers Grove, IL)

    1997-01-01

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

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

    DOE Patents [OSTI]

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

    1997-03-25

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

  14. Examination of the proposed conversion of the U.S. Navy nuclear fleet from highly enriched Uranium to low enriched Uranium

    E-Print Network [OSTI]

    McCord, Cameron (Cameron Liam)

    2013-01-01

    .The Treaty on the Non-Proliferation of Nuclear Weapons creates a loophole that allows a non-nuclear-weapon country to avoid international safeguards governing fissile materials if it claims that the materials will be used ...

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

    SciTech Connect (OSTI)

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

    1994-08-01

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

  16. Solvent-extraction and purification of uranium(VI) and molybdenum(VI) by tertiary amines from acid leach solutions

    SciTech Connect (OSTI)

    La Gamma, Ana M.G.; Becquart, Elena T.; Chocron, Mauricio

    2008-07-01

    Considering international interest in the yellow-cake price, Argentina is seeking to exploit new uranium ore bodies and processing plants. A study of similar plants would suggest that solvent- extraction with Alamine 336 is considered the best method for the purification and concentration of uranium present in leaching solutions. In order to study the purification of these leach liquors, solvent-extraction tests under different conditions were performed with simulated solutions which containing molybdenum and molybdenum-uranium mixtures. Preliminary extraction tests carried out on mill acid-leaching liquors are also presented. (authors)

  17. Helium on Venus: Implications for uranium and thorium

    E-Print Network [OSTI]

    Prather, MJ; Mcelroy, MB

    1983-01-01

    Implications for Uranium and Thorium Abstract. Helium isa wide range of uranium and thorium abundances. simi· lar toof crustal uranium and thorium. Studies of helium in Earth's

  18. Excess Uranium Inventory Management Plan | Department of Energy

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

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

  19. THE HIGH TEMPERATURE BEHAVIOR OF METALLIC INCLUSIONS IN URANIUM DIOXIDE.

    E-Print Network [OSTI]

    Yang, Rosa Lu.

    2010-01-01

    Products in Irradiated Uranium Dioxide," UKAEA Report AERE-OF METALLIC INCLUSIONS IN URANIUM DIOXIDE Rosa Lu Yang (Chemical State of Irradiated Uranium- Plutonium Oxide Fuel

  20. Occupational safety data and casualty rates for the uranium fuel cycle. [Glossaries

    SciTech Connect (OSTI)

    O'Donnell, F.R.; Hoy, H.C.

    1981-10-01

    Occupational casualty (injuries, illnesses, fatalities, and lost workdays) and production data are presented and used to calculate occupational casualty incidence rates for technologies that make up the uranium fuel cycle, including: mining, milling, conversion, and enrichment of uranium; fabrication of reactor fuel; transportation of uranium and fuel elements; generation of electric power; and transmission of electric power. Each technology is treated in a separate chapter. All data sources are referenced. All steps used to calculate normalized occupational casualty incidence rates from the data are presented. Rates given include fatalities, serious cases, and lost workdays per 100 man-years worked, per 10/sup 12/ Btu of energy output, and per other appropriate units of output.

  1. Gamma-spectrometric determination of 232U in uranium-bearing materials

    E-Print Network [OSTI]

    Jozsef Zsigrai; Cong Tam Nguyen; Andriy Berlizov

    2015-01-19

    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.

  2. Gamma-spectrometric determination of 232U in uranium-bearing materials

    E-Print Network [OSTI]

    Jozsef Zsigrai; Tam Cong Nguyen; Andrey Berlizov

    2015-08-07

    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.

  3. Gamma-spectrometric determination of 232U in uranium-bearing materials

    E-Print Network [OSTI]

    Zsigrai, Jozsef; Berlizov, Andriy

    2015-01-01

    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.

  4. Energy Department Selects Global Laser Enrichment for Future...

    Energy Savers [EERE]

    Site November 27, 2013 - 12:00pm Addthis Workers inspect cylinders containing depleted uranium hexafluoride. Workers inspect cylinders containing depleted uranium hexafluoride....

  5. Uranium hexafluoride bibliography

    SciTech Connect (OSTI)

    Burnham, S.L.

    1988-01-01

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

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

    SciTech Connect (OSTI)

    Busch, Robert D.; Bledsoe, Keith C

    2011-01-01

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

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

    SciTech Connect (OSTI)

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

    1980-10-01

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

  8. 2014 Uranium Marketing Annual Report

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

    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| Open Energy Informationmonthly gasoline price to fall toUranium Marketing Annual Report 2014 Uranium 201457 201425.

  9. 2014 Uranium Marketing Annual Report

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

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  10. 2014 Uranium Marketing Annual Report

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

    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| Open Energy Informationmonthly gasoline price to fall toUranium Marketing Annual Report 2014 Uranium 201457Feed

  11. 2014 Uranium Marketing Annual Report

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

    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| Open Energy Informationmonthly gasoline price to fall toUranium Marketing Annual Report 2014 Uranium

  12. 2014 Uranium Marketing Annual Report

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

    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| Open Energy Informationmonthly gasoline price to fall toUranium Marketing Annual Report 2014 Uranium17. Purchases of

  13. 2014 Uranium Marketing Annual Report

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

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  14. 2014 Uranium Marketing Annual Report

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

    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| Open Energy Informationmonthly gasoline price to fall toUranium Marketing Annual Report 2014 Uranium17.

  15. 2014 Uranium Marketing Annual Report

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

    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| Open Energy Informationmonthly gasoline price to fall toUranium Marketing Annual Report 20144. Uranium sellers to

  16. 2014 Uranium Marketing Annual Survey

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

    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| Open Energy Informationmonthly gasoline price to fall toUranium Marketing Annual Report 20144. Uranium sellers to57.

  17. Laser induced phosphorescence uranium analysis

    DOE Patents [OSTI]

    Bushaw, Bruce A. (Kennewick, WA)

    1986-01-01

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

  18. Laser induced phosphorescence uranium analysis

    DOE Patents [OSTI]

    Bushaw, B.A.

    1983-06-10

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

  19. Study of Chemical Changes in Uranium Oxyfluoride Particles Progress Report March - October 2009

    SciTech Connect (OSTI)

    Kips, R; Kristo, M; Hutcheon, I

    2009-11-22

    Nuclear forensics relies on the analysis of certain sample characteristics to determine the origin and history of a nuclear material. In the specific case of uranium enrichment facilities, it is the release of trace amounts of uranium hexafluoride (UF{sub 6}) gas - used for the enrichment of uranium - that leaves a process-characteristic fingerprint. When UF{sub 6} gas interacts with atmospheric moisture, uranium oxyfluoride particles or particle agglomerates are formed with sizes ranging from several microns down to a few tens of nanometers. These particles are routinely collected by safeguards organizations, such as the International Atomic Energy Agency (IAEA), allowing them to verify whether a facility is compliant with its declarations. Spectrometric analysis of uranium particles from UF{sub 6} hydrolysis has revealed the presence of both particles that contain fluorine, and particles that do not. It is therefore assumed that uranium oxyfluoride is unstable, and decomposes to form uranium oxide. Understanding the rate of fluorine loss in uranium oxyfluoride particles, and the parameters that control it, may therefore contribute to placing boundaries on the particle's exposure time in the environment. Expressly for the purpose of this study, we prepared a set of uranium oxyfluoride particles at the Institute for Reference Materials and Measurements (EU-JRC-IRMM) from a static release of UF{sub 6} in a humid atmosphere. The majority of the samples was stored in controlled temperature, humidity and lighting conditions. Single particles were characterized by a suite of micro-analytical techniques, including NanoSIMS, micro-Raman spectrometry (MRS), scanning (SEM) and transmission (TEM) electron microscopy, energy-dispersive X-ray spectrometry (EDX) and focused ion beam (FIB). The small particle size was found to be the main analytical challenge. The relative amount of fluorine, as well as the particle chemical composition and morphology were determined at different stages in the ageing process, and immediately after preparation. This report summarizes our most recent findings for each of the analytical techniques listed above, and provides an outlook on what remains to be resolved. Additional spectroscopic and mass spectrometric measurements were carried out at Pacific Northwest National Laboratory, but are not included in this summary.

  20. 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

    2007-01-01

    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