National Library of Energy BETA

Sample records for uranium fuel enrichment

  1. 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...

  2. Uranium enrichment

    SciTech Connect (OSTI)

    Not Available

    1991-04-01

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

  3. Uranium 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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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  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  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. 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.

  16. 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...

  17. 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...

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

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

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

  19. 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.

  20. 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.

  1. 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

  2. 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.

  3. 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.

  4. 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.

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

    Broader source: Energy.gov [DOE]

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

  6. 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 ...

  7. 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.

  8. 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.

  9. highly enriched uranium

    National Nuclear Security Administration (NNSA)

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  10. 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.

  11. 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.

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

    SciTech Connect (OSTI)

    Michael Tyacke; Dr. Igor Bolshinsky; Jeff Chamberlin

    2008-07-01

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

  13. Quadrilateral Cooperation on High-density Low-enriched Uranium Fuel

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity ofkandz-cm11 Outreach Home RoomPreservation of Fe(II) by Carbon-RichProton Delivery and Removal PurchasingProduction: Fact

  14. 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.

  15. 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 ...

  16. 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

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

    Energy Savers [EERE]

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

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

  19. 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 ...

  20. 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.

  1. 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.

  2. 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.

  3. 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 ...

  4. 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 ...

  5. 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 ...

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

    SciTech Connect (OSTI)

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

    1986-06-01

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

  7. 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.

  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. 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.

  10. Uranium enrichment management review: summary of analysis

    SciTech Connect (OSTI)

    Not Available

    1981-01-01

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

  11. 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.

  12. 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

  13. 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

  14. 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.

  15. 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.

  16. 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...

  17. 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...

  18. 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...

  19. 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.

  20. 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.

  1. 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).

  2. 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...

  3. 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.

  4. 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.

  5. 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.

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

    Office of Environmental Management (EM)

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

  7. 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.

  8. 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.

  9. 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.

  10. 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.

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

    SciTech Connect (OSTI)

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

    2013-07-01

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

  12. 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.

  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. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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

  19. 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.

  20. 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

  1. 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).

  2. 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

  3. 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

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  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. 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.

  10. 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.

  11. 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).

  12. 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.

  13. Fuel cycle optimization of thorium and uranium fueled PWR systems

    E-Print Network [OSTI]

    Garel, Keith Courtnay

    1977-01-01

    The burnup neutronics of uniform PWR lattices are examined with respect to reduction of uranium ore requirements with an emphasis on variation of the fuel-to-moderator ratio

  14. 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.

  15. 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.

  16. 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...

  17. 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.

  18. Nuclear Fuel Facts: Uranium | Department of Energy

    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 DeliciousMathematicsEnergyInterested Parties -DepartmentAvailable forSite |n t e OfficeResearch andFacts: Uranium

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

    SciTech Connect (OSTI)

    Parker, Frank L. [Vanderbilt University (United States)

    2012-07-01

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

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

    SciTech Connect (OSTI)

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

    2014-02-01

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

  1. 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.

  2. Colloids generation from metallic uranium fuel

    SciTech Connect (OSTI)

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

    2000-07-20

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

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

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

    Broader source: Energy.gov [DOE]

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

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  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. Feasibility study on AFR-100 fuel conversion from uranium-based fuel to thorium-based fuel

    SciTech Connect (OSTI)

    Heidet, F.; Kim, T.; Grandy, C. (Nuclear Engineering Division)

    2012-07-30

    Although thorium has long been considered as an alternative to uranium-based fuels, most of the reactors built to-date have been fueled with uranium-based fuel with the exception of a few reactors. The decision to use uranium-based fuels was initially made based on the technology maturity compared to thorium-based fuels. As a result of this experience, lot of knowledge and data have been accumulated for uranium-based fuels that made it the predominant nuclear fuel type for extant nuclear power. However, following the recent concerns about the extent and availability of uranium resources, thorium-based fuels have regained significant interest worldwide. Thorium is more abundant than uranium and can be readily exploited in many countries and thus is now seen as a possible alternative. As thorium-based fuel technologies mature, fuel conversion from uranium to thorium is expected to become a major interest in both thermal and fast reactors. In this study the feasibility of fuel conversion in a fast reactor is assessed and several possible approaches are proposed. The analyses are performed using the Advanced Fast Reactor (AFR-100) design, a fast reactor core concept recently developed by ANL. The AFR-100 is a small 100 MW{sub e} reactor developed under the US-DOE program relying on innovative fast reactor technologies and advanced structural and cladding materials. It was designed to be inherently safe and offers sufficient margins with respect to the fuel melting temperature and the fuel-cladding eutectic temperature when using U-10Zr binary metal fuel. Thorium-based metal fuel was preferred to other thorium fuel forms because of its higher heavy metal density and it does not need to be alloyed with zirconium to reduce its radiation swelling. The various approaches explored cover the use of pure thorium fuel as well as the use of thorium mixed with transuranics (TRU). Sensitivity studies were performed for the different scenarios envisioned in order to determine the best core performance characteristics for each of them. With the exception of the fuel type and enrichment, the reference AFR-100 core design characteristics were kept unchanged, including the general core layout and dimensions, assembly dimensions, materials and power rating. In addition, the mass of {sup 235}U required was kept within a reasonable range from that of the reference AFR-100 design. The core performance characteristics, kinetics parameters and reactivity feedback coefficients were calculated using the ANL suite of fast reactor analysis code systems. Orifice design calculations and the steady-state thermal-hydraulic analyses were performed using the SE2-ANL code. The thermal margins were evaluated by comparing the peak temperatures to the design limits for parameters such as the fuel melting temperature and the fuel-cladding eutectic temperature. The inherent safety features of AFR-100 cores proposed were assessed using the integral reactivity parameters of the quasi-static reactivity balance analysis. The design objectives and requirements, the computation methods used as well as a description of the core concept are provided in Section 2. The three major approaches considered are introduced in Section 3 and the neutronics performances of those approaches are discussed in the same section. The orifice zoning strategies used and the steady-state thermal-hydraulic performance are provided in Section 4. The kinetics and reactivity coefficients, including the inherent safety characteristics, are provided in Section 5, and the Conclusions in Section 6. Other scenarios studied and sensitivity studies are provided in the Appendix section.

  2. 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

  3. 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.

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

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

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  5. 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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  6. 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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  7. Highly Enriched Uranium Materials Facility, Major Design Changes

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

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  8. 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

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

    National Nuclear Security Administration (NNSA)

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

    National Nuclear Security Administration (NNSA)

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  11. Highly Enriched Uranium Materials Facility | National Nuclear Security

    National Nuclear Security Administration (NNSA)

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  12. Toxic Substances Control Act Uranium Enrichment Federal Facility Compliance

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

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

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

    SciTech Connect (OSTI)

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

    2012-07-01

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

  16. Neutron source, linear-accelerator fuel enricher and regenerator and associated methods

    DOE Patents [OSTI]

    Steinberg, Meyer (Huntington Station, NY); Powell, James R. (Shoreham, NY); Takahashi, Hiroshi (Setauket, NY); Grand, Pierre (Blue Point, NY); Kouts, Herbert (Brookhaven, NY)

    1982-01-01

    A device for producing fissile material inside of fabricated nuclear elements so that they can be used to produce power in nuclear power reactors. Fuel elements, for example, of a LWR are placed in pressure tubes in a vessel surrounding a liquid lead-bismuth flowing columnar target. A linear-accelerator proton beam enters the side of the vessel and impinges on the dispersed liquid lead-bismuth columns and produces neutrons which radiate through the surrounding pressure tube assembly or blanket containing the nuclear fuel elements. These neutrons are absorbed by the natural fertile uranium-238 elements and are transformed to fissile plutonium-239. The fertile fuel is thus enriched in fissile material to a concentration whereby they can be used in power reactors. After use in the power reactors, dispensed depleted fuel elements can be reinserted into the pressure tubes surrounding the target and the nuclear fuel regenerated for further burning in the power reactor.

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

    Broader source: Energy.gov [DOE]

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

  18. 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

  19. 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.

  20. Fuel Cycle Science & Technology | Nuclear Science | ORNL

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

    Radiochemical Separation & Processing Recycle & Waste Management Uranium Enrichment Used Nuclear Fuel Storage, Transportation, and Disposal Fusion Nuclear Science Isotope...

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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)

  6. 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.

  7. Maximum Fuel Utilization in Advanced Fast Reactors without Actinides Separation

    E-Print Network [OSTI]

    Heidet, Florent

    2010-01-01

    Albright, D. , Plutonium and highly enriched uranium, 1996 :and swelling in uranium-plutonium mixed nitride fuels.products and to extract plutonium or any other actinide from

  8. New Tool for Proliferation Resistance Evaluation Applied to Uranium and Thorium Fueled Fast Reactor Fuel Cycles 

    E-Print Network [OSTI]

    Metcalf, Richard R.

    2010-07-14

    reactor cycle and one scenario involves theft from a PUREX facility in a LWR cycle. The FBRFC was evaluated with uranium-plutonium fuel and a second time using thorium-uranium fuel. These diversion scenarios were tested with both uniform and expert weights...

  9. 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.

  10. 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.

  11. 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.

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

    SciTech Connect (OSTI)

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

    2013-10-01

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

  13. Depleted uranium as a backfill for nuclear fuel waste package

    DOE Patents [OSTI]

    Forsberg, C.W.

    1998-11-03

    A method is described for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package. 6 figs.

  14. Depleted uranium as a backfill for nuclear fuel waste package

    DOE Patents [OSTI]

    Forsberg, Charles W. (Oak Ridge, TN)

    1998-01-01

    A method for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package.

  15. 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.

  16. 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.

  17. 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

  18. 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.

  19. 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

  20. 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

  1. 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.

  2. Empirical modeling of uranium nitride fuels 

    E-Print Network [OSTI]

    Brozak, Daniel Edward

    1987-01-01

    of this work was to develop an irradiation performance data base for nitride fuels and to provide empirical modeling capabilities for fuel swelling and fission gas release in nitride fuels. The nitride fuels data base represents the most extensive effort... the formation of a empirical fit based upon consistent data. The forms of the recommended correlations for fuel swelling and fission gas release are shown on the following page; AV/V = a(Tb) (BUc)(FDd) (SDe) FGR = a(T )(BU )(FD ) where: AV/V FGR T BU FD...

  3. The concept of the use of recycled uranium for increasing the degree of security of export deliveries of fuel for light-water reactors

    SciTech Connect (OSTI)

    Alekseev, P. N.; Ivanov, E. A.; Nevinitsa, V. A.; Ponomarev-Stepnoi, N. N.; Rumyantsev, A. N.; Shmelev, V. M. [Russian Research Center Kurchatov Institute (Russian Federation); Borisevich, V. D.; Smirnov, A. Yu.; Sulaberidze, G. A. [National Nuclear Research University MEPhI (Russian Federation)

    2010-12-15

    The present paper deals with investigation of the possibilities for reducing the risk of proliferation of fissionable materials by means of increasing the degree of protection of fresh fuel intended for light-water reactors against unsanctioned use in the case of withdrawal of a recipient country of deliveries from IAEA safeguards. It is shown that the use of recycled uranium for manufacturing export nuclear fuel makes transfer of nuclear material removed from the fuel assemblies for weapons purposes difficult because of the presence of isotope {sup 232}U, whose content increases when one attempts to enrich uranium extracted from fresh fuel. In combination with restricted access to technologies for isotope separation by means of establishing international centers for uranium enrichment, this technical measure can significantly reduce the risk of proliferation associated with export deliveries of fuel made of low-enriched uranium. The assessment of a maximum level of contamination of nuclear material being transferred by isotope {sup 232}U for the given isotope composition of the initial fuel is obtained. The concept of further investigations of the degree of security of export deliveries of fuel assemblies with recycled uranium intended for light-water reactors is suggested.

  4. A Lifecycle Emissions Model (LEM): Lifecycle Emissions from Transportation Fuels, Motor Vehicles, Transportation Modes, Electricity Use, Heating and Cooking Fuels, and Materials

    E-Print Network [OSTI]

    Delucchi, Mark

    2003-01-01

    Motor-vehicle flows Uranium enrichment Agriculture Fuel production Nitrogen deposition Multi-modal emissions Corn-ethanol

  5. 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.

  6. A Neutronic Analysis of TRU Recycling in PWRs Loaded with MOX-UE Fuel (MOX with U-235 Enriched U Support)

    SciTech Connect (OSTI)

    G. Youinou; S. Bays

    2009-05-01

    This report presents the results of a study dealing with the homogeneous recycling of either Pu or Pu+Np or Pu+Np+Am or Pu+Np+Am+Cm in PWRs using MOX-UE fuel, i.e. standard MOX fuel with a U235 enriched uranium support instead of the standard tail uranium (0.25%) for standard MOX fuel. This approach allows to multirecycle Pu or TRU (Pu+MA) as long as U235 is available, by keeping the Pu or TRU content in the fuel constant and at a value ensuring a negative moderator void coefficient (i.e. the loss of the coolant brings imperatively the reactor to a subcritical state). Once this value is determined, the U235 enrichment of the MOX-UE fuel is adjusted in order to reach the target burnup (51 GWd/t in this study).

  7. 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.

  8. 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

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. Uranium chloride extraction of transuranium elements from LWR fuel

    DOE Patents [OSTI]

    Miller, William E. (Naperville, IL); Ackerman, John P. (Downers Grove, IL); Battles, James E. (Oak Forest, IL); Johnson, Terry R. (Wheaton, IL); Pierce, R. Dean (Naperville, IL)

    1992-01-01

    A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels containing rare earth and noble metal fission products as well as other fission products is disclosed. The oxide fuel is reduced with Ca metal in the presence of Ca chloride and a U-Fe alloy which is liquid at about 800.degree. C. to dissolve uranium metal and the noble metal fission product metals and transuranium actinide metals and rare earth fission product metals leaving Ca chloride having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein. The Ca chloride and CaO and the fission products contained therein are separated from the U-Fe alloy and the metal values dissolved therein. The U-Fe alloy having dissolved therein reduced metals from the spent nuclear fuel is contacted with a mixture of one or more alkali metal or alkaline earth metal halides selected from the class consisting of alkali metal or alkaline earth metal and Fe or U halide or a combination thereof to transfer transuranium actinide metals and rare earth metals to the halide salt leaving the uranium and some noble metal fission products in the U-Fe alloy and thereafter separating the halide salt and the transuranium metals dissolved therein from the U-Fe alloy and the metals dissolved therein.

  14. Uranium chloride extraction of transuranium elements from LWR fuel

    DOE Patents [OSTI]

    Miller, W.E.; Ackerman, J.P.; Battles, J.E.; Johnson, T.R.; Pierce, R.D.

    1992-08-25

    A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels containing rare earth and noble metal fission products as well as other fission products is disclosed. The oxide fuel is reduced with Ca metal in the presence of Ca chloride and a U-Fe alloy which is liquid at about 800 C to dissolve uranium metal and the noble metal fission product metals and transuranium actinide metals and rare earth fission product metals leaving Ca chloride having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein. The Ca chloride and CaO and the fission products contained therein are separated from the U-Fe alloy and the metal values dissolved therein. The U-Fe alloy having dissolved therein reduced metals from the spent nuclear fuel is contacted with a mixture of one or more alkali metal or alkaline earth metal halides selected from the class consisting of alkali metal or alkaline earth metal and Fe or U halide or a combination thereof to transfer transuranium actinide metals and rare earth metals to the halide salt leaving the uranium and some noble metal fission products in the U-Fe alloy and thereafter separating the halide salt and the transuranium metals dissolved therein from the U-Fe alloy and the metals dissolved therein. 1 figure.

  15. 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.

  16. 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.

  17. 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.

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

    SciTech Connect (OSTI)

    Rudisill, T; Fernando Fondeur, F

    2009-01-15

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

  19. Electrorefining process and apparatus for recovery of uranium and a mixture of uranium and plutonium from spent fuels

    DOE Patents [OSTI]

    Ackerman, John P. (Downers Grove, IL); Miller, William E. (Naperville, IL)

    1989-01-01

    An electrorefining process and apparatus for the recovery of uranium and a mixture of uranium and plutonium from spent fuel using an electrolytic cell having a lower molten cadmium pool containing spent nuclear fuel, an intermediate electrolyte pool, an anode basket containing spent fuel, and two cathodes, the first cathode composed of either a solid alloy or molten cadmium and the second cathode composed of molten cadmium. Using this cell, additional amounts of uranium and plutonium from the anode basket are dissolved in the lower molten cadmium pool, and then substantially pure uranium is electrolytically transported and deposited on the first alloy or molten cadmium cathode. Subsequently, a mixture of uranium and plutonium is electrotransported and deposited on the second molten cadmium cathode.

  20. Electrorefining process and apparatus for recovery of uranium and a mixture of uranium and plutonium from spent fuels

    DOE Patents [OSTI]

    Ackerman, J.P.; Miller, W.E.

    1987-11-05

    An electrorefining process and apparatus for the recovery of uranium and a mixture of uranium and plutonium from spent fuels is disclosed using an electrolytic cell having a lower molten cadmium pool containing spent nuclear fuel, an intermediate electrolyte pool, an anode basket containing spent fuels, two cathodes and electrical power means connected to the anode basket, cathodes and lower molten cadmium pool for providing electrical power to the cell. Using this cell, additional amounts of uranium and plutonium from the anode basket are dissolved in the lower molten cadmium pool, and then purified uranium is electrolytically transported and deposited on a first molten cadmium cathode. Subsequently, a mixture of uranium and plutonium is electrotransported and deposited on a second cathode. 3 figs.

  1. 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.

  2. 22.351 Systems Analysis of the Nuclear Fuel Cycle, Spring 2003

    E-Print Network [OSTI]

    Kazimi, Mujid S.

    In-depth technical and policy analysis of various options for the nuclear fuel cycle. Topics include uranium supply, enrichment fuel fabrication, in-core physics and fuel management of uranium, thorium and other fuel types, ...

  3. 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.

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

    SciTech Connect (OSTI)

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

    2010-04-01

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

  5. 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.

  6. Thermo-chemical Modelling of Uranium-free Nitride Fuels Mikael JOLKKONEN1;;y

    E-Print Network [OSTI]

    Haviland, David

    be applied to modelling of oxide and carbide fuels as well. In this paper we give a brief introductionThermo-chemical Modelling of Uranium-free Nitride Fuels Mikael JOLKKONEN1;Ă;y , Marco STREIT2 and accepted December 22, 2003) A production process for americium-bearing, uranium-free nitride fuels

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

    E-Print Network [OSTI]

    Romano, Paul K. (Paul Kollath)

    2009-01-01

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

  8. 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.

  9. 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...

  10. Synthesis of Uranium Trichloride for the Pyrometallurgical Processing of Used Nuclear Fuel

    SciTech Connect (OSTI)

    B.R. Westphal; J.C. Price; R.D. Mariani

    2011-11-01

    The pyroprocessing of used nuclear fuel via electrorefining requires the continued addition of uranium trichloride to sustain operations. Uranium trichloride is utilized as an oxidant in the system to allow separation of uranium metal from the minor actinides and fission products. The inventory of uranium trichloride had diminished to a point that production was necessary to continue electrorefiner operations. Following initial experimentation, cupric chloride was chosen as a reactant with uranium metal to synthesize uranium trichloride. Despite the variability in equipment and charge characteristics, uranium trichloride was produced in sufficient quantities to maintain operations in the electrorefiner. The results and conclusions from several experiments are presented along with a set of optimized operating conditions for the synthesis of uranium trichloride.

  11. 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.

  12. 22.251 / 22.351 Systems Analysis of the Nuclear Fuel Cycle, Fall 2005

    E-Print Network [OSTI]

    Kazimi, Mujid S.

    This course provides an in-depth technical and policy analysis of various options for the nuclear fuel cycle. Topics include uranium supply, enrichment fuel fabrication, in-core physics and fuel management of uranium, ...

  13. 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.

  14. Molten salt extraction of transuranic and reactive fission products from used uranium oxide fuel

    SciTech Connect (OSTI)

    Herrmann, Steven Douglas

    2014-05-27

    Used uranium oxide fuel is detoxified by extracting transuranic and reactive fission products into molten salt. By contacting declad and crushed used uranium oxide fuel with a molten halide salt containing a minor fraction of the respective uranium trihalide, transuranic and reactive fission products partition from the fuel to the molten salt phase, while uranium oxide and non-reactive, or noble metal, fission products remain in an insoluble solid phase. The salt is then separated from the fuel via draining and distillation. By this method, the bulk of the decay heat, fission poisoning capacity, and radiotoxicity are removed from the used fuel. The remaining radioactivity from the noble metal fission products in the detoxified fuel is primarily limited to soft beta emitters. The extracted transuranic and reactive fission products are amenable to existing technologies for group uranium/transuranic product recovery and fission product immobilization in engineered waste forms.

  15. Plutonium recovery from spent reactor fuel by uranium displacement

    DOE Patents [OSTI]

    Ackerman, J.P.

    1992-03-17

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

  16. Plutonium recovery from spent reactor fuel by uranium displacement

    DOE Patents [OSTI]

    Ackerman, John P. (Downers Grove, IL)

    1992-01-01

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

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

    DOE Patents [OSTI]

    Pruett, D.J.; McTaggart, D.R.

    1983-08-31

    Uranium and technetium in the product stream of the Purex process for recovery of uranium in spent nuclear fuel are separated by (1) contacting the aqueous Purex product stream with hydrazine to reduce Tc/sup +7/ therein to a reduced species, and (2) contacting said aqueous stream with an organic phase containing tributyl phosphate and an organic diluent to extract uranium from said aqueous stream into said organic phase.

  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. Determination of Uranium Metal Concentration in Irradiated Fuel Storage Basin Sludge Using Selective Dissolution

    SciTech Connect (OSTI)

    Delegard, Calvin H.; Sinkov, Sergey I.; Chenault, Jeffrey W.; Schmidt, Andrew J.; Welsh, Terri L.; Pool, Karl N.

    2014-03-01

    Uranium metal corroding in water-saturated sludges now held in the US Department of Energy Hanford Site K West irradiated fuel storage basin can create hazardous hydrogen atmospheres during handling, immobilization, or subsequent transport and storage. Knowledge of uranium metal concentration in sludge thus is essential to safe sludge management and process design, requiring an expeditious routine analytical method to detect uranium metal concentrations as low as 0.03 wt% in sludge even in the presence of 30 wt% or higher total uranium concentrations.

  1. Fuel-cycle facilities: preliminary safety and environmental information document. Volume VII

    SciTech Connect (OSTI)

    Not Available

    1980-01-01

    Information is presented concerning the mining and milling of uranium and thorium; uranium hexafluoride conversion; enrichment; fuel fabrication; reprocessing; storage options; waste disposal options; transportation; heavy-water-production facilities; and international fuel service centers.

  2. Y-12 fulfills major milestone in fuel conversion commitment for...

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

    The Y-12 National Security Complex recently fulfilled its commitment to provide low-enriched uranium as feedstock to fuel the "Safe LOW-POwer Kritical Experiment" (SLOWPOKE)...

  3. Development of Gd-Enriched Alloys for Spent Nuclear Fuel Applications--Part 1: Preliminary Characterization

    E-Print Network [OSTI]

    DuPont, John N.

    composition for any Gd level. Keywords gadolinium, neutron absorbing material, nuclear criticality safety support, (2) spent nuclear fuel geometry control, and (3) nuclear criticality safety. In additionDevelopment of Gd-Enriched Alloys for Spent Nuclear Fuel Applications--Part 1: Preliminary

  4. Optimization of a seed and blanket thorium-uranium fuel cycle for pressurized water reactors

    E-Print Network [OSTI]

    Wang, Dean, 1971-

    2003-01-01

    A heterogeneous LWR core design, which employs a thorium/uranium once through fuel cycle, is optimized for good economics, wide safety margins, minimal waste burden and high proliferation resistance. The focus is on the ...

  5. 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.

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

    SciTech Connect (OSTI)

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

    2011-10-01

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

  7. 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.

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

    SciTech Connect (OSTI)

    2013-07-01

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

  9. Evaluation of health effects in Sequoyah Fuels Corporation workers from accidental exposure to uranium hexafluoride

    SciTech Connect (OSTI)

    Fisher, D.R. (Pacific Northwest Lab., Richland, WA (USA)); Swint, M.J.; Kathren, R.L. (Hanford Environmental Health Foundation, Richland, WA (USA))

    1990-05-01

    Urine bioassay measurements for uranium and medical laboratory results were studied to determine whether there were any health effects from uranium intake among a group of 31 workers exposed to uranium hexafluoride (UF{sub 6}) and hydrolysis products following the accidental rupture of a 14-ton shipping cylinder in early 1986 at the Sequoyah Fuels Corporation uranium conversion facility in Gore, Oklahoma. Physiological indicators studied to detect kidney tissue damage included tests for urinary protein, casts and cells, blood, specific gravity, and urine pH, blood urea nitrogen, and blood creatinine. We concluded after reviewing two years of follow-up medical data that none of the 31 workers sustained any observable health effects from exposure to uranium. The early excretion of uranium in urine showed more rapid systemic uptake of uranium from the lung than is assumed using the International Commission on Radiological Protection (ICRP) Publication 30 and Publication 54 models. The urinary excretion data from these workers were used to develop an improved systemic recycling model for inhaled soluble uranium. We estimated initial intakes, clearance rates, kidney burdens, and resulting radiation doses to lungs, kidneys, and bone surfaces. 38 refs., 10 figs., 7 tabs.

  10. 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.

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

    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 DeliciousMathematicsEnergyInterested Parties -Department of EnergyNEW1 NEPAOctoberofNNMCABDepartment

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

    SciTech Connect (OSTI)

    Taylor-Pashow, K.

    2011-06-08

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

  13. Supplement Analysis ? Spent Nuclear Fuel and SRS H-Canyon Operations

    Energy Savers [EERE]

    DOEEIS-0218-SA-07 SUPPLEMENT ANALYSIS FOR THE FOREIGN RESEARCH REACTOR SPENT NUCLEAR FUEL ACCEPTANCE PROGRAM Highly Enriched Uranium Target Residue Material Transportation U.S....

  14. NNSA Authorizes Start-Up of Highly Enriched Uranium Materials Facility at

    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.MOX Adventure6145836/%2A

  15. Y-12 and the ÂŤsuper enriched Uranium 235ÂŽ

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

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

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustmentsShirley Ann Jackson About1996HowFOAShowing YouNeedofDepartment of EnergyEducation

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

    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 GasAdjustmentsShirleyEnergy A plug-in electricLaboratory | DepartmentDOE ZeroofBatteriesHybridNational

  18. 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.

  19. 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.

  20. Diesel engine experiments with oxygen enrichment, water addition and lower-grade fuel

    SciTech Connect (OSTI)

    Sekar, R.R.; Marr, W.W.; Cole, R.L.; Marciniak, T.J. ); Schaus, J.E. )

    1990-01-01

    The concept of oxygen enriched air applied to reciprocating engines is getting renewed attention in the context of the progress made in the enrichment methods and the tougher emissions regulations imposed on diesel and gasoline engines. An experimental project was completed in which a direct injection diesel engine was tested with intake oxygen levels of 21% -- 35%. Since an earlier study indicated that it is necessary to use a cheaper fuel to make the concept economically attractive, a less refined fuel was included in the test series. Since a major objection to the use of oxygen enriched combustion air had been the increase in NO{sub x} emissions, a method must be found to reduce NO{sub x}. Introduction of water into the engine combustion process was included in the tests for this purpose. Fuel emulsification with water was the means used here even though other methods could also be used. The teat data indicated a large increase in engine power density, slight improvement in thermal efficiency, significant reductions in smoke and particulate emissions and NO{sub x} emissions controllable with the addition of water. 15 refs., 10 figs., 2 tabs.

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

    SciTech Connect (OSTI)

    Sean M. McDeavitt

    2011-04-29

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

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

    SciTech Connect (OSTI)

    Chodak, P. III

    1996-05-01

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

  3. Mr. William f. Crow, Acting Director . Uranium Fuel Licensing Branch

    Office of Legacy Management (LM)

    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-SessionsSouth DakotaRobbins and MyersHr.EvaluationJune~of theOfll s'_665,45

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

    E-Print Network [OSTI]

    Ensor, Brendan M. (Brendan Melvin)

    2012-01-01

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

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

    SciTech Connect (OSTI)

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

    1980-01-01

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

  6. Beyond Basic Target Enrichment: New Tools to Fuel Your NGS Research ( 7th Annual SFAF Meeting, 2012)

    ScienceCinema (OSTI)

    Carter, Jennifer [Agilent

    2013-03-22

    Jennifer Carter on "Beyond Basic Target Enrichment: New Tools to fuel your NGS Research" at the 2012 Sequencing, Finishing, Analysis in the Future Meeting held June 5-7, 2012 in Santa Fe, New Mexico.

  7. Spatial correction factors for YALINA Booster facility loaded with medium and low enriched fuels

    SciTech Connect (OSTI)

    Talamo, A.; Gohar, Y. [Argonne National Laboratory, 9700 S. Cass Ave, Argonne, IL 60439 (United States); Bournos, V.; Fokov, Y.; Kiyavitskaya, H.; Routkovskaya, C. [Joint Inst. for Power and Nuclear Research-Sosny, 99 Academician A.K.Krasin Str, Minsk 220109 (Belarus)

    2012-07-01

    The Bell and Glasstone spatial correction factor is used in analyses of subcritical assemblies to correct the experimental reactivity as function of the detector position. Besides the detector position, several other parameters affect the correction factor: the energy weighting function of the detector, the detector size, the energy-angle distribution of source neutrons, and the reactivity of the subcritical assembly. This work focuses on the dependency of the correction factor on the detector material and it investigates the YALINA Booster subcritical assembly loaded with medium (36%) and low (10%) enriched fuels. (authors)

  8. 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.

  9. 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.

  10. 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...

  11. Standard test method for atom percent fission in uranium and plutonium fuel (Neodymium-148 Method)

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    1996-01-01

    1.1 This test method covers the determination of stable fission product 148Nd in irradiated uranium (U) fuel (with initial plutonium (Pu) content from 0 to 50 %) as a measure of fuel burnup (1-3). 1.2 It is possible to obtain additional information about the uranium and plutonium concentrations and isotopic abundances on the same sample taken for burnup analysis. If this additional information is desired, it can be obtained by precisely measuring the spike and sample volumes and following the instructions in Test Method E267. 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

  12. Purification of uranium alloys by differential solubility of oxides and production of purified fuel precursors

    DOE Patents [OSTI]

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

    1997-01-01

    A method for purifying metallic alloys of uranium for use as nuclear reactor fuels in which the metal alloy is first converted to an oxide and then dissolved in nitric acid. Initial removal of metal oxide impurities not soluble in nitric acid is accomplished by filtration or other physical means. Further purification can be accomplished by carbonate leaching of uranyl ions from the partially purified solution or using traditional methods such as solvent extraction.

  13. Purification of uranium alloys by differential solubility of oxides and production of purified fuel precursors

    DOE Patents [OSTI]

    McLean, W. II; Miller, P.E.

    1997-12-16

    A method is described for purifying metallic alloys of uranium for use as nuclear reactor fuels in which the metal alloy is first converted to an oxide and then dissolved in nitric acid. Initial removal of metal oxide impurities not soluble in nitric acid is accomplished by filtration or other physical means. Further purification can be accomplished by carbonate leaching of uranyl ions from the partially purified solution or using traditional methods such as solvent extraction. 3 figs.

  14. 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.

  15. 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.

  16. 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)9 FederalRivers andMEDA Station3/%2A ¡BLM Public

  17. Production of small uranium dioxide microspheres for cermet nuclear fuel using the internal gelation process

    SciTech Connect (OSTI)

    Collins, Robert T; Collins, Jack Lee; Hunt, Rodney Dale; Ladd-Lively, Jennifer L; Patton, Kaara K; Hickman, Robert

    2014-01-01

    The U.S. National Aeronautics and Space Administration (NASA) is developing a uranium dioxide (UO2)/tungsten cermet fuel for potential use as the nuclear cryogenic propulsion stage (NCPS). The first generation NCPS is expected to be made from dense UO2 microspheres with diameters between 75 and 150 m. Previously, the internal gelation process and a hood-scale apparatus with a vibrating nozzle were used to form gel spheres, which became UO2 kernels with diameters between 350 and 850 m. For the NASA spheres, the vibrating nozzle was replaced with a custom designed, two-fluid nozzle to produce gel spheres in the desired smaller size range. This paper describes the operational methodology used to make 3 kg of uranium oxide microspheres.

  18. 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.

  19. Methodology for comparing the health effects of electricity generation from uranium and coal fuels

    SciTech Connect (OSTI)

    Rhyne, W.R.; El-Bassioni, A.A.

    1981-12-08

    A methodology was developed for comparing the health risks of electricity generation from uranium and coal fuels. The health effects attributable to the construction, operation, and decommissioning of each facility in the two fuel cycle were considered. The methodology is based on defining (1) requirement variables for the materials, energy, etc., (2) effluent variables associated with the requirement variables as well as with the fuel cycle facility operation, and (3) health impact variables for effluents and accidents. The materials, energy, etc., required for construction, operation, and decommissioning of each fuel cycle facility are defined as primary variables. The materials, energy, etc., needed to produce the primary variable are defined as secondary requirement variables. Each requirement variable (primary, secondary, etc.) has associated effluent variables and health impact variables. A diverging chain or tree is formed for each primary variable. Fortunately, most elements reoccur frequently to reduce the level of analysis complexity. 6 references, 11 figures, 6 tables.

  20. 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.

  1. 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.

  2. 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.

  3. 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.

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

    SciTech Connect (OSTI)

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

    1999-02-01

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

  5. 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).

  6. Detailed Destructive Post-Irradiation Examinations of Mixed Uranium and Plutonium Oxide Fuel

    SciTech Connect (OSTI)

    Delashmitt, Jeffrey {Jeff} S [ORNL; Keever, Tamara {Tammy} Jo [ORNL; Smith, Rob R [ORNL] [ORNL; Hexel, Cole R [ORNL] [ORNL; Ilgner, Ralph H [ORNL] [ORNL

    2010-01-01

    The United States Department of Energy (DOE) Fissile Materials Disposition Program (FMDP) is pursuing disposal of surplus weapons-usable plutonium by reactor irradiation as the fissile constituent of MOX fuel. Lead test assemblies (LTAs) have been irradiated for approximately 36 months in Duke Energy's Catawba-1 nuclear power plant (NPP). Per the mixed oxide (MOX) fuel topical report, approved by the U.S. Nuclear Regulatory Commission (NRC), destructive post-irradiation examinations (PIEs) are to be performed on second cycle rods (irradiated to an average burnup of approximately 45 GWd/MTHM). The Radiochemical Analysis Group (RAG) at Oak Ridge National Laboratory (ORNL) is currently performing the detailed destructive post-irradiation examinations (PIE) on four of the mixed uranium and plutonium oxide fuel rods. The analytical process involves dissolution of designated fuel segments in a shielded hot cell for high precision quantification of select fission products and actinide isotopes employing isotope dilution mass spectrometry (IDMS) among other analyses. The hot cell dissolution protocol to include the collection and subsequent alkaline fusion digestion of the fuel's acid resistant metallic particulates will be presented. Although the IDMS measurements of the fission products and actinide isotopes will not be completed by the time of the 51st INMM meeting, the setup and testing of the HPLC chromatographic separations in preparation for these measurements will be discussed.

  7. 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.

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

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

    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 RankADVANCED MANUFACTURINGEnergy BillsNo. 195 - Oct.7, 2015Verizon and VerizonCells: S. DEPARTMENTonReport on

  9. Acoustic Measurements of the Elastic Properties and Quality of Plutonium and Uranium Based Oxide Fuels

    E-Print Network [OSTI]

    Acoustic Measurements of the Elastic Properties and Quality of Plutonium and Uranium Based Oxide ranged from depleted uranium oxide to mixtures of plutonium and depleted uranium oxide (MOX) and mixed.8Pu0.2 92 121 152.6 63.3 281Plutonium Futures -- The Science 2010, September 19-23, 2010, Keystone, CO

  10. 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

  11. Standard guide for pyrophoricity/combustibility testing in support of pyrophoricity analyses of metallic uranium spent nuclear fuel

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2007-01-01

    1.1 This guide covers testing protocols for testing the pyrophoricity/combustibility characteristics of metallic uranium-based spent nuclear fuel (SNF). The testing will provide basic data for input into more detailed computer codes or analyses of thermal, chemical, and mechanical SNF responses. These analyses would support the engineered barrier system (EBS) design bases and safety assessment of extended interim storage facilities and final disposal in a geologic repository. The testing also could provide data related to licensing requirements for the design and operation of a monitored retrievable storage facility (MRS) or independent spent fuel storage installation (ISFSI). 1.2 This guide describes testing of metallic uranium and metallic uranium-based SNF in support of transportation (in accordance with the requirements of 10CFR71), interim storage (in accordance with the requirements of 10CFR72), and geologic repository disposal (in accordance with the requirements of 10CFR60/63). The testing described ...

  12. Transient fission-gas behavior in uranium nitride fuel under proposed space applications. Doctoral thesis

    SciTech Connect (OSTI)

    Deforest, D.L.

    1991-12-01

    In order to investigate whether fission gas swelling and release would be significant factors in a space based nuclear reactor operating under the Strategic Defense Initiative (SDI) program, the finite element program REDSTONE (Routine For Evaluating Dynamic Swelling in Transient Operational Nuclear Environments) was developed to model the 1-D, spherical geometry diffusion equations describing transient fission gas behavior in a single uranium nitride fuel grain. The equations characterized individual bubbles, rather than bubble groupings. This limits calculations to those scenarios where low temperatures, low burnups, or both were present. Instabilities in the bubble radii calculations forced the implementation of additional constraints limiting the bubble sizes to minimum and maximum (equilibrium) radii. The validity of REDSTONE calculations were checked against analytical solutions for internal consistency and against experimental studies for agreement with swelling and release results.

  13. 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.

  14. Fuel bundle design for enhanced usage of plutonium fuel

    DOE Patents [OSTI]

    Reese, Anthony P. (San Jose, CA); Stachowski, Russell E. (Fremont, CA)

    1995-01-01

    A nuclear fuel bundle includes a square array of fuel rods each having a concentration of enriched uranium and plutonium. Each rod of an interior array of the rods also has a concentration of gadolinium. The interior array of rods is surrounded by an exterior array of rods void of gadolinium. By this design, usage of plutonium in the nuclear reactor is enhanced.

  15. Pyroprocessing of fast flux test facility nuclear fuel

    SciTech Connect (OSTI)

    Westphal, B.R.; Wurth, L.A.; Fredrickson, G.L.; Galbreth, G.G.; Vaden, D.; Elliott, M.D.; Price, J.C.; Honeyfield, E.M.; Patterson, M.N. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID, 83415 (United States)

    2013-07-01

    Used nuclear fuel from the Fast Flux Test Facility (FFTF) was recently transferred to the Idaho National Laboratory and processed by pyroprocessing in the Fuel Conditioning Facility. Approximately 213 kg of uranium from sodium-bonded metallic FFTF fuel was processed over a one year period with the equipment previously used for the processing of EBR-II used fuel. The peak burnup of the FFTF fuel ranged from 10 to 15 atom% for the 900+ chopped elements processed. Fifteen low-enriched uranium ingots were cast following the electrorefining and distillation operations to recover approximately 192 kg of uranium. A material balance on the primary fuel constituents, uranium and zirconium, during the FFTF campaign will be presented along with a brief description of operating parameters. Recoverable uranium during the pyroprocessing of FFTF nuclear fuel was greater than 95% while the purity of the final electro-refined uranium products exceeded 99%. (authors)

  16. Pyroprocessing of Fast Flux Test Facility Nuclear Fuel

    SciTech Connect (OSTI)

    B.R. Westphal; G.L. Fredrickson; G.G. Galbreth; D. Vaden; M.D. Elliott; J.C. Price; E.M. Honeyfield; M.N. Patterson; L. A. Wurth

    2013-10-01

    Used nuclear fuel from the Fast Flux Test Facility (FFTF) was recently transferred to the Idaho National Laboratory and processed by pyroprocessing in the Fuel Conditioning Facility. Approximately 213 kg of uranium from sodium-bonded metallic FFTF fuel was processed over a one year period with the equipment previously used for the processing of EBR-II used fuel. The peak burnup of the FFTF fuel ranged from 10 to 15 atom% for the 900+ chopped elements processed. Fifteen low-enriched uranium ingots were cast following the electrorefining and distillation operations to recover approximately 192 kg of uranium. A material balance on the primary fuel constituents, uranium and zirconium, during the FFTF campaign will be presented along with a brief description of operating parameters. Recoverable uranium during the pyroprocessing of FFTF nuclear fuel was greater than 95% while the purity of the final electrorefined uranium products exceeded 99%.

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

    SciTech Connect (OSTI)

    Biswas, D; Mennerdahl, D

    2008-06-23

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

  18. 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.

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

    SciTech Connect (OSTI)

    Lewis, K.D.

    1994-09-01

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

  20. 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.

  1. 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

  2. Loss-of-flow analysis of an unfinned, graded fuel meat, LEU monolithic U-10Mo fuel design in support of the MITR-II fuel conversion

    E-Print Network [OSTI]

    Don, Sarah M

    2014-01-01

    In order to satisfy requirements of the Global Threat Reduction Initiative (GTRI), the 6 MW MIT Research Reactor (MITR-II) is to convert from the current 93%-enr 235U highly-enriched uranium (HEU) fuel to the low-enriched ...

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

    DOE Patents [OSTI]

    Friedman, Horace A. (Oak Ridge, TN)

    1985-01-01

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

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

    DOE Patents [OSTI]

    Friedman, H.A.

    1984-06-13

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

  5. 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.

  6. 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

  7. Calculation of Design Parameters for an Equilibrium LEU Core in the NBSR using a U7Mo Dispersion Fuel

    SciTech Connect (OSTI)

    Hanson A. L.; Diamond D.

    2014-06-30

    A plan is being developed for the conversion of the NIST research reactor (NBSR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. The LEU fuel may be a monolithic foil (LEUm) of U10Mo (10% molybdenum by weight in an alloy with uranium) or a dispersion of U7Mo in aluminum (LEUd). A previous report provided neutronic calculations for the LEUm fuel and this report presents the neutronics parameters for the LEUd fuel. The neutronics parameters for the LEUd fuel are compared to those previously obtained for the present HEU fuel and the proposed LEUm fuel. The results show no significant differences between the LEUm and the LEUd other than the LEUd fuel requires slightly less uranium than the LEUm fuel due to less molybdenum being present. The calculations include kinetics parameters, reactivity coefficients, reactivity worths of control elements and abnormal configurations, and power distributions under normal operation and with misloaded fuel elements.

  8. 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.

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

    SciTech Connect (OSTI)

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

    2004-10-06

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

  10. Thermo-chemical Modelling of Uranium-free Nitride Fuels Mikael JOLKKONEN1;;y

    E-Print Network [OSTI]

    predict that the immediate product of the typical decarburisation step is not methane, but hydrogen is a mixture of trans- uranium mononitrides and zirconium mononitride, forming a homogenous solid solution over to be irradiated in the Advanced Test Reactor in Idaho, as well as in Phenix, there is a need for optimisation

  11. 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

  12. 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 200şC 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.

  13. 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

  14. 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.

  15. 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.

  16. World nuclear fuel cycle requirements 1991

    SciTech Connect (OSTI)

    Not Available

    1991-10-10

    The nuclear fuel cycle consists of mining and milling uranium ore, processing the uranium into a form suitable for generating electricity, burning'' the fuel in nuclear reactors, and managing the resulting spent nuclear fuel. This report presents projections of domestic and foreign requirements for natural uranium and enrichment services as well as projections of discharges of spent nuclear fuel. These fuel cycle requirements are based on the forecasts of future commercial nuclear power capacity and generation published in a recent Energy Information Administration (EIA) report. Also included in this report are projections of the amount of spent fuel discharged at the end of each fuel cycle for each nuclear generating unit in the United States. The International Nuclear Model is used for calculating the projected nuclear fuel cycle requirements. 14 figs., 38 tabs.

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

    SciTech Connect (OSTI)

    Kovalishin, A. A.; Prosyolkov, V. N.; Sidorenko, V. D.; Stogov, Yu. V.

    2014-12-15

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

  18. Fuel bundle design for enhanced usage of plutonium fuel

    DOE Patents [OSTI]

    Reese, A.P.; Stachowski, R.E.

    1995-08-08

    A nuclear fuel bundle includes a square array of fuel rods each having a concentration of enriched uranium and plutonium. Each rod of an interior array of the rods also has a concentration of gadolinium. The interior array of rods is surrounded by an exterior array of rods void of gadolinium. By this design, usage of plutonium in the nuclear reactor is enhanced. 10 figs.

  19. 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.

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

    SciTech Connect (OSTI)

    Not Available

    1980-01-01

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

  1. 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.

  2. Methodology and a preliminary data base for examining the health risks of electricity generation from uranium and coal fuels

    SciTech Connect (OSTI)

    El-Bassioni, A.A.

    1980-08-01

    An analytical model was developed to assess and examine the health effects associated with the production of electricity from uranium and coal fuels. The model is based on a systematic methodology that is both simple and easy to check, and provides details about the various components of health risk. A preliminary set of data that is needed to calculate the health risks was gathered, normalized to the model facilities, and presented in a concise manner. Additional data will become available as a result of other evaluations of both fuel cycles, and they should be included in the data base. An iterative approach involving only a few steps is recommended for validating the model. After each validation step, the model is improved in the areas where new information or increased interest justifies such upgrading. Sensitivity analysis is proposed as the best method of using the model to its full potential. Detailed quantification of the risks associated with the two fuel cycles is not presented in this report. The evaluation of risks from producing electricity by these two methods can be completed only after several steps that address difficult social and technical questions. Preliminary quantitative assessment showed that several factors not considered in detail in previous studies are potentially important. 255 refs., 21 figs., 179 tabs.

  3. The fuel cycle economics of improved uranium utilization in light water reactors

    E-Print Network [OSTI]

    Abbaspour, Ali Tehrani

    A simple fuel cycle cost model has been formulated, tested satisfactorily (within better than 3% for a wide range of cases)

  4. Power Surge: Uranium alloy fuel for TerraPower | Y-12 National Security

    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 wouldMass mapSpeedingProgram Guidelines This document w w w.pv - te ch.orgPowerComplex Power

  5. 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.

  6. 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.

  7. Beryllium Impregnation of Uranium Fuel: Thermal Modeling of Cylindrical Objects for Efficiency Evaluation 

    E-Print Network [OSTI]

    Lynn, Nicholas

    2011-08-04

    With active research projects related to nuclear waste immobilization and high conductivity nuclear fuels, a thermal model has been developed to simulate the temperature profile within a heat generating cylinder in order to imitate the behavior...

  8. Silicon carbide performance as cladding for advanced uranium and thorium fuels for light water reactors

    E-Print Network [OSTI]

    Sukjai, Yanin

    2014-01-01

    There has been an ongoing interest in replacing the fuel cladding zirconium-based alloys by other materials to reduce if not eliminate the autocatalytic and exothermic chemical reaction with water and steam at above 1,200 ...

  9. 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.

  10. Oxidative alteration of spent fuel in a silica-rich environment...

    Office of Scientific and Technical Information (OSTI)

    NUCLEAR FACILITIES; DISSOLUTION; HYDRATES; PERFORMANCE; SILICATES; SPENT FUELS; THERMODYNAMICS; URANINITES; URANIUM CARBONATES; URANIUM DEPOSITS; URANIUM ORES; URANYL SILICATES;...

  11. 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.

  12. Laser shockwave technique for characterization of nuclear fuel plate interfaces

    SciTech Connect (OSTI)

    Perton, M.; Levesque, D.; Monchalin, J.-P.; Lord, M. [National Research Council Canada, 75 de Mortagne Blvd, Boucherville, Quebec, J4B 6Y4 (Canada); Smith, J. A.; Rabin, B. H. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States)

    2013-01-25

    The US National Nuclear Security Agency is tasked with minimizing the worldwide use of high-enriched uranium. One aspect of that effort is the conversion of research reactors to monolithic fuel plates of low-enriched uranium. The manufacturing process includes hot isostatic press bonding of an aluminum cladding to the fuel foil. The Laser Shockwave Technique (LST) is here evaluated for characterizing the interface strength of fuel plates using depleted Uranium/Mo foils. LST is a non-contact method that uses lasers for the generation and detection of large amplitude acoustic waves and is therefore well adapted to the quality assurance of this process. Preliminary results show a clear signature of well-bonded and debonded interfaces and the method is able to classify/rank the bond strength of fuel plates prepared under different HIP conditions.

  13. Laser Shockwave Technique For Characterization Of Nuclear Fuel Plate Interfaces

    SciTech Connect (OSTI)

    James A. Smith; Barry H. Rabin; Mathieu Perton; Daniel Lévesque; Jean-Pierre Monchalin; Martin Lord

    2012-07-01

    The US National Nuclear Security Agency is tasked with minimizing the worldwide use of high-enriched uranium. One aspect of that effort is the conversion of research reactors to monolithic fuel plates of low-enriched uranium. The manufacturing process includes hot isostatic press bonding of an aluminum cladding to the fuel foil. The Laser Shockwave Technique (LST) is here evaluated for characterizing the interface strength of fuel plates using depleted Uranium/Mo foils. LST is a non-contact method that uses lasers for the generation and detection of large amplitude acoustic waves and is therefore well adapted to the quality assurance of this process. Preliminary results show a clear signature of well-bonded and debonded interfaces and the method is able to classify/rank the bond strength of fuel plates prepared under different HIP conditions.

  14. 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.

  15. 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.

  16. Atomic Diffusion in the Uranium-50wt% Zirconium Nuclear Fuel System 

    E-Print Network [OSTI]

    Eichel, Daniel

    2013-06-17

    (software package) EPMA Electron probe microanalysis FCML Fuel Cycles and Materials Laboratory FGR Fission gas release HCP Hexagonal close-packed (crystal structure) L? , M? Nomenclature for atomic electron transition and emission of associated... characteristic x-ray LIF Lithium Fluoride (diffracting crystal) LWR Light water reactor MRF Materials Research Furnace vi Nwk### Diffusion sample labeling nomenclature, in which N is the number of weeks in the annealing time and ### is the temperature...

  17. Secretary Chu, HEUMF and Dr. Chris Keim - Or: History continues with Highly Enriched Uranium facility (title as it appeared in The Oak Ridger)

    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, High-ThroughputUpcomingmagnetoresistanceand Government Reform | Department of

  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. 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)

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

    SciTech Connect (OSTI)

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

    2013-07-01

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

  1. 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.

  2. Vitrification of HLW Produced by Uranium/Molybdenum Fuel Reprocessing in COGEMA's Cold Crucible Melter

    SciTech Connect (OSTI)

    Do Quang, R.; Petitjean, V.; Hollebecque, F.; Pinet, O.; Flament, T.; Prod'homme, A.

    2003-02-25

    The performance of the vitrification process currently used in the La Hague commercial reprocessing plants has been continuously improved during more than ten years of operation. In parallel COGEMA (industrial Operator), the French Atomic Energy Commission (CEA) and SGN (respectively COGEMA's R&D provider and Engineering) have developed the cold crucible melter vitrification technology to obtain greater operating flexibility, increased plant availability and further reduction of secondary waste generated during operations. The cold crucible is a compact water-cooled melter in which the radioactive waste and the glass additives are melted by direct high frequency induction. The cooling of the melter produces a solidified glass layer that protects the melter's inner wall from corrosion. Because the heat is transferred directly to the melt, high operating temperatures can be achieved with no impact on the melter itself. COGEMA plans to implement the cold crucible technology to vitrify high level liquid waste from reprocessed spent U-Mo-Sn-Al fuel (used in gas cooled reactor). The cold crucible was selected for the vitrification of this particularly hard-to-process waste stream because it could not be reasonably processed in the standard hot induction melters currently used at the La Hague vitrification facilities : the waste has a high molybdenum content which makes it very corrosive and also requires a special high temperature glass formulation to obtain sufficiently high waste loading factors (12 % in molybdenum). A special glass formulation has been developed by the CEA and has been qualified through lab and pilot testing to meet standard waste acceptance criteria for final disposal of the U-Mo waste. The process and the associated technologies have been also being qualified on a full-scale prototype at the CEA pilot facility in Marcoule. Engineering study has been integrated in parallel in order to take into account that the Cold Crucible should be installed remotely in one of the R7 vitrification cell. This paper will present the results obtained in the framework of these qualification programs.

  3. 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}).

  4. 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.

  5. Designing a New Fuel for HFIR-Performance Parameters for LEU Core Configurations

    SciTech Connect (OSTI)

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

    2009-01-01

    An engineering design study for a fuel that would enable the conversion of the High Flux Isotope Reactor from highly enriched uranium to low enriched uranium fuel is ongoing as part of an effort sponsored by the U.S. Department of Energy's National Nuclear Security Administration through the Global Threat Reduction Initiative. Given the unique fuel and core design and high power density of the reactor and the requirement that the impact of the fuel change on the core performance and operation be minimal, this conversion study presents a complex and challenging task, requiring improvements in the computational models currently used to support the operation of the reactor and development of new models that would take advantage of newly available simulation methods and tools. The computational models used to search for a fuel design that would meet the requirements for the conversion study and the results obtained with these models are presented and discussed. Estimates of relevant reactor performance parameters for the low enriched uranium fuel core are presented and compared to the corresponding data for the currently operating highly enriched uranium fuel core.

  6. 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.

  7. An assessment of the attractiveness of material associated with thorium/uranium and uranium closed fuel cycles from a safeguards perspective

    SciTech Connect (OSTI)

    Bathke, Charles Gary [Los Alamos National Laboratory; Wallace, Richard K [Los Alamos National Laboratory; Hase, Kevin R [Los Alamos National Laboratory; Sleaford, Brad W [LLNL; Ebbinghaus, Bartley B [LLNL; Collins, Brian W [PNNL; Bradley, Keith S [LLNL; Prichard, Andrew W [PNNL; Smith, Brian W [PNNL

    2010-01-01

    This paper reports the continued evaluation of the attractiveness of materials mixtures containing special nuclear materials (SNM) associated with various proposed nuclear fuel cycles. Specifically, this paper examines two closed fuel cycles. The first fuel cycle examined is a thorium fuel cycle in which a pressurized heavy water reactor (PHWR) is fueled with mixtures of plutonium/thorium and {sup 233}U/thorium. The used fuel is then reprocessed using the THOREX process and the actinides are recycled. The second fuel cycle examined consists of conventional light water reactors (LWR) whose fuel is reprocessed for actinides that are then fed to and recycled until consumed in fast-spectrum reactors: fast reactors and accelerator driven systems (ADS). As reprocessing of LWR fuel has already been examined, this paper will focus on the reprocessing of the scheme's fast-spectrum reactors' fuel. This study will indicate what is required to render these materials as having low utility for use in nuclear weapons. Nevertheless, the results of this paper suggest that all reprocessing products evaluated so far need to be rigorously safeguarded and provided high levels of physical protection. These studies were performed at the request of the United States Department of Energy (DOE). The methodology and key findings will be presented.

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

    SciTech Connect (OSTI)

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

    2012-05-02

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

  9. Uranium 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.

  10. Assessment of Nuclear Fuels using Radiographic Thickness Measurement Method

    SciTech Connect (OSTI)

    Muhammad Abir; Fahima Islam; Hyoung Koo Lee; Daniel Wachs

    2014-11-01

    The Convert branch of the National Nuclear Security Administration (NNSA) Global Threat Reduction Initiative (GTRI) focuses on the development of high uranium density fuels for research and test reactors for nonproliferation. This fuel is aimed to convert low density high enriched uranium (HEU) based fuel to high density low enriched uranium (LEU) based fuel for high performance research reactors (HPRR). There are five U.S. reactors that fall under the HPRR category, including: the Massachusetts Institute of Technology Reactor (MITR), the National Bureau of Standards Reactor (NBSR), the Missouri University Research Reactor (UMRR), the Advanced Test Reactor (ATR), and the High Flux Isotope Reactor (HFIR). U-Mo alloy fuel phase in the form of either monolithic or dispersion foil type fuels, such as ATR Full-size In center flux trap Position (AFIP) and Reduced Enrichment for Research and Test Reactor (RERTR), are being designed for this purpose. The fabrication process1 of RERTR is susceptible to introducing a variety of fuel defects. A dependable quality control method is required during fabrication of RERTR miniplates to maintain the allowable design tolerances, therefore evaluating and analytically verifying the fabricated miniplates for maintaining quality standards as well as safety. The purpose of this work is to analyze the thickness of the fabricated RERTR-12 miniplates using non-destructive technique to meet the fuel plate specification for RERTR fuel to be used in the ATR.

  11. EVALUATION OF CORE PHYSICS ANALYSIS METHODS FOR CONVERSION OF THE INL ADVANCED TEST REACTOR TO LOW-ENRICHMENT FUEL

    SciTech Connect (OSTI)

    Mark DeHart; Gray S. Chang

    2012-04-01

    Computational neutronics studies to support the possible conversion of the ATR to LEU are underway. Simultaneously, INL is engaged in a physics methods upgrade project to put into place modern computational neutronics tools for future support of ATR fuel cycle and experiment analysis. A number of experimental measurements have been performed in the ATRC in support of the methods upgrade project, and are being used to validate the new core physics methods. The current computational neutronics work is focused on performance of scoping calculations for the ATR core loaded with a candidate LEU fuel design. This will serve as independent confirmation of analyses that have been performed previously, and will evaluate some of the new computational methods for analysis of a candidate LEU fuel for ATR.

  12. IRRADIATION PERFORMANCE OF U-Mo MONOLITHIC FUEL

    SciTech Connect (OSTI)

    M.K. Meyer; J. Gan; J.-F. Jue; D.D. Keiser; E. Perez; A. Robinson; D.M. Wachs; N. Woolstenhulme; G.L. Hofman; Y.-S. Kim

    2014-04-01

    High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. UMo alloys represent the best known tradeoff in these properties. Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.

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

    SciTech Connect (OSTI)

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

    2008-06-13

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

  14. 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.

  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 2014 Uranium 201457

  16. 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

  17. 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

  18. 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

  19. 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

  20. 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.

  1. 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

  2. 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.

  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. Projections of Full-Fuel-Cycle Energy and Emissions Metrics

    E-Print Network [OSTI]

    Coughlin, Katie

    2013-01-01

    comprises the mining and milling of uranium ore, conversionof Uranium-235 by fuel type and production stage. Mining and

  7. Assessment for advanced fuel cycle options in CANDU

    SciTech Connect (OSTI)

    Morreale, A.C.; Luxat, J.C. [McMaster University, 1280 Main St. W. Hamilton, Ontario, L8S 4L7 (Canada); Friedlander, Y. [AMEC-NSS Ltd., 700 University Ave. 4th Floor, Toronto, Ontario, M5G 1X6 (Canada)

    2013-07-01

    The possible options for advanced fuel cycles in CANDU reactors including actinide burning options and thorium cycles were explored and are feasible options to increase the efficiency of uranium utilization and help close the fuel cycle. The actinide burning TRUMOX approach uses a mixed oxide fuel of reprocessed transuranic actinides from PWR spent fuel blended with natural uranium in the CANDU-900 reactor. This system reduced actinide content by 35% and decreased natural uranium consumption by 24% over a PWR once through cycle. The thorium cycles evaluated used two CANDU-900 units, a generator and a burner unit along with a driver fuel feedstock. The driver fuels included plutonium reprocessed from PWR, from CANDU and low enriched uranium (LEU). All three cycles were effective options and reduced natural uranium consumption over a PWR once through cycle. The LEU driven system saw the largest reduction with a 94% savings while the plutonium driven cycles achieved 75% savings for PWR and 87% for CANDU. The high neutron economy, online fuelling and flexible compact fuel make the CANDU system an ideal reactor platform for many advanced fuel cycles.

  8. The use of U/sub 3/Si/sub 2/ dispersed in aluminum in plate-type fuel elements for research and test reactors

    SciTech Connect (OSTI)

    Snelgrove, J.L.; Domagala, R.F.; Hofman, G.L.; Wiencek, T.C.; Copeland, G.L.; Hobbs, R.W.; Senn, R.L.

    1987-10-01

    A high-density fuel based on U/sub 3/Si/sub 2/ dispersed in aluminum has been developed and tested for use in converting plate-type research and test reactors from the use of highly enriched uranium to the use of low-enriched uranium. Results of preirradiation testing and the irradiation and postirradiation examination of miniature fuel plates and full-sized fuel elements are summarized. Swelling of the U/sub 3/Si/sub 2/ fuel particles is a linear function of the fission density in the particle to well beyond the fission density achievable in low-enriched fuels. U/sub 3/Si/sub 2/ particle swelling rate is approximately the same as that of the commonly used UAl/sub x/ fuel particle. The presence of minor amounts of U/sub 3/Si or uranium solid solution in the fuel result in greater, but still acceptable, fuel swelling. Blister threshold temperatures are at least as high as those of currently used fuels. An exothermic reaction occurs near the aluminum melting temperature, but the measured energy releases were low enough not to substantially worsen the consequences of an accident. U/sub 3/Si/sub 2/-aluminum dispersion fuel with uranium densities up to at least 4.8 Mg/m/sup 3/ is a suitable LEU fuel for typical plate-type research and test reactors. 42 refs., 28 figs., 7 tabs.

  9. 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...

  10. 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...

  11. 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...

  12. 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...

  13. 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...

  14. 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...

  15. Incorporation of Hydride Nuclear Fuels in Commercial Light Water Reactors

    E-Print Network [OSTI]

    Terrani, Kurt Amir

    2010-01-01

    Uranium in the fuel remains metallic since the equilibriumtype of hydride fuel consists of metallic uranium particlesalloying the metallic components of the fuel followed by a

  16. Efficacy of a solution-based approach for making sodalite waste forms for an oxide reduction salt utilized in the reprocessing of used uranium oxide fuel

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

    Riley, Brian J.; Pierce, David A.; Frank, Steven M.; Matyáš, Josef; Burns, Carolyne A.

    2015-04-01

    This paper describes the various approaches attempted to make solution-derived sodalite with a LiCl-Li2O oxide reduction salt used to dissolve used uranium oxide fuel so the uranium can be recovered and recycled. The approaches include modified sol-gel and solutionbased synthesis processes. As-made products were mixed with 5 and 10 mass% of a Na2O-B2O3- SiO2 glass binder and these, along with product without a binder, were heated using either a cold-press-and-sinter method or hot uniaxial pressing. The results demonstrate the limitation of sodalite yield due to the fast intermediate reactions between Na+ and Cl- to form halite in solution and Li2Omore »and SiO2 to form lithium silicates (e.g., Li2SiO3 or Li2Si2O5) in the calcined and sintered pellets. The results show that pellets can be made with high sodalite fractions in the crystalline product (~92 mass%) and low porosities using a solution-based approach and this LiCl-Li2O salt but that the incorporation of Li into the sodalite is low.« less

  17. Modeling the Influence of Interaction Layer Formation on Thermal Conductivity of U–Mo Dispersion Fuel

    SciTech Connect (OSTI)

    Burkes, Douglas; Casella, Andrew M.; Huber, Tanja K.

    2015-01-01

    The Global Threat Reduction Initiative Program continues to develop existing and new plate- and rod-type research and test reactor fuels with maximum attainable uranium loadings capable of potentially converting a number of the world’s remaining high-enriched uranium fueled reactors to low-enriched uranium fuel. Currently, the program is focused on assisting with the development and qualification of an even higher density fuel type consisting of a uranium-molybdenum (U-Mo) alloy dispersed in an aluminum matrix. Thermal conductivity is an important consideration in determining the operational temperature of the fuel plate and can be influenced by interaction layer formation between the fuel and matrix, porosity that forms during fabrication of the fuel plates, and upon the concentration of the dispersed phase within the matrix. This paper develops and validates a simple model to study the influence of interaction layer formation and conductivity, fuel particle size, and volume fraction of fuel dispersed in the matrix on the effective conductivity of the composite. The model shows excellent agreement with results previously presented in the literature. In particular, the thermal conductivity of the interaction layer does not appear to be important in determining the overall conductivity of the composite, while formation of the interaction layer and subsequent consumption of the matrix reveals a rather significant effect. The effective thermal conductivity of the composite can be influenced by the fuel particle distribution by minimizing interaction layer formation and preserving the higher thermal conductivity matrix.

  18. 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".

  19. 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.

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

    E-Print Network [OSTI]

    Matthews, Isaac A

    2010-01-01

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

  1. Effects of EGR, water/N2/CO2 injection and oxygen enrichment on the availability destroyed due to combustion for a range of conditions and fuels 

    E-Print Network [OSTI]

    Sivadas, Hari Shanker

    2009-06-02

    This study was directed at examining the effects of exhaust gas recirculation (EGR), water/N2/CO2 injections and oxygen enrichment on availability destroyed because of combustion in simple systems like those of constant ...

  2. Design and Testing of Prototypic Elements Containing Monolithic Fuel

    SciTech Connect (OSTI)

    N.E. Woolstenhulme; M.K. Meyer; D.M. Wachs

    2011-10-01

    The US fuel development team has performed numerous irradiation tests on small to medium sized specimens containing low enriched uranium fuel designs. The team is now focused on qualification and demonstration of the uranium-molybdenum Base Monolithic Design and has entered the next generation of testing with the design and irradiation of prototypic elements which contain this fuel. The designs of fuel elements containing monolithic fuel, such as AFIP-7 (which is currently under irradiation) and RERTR-FE (which is currently under fabrication), are appropriate progressions relative to the technology life cycle. The culmination of this testing program will occur with the design, fabrication, and irradiation of demonstration products to include the base fuel demonstration and design demonstration experiments. Future plans show that design, fabrication, and testing activities will apply the rigor needed for a demonstration campaign.

  3. Environmental assessment for the demonstration of uranium-atomic vapor laser isotope separation (U-AVLIS) at Lawrence Livermore National Laboratory

    SciTech Connect (OSTI)

    Not Available

    1991-05-01

    The U.S. Department of Energy (DOE), Office of Nuclear Energy, proposes to use full-scale lasers and separators to demonstrate uranium enrichment as part of the national Uranium-Atomic Vapor Laser Isotope Separation (U-AVLIS) Program. Demonstration of uranium enrichment is planned to be conducted in Building 490 of the Lawrence Livermore National Laboratory (LLNL), near Livermore, California in 1991 and 1992. The collective goal of the U-AVLIS Program is to develop and demonstrate an integrated technology for low-cost enrichment of uranium for nuclear reactor fuel. Alternatives to the proposed LLNL demonstration activity are no action, use of alternative LLNL facilities, and use of an alternative DOE site. This EA describes the existing LLNL environment and surroundings that could be impacted by the proposed action. Potential impacts to on- site and off-site environments predicted during conduct of the Uranium Demonstration System (UDS) at LLNL and alternative actions are reported in this EA. The analysis covers routine activities and potential accidents. 81 refs., 8 figs., 6 tabs.

  4. 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)

  5. 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.

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

    SciTech Connect (OSTI)

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

    2013-03-01

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

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

    SciTech Connect (OSTI)

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

    2014-03-01

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

  8. Depleted uranium dioxide melting in cold crucible melter and production of granules from the melt for use in casks for spent nuclear fuel and radioactive wastes

    SciTech Connect (OSTI)

    Gotovchikov, V.T.; Seredenko, V.A.; Shatalov, V.V.; Mironov, B.S.; Kaplenkov, V.N.; Seredenko, A.V.; Saranchin, V.K.; Shulgin, A.S. [All-Russian Research Institute of Chemical Technology (ARRICT), Moscow (Russian Federation); Haire, M.J.; Forsberg, C.W. [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)

    2007-07-01

    This paper describes the results of a joint research program between the Russian Research Institute of Chemical Technology and Oak Ridge National Laboratory in the United States to develop new radiation shielding materials for use in the construction of casks for spent nuclear fuel (SNF) and radioactive wastes. Research and development is underway to develop SNF storage, transport, and disposal casks using shielding made with two new depleted uranium dioxide (DUO{sub 2}) materials: a DUO{sub 2}-steel cermet, and, DUCRETE with DUAGG (DUO{sub 2} aggregate). Melting the DUO{sub 2} and allowing it to freeze will produce a near 100% theoretical density product and assures that the product produces no volatile materials upon subsequent heating. Induction cold-crucible melters (ICCM) are being developed for this specific application. An ICCM is, potentially, a high throughput low-cost process. Schematics of a pilot facility were developed for the production of molten DUO{sub 2} from DU{sub 3}O{sub 8} to produce granules <1 mm in diameter in a continuous mode of operation. Thermodynamic analysis was conducted for uranium-oxygen system in the temperature range from 300 to 4000 K in various gas mediums. Temperature limits of stability for various uranium oxides were determined. Experiments on melting DUO{sub 2} were carried out in a high frequency ICCM in a cold crucible with a 120 mm in diameter. The microstructure of molten DUO{sub 2} was studied and lattice parameters were determined. It was experimentally proved, and validated by X-ray analysis, that an opportunity exists to produce molten DUO{sub 2} from mixed oxides (primarily DU{sub 3}O{sub 8}) by reduction melting in ICCM. This will allow using DU{sub 3}O{sub 8} directly to make DUO{sub 2}-a separate unit operation to produce UO{sub 2} feed material is not needed. Experiments were conducted concerning the addition of alloying components, gadolinium et al. oxides, into the DUO{sub 2} melt while in the crucible. These additives improve neutron and gamma radiation shielding and operation properties of the final solids. Cermet samples of 50 wt % DUO{sub 2} were produced. (authors)

  9. 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.

  10. Start-up fuel and power flattening of sodium-cooled candle core

    SciTech Connect (OSTI)

    Takaki, Naoyuki; Sagawa, Yu; Umino, Akitake [Department of Nuclear Safety Engineering, Tokyo City University 1-28-1 Tamazutsumi, Setagaya, Tokyo 158-8557 (Japan); Sekimoto, Hiroshi [University of California, Berkeley, CA 94720 (United States)

    2013-07-01

    The hard neutron spectrum and unique power shape of CANDLE enable its distinctive performances such as achieving high burnup more than 30% and exempting necessity of both enrichment and reprocessing. On the other hand, they also cause several challenging problems. One is how the initial fuel can be prepared to start up the first CANDLE reactor because the equilibrium fuel composition that enables stable CANDLE burning is complex both in axial and radial directions. Another prominent problem is high radial power peaking factor that worsens averaged burnup, namely resource utilization factor in once-through mode and shorten the life time of structure materials. The purposes of this study are to solve these two problems. Several ideas for core configurations and startup fuel using single enrichment uranium and iron as a substitute of fission products are studied. As a result, it is found that low enriched uranium is applicable to ignite the core but all concepts examined here exceeded heat limits. Adjustment in enrichment and height of active and burnt zone is opened for future work. Sodium duct assemblies and thorium fuel assemblies loaded in the center region are studied as measures to reduce radial power peaking factor. Replacing 37 fuels by thorium fuel assemblies in the zeroth to third row provides well-balanced performance with flattened radial power distribution. The CANDLE core loaded with natural uranium in the outer and thorium in the center region achieved 35.6% of averaged burnup and 7.0 years of cladding life time owing to mitigated local fast neutron irradiation at the center. Using thorium with natural or depleted uranium in CANDLE reactor is also beneficial to diversifying fission resource and extending available term of fission energy without expansion of needs for enrichment and reprocessing.

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

    E-Print Network [OSTI]

    Richard, Joshua (Joshua Glenn)

    2012-01-01

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

  12. World nuclear fuel cycle requirements 1990

    SciTech Connect (OSTI)

    Not Available

    1990-10-26

    This analysis report presents the projected requirements for uranium concentrate and uranium enrichment services to fuel the nuclear power plants expected to be operating under three nuclear supply scenarios. Two of these scenarios, the Lower Reference and Upper Reference cases, apply to the United States, Canada, Europe, the Far East, and other countries with free market economies (FME countries). A No New Orders scenario is presented only for the United States. These nuclear supply scenarios are described in Commercial Nuclear Power 1990: Prospects for the United States and the World (DOE/EIA-0438(90)). This report contains an analysis of the sensitivities of the nuclear fuel cycle projections to different levels and types of projected nuclear capacity, different enrichment tails assays, higher and lower capacity factors, changes in nuclear fuel burnup levels, and other exogenous assumptions. The projections for the United States generally extend through the year 2020, and the FME projections, which include the United States, are provided through 2010. The report also presents annual projections of spent nuclear fuel discharges and inventories of spent fuel. Appendix D includes domestic spent fuel projections through the year 2030 for the Lower and Upper Reference cases and through 2040, the last year in which spent fuel is discharged, for the No New Orders case. These disaggregated projections are provided at the request of the Department of Energy's Office of Civilian Radioactive Waste Management.

  13. Irradiation behavior of the interaction product of U-Mo fuel particle dispersion in an Al matrix.

    SciTech Connect (OSTI)

    Kim, Y.S.; Hofman, G. (Nuclear Engineering Division)

    2012-06-01

    Irradiation performance of U-Mo fuel particles dispersed in Al matrix is stable in terms of fuel swelling and is suitable for the conversion of research and test reactors from highly enriched uranium (HEU) to low enriched uranium (LEU). However, tests of the fuel at high temperatures and high burnups revealed obstacles caused by the interaction layers forming between the fuel particle and matrix. In some cases, fission gas filled pores grow and interconnect in the interdiffusion layer resulting in fuel plate failure. Postirradiation observations are made to examine the behavior of the interdiffusion layers. The interdiffusion layers show a fluid-like behavior characteristic of amorphous materials. In the amorphous interdiffusion layers, fission gas diffusivity is high and the material viscosity is low so that the fission gas pores readily form and grow. Based on the observations, a pore formation mechanism is proposed and potential remedies to suppress the pore growth are also introduced.

  14. 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.

  15. Integration of the AVLIS (atomic vapor laser isotopic separation) process into the nuclear fuel cycle. [Effect of AVLIS feed requirements on overall fuel cycle

    SciTech Connect (OSTI)

    Hargrove, R.S.; Knighton, J.B.; Eby, R.S.; Pashley, J.H.; Norman, R.E.

    1986-08-01

    AVLIS RD and D efforts are currently proceeding toward full-scale integrated enrichment demonstrations in the late 1980's and potential plant deployment in the mid 1990's. Since AVLIS requires a uranium metal feed and produces an enriched uranium metal product, some change in current uranium processing practices are necessitated. AVLIS could operate with a UF/sub 6/-in UF/sub 6/-out interface with little effect to the remainder of the fuel cycle. This path, however, does not allow electric utility customers to realize the full potential of low cost AVLIS enrichment. Several alternative processing methods have been identified and evaluated which appear to provide opportunities to make substantial cost savings in the overall fuel cycle. These alternatives involve varying levels of RD and D resources, calendar time, and technical risk to implement and provide these cost reduction opportunities. Both feed conversion contracts and fuel fabricator contracts are long-term entities. Because of these factors, it is not too early to start planning and making decisions on the most advantageous options so that AVLIS can be integrated cost effectively into the fuel cycle. This should offer economic opportunity to all parties involved including DOE, utilities, feed converters, and fuel fabricators. 10 refs., 11 figs., 2 tabs.

  16. 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.

  17. Advanced Fuel Cycle Cost Basis

    SciTech Connect (OSTI)

    D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert; E. Schneider

    2008-03-01

    This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 25 cost modules—23 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste.

  18. Advanced Fuel Cycle Cost Basis

    SciTech Connect (OSTI)

    D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert

    2007-04-01

    This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 26 cost modules—24 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, and high-level waste.

  19. Advanced Fuel Cycle Cost Basis

    SciTech Connect (OSTI)

    D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert; E. Schneider

    2009-12-01

    This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 25 cost modules—23 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste.

  20. 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.+1

  1. 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. Purchases6a.

  2. Measures of the environmental footprint of the front end of the nuclear fuel cycle

    SciTech Connect (OSTI)

    E. Schneider; B. Carlsen; E. Tavrides; C. van der Hoeven; U. Phathanapirom

    2013-11-01

    Previous estimates of environmental impacts associated with the front end of the nuclear fuel cycle (FEFC) have focused primarily on energy consumption and CO2 emissions. Results have varied widely. This work builds upon reports from operating facilities and other primary data sources to build a database of front end environmental impacts. This work also addresses land transformation and water withdrawals associated with the processes of the FEFC. These processes include uranium extraction, conversion, enrichment, fuel fabrication, depleted uranium disposition, and transportation. To allow summing the impacts across processes, all impacts were normalized per tonne of natural uranium mined as well as per MWh(e) of electricity produced, a more conventional unit for measuring environmental impacts that facilitates comparison with other studies. This conversion was based on mass balances and process efficiencies associated with the current once-through LWR fuel cycle. Total energy input is calculated at 8.7 x 10- 3 GJ(e)/MWh(e) of electricity and 5.9 x 10- 3 GJ(t)/MWh(e) of thermal energy. It is dominated by the energy required for uranium extraction, conversion to fluoride compound for subsequent enrichment, and enrichment. An estimate of the carbon footprint is made from the direct energy consumption at 1.7 kg CO2/MWh(e). Water use is likewise dominated by requirements of uranium extraction, totaling 154 L/MWh(e). Land use is calculated at 8 x 10- 3 m2/MWh(e), over 90% of which is due to uranium extraction. Quantified impacts are limited to those resulting from activities performed within the FEFC process facilities (i.e. within the plant gates). Energy embodied in material inputs such as process chemicals and fuel cladding is identified but not explicitly quantified in this study. Inclusion of indirect energy associated with embodied energy as well as construction and decommissioning of facilities could increase the FEFC energy intensity estimate by a factor of up to 2.

  3. Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors

    E-Print Network [OSTI]

    Galvez, Cristhian

    2011-01-01

    considers the usage of low enriched uranium fuel. A secondkernels are made from Low Enriched Uranium (LEU) in the form

  4. Efficacy of a Solution-Based Approach for Making Sodalite Waste Forms for an Oxide Reduction Salt Utilized in the Reprocessing of Used Uranium Oxide Fuel

    SciTech Connect (OSTI)

    Riley, Brian J.; Pierce, David A.; Frank, Steven M.; Matyas, Josef; Burns, Carolyn A.

    2015-04-01

    This paper describes various approaches for making sodalite with a LiCl-Li2O oxide reduction salt used to recover uranium from used oxide fuel. The approaches include sol-gel and solution-based synthesis processes. As-made products were mixed with 5 and 10 mass% of a Na2O-B2O3-SiO2 glass binder and these, along with product without a binder, were heated using either a cold-press-and-sinter method or hot uniaxial pressing. The results demonstrate the limitation of sodalite yield due to the fast intermediate reactions between Na+ and Cl- to form halite in solution and Li2O and SiO2 to form lithium silicates (e.g., Li2SiO3 or Li2Si2O5) in the calcined and sintered pellets. The results show that pellets can be made with high sodalite fractions (~92 mass%) and low porosities using a solution-based approach and this LiCl-Li2O salt.

  5. Uranium Recovery from Seawater: Development of Fiber Adsorbents Prepared via Atom-Transfer Radical Polymerization

    SciTech Connect (OSTI)

    Saito, Tomonori; Brown, Suree; Chatterjee, Sabornie; Kim, Jungseung; Tsouris, Constantinos; Mayes, Richard; Kuo, Li-Jung; Gill, Gary A.; Oyola, Yatsandra; Janke, C.; Dai, Sheng

    2014-07-09

    Uranium exists uniformly at a concentration of ~3.3 ppb in seawater. The extraction of uranium from seawater presents a very attractive alternative source of uranium for nuclear fuel needs.

  6. 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.

  7. Differential Die-Away Instrument: Report on Fuel Assembly Mock-up Measurements with Neutron Generator

    SciTech Connect (OSTI)

    Goodsell, Alison Victoria; Swinhoe, Martyn Thomas; Henzl, Vladimir; Rael, Carlos D.; Desimone, David J.

    2014-09-18

    Fresh fuel experiments for the differential die-away (DDA) project were performed using a DT neutron generator, a 15x15 PWR fuel assembly, and nine 3He detectors in a water tank inside of a shielded cell at Los Alamos National Laboratory (LANL). Eight different fuel enrichments were created using low enriched (LEU) and depleted uranium (DU) dioxide fuel rods. A list-mode data acquisition system recorded the time-dependent signal and analysis of the DDA signal die-away time was performed. The die-away time depended on the amount of fissile material in the fuel assembly and the position of the detector. These experiments were performed in support of the spent nuclear fuel Next Generation Safeguards Initiative DDA project. Lessons learned from the fresh fuel DDA instrument experiments and simulations will provide useful information to the spent fuel project.

  8. 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.

  9. Development and validation of capabilities to measure thermal properties of layered monolithic U-Mo alloy plate-type fuel

    SciTech Connect (OSTI)

    Burkes, Douglas; Casella, Andrew M.; Buck, Edgar C.; Casella, Amanda J.; Edwards, Matthew K.; MacFarlan, Paul J.; Pool, Karl N.; Smith, Frances N.; Steen, Franciska H.

    2014-07-19

    The uranium-molybdenum (U-Mo) alloy in a monolithic form has been proposed as one fuel design capable of converting some of the world’s highest power research reactors from the use of high enriched uranium (HEU) to low enriched uranium (LEU). One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of thermal conductivity behavior of the fuel system as a function of temperature and expected irradiation conditions. The purpose of this paper is to verify and validate the functionality of equipment methods installed in hot cells for eventual measurements on irradiated uranium-molybdenum (U-Mo) monolithic fuel specimens, procedures to operate the equipment, and models to extract the desired thermal properties. The results presented here demonstrate the adequacy of the equipment, procedures and models that have been developed for this purpose based on measurements conducted on surrogate depleted uranium-molybdenum (DU-Mo) alloy samples containing a zirconium diffusion barrier and clad in aluminum alloy 6061 (AA6061). The results are in excellent agreement with thermal property data reported in the literature for similar U-Mo alloys as a function of temperature.

  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. Uranium purchased by

  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. Uranium purchased byb.

  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. Uranium purchased byb.S2.

  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. Uranium purchased

  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. Uranium purchasedb.

  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. Uranium purchasedb.4.

  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. Uranium purchasedb.4..

  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. Uranium purchasedb.4..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. Uranium

  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. Deliveries of

  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. Deliveries of4.

  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. Deliveries

  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. Deliveries6.

  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. Deliveries6.7.

  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.

  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. Foreign

  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. Uranium3.9. Foreign.

  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. Uranium3.9. Foreign.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. Uranium3.9. Foreign.0.1.

  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. Uranium3.9.

  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. Uranium3.9.3. Inventories

  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. Uranium3.9.3.

  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. Uranium3.9.3.3.

  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. Uranium3.9.3.3.b.

  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. Uranium3.9.3.3.b.8.

  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. Uranium3.9.3.3.b.8.9.

  16. 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

  17. RADIOACTIVE WASTE STREAMS FROM VARIOUS POTENTIAL NUCLEAR FUEL CYCLE OPTIONS

    SciTech Connect (OSTI)

    Nick Soelberg; Steve Piet

    2010-11-01

    Five fuel cycle options, about which little is known compared to more commonly known options, have been studied in the past year for the United States Department of Energy. These fuel cycle options, and their features relative to uranium-fueled light water reactor (LWR)-based fuel cycles, include: • Advanced once-through reactor concepts (Advanced Once-Through, or AOT) – intended for high uranium utilization and long reactor operating life, use depleted uranium in some cases, and avoid or minimize used fuel reprocessing • Fission-fusion hybrid (FFH) reactor concepts – potential variations are intended for high uranium or thorium utilization, produce fissile material for use in power generating reactors, or transmute transuranic (TRU) and some radioactive fission product (FP) isotopes • High temperature gas reactor (HTGR) concepts - intended for high uranium utilization, high reactor thermal efficiencies; they have unique fuel designs • Molten salt reactor (MSR) concepts – can breed fissile U-233 from Th fuel and avoid or minimize U fuel enrichment, use on-line reprocessing of the used fuel, produce lesser amounts of long-lived, highly radiotoxic TRU elements, and avoid fuel assembly fabrication • Thorium/U-233 fueled LWR (Th/U-233) concepts – can breed fissile U-233 from Th fuel and avoid or minimize U fuel enrichment, and produce lesser amounts of long-lived, highly radiotoxic TRU elements. These fuel cycle options could result in widely different types and amounts of used or spent fuels, spent reactor core materials, and waste streams from used fuel reprocessing, such as: • Highly radioactive, high-burnup used metal, oxide, or inert matrix U and/or Th fuels, clad in Zr, steel, or composite non-metal cladding or coatings • Spent radioactive-contaminated graphite, SiC, carbon-carbon-composite, metal, and Be reactor core materials • Li-Be-F salts containing U, TRU, Th, and fission products • Ranges of separated or un-separated activation products, fission products, and actinides. Waste forms now used or studied for used LWR fuels can be used for some of these waste streams – but some waste forms may need to be developed for unique waste streams.

  18. Storage and disposition of weapons usable fissile materials (FMD) PEIS: Blending of U-233 to {lt}12% or {lt}5% enrichment at the Idaho National Engineering Laboratory. Data report, Draft: Version 1

    SciTech Connect (OSTI)

    Shaber, E.L.

    1995-08-01

    Uranium-233 (U-233), a uranium isotope, is a fissionable material capable of fueling nuclear reactors or being utilized in the manufacturing of nuclear weapons. As such, it is controlled as a special nuclear material. The Idaho National Engineering Laboratory (INEL) and Oak Ridge National Laboratory (ORNL) currently store the Department of Energy`s (DOE`s) supply of unirradiated U-233 fuel materials. Irradiated U-233 is covered by the national spent nuclear fuel (SNF) program and is not in the scope of this report. The U-233 stored at ORNL is relatively pure uranium oxide in the form of powder or monolithic solids. This material is currently stored in stainless steel canisters of variable lengths measuring about 3 inches in diameter. The ORNL material enrichment varies with some material containing considerable amounts of U-235. The INEL material is fuel from the Light Water Breeder Reactor (LWBR) Program and consists of enriched uranium and thorium oxides in zircaloy cladding. The DOE inventory of U-233 contains trace quantities of U-232, and daughter products from the decay of U-232 and U-233, resulting in increased radioactivity over time. These increased levels of radioactivity generally result in the need for special handling considerations.

  19. Letter Report: Looking Ahead at Nuclear Fuel Resources

    SciTech Connect (OSTI)

    J. Stephen Herring

    2013-09-01

    The future of nuclear energy and its ability to fulfill part of the world’s energy needs for centuries to come depend on a reliable input of nuclear fuel, either thorium or uranium. Obviously, the present nuclear fuel cycle is completely dependent on uranium. Future thorium cycles will also depend on 235U or fissile isotopes separated from used fuel to breed 232Th into fissile 233U. This letter report discusses several emerging areas of scientific understanding and technology development that will clarify and enable assured supplies of uranium and thorium well into the future. At the most fundamental level, the nuclear energy community needs to appreciate the origins of uranium and thorium and the processes of planetary accretion by which those materials have coalesced to form the earth and other planets. Secondly, the studies of geophysics and geochemistry are increasing understanding of the processes by which uranium and thorium are concentrated in various locations in the earth’s crust. Thirdly, the study of neutrinos and particularly geoneutrinos (neutrinos emitted by radioactive materials within the earth) has given an indication of the overall global inventories of uranium and thorium, though little indication for those materials’ locations. Crustal temperature measurements have also given hints of the vertical distribution of radioactive heat sources, primarily 238U and 232Th, within the continental crust. Finally, the evolving technologies for laser isotope separation are indicating methods for reducing the energy input to uranium enrichment but also for tailoring the isotopic vectors of fuels, burnable poisons and structural materials, thereby adding another tool for dealing with long-term waste management.

  20. 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.

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

    SciTech Connect (OSTI)

    NONE

    2000-12-01

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

  2. 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.

  3. 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

  4. 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.

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

    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 DeliciousMathematics And StatisticsProgramof-SA-02:Innovative EnergyMexico Emergency Management Exercisethe Oak Ridge

  6. 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.

  7. 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.

  8. fuel

    National Nuclear Security Administration (NNSA)

    4%2A en Cheaper catalyst may lower fuel costs for hydrogen-powered cars http:www.nnsa.energy.govblogcheaper-catalyst-may-lower-fuel-costs-hydrogen-powered-cars

  9. Commercial nuclear fuel from U.S. and Russian surplus defense inventories: Materials, policies, and market effects

    SciTech Connect (OSTI)

    NONE

    1998-05-01

    Nuclear materials declared by the US and Russian governments as surplus to defense programs are being converted into fuel for commercial nuclear reactors. This report presents the results of an analysis estimating the market effects that would likely result from current plans to commercialize surplus defense inventories. The analysis focuses on two key issues: (1) the extent by which traditional sources of supply, such as production from uranium mines and enrichment plants, would be displaced by the commercialization of surplus defense inventories or, conversely, would be required in the event of disruptions to planned commercialization, and (2) the future price of uranium considering the potential availability of surplus defense inventories. Finally, the report provides an estimate of the savings in uranium procurement costs that could be realized by US nuclear power generating companies with access to competitively priced uranium supplied from surplus defense inventories.

  10. Irradiation Experiment Conceptual Design Parameters for NBSR Fuel Conversion

    SciTech Connect (OSTI)

    Brown, N. R.; Brown, N. R.; Baek, J. S; Hanson, A. L.; Cuadra, A.; Cheng, L. Y.; Diamond, D. J.

    2014-04-30

    It has been proposed to convert the National Institute of Standards and Technology (NIST) research reactor, known as the NBSR, from high-enriched uranium (HEU) fuel to low-Enriched uranium (LEU) fuel. The motivation to convert the NBSR to LEU fuel is to reduce the risk of proliferation of special nuclear material. This report is a compilation of relevant information from recent studies related to the proposed conversion using a metal alloy of LEU with 10 w/o molybdenum. The objective is to inform the design of the mini-plate and full-size-Plate irradiation experiments that are being planned. This report provides relevant dimensions of the fuel elements, and the following parameters at steady state: average and maximum fission rate density and fission density, fuel temperature distribution for the plate with maximum local temperature, and two-dimensional heat flux profiles of fuel plates with high power densities. The latter profiles are given for plates in both the inner and outer core zones and for cores with both fresh and depleted shim arms (reactivity control devices). A summary of the methodology to obtain these results is presented. Fuel element tolerance assumptions and hot channel factors used in the safety analysis are also given.

  11. Experiments in anodic film effects during electrorefining of scrap U-10Mo fuels in support of modeling efforts

    SciTech Connect (OSTI)

    Van Kleeck, M. [School of Nuclear Engineering, Purdue University, West Lafayette, IN 47907 (United States); Chemical Sciences and Engineering Division, Argonne National Laboratory, Argonne, IL 60439 (United States); Willit, J.; Williamson, M.A. [Chemical Sciences and Engineering Division, Argonne National Laboratory, Argonne, IL 60439 (United States); Fentiman, A.W. [School of Nuclear Engineering, Purdue University, West Lafayette, IN 47907 (United States)

    2013-07-01

    A monolithic uranium molybdenum alloy clad in zirconium has been proposed as a low enriched uranium (LEU) fuel option for research and test reactors, as part of the Reduced Enrichment for Research and Test Reactors program. Scrap from the fuel's manufacture will contain a significant portion of recoverable LEU. Pyroprocessing has been identified as an option to perform this recovery. A model of a pyroprocessing recovery procedure has been developed to assist in refining the LEU recovery process and designing the facility. Corrosion theory and a two mechanism transport model were implemented on a Mat-Lab platform to perform the modeling. In developing this model, improved anodic behavior prediction became necessary since a dense uranium-rich salt film was observed at the anode surface during electrorefining experiments. Experiments were conducted on uranium metal to determine the film's character and the conditions under which it forms. The electro-refiner salt used in all the experiments was eutectic LiCl/KCl containing UCl{sub 3}. The anodic film material was analyzed with ICP-OES to determine its composition. Both cyclic voltammetry and potentiodynamic scans were conducted at operating temperatures between 475 and 575 C. degrees to interrogate the electrochemical behavior of the uranium. The results show that an anodic film was produced on the uranium electrode. The film initially passivated the surface of the uranium on the working electrode. At high over potentials after a trans-passive region, the current observed was nearly equal to the current observed at the initial active level. Analytical results support the presence of K{sub 2}UCl{sub 6} at the uranium surface, within the error of the analytical method.

  12. Theory of Antineutrino Monitoring of Burning MOX Plutonium Fuels

    E-Print Network [OSTI]

    A. C. Hayes; H. R. Trellue; Michael Martin Nieto; W. B. WIlson

    2011-10-03

    This letter presents the physics and feasibility of reactor antineutrino monitoring to verify the burnup of plutonium loaded in the reactor as a Mixed Oxide (MOX) fuel. It examines the magnitude and temporal variation in the antineutrino signals expected for different MOX fuels, for the purposes of nuclear accountability and safeguards. The antineutrino signals from reactor-grade and weapons-grade MOX are shown to be distinct from those from burning low enriched uranium. Thus, antineutrino monitoring could be used to verify the destruction of plutonium in reactors, though verifying the grade of the plutonium being burned is found to be more challenging.

  13. Theory of Antineutrino Monitoring of Burning MOX Plutonium Fuels

    E-Print Network [OSTI]

    Hayes, A C; Nieto, Michael Martin; WIlson, W B

    2011-01-01

    This letter presents the physics and feasibility of reactor antineutrino monitoring to verify the burnup of plutonium loaded in the reactor as a Mixed Oxide (MOX) fuel. It examines the magnitude and temporal variation in the antineutrino signals expected for different MOX fuels, for the purposes of nuclear accountability and safeguards. The antineutrino signals from reactor-grade and weapons-grade MOX are shown to be distinct from those from burning low enriched uranium. Thus, antineutrino monitoring could be used to verify the destruction of plutonium in reactors, though verifying the grade of the plutonium being burned is found to be more challenging.

  14. fuel

    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)9 Federal RegisterStorm1 3446 YEAR/%2Afissile4/%2A en

  15. Moving toward multilateral mechanisms for the fuel cycle

    SciTech Connect (OSTI)

    Panasyuk,A.; Rosenthal,M.; Efremov, G. V.

    2009-04-17

    Multilateral mechanisms for the fuel cycle are seen as a potentially important way to create an industrial infrastructure that will support a renaissance and at the same time not contribute to the risk of nuclear proliferation. In this way, international nuclear fuel cycle centers for enrichment can help to provide an assurance of supply of nuclear fuel that will reduce the likelihood that individual states will pursue this sensitive technology, which can be used to produce nuclear material directly usable nuclear weapons. Multinational participation in such mechanisms can also potentially promote transparency, build confidence, and make the implementation of IAEA safeguards more effective or more efficient. At the same time, it is important to ensure that there is no dissemination of sensitive technology. The Russian Federation has taken a lead role in this area by establishing an International Uranium Enrichment Center (IUEC) for the provision of enrichment services at its uranium enrichment plant located at the Angarsk Electrolysis Chemical Complex (AECC). This paper describes how the IUEe is organized, who its members are, and the steps that it has taken both to provide an assured supply of nuclear fuel and to ensure protection of sensitive technology. It also describes the relationship between the IUEC and the IAEA and steps that remain to be taken to enhance its assurance of supply. Using the IUEC as a starting point for discussion, the paper also explores more generally the ways in which features of such fuel cycle centers with multinational participation can have an impact on safeguards arrangements, transparency, and confidence-building. Issues include possible lAEA safeguards arrangements or other links to the IAEA that might be established at such fuel cycle centers, impact of location in a nuclear weapon state, and the transition by the IAEA to State Level safeguards approaches.

  16. Irradiation Experiment Conceptual Design Parameters for NBSR Fuel Conversion

    SciTech Connect (OSTI)

    Brown N. R.; Brown,N.R.; Baek,J.S; Hanson, A.L.; Cuadra,A.; Cheng,L.Y.; Diamond, D.J.

    2013-03-31

    It has been proposed to convert the National Institute of Standards and Technology (NIST) research reactor, known as the NBSR, from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. The motivation to convert the NBSR to LEU fuel is to reduce the risk of proliferation of special nuclear material. This report is a compilation of relevant information from recent studies related to the proposed conversion using a metal alloy of LEU with 10 w/o molybdenum. The objective is to inform the design of the mini-plate and full-size plate irradiation experiments that are being planned. This report provides relevant dimensions of the fuel elements, and the following parameters at steady state: average and maximum fission rate density and fission density, fuel temperature distribution for the plate with maximum local temperature, and two-dimensional heat flux profiles of fuel plates with high power densities. . The latter profiles are given for plates in both the inner and outer core zones and for cores with both fresh and depleted shim arms (reactivity control devices). In addition, a summary of the methodology to obtain these results is presented.

  17. Westinghouse VANTAGE+ fuel assembly to meet future PWR operating requirements

    SciTech Connect (OSTI)

    Doshi, P.K.; Chapin, D.L.; Scherpereel, L.R.

    1988-01-01

    Many utilities operating pressurized water reactors (PWRs) are implementing longer reload cycles. Westinghouse is addressing this trend with fuel products that increase fuel utilization through higher discharge burnups. Higher burnup helps to offset added enriched uranium costs necessary to enable the higher energy output of longer cycles. Current fuel products have burnup capabilities in the area of 40,000 MWd/tonne U or more. There are three main phenomena that must be addressed to achieve even higher burnup levels: accelerated cladding, waterside corrosion, and hydriding; increased fission gas production; and fuel rod growth. Long cycle lengths also require efficient burnable absorbers to control the excess reactivity associated with increased fuel enrichment while maintaining a low residual absorber penalty at the end of cycle. Westinghouse VANTAGE + PWR fuel incorporates features intended to enhance fuel performance at very high burnups, including advances in the three basic elements of the fuel assembly: fuel cladding, fuel rod, and fuel assembly skeleton. ZIRLO {sup TM} cladding, an advanced Zircaloy cladding that contains niobium, offers a significant improvement in corrosion resistance relative to Zircaloy-4. Another important Westinghouse PWR fuel feature that facilitates long cycles is the zirconium diboride integral fuel burnable absorber (ZrB{sub 2}IFBA).

  18. 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.

  19. 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...

  20. Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

    SciTech Connect (OSTI)

    Jennifer Lyons; Wade R. Marcum; Mark D. DeHart; Sean R. Morrell

    2014-01-01

    The Advanced Test Reactor (ATR), under the Reduced Enrichment for Research and Test Reactors (RERTR) Program and the Global Threat Reduction Initiative (GTRI), is conducting feasibility studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low enriched uranium (LEU) composition. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary analysis of TRIGA® fuel within the current ATR fuel envelopes and compares it to the functional requirements delineated by the Naval Reactors Program, which includes: greater than 4.8E+14 fissions/s/g of 235U, a fast to thermal neutron flux ratio that is less than 5% deviation of its current value, a constant cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other parameters outside those put forth by the Naval Reactors Program which are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time.

  1. Fuel

    SciTech Connect (OSTI)

    NONE

    1999-10-01

    Two subjects are covered in this section. They are: (1) Health effects of possible contamination at Paducah Gaseous Diffusion Plant to be studied; and (2) DOE agrees on test of MOX fuel in Canada.

  2. Neutronic Analyses for HEU to LEU fuel conversion of the Massachusetts Institute of Technology.

    SciTech Connect (OSTI)

    Wilson, E. H.; Newton, T. H.; Bergeron, A.; Horelik, N.; Stevens, J. G (Nuclear Engineering Division); ( NS)

    2011-03-02

    The Massachusetts Institute of Technology (MIT) reactor (MITR-II), based in Cambridge, Massachusetts, is a research reactor designed primarily for experiments using neutron beam and in-core irradiation facilities. It delivers a neutron flux comparable to current LWR power reactors in a compact 6 MW core using Highly Enriched Uranium (HEU) fuel. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context, most research and test reactors both domestic and international have started a program of conversion to the use of Low Enriched Uranium (LEU) fuel. A new type of LEU fuel based on a mixture of uranium and molybdenum (UMo) is expected to allow the conversion of compact high performance reactors like the MITR-II. This report presents the results of steady state neutronic safety analyses for conversion of MITR-II from the use of HEU fuel to the use of U-Mo LEU fuel. The objective of this work was to demonstrate that the safety analyses meet current requirements for an LEU core replacement of MITR-II.

  3. 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

  4. 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

  5. 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.

  6. U.S. And Russia Complete Nuclear Security Upgrades Under Bratislava...

    Office of Environmental Management (EM)

    research reactors internationally fueled with highly enriched uranium (HEU) to low enriched uranium fuel and to return all Russian-origin HEU fresh and spent nuclear fuel stored...

  7. Effects of thermal treatment on the co-rolled U-Mo fuel foils

    SciTech Connect (OSTI)

    Dennis D. Keiser, Jr.; Tammy L. Trowbridge; Cynthia R. Breckenridge; Brady L. Mackowiak; Glenn A. Moore; Barry H. Rabin; Mitchell K. Meyer

    2014-11-01

    A monolithic fuel type is being developed to convert US high performance research and test reactors such as Advanced Test Reactor (ATR) at Idaho National Laboratory from highly enriched uranium (HEU) to low-enriched uranium (LEU). The interaction between the cladding and the U-Mo fuel meat during fuel fabrication and irradiation is known to have negative impacts on fuel performance, such as mechanical integrity and dimensional stability. In order to eliminate/minimize the direct interaction between cladding and fuel meat, a thin zirconium diffusion barrier was introduced between the cladding and U-Mo fuel meat through a co-rolling process. A complex interface between the zirconium and U-Mo was developed during the co-rolling process. A predictable interface between zirconium and U-Mo is critical to achieve good fuel performance since the interfaces can be the weakest link in the monolithic fuel system. A post co-rolling annealing treatment is expected to create a well-controlled interface between zirconium and U-Mo. A systematic study utilizing post co-rolling annealing treatment has been carried out. Based on microscopy results, the impacts of the annealing treatment on the interface between zirconium and U-Mo will be presented and an optima annealing treatment schedule will be suggested. The effects of the annealing treatment on the fuel performance will also be discussed.

  8. 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.

  9. Performance of Thorium-Based Mixed-Oxide Fuels for the Consumption of Plutonium in Current and Advanced Reactors

    SciTech Connect (OSTI)

    Weaver, Kevan D.; Herring, J. Stephen [Idaho National Engineering and Environmental Laboratory (United States)

    2003-07-15

    A renewed interest in thorium-based fuels has arisen lately based on the need for proliferation resistance, longer fuel cycles, higher burnup, and improved waste form characteristics. Recent studies have been directed toward homogeneously mixed, heterogeneously mixed, and seed-and-blanket thorium-uranium oxide fuel cycles that rely on 'in situ' use of the bred-in {sup 233}U. However, due to the higher initial enrichment required to achieve acceptable burnups, these fuels are encountering economic constraints. Thorium can nevertheless play a large role in the nuclear fuel cycle, particularly in the reduction of plutonium inventories. While uranium-based mixed-oxide (MOX) fuel will decrease the amount of plutonium in discharged fuel, the reduction is limited due to the breeding of more plutonium (and higher actinides) from the {sup 238}U. Here, we present calculational results and a comparison of the potential burnup of a thorium-based and uranium-based mixed-oxide fuel in a light water reactor. Although the uranium-based fuels outperformed the thorium-based fuels in achievable burnup, a depletion comparison of the initially charged plutonium (both reactor and weapons grade) showed that the thorium-based fuels outperformed the uranium-based fuels by more that a factor of 2, where >70% of the total plutonium in the thorium-based fuel is consumed during the cycle. This is significant considering that the achievable burnups of the thorium-based fuels were 1.4 to 4.6 times less than the uranium-based fuels for similar plutonium enrichments. For equal specific burnups of {approx}60 MWd/kg (i.e., using variable plutonium weight percentages to give the desired burnup), the thorium-based fuels still outperform the uranium-based fuels by more than a factor of 2, where the total plutonium consumption in a three-batch, 18-month cycle was 60 to 70%. This is fairly significant considering that 10 to 15% (by weight) more plutonium is needed in the thorium-based fuels as compared to the uranium-based fuels to achieve these burnups.Furthermore, thorium-based fuels could also be used as a strategy for reducing the amount of long-lived nuclides (including the minor actinides) and thus the radiotoxicity in spent nuclear fuel. Although the breeding of {sup 233}U is a concern, the presence of {sup 232}U and its daughter products (namely {sup 208}Tl) can aid in making this fuel self-protecting, and/or enough {sup 238}U can be added to denature the fissile uranium. From these calculations, it appears that thorium-based fuel for plutonium incineration is superior when compared to uranium-based fuel and should be considered as an alternative to traditional MOX in both current and future/advanced reactor designs.

  10. Performance of Thorium-Based Mixed Oxide Fuels for the Consumption of Plutonium in Current and Advanced Reactors

    SciTech Connect (OSTI)

    Weaver, Kevan Dean; Herring, James Stephen

    2003-07-01

    A renewed interest in thorium-based fuels has arisen lately based on the need for proliferation resistance, longer fuel cycles, higher burnup, and improved waste form characteristics. Recent studies have been directed toward homogeneously mixed, heterogeneously mixed, and seed-and-blanket thorium-uranium oxide fuel cycles that rely on "in situ" use of the bred-in 233U. However, due to the higher initial enrichment required to achieve acceptable burnups, these fuels are encountering economic constraints. Thorium can nevertheless play a large role in the nuclear fuel cycle, particularly in the reduction of plutonium inventories. While uranium-based mixed-oxide (MOX) fuel will decrease the amount of plutonium in discharged fuel, the reduction is limited due to the breeding of more plutonium (and higher actinides) from the 238U. Here, we present calculational results and a comparison of the potential burnup of a thorium-based and uranium-based mixed-oxide fuel in a light water reactor. Although the uranium-based fuels outperformed the thorium-based fuels in achievable burnup, a depletion comparison of the initially charged plutonium (both reactor and weapons grade) showed that the thorium-based fuels outperformed the uranium-based fuels by more that a factor of 2, where >70% of the total plutonium in the thorium-based fuel is consumed during the cycle. This is significant considering that the achievable burnups of the thorium-based fuels were 1.4 to 4.6 times less than the uranium-based fuels for similar plutonium enrichments. For equal specific burnups of ~60 MWd/kg (i.e., using variable plutonium weight percentages to give the desired burnup), the thorium-based fuels still outperform the uranium-based fuels by more than a factor of 2, where the total plutonium consumption in a three-batch, 18-month cycle was 60 to 70%. This is fairly significant considering that 10 to 15% (by weight) more plutonium is needed in the thorium-based fuels as compared to the uranium-based fuels to achieve these burnups. Furthermore, thorium-based fuels could also be used as a strategy for reducing the amount of long-lived nuclides (including the minor actinides) and thus the radiotoxicity in spent nuclear fuel. Although the breeding of 233U is a concern, the presence of 232U and its daughter products (namely 208Tl) can aid in making this fuel self-protecting, and/or enough 238U can be added to denature the fissile uranium. From these calculations, it appears that thorium-based fuel for plutonium incineration is superior when compared to uranium-based fuel and should be considered as an alternative to traditional MOX in both current and future/advanced reactor designs.

  11. 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...

  12. Performance of Thorium-Based Mixed Oxide Fuels for the Consumption of Plutonium and Minor Actinides in Current and Advanced Reactors

    SciTech Connect (OSTI)

    Weaver, Kevan Dean; Herring, James Stephen

    2002-06-01

    A renewed interest in thorium-based fuels has arisen lately based on the need for proliferation resistance, longer fuel cycles, higher burnup and improved wasteform characteristics. Recent studies have been directed toward homogeneously mixed, heterogeneously mixed, and seed-and-blanket thorium-uranium fuel cycles that rely on "in situ" use of the bred-in U-233. However, due to the higher initial enrichment required to achieve acceptable burnups, these fuels are encountering economic constraints. Thorium can nevertheless play a large role in the nuclear fuel cycle; particularly in the reduction of plutonium. While uranium-based mixedoxide (MOX) fuel will decrease the amount of plutonium, the reduction is limited due to the breeding of more plutonium (and higher actinides) from the U-238. Here we present calculational results and a comparison of the potential burnup of a thorium-based and uranium-based mixed oxide fuel in a light water reactor (LWR). Although the uranium-based fuels outperformed the thorium-based fuels in achievable burnup, a depletion comparison of the initially charged plutonium (both reactor and weapons grade) showed that the thorium-based fuels outperformed the uranium-based fuels by more that a factor of 2; where more than 70% of the total plutonium in the thorium-based fuel is consumed during the cycle. This is significant considering that the achievable burnup of the thorium-based fuels were 1.4 to 4.6 times less than the uranium-based fuels. Furthermore, use of a thorium-based fuel could also be used as a strategy for reducing the amount of long-lived nuclides (including the minor actinides), and thus the radiotoxicity in spent nuclear fuel. Although the breeding of U-233 is a concern, the presence of U-232 and its daughter products can aid in making this fuel self-protecting, and/or enough U-238 can be added to denature the fissile uranium. From these calculations, it appears that thorium-based fuel for plutonium incineration is superior as compared to uranium-based fuel, and should be considered as an alternative to traditional MOX in both current and future reactor designs.

  13. User Guide for VISION 3.4.7 (Verifiable Fuel Cycle Simulation) Model

    SciTech Connect (OSTI)

    Jacob J. Jacobson; Robert F. Jeffers; Gretchen E. Matthern; Steven J. Piet; Wendell D. Hintze

    2011-07-01

    The purpose of this document is to provide a guide for using the current version of the Verifiable Fuel Cycle Simulation (VISION) model. This is a complex model with many parameters and options; the user is strongly encouraged to read this user guide before attempting to run the model. This model is an R&D work in progress and may contain errors and omissions. It is based upon numerous assumptions. This model is intended to assist in evaluating 'what if' scenarios and in comparing fuel, reactor, and fuel processing alternatives at a systems level. The model is not intended as a tool for process flow and design modeling of specific facilities nor for tracking individual units of fuel or other material through the system. The model is intended to examine the interactions among the components of a fuel system as a function of time varying system parameters; this model represents a dynamic rather than steady-state approximation of the nuclear fuel system. VISION models the nuclear cycle at the system level, not individual facilities, e.g., 'reactor types' not individual reactors and 'separation types' not individual separation plants. Natural uranium can be enriched, which produces enriched uranium, which goes into fuel fabrication, and depleted uranium (DU), which goes into storage. Fuel is transformed (transmuted) in reactors and then goes into a storage buffer. Used fuel can be pulled from storage into either separation or disposal. If sent to separations, fuel is transformed (partitioned) into fuel products, recovered uranium, and various categories of waste. Recycled material is stored until used by its assigned reactor type. VISION is comprised of several Microsoft Excel input files, a Powersim Studio core, and several Microsoft Excel output files. All must be co-located in the same folder on a PC to function. You must use Powersim Studio 8 or better. We have tested VISION with the Studio 8 Expert, Executive, and Education versions. The Expert and Education versions work with the number of reactor types of 3 or less. For more reactor types, the Executive version is currently required. The input files are Excel2003 format (xls). The output files are macro-enabled Excel2007 format (xlsm). VISION 3.4 was designed with more flexibility than previous versions, which were structured for only three reactor types - LWRs that can use only uranium oxide (UOX) fuel, LWRs that can use multiple fuel types (LWR MF), and fast reactors. One could not have, for example, two types of fast reactors concurrently. The new version allows 10 reactor types and any user-defined uranium-plutonium fuel is allowed. (Thorium-based fuels can be input but several features of the model would not work.) The user identifies (by year) the primary fuel to be used for each reactor type. The user can identify for each primary fuel a contingent fuel to use if the primary fuel is not available, e.g., a reactor designated as using mixed oxide fuel (MOX) would have UOX as the contingent fuel. Another example is that a fast reactor using recycled transuranic (TRU) material can be designated as either having or not having appropriately enriched uranium oxide as a contingent fuel. Because of the need to study evolution in recycling and separation strategies, the user can now select the recycling strategy and separation technology, by year.

  14. 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...

  15. 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.

  16. Measures of the Environmental Footprint of the Front End of the Nuclear Fuel Cycle

    SciTech Connect (OSTI)

    Brett Carlsen; Emily Tavrides; Erich Schneider

    2010-08-01

    Previous estimates of environmental impacts associated with the front end of the nuclear fuel cycle have focused primarily on energy consumption and CO2 emissions. Results have varied widely. Section 2 of this report provides a summary of historical estimates. This study revises existing empirical correlations and their underlying assumptions to fit to a more complete set of existing data. This study also addresses land transformation, water withdrawals, and occupational and public health impacts associated with the processes of the front end of the once-through nuclear fuel cycle. These processes include uranium mining, milling, refining, conversion, enrichment, and fuel fabrication. Metrics are developed to allow environmental impacts to be summed across the full set of front end processes, including transportation and disposition of the resulting depleted uranium.

  17. 10 CFR 830 Major Modification Determination for Advanced Test Reactor LEU Fuel Conversion

    SciTech Connect (OSTI)

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

    2012-05-01

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

  18. 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.

  19. 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.

  20. 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.

  1. Resource intensities of the front end of the nuclear fuel cycle

    SciTech Connect (OSTI)

    Schneider, E.; Phathanapirom, U.; Eggert, R.; Collins, J.

    2013-07-01

    This paper presents resource intensities, including direct and embodied energy consumption, land and water use, associated with the processes comprising the front end of the nuclear fuel cycle. These processes include uranium extraction, conversion, enrichment, fuel fabrication and depleted uranium de-conversion. To the extent feasible, these impacts are calculated based on data reported by operating facilities, with preference given to more recent data based on current technologies and regulations. All impacts are normalized per GWh of electricity produced. Uranium extraction is seen to be the most resource intensive front end process. Combined, the energy consumed by all front end processes is equal to less than 1% of the electricity produced by the uranium in a nuclear reactor. Land transformation and water withdrawals are calculated at 8.07 m{sup 2} /GWh(e) and 1.37x10{sup 5} l/GWh(e), respectively. Both are dominated by the requirements of uranium extraction, which accounts for over 70% of land use and nearly 90% of water use.

  2. Molten-Salt Depleted-Uranium Reactor

    E-Print Network [OSTI]

    Dong, Bao-Guo; Gu, Ji-Yuan

    2015-01-01

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

  3. 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...

  4. 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...

  5. 2014 Domestic Uranium Production 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: AlternativeMonthly","10/2015"Monthly","10/2015" ,"Release7 Relative Standard Errors for Relative StandardCensusp2. U.S. uranium

  6. Spectroscopic Evidence for Uranium Bearing Precipitates in Vadose Zone

    E-Print Network [OSTI]

    of past nuclear fuel fabrication processes, uranium (U) has been recognized as one of the most widespreadHanfordsitesthatreceivedU-containingwastesduring its mission of Pu production between 1940 and 1990. Unirradiated fuel rod wastes were disposed to the 300 Area that included copper-uranium-nitric acid solutions and dissolved aluminum cladding (basic

  7. 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...

  8. 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.)

  9. 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.

  10. Accelerator-driven transmutation of spent fuel elements

    DOE Patents [OSTI]

    Venneri, Francesco (Los Alamos, NM); Williamson, Mark A. (Los Alamos, NM); Li, Ning (Los Alamos, NM)

    2002-01-01

    An apparatus and method is described for transmuting higher actinides, plutonium and selected fission products in a liquid-fuel subcritical assembly. Uranium may also be enriched, thereby providing new fuel for use in conventional nuclear power plants. An accelerator provides the additional neutrons required to perform the processes. The size of the accelerator needed to complete fuel cycle closure depends on the neutron efficiency of the supported reactors and on the neutron spectrum of the actinide transmutation apparatus. Treatment of spent fuel from light water reactors (LWRs) using uranium-based fuel will require the largest accelerator power, whereas neutron-efficient high temperature gas reactors (HTGRs) or CANDU reactors will require the smallest accelerator power, especially if thorium is introduced into the newly generated fuel according to the teachings of the present invention. Fast spectrum actinide transmutation apparatus (based on liquid-metal fuel) will take full advantage of the accelerator-produced source neutrons and provide maximum utilization of the actinide-generated fission neutrons. However, near-thermal transmutation apparatus will require lower standing

  11. Decommissioning of uranium mines in Canada

    SciTech Connect (OSTI)

    Zgola, M.B. [Atomic Energy Control Board, Ottawa, Ontario (Canada)

    1996-12-31

    The Atomic Energy Control Board (AECB) regulates the nuclear fuel cycle in Canada. This paper overviews the nature and function of the AECB; discusses its {open_quotes}site-specific{close_quotes} approach to regulating the decommissioning of uranium mining facilities; catalogues the location and status of inactive uranium tailings impoundments in Canada; and, summarizes the decommissioning work at the licensed Elliot Lake tailings impoundments.

  12. Reduced-reactivity-swing LEU fuel cycle analyses for HFR Petten

    SciTech Connect (OSTI)

    Deen, J.R.; Snelgrove, J.L.

    1985-01-01

    The primary objective of these low enriched uranium (LEU) fuel cycle analyses was to effect at least a 33% reduction in the reactivity swing now experienced in the high enriched uranium (HEU) cycle while minimizing increases in /sup 235/U loading and power peaking. All LEU equilibrium fuel cycle calculations were performed using either a 19- or 20-plate fuel element with 0.76-mm-thick meat and 0.5- or 0.6-mm-thick Cd wires as burnable absorbers and 16- or 17-plate control rod fuel followers with 0.76-mm-thick meat. Burnup-dependent microscopic cross sections were used for all heavy metals and fission products. A three-dimensional model was used to account for the effect of partially inserted control rods upon burnup profiles of fuel and of burnable absorbers and upon power peaking. The equilibrium cycle reactivity swing (or, equivalently control rod movement) was reduced by 50% using LEU fuel with U meat densities <4.8 Mg/m/sup 3/. 6 refs., 4 tabs.

  13. Analysis of palladium coatings to remove hydrogen isotopes from zirconium fuel rods in Canada deuterium uranium-pressurized heavy water reactors; Thermal and neutron diffusion effects

    SciTech Connect (OSTI)

    Stokes, C.L.; Buxbaum, R.E. )

    1992-05-01

    This paper reports that, in pressurized heavy water nuclear reactors of the type standardly used in Canada (Canada deuterium uranium-pressurized heavy water reactors), the zirconium alloy pressure tubes of the core absorb deuterium produced by corrosion reactions. This deuterium weakens the tubes through hydrogen embrittlement. Thin palladium coatings on the outside of the zirconium are analyzed as a method for deuterium removal. This coating is expected to catalyze the reaction D{sub 2} + 1/2O{sub 2} {r reversible} D{sub 2}O when O{sub 2} is added to the annular (insulating) gas in the tubes. Major reductions in the deuterium concentration and, hence, hydrogen embrittlement are predicted. Potential problems such as plating the tube geometry, neutron absorption, catalyst deactivation, radioactive waste production, and oxygen corrosion are shown to be manageable. Also, a simple set of equations are derived to calculate the effect on diffusion caused by neutron interactions. Based on calculations of ordinary and neutron flux induced diffusion, a palladium coating of 1 {times} 10{sup {minus}6} m is recommended. This would cost approximately $60,000 per reactor unit and should more than double reactor lifetime. Similar coatings and similar interdiffusion calculations might have broad applications.

  14. MOX Fuel Presentation to Duke Board of Directors

    National Nuclear Security Administration (NNSA)

    PuO 2 with 95% depleted UO 2 - Like LEU fuel pellets, MOX fuel pellets are primarily uranium * Fission power comes primarily from plutonium (Pu 239 ) instead of uranium (U 235 )...

  15. 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.

  16. 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.

  17. 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%.

  18. 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.

  19. 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)

  20. 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...

  1. 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.

  2. 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...

  3. 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...

  4. 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...

  5. 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...

  6. 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...

  7. 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...

  8. 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

  9. 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,

  10. 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...

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

    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-effectWorking With WIPPfinalUnexpectedofWyko NT33004. Uranium4. UraniumUraniumTrack

  12. 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

  13. The Uranium Institute 24th Annual Symposium

    E-Print Network [OSTI]

    Laughlin, Robert B.

    -239 for use in subsequent reactors. A fast neutron reactor is capable of producing more plutonium fuel than the uranium fuel it burns, leading to a breeder reactor. In addition, if the reactor is a fast with half lives of 30 years or less. The fast neutron reactor of preference was to be cooled with liquid

  14. Methods for Investigating Gas Bubble Formation in Uranium-Zirconium Alloys 

    E-Print Network [OSTI]

    Mews, Kathryn Ann Wright

    2013-05-06

    Uranium-zirconium alloy nuclear fuels have many advantages as compared with ceramic fuels, especially for fast reactor systems. However, metallic fuels aren’t currently used in commercial power production due in part to issues with fuel swelling...

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

    SciTech Connect (OSTI)

    Pereira, Mario M.; Slonecker, Bruce D.

    2012-06-01

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

  16. Environmental Assessment of Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Not Available

    1994-04-01

    The Department of Energy has completed the Environmental Assessment (EA) of Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel and issued a Finding of No Significant Impact (FONSI) for the proposed action. The EA and FONSI are enclosed for your information. The Department has decided to accept a limited number of spent nuclear fuel elements (409 elements) containing uranium that was enriched in the United States from eight research reactors in Austria, Denmark, Germany, Greece, the Netherlands, Sweden, and Switzerland. This action is necessary to maintain the viability of a major US nuclear weapons nonproliferation program to limit or eliminate the use of highly enriched uranium in civil programs. The purpose of the EA is to maintain the cooperation of the foreign research reactor operators with the nonproliferation program while a more extensive Environmental Impact Statement (EIS) is prepared on a proposed broader policy involving the acceptance of up to 15,000 foreign research reactor spent fuel elements over a 10 to 15 year period. Based on an evaluation of transport by commercial container liner or chartered vessel, five eastern seaboard ports, and truck and train modes of transporting the spent fuel overland to the Savannah River Sits, the Department has concluded that no significant impact would result from any combination of port and made of transport. In addition, no significant impacts were found from interim storage of spent fuel at the Savannah River Site.

  17. 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.

  18. Improving Natural Uranium Utilization By Using Thorium in Low Moderation PWRs - A Preliminary Neutronic Scoping Study

    SciTech Connect (OSTI)

    Gilles Youinou; Ignacio Somoza

    2010-10-01

    The Th-U fuel cycle is not quite self-sustainable when used in water-cooled reactors and with fuel burnups higher than a few thousand of MWd/t characteristic of CANDU reactors operating with a continuous refueling. For the other industrially mature water-cooled reactors (i.e. PWRs and BWRs) it is economically necessary that the fuel has enough reactivity to reach fuel burnups of the order of a few tens of thousand of MWd/t. In this particular case, an additional input of fissile material is necessary to complement the bred fissile U-233. This additional fissile material could be included in the form of Highly Enriched Uranium (HEU) at the fabrication of the Th-U fuel. The objective of this preliminary neutronic scoping study is to determine (1) how much HEU and, consequently, how much natural uranium is necessary in such Th-U fuel cycle with U recycling and (2) how much TRansUranics (TRU=Pu, Np, Am and Cm) are produced. These numbers are then compared with those of a standard UO2 PWR. The thorium reactors considered have a homogeneous hexagonal lattice made up of the same (Th-U)O2 pins. Furthermore, at this point, we are not considering the use of blankets inside or outside the core. The lattice pitch has been varied to estimate the effect of the water-to-fuel volume ratio, and light water as well as heavy water have been considered. For most cases, an average burnup at discharge of 45,000 MWd/t has been considered.

  19. Validation of ATR Fission Power Deposition Fraction in HEU and LEU Fuel Plates

    SciTech Connect (OSTI)

    G. S. Chang

    2008-09-01

    The Advanced Test Reactor (ATR) is a high power (250 MW), high neutron flux research reactor operating in the United States. Powered with highly enriched uranium (HEU), the ATR has a maximum unperturbed thermal neutron flux rating of 1.0 x 1015 n/cm2–s. Because of its high power and large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. A detailed plate-by-plate MCNP ATR full core model has been developed and validated for the low-enriched uranium (LEU) fuel conversion feasibility study. Using this model, an analysis has been performed to determine the LEU density and U-235 enrichment required in the fuel meat to yield equivalent K-eff versus effective full power days (EFPDs) between the HEU and LEU cores. This model has also been used to optimize U-235 content of the LEU core, minimizing the differences in K-eff and heat flux profile between the HEU and LEU cores at 115 MW total core power for 125 EFPDs. The LEU core conversion feasibility study evaluated foil type (U-10Mo) fuel with the LEU reference design of 19.7 wt% U-235 enrichment. The LEU reference design has a fixed fuel meat thickness of 0.330 mm and can sustain the same operating cycle length as the HEU fuel. Heat flux and fission power density are parameters that are proportional to the fraction of fission power deposited in fuel. Thus, the accurate determination of the fraction of fission power deposited in the fuel is important to ATR nuclear safety. In this work, a new approach was developed and validated, the Tally Fuel Cells Only (TFCO) method. This method calculates and compares the fission power deposition fraction between HEU and LEU fuel plates. Due to the high density of the U-10Mo LEU fuel, the fission ?-energy deposition fraction is 37.12%, which is larger than the HEU’s ?-energy deposition fraction of 19.7%. As a result, the fuel decay heat cooling will need to be improved. During the power operation, the total fission energy (200 MeV per fission) deposition fraction of LEU and HEU are 90.9% and 89.1%, respectively.

  20. Oxygen Enriched Combustion System Performance Study 

    E-Print Network [OSTI]

    Chen, S. L.; Kwan, Y.; Abele, A. R.; Silver, L. S.; Kobayashi, H.

    1987-01-01

    i ndustri al furnaces, such as gl ass melting furnaces, appear to be the most promising appli cations for oxygen enriched combustion. In these applications, the principal energy savings results from minimizing the fuel energy required to heat... with non-water cooled refractory burner til es. The Oxytherm Burner was developed jointly by Maxon and Corning Glass for application of oxygen/fuel combustion in glass furnaces. This burner is also a non-water cooled refractory design with a specially...

  1. Fuel Cycle System Analysis Handbook

    SciTech Connect (OSTI)

    Steven J. Piet; Brent W. Dixon; Dirk Gombert; Edward A. Hoffman; Gretchen E. Matthern; Kent A. Williams

    2009-06-01

    This Handbook aims to improve understanding and communication regarding nuclear fuel cycle options. It is intended to assist DOE, Campaign Managers, and other presenters prepare presentations and reports. When looking for information, check here. The Handbook generally includes few details of how calculations were performed, which can be found by consulting references provided to the reader. The Handbook emphasizes results in the form of graphics and diagrams, with only enough text to explain the graphic, to ensure that the messages associated with the graphic is clear, and to explain key assumptions and methods that cause the graphed results. Some of the material is new and is not found in previous reports, for example: (1) Section 3 has system-level mass flow diagrams for 0-tier (once-through), 1-tier (UOX to CR=0.50 fast reactor), and 2-tier (UOX to MOX-Pu to CR=0.50 fast reactor) scenarios - at both static and dynamic equilibrium. (2) To help inform fast reactor transuranic (TRU) conversion ratio and uranium supply behavior, section 5 provides the sustainable fast reactor growth rate as a function of TRU conversion ratio. (3) To help clarify the difference in recycling Pu, NpPu, NpPuAm, and all-TRU, section 5 provides mass fraction, gamma, and neutron emission for those four cases for MOX, heterogeneous LWR IMF (assemblies mixing IMF and UOX pins), and a CR=0.50 fast reactor. There are data for the first 10 LWR recycle passes and equilibrium. (4) Section 6 provides information on the cycle length, planned and unplanned outages, and TRU enrichment as a function of fast reactor TRU conversion ratio, as well as the dilution of TRU feedstock by uranium in making fast reactor fuel. (The recovered uranium is considered to be more pure than recovered TRU.) The latter parameter impacts the required TRU impurity limits specified by the Fuels Campaign. (5) Section 7 provides flows for an 800-tonne UOX separation plant. (6) To complement 'tornado' economic uncertainty diagrams, which show at a glance combined uncertainty information, section 9.2 has a new set of simpler graphs that show the impact on fuel cycle costs for once through, 1-tier, and 2-tier scenarios as a function of key input parameters.

  2. 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...

  3. Fluid-Structure Interaction Modeling of High-Aspect Ratio Nuclear Fuel Plates Using COMSOL

    SciTech Connect (OSTI)

    Curtis, Franklin G [ORNL] [ORNL; Ekici, Kivanc [ORNL] [ORNL; Freels, James D [ORNL] [ORNL

    2013-01-01

    The High Flux Isotope Reactor at the Oak Ridge National Lab is in the research stage of converting its fuel from high-enriched uranium to low-enriched uranium. Due to different physical properties of the new fuel and changes to the internal fuel plate design, the current safety basis must be re-evaluated through rigorous computational analyses. One of the areas being explored is the fluid-structure interaction phenomenon due to the interaction of thin fuel plates (50 mils thickness) and the cooling fluid (water). Detailed computational fluid dynamics and fluid-structure interaction simulations have only recently become feasible due to improved numerical algorithms and advancements in computing technology. For many reasons including the already built-in fluid-structure interaction module, COMSOL has been chosen for this complex problem. COMSOL's ability to solve multiphysics problems using a fully-coupled and implicit solution algorithm is crucial in obtaining a stable and accurate solution. Our initial findings show that COMSOL can accurately model such problems due to its ability to closely couple the fluid dynamics and the structural dynamics problems.

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

    National Nuclear Security Administration (NNSA)

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

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

    National Nuclear Security Administration (NNSA)

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

  7. Determining Reactor Flux from Xenon-136 and Cesium-135 in Spent Fuel

    E-Print Network [OSTI]

    A. C. Hayes; Gerard Jungman

    2012-05-30

    The ability to infer the reactor flux from spent fuel or seized fissile material would enhance the tools of nuclear forensics and nuclear nonproliferation significantly. We show that reactor flux can be inferred from the ratios of xenon-136 to xenon-134 and cesium-135 to cesium-137. If the average flux of a reactor is known, the flux inferred from measurements of spent fuel could help determine whether that spent fuel was loaded as a blanket or close to the mid-plane of the reactor. The cesium ratio also provides information on reactor shutdowns during the irradiation of fuel, which could prove valuable for identifying the reactor in question through comparisons with satellite reactor heat monitoring data. We derive analytic expressions for these correlations and compare them to experimental data and to detailed reactor burn simulations. The enrichment of the original uranium fuel affects the correlations by up to 3 percent, but only at high flux.

  8. Objectives, Strategies, and Challenges for the Advanced Fuel Cycle Initiative

    SciTech Connect (OSTI)

    Steven Piet; Brent Dixon; David Shropshire; Robert Hill; Roald Wigeland; Erich Schneider; J. D. Smith

    2005-04-01

    This paper will summarize the objectives, strategies, and key chemical separation challenges for the Advanced Fuel Cycle Initiative (AFCI). The major objectives are as follows: Waste management - defer the need for a second geologic repository for a century or more, Proliferation resistance - be more resistant than the existing PUREX separation technology or uranium enrichment, Energy sustainability - turn waste management liabilities into energy source assets to ensure that uranium ore resources do not become a constraint on nuclear power, and Systematic, safe, and economic management of the entire fuel cycle. There are four major strategies for the disposal of civilian spent fuel: Once-through - direct disposal of all discharged nuclear fuel, Limited recycle - recycle transuranic elements once and then direct disposal, Continuous recycle - recycle transuranic elements repeatedly, and Sustained recycle - same as continuous except previously discarded depleted uranium is also recycled. The key chemical separation challenges stem from the fact that the components of spent nuclear fuel vary greatly in their influence on achieving program objectives. Most options separate uranium to reduce the weight and volume of waste and the number and cost of waste packages that require geologic disposal. Separated uranium can also be used as reactor fuel. Most options provide means to recycle transuranic (TRU) elements - plutonium (Pu), neptunium (Np), americium (Am), curium (Cm). Plutonium must be recycled to obtain repository, proliferation, and energy recovery benefits. U.S. non-proliferation policy forbids separation of plutonium by itself; therefore, one or more of the other transuranic elements must be kept with the plutonium; neptunium is considered the easiest option. Recycling neptunium also provides repository benefits. Americium recycling is also required to obtain repository benefits. At the present time, curium recycle provides relatively little benefit; indeed, recycling curium in thermal reactors would significantly increase the hazard (hence cost) of the resulting fuel. Most options separate short-lived fission products cesium and strontium to allow them to decay in separate storage facilities tailored to that need, rather than complicate long-term geologic disposal. This can also reduce the number and cost of waste packages requiring geologic disposal. These savings are balanced by costs for separation and recycle systems. Several long-lived fission products, such as technetium-99 and iodine-129 go to geologic disposal in improved waste forms, recognizing that transmutation of these isotopes would be a slow process; however, the program has not precluded their transmutation as a future alternative.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. Nuclear-fuel-cycle risk assessment: descriptions of representative non-reactor facilities. Sections 1-14

    SciTech Connect (OSTI)

    Schneider, K.J.

    1982-09-01

    The Fuel Cycle Risk Assessment Program was initiated to provide risk assessment methods for assistance in the regulatory process for nuclear fuel cycle facilities other than reactors. This report, the first from the program, defines and describes fuel cycle elements that are being considered in the program. One type of facility (and in some cases two) is described that is representative of each element of the fuel cycle. The descriptions are based on real industrial-scale facilities that are current state-of-the-art, or on conceptual facilities where none now exist. Each representative fuel cycle facility is assumed to be located on the appropriate one of four hypothetical but representative sites described. The fuel cycles considered are for Light Water Reactors with once-through flow of spent fuel, and with plutonium and uranium recycle. Representative facilities for the following fuel cycle elements are described for uranium (or uranium plus plutonium where appropriate): mining, milling, conversion, enrichment, fuel fabrication, mixed-oxide fuel refabrication, fuel reprocessing, spent fuel storage, high-level waste storage, transuranic waste storage, spent fuel and high-level and transuranic waste disposal, low-level and intermediate-level waste disposal, and transportation. For each representative facility the description includes: mainline process, effluent processing and waste management, facility and hardware description, safety-related information and potential alternative concepts for that fuel cycle element. The emphasis of the descriptive material is on safety-related information. This includes: operating and maintenance requirements, input/output of major materials, identification and inventories of hazardous materials (particularly radioactive materials), unit operations involved, potential accident driving forces, containment and shielding, and degree of hands-on operation.

  14. Issues in the use of Weapons-Grade MOX Fuel in VVER-1000 Nuclear Reactors: Comparison of UO2 and MOX Fuels

    SciTech Connect (OSTI)

    Carbajo, J.J.

    2005-05-27

    The purpose of this report is to quantify the differences between mixed oxide (MOX) and low-enriched uranium (LEU) fuels and to assess in reasonable detail the potential impacts of MOX fuel use in VVER-1000 nuclear power plants in Russia. This report is a generic tool to assist in the identification of plant modifications that may be required to accommodate receiving, storing, handling, irradiating, and disposing of MOX fuel in VVER-1000 reactors. The report is based on information from work performed by Russian and U.S. institutions. The report quantifies each issue, and the differences between LEU and MOX fuels are described as accurately as possible, given the current sources of data.

  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. GLOBAL MONITORING OF URANIUM HEXIFLORIDE CYLINDERS NEXT STEPS IN DEVELOPMENT OF AN ACTION PLAN

    SciTech Connect (OSTI)

    Hanks, D.

    2010-06-09

    Over 40 industrial facilities world-wide use standardized uranium hexafluoride (UF{sub 6}) cylinders for transport, storage and in-process receiving in support of uranium conversion, enrichment and fuel fabrication processes. UF{sub 6} is processed and stored in the cylinders, with over 50,000 tU of UF{sub 6} transported each year in these International Organization for Standardization (ISO) qualified containers. Although each cylinder is manufactured to an ISO standard that calls for a nameplate with the manufacturer's identification number (ID) and the owner's serial number engraved on it, these can be quite small and difficult to read. Recognizing that each facility seems to use a different ID, a cylinder can have several different numbers recorded on it by means of metal plates, sticky labels, paint or even marker pen as it travels among facilities around the world. The idea of monitoring movements of UF{sub 6} cylinders throughout the global uranium fuel cycle has become a significant issue among industrial and safeguarding stakeholders. Global monitoring would provide the locations, movements, and uses of cylinders in commercial nuclear transport around the world, improving the efficiency of industrial operations while increasing the assurance that growing nuclear commerce does not result in the loss or misuse of cylinders. It should be noted that a unique ID (UID) attached to a cylinder in a verifiable manner is necessary for safeguarding needs and ensuring positive ID, but not sufficient for an effective global monitoring system. Modern technologies for tracking and inventory control can pair the UID with sensors and secure data storage for content information and complete continuity of knowledge over the cylinder. This paper will describe how the next steps in development of an action plan for employing a global UF{sub 6} cylinder monitoring network could be cultivated using four primary UID functions - identification, tracking, controlling, and accounting.

  17. 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.

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

    SciTech Connect (OSTI)

    Not Available

    1980-01-01

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

  19. Reduced enrichment for research and test reactors: Proceedings

    SciTech Connect (OSTI)

    Not Available

    1993-07-01

    The 15th annual Reduced Enrichment for Research and Test Reactors (RERTR) international meeting was organized by Ris{o} National Laboratory in cooperation with the International Atomic Energy Agency and Argonne National Laboratory. The topics of the meeting were the following: National Programs, Fuel Fabrication, Licensing Aspects, States of Conversion, Fuel Testing, and Fuel Cycle. Individual papers have been cataloged separately.

  20. 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 ...