Sample records for average fuel enrichment

  1. Thermal breeder fuel enrichment zoning

    DOEpatents

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

    1992-01-01

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

  2. Fuel rod with annular nuclear fuel pellets having same U-235 enrichment and different annulus sizes for graduated enrichment loading

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Mildrum, C.M.

    1987-08-18

    A fuel rod is described for a nuclear reactor fuel assembly, comprising: (a) a hollow cladding tube; (b) a pair of end plugs connected to and sealing the cladding tube at opposite ends thereof; (c) a plurality of fuel pellets contained on the tube and being composed of fissile material having a single enrichment the value of which is at the level of the maximum enrichment loading of the rod, the pellets having provided in a stack having one end disposed adjacent to one of the end plugs and an opposite end disposed remote from the other of the endmore » plugs; and (d) a plenum spring disposed in the tube between the other end plug and the opposite end of the pellet stack for retaining the pellets in a stack form; (e) at least some of the fuel pellets having an annular configuration and at least other of the fuel pellets having a solid configuration; (f) each of some of the annular fuel pellets having an annulus of a first size; (e) each of other of the annual fuel pellets having an annulus of a second size different from the first size, whereby graduation of axial enrichment loading is provided between the annual fuel pellets of the fuel rod.« less

  3. Hydrogen-enriched fuels

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Roser, R.

    1998-08-01

    NRG Technologies, Inc. is attempting to develop hardware and infrastructure that will allow mixtures of hydrogen and conventional fuels to become viable alternatives to conventional fuels alone. This commercialization can be successful if the authors are able to achieve exhaust emission levels of less than 0.03 g/kw-hr NOx and CO; and 0.15 g/kw-hr NMHC at full engine power without the use of exhaust catalysts. The major barriers to achieving these goals are that the lean burn regimes required to meet exhaust emissions goals reduce engine output substantially and tend to exhibit higher-than-normal total hydrocarbon emissions. Also, hydrogen addition to conventionalmore » fuels increases fuel cost, and reduces both vehicle range and engine output power. Maintaining low emissions during transient driving cycles has not been demonstrated. A three year test plan has been developed to perform the investigations into the issues described above. During this initial year of funding research has progressed in the following areas: (a) a cost effective single-cylinder research platform was constructed; (b) exhaust gas speciation was performed to characterize the nature of hydrocarbon emissions from hydrogen-enriched natural gas fuels; (c) three H{sub 2}/CH{sub 4} fuel compositions were analyzed using spark timing and equivalence ratio sweeping procedures and finally; (d) a full size pick-up truck platform was converted to run on HCNG fuels. The testing performed in year one of the three year plan represents a baseline from which to assess options for overcoming the stated barriers to success.« less

  4. 49 CFR 537.9 - Determination of fuel economy values and average fuel economy.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 49 Transportation 6 2014-10-01 2014-10-01 false Determination of fuel economy values and average fuel economy. 537.9 Section 537.9 Transportation Other Regulations Relating to Transportation... ECONOMY REPORTS § 537.9 Determination of fuel economy values and average fuel economy. (a) Vehicle...

  5. 49 CFR 537.9 - Determination of fuel economy values and average fuel economy.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 49 Transportation 6 2012-10-01 2012-10-01 false Determination of fuel economy values and average fuel economy. 537.9 Section 537.9 Transportation Other Regulations Relating to Transportation... ECONOMY REPORTS § 537.9 Determination of fuel economy values and average fuel economy. (a) Vehicle...

  6. 49 CFR 537.9 - Determination of fuel economy values and average fuel economy.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 49 Transportation 6 2013-10-01 2013-10-01 false Determination of fuel economy values and average fuel economy. 537.9 Section 537.9 Transportation Other Regulations Relating to Transportation... ECONOMY REPORTS § 537.9 Determination of fuel economy values and average fuel economy. (a) Vehicle...

  7. 49 CFR 537.9 - Determination of fuel economy values and average fuel economy.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 49 Transportation 6 2011-10-01 2011-10-01 false Determination of fuel economy values and average fuel economy. 537.9 Section 537.9 Transportation Other Regulations Relating to Transportation... ECONOMY REPORTS § 537.9 Determination of fuel economy values and average fuel economy. (a) Vehicle...

  8. 40 CFR 600.510-08 - Calculation of average fuel economy.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 40 Protection of Environment 31 2012-07-01 2012-07-01 false Calculation of average fuel economy...) ENERGY POLICY FUEL ECONOMY AND GREENHOUSE GAS EXHAUST EMISSIONS OF MOTOR VEHICLES Procedures for Determining Manufacturer's Average Fuel Economy and Manufacturer's Average Carbon-Related Exhaust Emissions...

  9. 40 CFR 600.510-08 - Calculation of average fuel economy.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 40 Protection of Environment 31 2013-07-01 2013-07-01 false Calculation of average fuel economy...) ENERGY POLICY FUEL ECONOMY AND GREENHOUSE GAS EXHAUST EMISSIONS OF MOTOR VEHICLES Procedures for Determining Manufacturer's Average Fuel Economy and Manufacturer's Average Carbon-Related Exhaust Emissions...

  10. 76 FR 67765 - Notice of Availability of Uranium Enrichment Fuel Cycle Facility's Inspection Reports Regarding...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-11-02

    ... Uranium Enrichment Fuel Cycle Facility's Inspection Reports Regarding Louisiana Energy Services, National..., Uranium Enrichment Branch, Division of Fuel Cycle Safety and Safeguards, Office of Nuclear Material Safety... Commission. Brian W. Smith, Chief, Uranium Enrichment Branch, Division of Fuel Cycle Safety and Safeguards...

  11. 75 FR 44817 - Notice of Availability of Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-07-29

    ... Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services, National... Enrichment Branch, Division of Fuel Cycle Safety and Safeguards, Office of Nuclear Material Safety and... Enrichment Branch, Division of Fuel Cycle Safety and Safeguards, Office of Nuclear Material Safety and...

  12. 40 CFR 600.510-12 - Calculation of average fuel economy and average carbon-related exhaust emissions.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 29 2010-07-01 2010-07-01 false Calculation of average fuel economy and average carbon-related exhaust emissions. 600.510-12 Section 600.510-12 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) ENERGY POLICY FUEL ECONOMY AND CARBON-RELATED EXHAUST EMISSIONS OF MOTOR VEHICLES Fuel Economy Regulation...

  13. Maximizing power generation from dark fermentation effluents in microbial fuel cell by selective enrichment of exoelectrogens and optimization of anodic operational parameters.

    PubMed

    Varanasi, Jhansi L; Sinha, Pallavi; Das, Debabrata

    2017-05-01

    To selectively enrich an electrogenic mixed consortium capable of utilizing dark fermentative effluents as substrates in microbial fuel cells and to further enhance the power outputs by optimization of influential anodic operational parameters. A maximum power density of 1.4 W/m 3 was obtained by an enriched mixed electrogenic consortium in microbial fuel cells using acetate as substrate. This was further increased to 5.43 W/m 3 by optimization of influential anodic parameters. By utilizing dark fermentative effluents as substrates, the maximum power densities ranged from 5.2 to 6.2 W/m 3 with an average COD removal efficiency of 75% and a columbic efficiency of 10.6%. A simple strategy is provided for selective enrichment of electrogenic bacteria that can be used in microbial fuel cells for generating power from various dark fermentative effluents.

  14. 77 FR 65729 - Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services LLC...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-10-30

    ... NUCLEAR REGULATORY COMMISSION [Docket No. 70-3103; NRC-2010-0264] Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services LLC, National Enrichment Facility, Eunice..., Chief, Uranium Enrichment Branch, Division of Fuel Cycle Safety and Safeguards, Office of Nuclear...

  15. 49 CFR 537.9 - Determination of fuel economy values and average fuel economy.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 49 Transportation 6 2010-10-01 2010-10-01 false Determination of fuel economy values and average fuel economy. 537.9 Section 537.9 Transportation Other Regulations Relating to Transportation (Continued) NATIONAL HIGHWAY TRAFFIC SAFETY ADMINISTRATION, DEPARTMENT OF TRANSPORTATION AUTOMOTIVE FUEL ECONOMY REPORTS § 537.9 Determination of fuel...

  16. 78 FR 23312 - Uranium Enrichment Fuel Cycle Inspection Reports Regarding Louisiana Energy Services, National...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-04-18

    ... NUCLEAR REGULATORY COMMISSION [Docket No. 70-3103; NRC-2010-0264] Uranium Enrichment Fuel Cycle Inspection Reports Regarding Louisiana Energy Services, National Enrichment Facility, Eunice, New Mexico..., Division of Fuel Cycle Safety, and Safeguards Office of Nuclear Material Safety, and Safeguards. [FR Doc...

  17. 40 CFR 600.510-08 - Calculation of average fuel economy.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... Regulations for Model Year 1978 Passenger Automobiles and for 1979 and Later Model Year Automobiles (Light Trucks and Passenger Automobiles)-Procedures for Determining Manufacturer's Average Fuel Economy and...) Average fuel economy will be calculated to the nearest 0.1 mpg for the classes of automobiles identified...

  18. 40 CFR 600.510-86 - Calculation of average fuel economy.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... Regulations for Model Year 1978 Passenger Automobiles and for 1979 and Later Model Year Automobiles (Light Trucks and Passenger Automobiles)-Procedures for Determining Manufacturer's Average Fuel Economy and...) Average fuel economy will be calculated to the nearest 0.1 mpg for the classes of automobiles identified...

  19. 40 CFR 600.510-93 - Calculation of average fuel economy.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... Regulations for Model Year 1978 Passenger Automobiles and for 1979 and Later Model Year Automobiles (Light Trucks and Passenger Automobiles)-Procedures for Determining Manufacturer's Average Fuel Economy and...) Average fuel economy will be calculated to the nearest 0.1 mpg for the classes of automobiles identified...

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

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Parker, Frank L.

    2012-07-01

    Current fuel cycles offer a number of opportunities for access to plutonium, opportunities to create highly enriched uranium and access highly radioactive wastes to create nuclear weapons and 'dirty' bombs. The non-proliferating fuel cycle however eliminates or reduces such opportunities and access by eliminating the mining, milling and enrichment of uranium. The non-proliferating fuel cycle also reduces the production of plutonium per unit of energy created, eliminates reprocessing and the separation of plutonium from the spent fuel and the creation of a stream of high-level waste. It further simplifies the search for land based deep geologic repositories and interim storagemore » 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

  1. High shear enrichment improves the performance of the anodophilic microbial consortium in a microbial fuel cell

    PubMed Central

    Pham, Hai The; Boon, Nico; Aelterman, Peter; Clauwaert, Peter; De Schamphelaire, Liesje; Van Oostveldt, Patrick; Verbeken, Kim; Rabaey, Korneel; Verstraete, Willy

    2008-01-01

    Summary In many microbial bioreactors, high shear rates result in strong attachment of microbes and dense biofilms. In this study, high shear rates were applied to enrich an anodophilic microbial consortium in a microbial fuel cell (MFC). Enrichment at a shear rate of about 120 s−1 resulted in the production of a current and power output two to three times higher than those in the case of low shear rates (around 0.3 s−1). Biomass and biofilm analyses showed that the anodic biofilm from the MFC enriched under high shear rate conditions, in comparison with that under low shear rate conditions, had a doubled average thickness and the biomass density increased with a factor 5. The microbial community of the former, as analysed by DGGE, was significantly different from that of the latter. The results showed that enrichment by applying high shear rates in an MFC can result in a specific electrochemically active biofilm that is thicker and denser and attaches better, and hence has a better performance. PMID:21261869

  2. Manufacturer's Policies Concerning Average Fuel Economy Standards

    DOT National Transportation Integrated Search

    1979-01-01

    The National Highway Traffic Safety Administration (NHTSA) has been given the responsibility for implementing the average fuel economy standards for passenger automobiles mandated by the Energy Policy and Conservation Act (P.L. 94-163). The standards...

  3. Experimental evaluation of oxygen-enriched air and emulsified fuels in a six-cylinder diesel engine

    NASA Astrophysics Data System (ADS)

    Sekar, R. R.; Marr, W. W.; Cole, R. L.; Marciniak, T. J.; Longman, D. E.

    1993-01-01

    The objectives of this investigation are to (1) determine the technical feasibility of using oxygen-enriched air to increase the efficiency of and reduce emissions from diesel engines, (2) examine the effects of water-emulsified fuel on the formation of nitrogen oxides in oxygen-enriched combustion, and (3) investigate the use of lower-grade fuels in high-speed diesel engines by emulsifying the fuel with water. These tests, completed on a Caterpillar model 3406B, six-cylinder engine are a scale-up from previous, single-cylinder-engine tests. The engine was tested with (1) intake-air oxygen levels up to 30%, (2) water content up to 20% of the fuel, (3) three fuel-injection timings, and (4) three fuel-flow rates (power levels). The Taguchi technique for experimental design was used to minimize the number of experimental points in the test matrix. Four separate test matrices were run to cover two different fuel-flow-rate strategies and two different fuels (No. 2 diesel and No. 6 diesel). A liquid-oxygen tank located outside the test cell supplied the oxygen for the tests. The only modification of the engine was installation of a pressure transducer in one cylinder. All tests were run at 1800 rpm, which corresponds to the synchronous speed of a 60-Hz generator. Test results show that oxygen enrichment results in power increases of 50% or more while significantly decreasing the levels of smoke and particulates emitted. The increase in power was accompanied by a small increase in thermal efficiency. Maximum engine power was limited by the test-cell dynamometer capacity and the capacity of the fuel-injection pump. Oxygen enrichment increases nitrogen-oxide emissions significantly. No adverse effects of oxygen enrichment on the turbocharger were observed. The engine operated successfully with No. 6 fuel, but it operated at a lower thermal efficiency and emitted more smoke and particulates than with No. 2 fuel.

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

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Renfro, David G; Chandler, David; Cook, David Howard

    2014-11-01

    Engineering design studies of the feasibility of conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL) as part of an effort sponsored by the U.S. Department of Energy s Global Threat Reduction Initiative (GTRI)/Reduced Enrichment for Research and Test Reactors (RERTR) program. The fuel type selected by the program for the conversion of the five high-power research reactors in the U.S. that still use HEU fuel is a new U-Mo monolithic fuel. Studies by ORNL have previously indicated that HFIR can be successfully convertedmore » 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

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

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Renfro, David; Chandler, David; Cook, David

    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 usingmore » 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

  6. 40 CFR 600.510-08 - Calculation of average fuel economy.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 29 2010-07-01 2010-07-01 false Calculation of average fuel economy. 600.510-08 Section 600.510-08 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) ENERGY POLICY FUEL ECONOMY AND CARBON-RELATED EXHAUST EMISSIONS OF MOTOR VEHICLES Fuel Economy Regulations for Model Year 1978 Passenger Automobiles...

  7. 40 CFR 600.510-86 - Calculation of average fuel economy.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 29 2010-07-01 2010-07-01 false Calculation of average fuel economy. 600.510-86 Section 600.510-86 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) ENERGY POLICY FUEL ECONOMY AND CARBON-RELATED EXHAUST EMISSIONS OF MOTOR VEHICLES Fuel Economy Regulations for Model Year 1978 Passenger Automobiles...

  8. 40 CFR 600.510-93 - Calculation of average fuel economy.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 29 2010-07-01 2010-07-01 false Calculation of average fuel economy. 600.510-93 Section 600.510-93 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) ENERGY POLICY FUEL ECONOMY AND CARBON-RELATED EXHAUST EMISSIONS OF MOTOR VEHICLES Fuel Economy Regulations for Model Year 1978 Passenger Automobiles...

  9. 49 CFR 525.11 - Termination of exemption; amendment of alternative average fuel economy standard.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... average fuel economy standard. 525.11 Section 525.11 Transportation Other Regulations Relating to... EXEMPTIONS FROM AVERAGE FUEL ECONOMY STANDARDS § 525.11 Termination of exemption; amendment of alternative average fuel economy standard. (a) Any exemption granted under this part for an affected model year does...

  10. 49 CFR 525.11 - Termination of exemption; amendment of alternative average fuel economy standard.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... average fuel economy standard. 525.11 Section 525.11 Transportation Other Regulations Relating to... EXEMPTIONS FROM AVERAGE FUEL ECONOMY STANDARDS § 525.11 Termination of exemption; amendment of alternative average fuel economy standard. (a) Any exemption granted under this part for an affected model year does...

  11. 49 CFR 525.11 - Termination of exemption; amendment of alternative average fuel economy standard.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... average fuel economy standard. 525.11 Section 525.11 Transportation Other Regulations Relating to... EXEMPTIONS FROM AVERAGE FUEL ECONOMY STANDARDS § 525.11 Termination of exemption; amendment of alternative average fuel economy standard. (a) Any exemption granted under this part for an affected model year does...

  12. 49 CFR 525.11 - Termination of exemption; amendment of alternative average fuel economy standard.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... average fuel economy standard. 525.11 Section 525.11 Transportation Other Regulations Relating to... EXEMPTIONS FROM AVERAGE FUEL ECONOMY STANDARDS § 525.11 Termination of exemption; amendment of alternative average fuel economy standard. (a) Any exemption granted under this part for an affected model year does...

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

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Sekar, R.R.; Marr, W.W.; Cole, R.L.

    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 airmore » 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.« less

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

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    K. J. Allen; I. Bolshinsky; L. L. Biro

    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 Horiamore » 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.« less

  15. Uncertainties in Estimates of Fleet Average Fuel Economy : A Statistical Evaluation

    DOT National Transportation Integrated Search

    1977-01-01

    Research was performed to assess the current Federal procedure for estimating the average fuel economy of each automobile manufacturer's new car fleet. Test vehicle selection and fuel economy estimation methods were characterized statistically and so...

  16. Steady-State Thermal-Hydraulics Analyses for the Conversion of BR2 to Low Enriched Uranium Fuel

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Licht, J.; Bergeron, A.; Dionne, B.

    The code PLTEMP/ANL version 4.2 was used to perform the steady-state thermal-hydraulic analyses of the BR2 research reactor for conversion from Highly-Enriched to Low Enriched Uranium fuel (HEU and LEU, respectively). Calculations were performed to evaluate different fuel assemblies with respect to the onset of nucleate boiling (ONB), flow instability (FI), critical heat flux (CHF) and fuel temperature at beginning of cycle conditions. The fuel assemblies were characteristic of fresh fuel (0% burnup), highest heat flux (16% burnup), highest power (32% burnup) and highest burnup (46% burnup). Results show that the high heat flux fuel element is limiting for ONB,more » FI, and CHF, for both HEU and LEU fuel, but that the high power fuel element produces similar margin in a few cases. The maximum fuel temperature similarly occurs in both the high heat flux and high power fuel assemblies for both HEU and LEU fuel. A sensitivity study was also performed to evaluate the variation in fuel temperature due to uncertainties in the thermal conductivity degradation associated with burnup.« less

  17. National Highway Traffic Safety Administration Corporate Average Fuel Economy (CAFE) Standards

    DOT National Transportation Integrated Search

    2003-01-01

    The National Highway Traffic Safety Administration (NHTSA) must set Corporate Average Fuel Economy (CAFE) standards for light trucks. This was authorized by the Energy Policy and Conservation Act, which added Title V: Imporving Automotive Fuel Effici...

  18. Processing of irradiated, enriched uranium fuels at the Savannah River Plant

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Hyder, M L; Perkins, W C; Thompson, M C

    Uranium fuels containing /sup 235/U at enrichments from 1.1% to 94% are processed and recovered, along with neptunium and plutonium byproducts. The fuels to be processed are dissolved in nitric acid. Aluminum-clad fuels are disssolved using a mercury catalyst to give a solution rich in aluminum. Fuels clad in more resistant materials are dissolved in an electrolytic dissolver. The resulting solutions are subjected to head-end treatment, including clarification and adjustment of acid and uranium concentration before being fed to solvent extraction. Uranium, neptunium, and plutonium are separated from fission products and from one another by multistage countercurrent solvent extraction withmore » dilute tri-n-butyl phosphate in kerosene. Nitric acid is used as the salting agent in addition to aluminum or other metal nitrates present in the feed solution. Nuclear safety is maintained through conservative process design and the use of monitoring devices as secondary controls. The enriched uranium is recovered as a dilute solution and shipped off-site for further processing. Neptunium is concentrated and sent to HB-Line for recovery from solution. The relatively small quantities of plutonium present are normally discarded in aqueous waste, unless the content of /sup 238/Pu is high enough to make its recovery desirable. Most of the /sup 238/Pu can be recovered by batch extraction of the waste solution, purified by counter-current solvent extraction, and converted to oxide in HB-Line. By modifying the flowsheet, /sup 239/Pu can be recovered from low-enriched uranium in the extraction cycle; neptunium is then not recovered. The solvent is subjected to an alkaline wash before reuse to remove degraded solvent and fission products. The aqueous waste is concentrated and partially deacidified by evaporation before being neutralized and sent to the waste tanks; nitric acid from the overheads is recovered for reuse.« less

  19. Enhanced Low-Enriched Uranium Fuel Element for the Advanced Test Reactor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Pope, M. A.; DeHart, M. D.; Morrell, S. R.

    2015-03-01

    Under the current US Department of Energy (DOE) policy and planning scenario, the Advanced Test Reactor (ATR) and its associated critical facility (ATRC) will be reconfigured to operate on low-enriched uranium (LEU) fuel. This effort has produced a conceptual design for an Enhanced LEU Fuel (ELF) element. This fuel features monolithic U-10Mo fuel foils and aluminum cladding separated by a thin zirconium barrier. As with previous iterations of the ELF design, radial power peaking is managed using different U-10Mo foil thicknesses in different plates of the element. The lead fuel element design, ELF Mk1A, features only three fuel meat thicknesses,more » a reduction from the previous iterations meant to simplify manufacturing. Evaluation of the ELF Mk1A fuel design against reactor performance requirements is ongoing, as are investigations of the impact of manufacturing uncertainty on safety margins. The element design has been evaluated in what are expected to be the most demanding design basis accident scenarios and has met all initial thermal-hydraulic criteria.« less

  20. Study on Characteristics of Co-firing Ammonia/Methane Fuels under Oxygen Enriched Combustion Conditions

    NASA Astrophysics Data System (ADS)

    Xiao, Hua; Wang, Zhaolin; Valera-Medina, Agustin; Bowen, Philip J.

    2018-06-01

    Having a background of utilising ammonia as an alternative fuel for power generation, exploring the feasibility of co-firing ammonia with methane is proposed to use ammonia to substitute conventional natural gas. However, improvement of the combustion of such fuels can be achieved using conditions that enable an increase of oxygenation, thus fomenting the combustion process of a slower reactive molecule as ammonia. Therefore, the present study looks at oxygen enriched combustion technologies, a proposed concept to improve the performance of ammonia/methane combustion. To investigate the characteristics of ammonia/methane combustion under oxygen enriched conditions, adiabatic burning velocity and burner stabilized laminar flame emissions were studied. Simulation results show that the oxygen enriched method can help to significantly enhance the propagation of ammonia/methane combustion without changing the emission level, which would be quite promising for the design of systems using this fuel for practical applications. Furthermore, to produce low computational-cost flame chemistry for detailed numerical analyses for future combustion studies, three reduced combustion mechanisms of the well-known Konnov's mechanism were compared in ammonia/methane flame simulations under practical gas turbine combustor conditions. Results show that the reduced reaction mechanisms can provide good results for further analyses of oxygen enriched combustion of ammonia/methane. The results obtained in this study also allow gas turbine designers and modellers to choose the most suitable mechanism for further combustion studies and development.

  1. 49 CFR 525.11 - Termination of exemption; amendment of alternative average fuel economy standard.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 49 Transportation 6 2010-10-01 2010-10-01 false Termination of exemption; amendment of alternative average fuel economy standard. 525.11 Section 525.11 Transportation Other Regulations Relating to Transportation (Continued) NATIONAL HIGHWAY TRAFFIC SAFETY ADMINISTRATION, DEPARTMENT OF TRANSPORTATION EXEMPTIONS FROM AVERAGE FUEL ECONOMY STANDARDS...

  2. 41 CFR 102-34.55 - Are there fleet average fuel economy standards we must meet?

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... fuel economy standards we must meet? 102-34.55 Section 102-34.55 Public Contracts and Property... average fuel economy standards we must meet? (a) Yes. 49 U.S.C. 32917 and Executive Order 12375 require that each executive agency meet the fleet average fuel economy standards in place as of January 1 of...

  3. 41 CFR 102-34.55 - Are there fleet average fuel economy standards we must meet?

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... fuel economy standards we must meet? 102-34.55 Section 102-34.55 Public Contracts and Property... average fuel economy standards we must meet? (a) Yes. 49 U.S.C. 32917 and Executive Order 12375 require that each executive agency meet the fleet average fuel economy standards in place as of January 1 of...

  4. 41 CFR 102-34.55 - Are there fleet average fuel economy standards we must meet?

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... fuel economy standards we must meet? 102-34.55 Section 102-34.55 Public Contracts and Property... average fuel economy standards we must meet? (a) Yes. 49 U.S.C. 32917 and Executive Order 12375 require that each executive agency meet the fleet average fuel economy standards in place as of January 1 of...

  5. 41 CFR 102-34.55 - Are there fleet average fuel economy standards we must meet?

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... fuel economy standards we must meet? 102-34.55 Section 102-34.55 Public Contracts and Property... average fuel economy standards we must meet? (a) Yes. 49 U.S.C. 32917 and Executive Order 12375 require that each executive agency meet the fleet average fuel economy standards in place as of January 1 of...

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

    DOEpatents

    Steinberg, Meyer; Powell, James R.; Takahashi, Hiroshi; Grand, Pierre; Kouts, Herbert

    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.

  7. 41 CFR 102-34.60 - How do we calculate the average fuel economy for Government motor vehicles?

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... average fuel economy for Government motor vehicles? 102-34.60 Section 102-34.60 Public Contracts and... How do we calculate the average fuel economy for Government motor vehicles? You must calculate the average fuel economy for Government motor vehicles as follows: (a) Because there are so many motor vehicle...

  8. 41 CFR 102-34.60 - How do we calculate the average fuel economy for Government motor vehicles?

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... average fuel economy for Government motor vehicles? 102-34.60 Section 102-34.60 Public Contracts and... How do we calculate the average fuel economy for Government motor vehicles? You must calculate the average fuel economy for Government motor vehicles as follows: (a) Because there are so many motor vehicle...

  9. 41 CFR 102-34.60 - How do we calculate the average fuel economy for Government motor vehicles?

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... average fuel economy for Government motor vehicles? 102-34.60 Section 102-34.60 Public Contracts and... How do we calculate the average fuel economy for Government motor vehicles? You must calculate the average fuel economy for Government motor vehicles as follows: (a) Because there are so many motor vehicle...

  10. 41 CFR 102-34.60 - How do we calculate the average fuel economy for Government motor vehicles?

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... average fuel economy for Government motor vehicles? 102-34.60 Section 102-34.60 Public Contracts and... How do we calculate the average fuel economy for Government motor vehicles? You must calculate the average fuel economy for Government motor vehicles as follows: (a) Because there are so many motor vehicle...

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

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Kelly Cummins; Igor Bolshinsky; Ken Allen

    2009-07-01

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

  12. Enrichment Zoning Options for the Small Nuclear Rocket Engine (SNRE)

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Bruce G. Schnitzler; Stanley K. Borowski

    2010-07-01

    the 564 fuel elements within 1% of the average element power. Results for this and alternate enrichment zoning options for the SNRE are compared.« less

  13. Evaluation of a uranium zirconium hydride fuel rod option for conversion of the MIT research reactor (MITR) from highly-enriched uranium to low-enriched uranium

    DOE PAGES

    Dunn, F. E.; Wilson, E. H.; Feldman, E. E.; ...

    2017-03-23

    The conversion of the Massachusetts Institute of Technology Reactor (MITR) from the use of highly-enriched uranium (HEU) fuel-plate assemblies to low-enriched uranium (LEU) by replacing the HEU fuel plates with specially designed General Atomics (GA) uranium zirconium hydride (UZrH) LEU fuel rods is evaluated in this paper. The margin to critical heat flux (CHF) in the core, which is cooled by light water at low pressure, is evaluated analytically for steady-state operation. A form of the Groeneveld CHF lookup table method is used and described in detail. A CHF ratio of 1.41 was found in the present analysis at 10more » MW with engineering hot channel factors included. Therefore, the nominal reactor core power, and neutron flux performance, would need to be reduced by at least 25% in order to meet the regulatory requirement of a minimum CHF ratio of 2.0.« less

  14. Evaluation of a uranium zirconium hydride fuel rod option for conversion of the MIT research reactor (MITR) from highly-enriched uranium to low-enriched uranium

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Dunn, F. E.; Wilson, E. H.; Feldman, E. E.

    The conversion of the Massachusetts Institute of Technology Reactor (MITR) from the use of highly-enriched uranium (HEU) fuel-plate assemblies to low-enriched uranium (LEU) by replacing the HEU fuel plates with specially designed General Atomics (GA) uranium zirconium hydride (UZrH) LEU fuel rods is evaluated in this paper. The margin to critical heat flux (CHF) in the core, which is cooled by light water at low pressure, is evaluated analytically for steady-state operation. A form of the Groeneveld CHF lookup table method is used and described in detail. A CHF ratio of 1.41 was found in the present analysis at 10more » MW with engineering hot channel factors included. Therefore, the nominal reactor core power, and neutron flux performance, would need to be reduced by at least 25% in order to meet the regulatory requirement of a minimum CHF ratio of 2.0.« less

  15. A lithotrophic microbial fuel cell operated with pseudomonads-dominated iron-oxidizing bacteria enriched at the anode

    PubMed Central

    Nguyen, Thuy Thu; Luong, Tha Thanh Thi; Tran, Phuong Hoang Nguyen; Bui, Ha Thi Viet; Nguyen, Huy Quang; Dinh, Hang Thuy; Kim, Byung Hong; Pham, Hai The

    2015-01-01

    In this study, we attempted to enrich neutrophilic iron bacteria in a microbial fuel cell (MFC)-type reactor in order to develop a lithotrophic MFC system that can utilize ferrous iron as an inorganic electron donor and operate at neutral pHs. Electrical currents were steadily generated at an average level of 0.6 mA (or 0.024 mA cm–2 of membrane area) in reactors initially inoculated with microbial sources and operated with 20 mM Fe2+ as the sole electron donor and 10 ohm external resistance; whereas in an uninoculated reactor (the control), the average current level only reached 0.2 mA (or 0.008 mA cm–2 of membrane area). In an inoculated MFC, the generation of electrical currents was correlated with increases in cell density of bacteria in the anode suspension and coupled with the oxidation of ferrous iron. Cultivation-based and denaturing gradient gel electrophoresis analyses both show the dominance of some Pseudomonas species in the anode communities of the MFCs. Fluorescent in-situ hybridization results revealed significant increases of neutrophilic iron-oxidizing bacteria in the anode community of an inoculated MFC. The results, altogether, prove the successful development of a lithotrophic MFC system with iron bacteria enriched at its anode and suggest a chemolithotrophic anode reaction involving some Pseudomonas species as key players in such a system. The system potentially offers unique applications, such as accelerated bioremediation or on-site biodetection of iron and/or manganese in water samples. PMID:25712332

  16. Development of Nitride Coating Using Atomic Layer Deposition for Low-Enriched Uranium Fuel Powder

    NASA Astrophysics Data System (ADS)

    Bhattacharya, Sumit

    High-performance research reactors require fuel that operates at high specific power and can withstand high fission density, but at relatively low temperatures. The design of the research reactor fuels is done for efficient heat emission, and consists of assemblies of thin-plates cladding made from aluminum alloy. The low-enriched fuels (LEU) were developed for replacing high-enriched fuels (HEU) for these reactors necessitates a significantly increased uranium density in the fuel to counterbalance the decrease in enrichment. One of the most promising new fuel candidate is U-Mo alloy, in a U-Mo/Al dispersion fuel form, due to its high uranium loading as well as excellent irradiation resistance performance, is being developed extensively to convert from HEU fuel to LEU fuel for high-performance research reactors. However, the formation of an interaction layer (IL) between U-Mo particles and the Al matrix, and the associated pore formation, under high heat flux and high burnup conditions, degrade the irradiation performance of the U-Mo/Al dispersion fuel. From the recent tests results accumulated from the surface engineering of low enriched uranium fuel (SELENIUM) and MIR reactor displayed that a surface barrier coating like physical vapor deposited (PVD) zirconium nitride (ZrN) can significantly reduce the interaction layer. The barrier coating performed well at low burn up but above a fluence rate of 5x 1021 ions/cm2 the swelling reappeared due to formation interaction layer. With this result in mind the objective of this research was to develop an ultrathin ZrN coating over particulate uranium-molybdenum nuclear fuel using a modified savannah 200 atomic layer deposition (ALD) system. This is done in support of the US Department of Energy's (DOE) effort to slow down the interaction at fluence rate and reach higher burn up for high power research reactor. The low-pressure Savannah 200 ALD system is modified to be designed as a batch powder coating system using the

  17. Department of Transportation, National Highway Traffic Safety Administration : light truck average fuel economy standard, model year 1999

    DOT National Transportation Integrated Search

    1997-04-18

    Section 32902(a) of title 49, United States Code, requires the Secretary of Transportation to prescribe by regulation, at least 18 months in advance of each model year, average fuel economy standards (known as "Corporate Average Fuel Economy" or "CAF...

  18. Enhancing clostridial acetone-butanol-ethanol (ABE) production and improving fuel properties of ABE-enriched biodiesel by extractive fermentation with biodiesel.

    PubMed

    Li, Qing; Cai, Hao; Hao, Bo; Zhang, Congling; Yu, Ziniu; Zhou, Shengde; Chenjuan, Liu

    2010-12-01

    The extractive acetone-butanol-ethanol (ABE) fermentations of Clostridium acetobutylicum were evaluated using biodiesel as the in situ extractant. The biodiesel preferentially extracted butanol, minimized product inhibition, and increased production of butanol (from 11.6 to 16.5 g L⁻¹) and total solvents (from 20.0 to 29.9 g L⁻¹) by 42% and 50%, respectively. The fuel properties of the ABE-enriched biodiesel obtained from the extractive fermentations were analyzed. The key quality indicators of diesel fuel, such as the cetane number (increased from 48 to 54) and the cold filter plugging point (decreased from 5.8 to 0.2 °C), were significantly improved for the ABE-enriched biodiesel. Thus, the application of biodiesel as the extractant for ABE fermentation would increase ABE production, bypass the energy intensive butanol recovery process, and result in an ABE-enriched biodiesel with improved fuel properties.

  19. Loading blended, low-enriched uranium fuel in browns ferry units 2 and 3

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Brown, C.; Eichenberg, T.; Haun, J.

    2006-07-01

    This paper summarizes fuel and cycle design results for the Tennessee Valley Authority (TVA) / Dept. of Energy (DOE) program to burn blended, low-enriched uranium (BLEU) material in the Browns Ferry Nuclear Units 2 and 3. The BLEU material typically has about 60 times the allowed limit of U-236 in what would be defined as commercial, i.e., virgin, uranium. U-236 in particular is a strong neutron absorber. Also included is a comparison of cycles using commercial uranium versus BLEU to determine the impact on key core design parameters of the high U-236 content in the BLEU. Finally, there is amore » short discussion of the economic advantages of BLEU fuel. (authors)« less

  20. The Military Significance of Small Uranium Enrichment Facilities Fed with Low-Enrichment Uranium (Redacted)

    DTIC Science & Technology

    1969-12-01

    a five-year supply of enriched uranium for reactor fuel . Nevertheless, it seems clear that some foreign enrichment developments are approaching a...produc- tion of fissile material could powerfully influence the assessment of risks and benefits of a nuclear weapons development program . Since... program is likely to include the production of its own relatively pure fissile plutonium. This would involve more rapid cycling and reprocessing of fuel

  1. DIissolution of low enriched uranium from the experimental breeder reactor-II fuel stored at the Idaho National Laboratory

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Daniel, G.; Rudisill, T.; Almond, P.

    The Idaho National Laboratory (INL) is actively engaged in the development of electrochemical processing technology for the treatment of fast reactor fuels using irradiated fuel from the Experimental Breeder Reactor-II (EBR-II) as the primary test material. The research and development (R&D) activities generate a low enriched uranium (LEU) metal product from the electrorefining of the EBR-II fuel and the subsequent consolidation and removal of chloride salts by the cathode processor. The LEU metal ingots from past R&D activities are currently stored at INL awaiting disposition. One potential disposition pathway is the shipment of the ingots to the Savannah River Sitemore » (SRS) for dissolution in H-Canyon. Carbon steel cans containing the LEU metal would be loaded into reusable charging bundles in the H-Canyon Crane Maintenance Area and charged to the 6.4D or 6.1D dissolver. The LEU dissolution would be accomplished as the final charge in a dissolver batch (following the dissolution of multiple charges of spent nuclear fuel (SNF)). The solution would then be purified and the 235U enrichment downblended to allow use of the U in commercial reactor fuel. To support this potential disposition path, the Savannah River National Laboratory (SRNL) developed a dissolution flowsheet for the LEU using samples of the material received from INL.« less

  2. Impact of thermal spectrum small modular reactors on performance of once-through nuclear fuel cycles with low-enriched uranium

    DOE PAGES

    Brown, Nicholas R.; Worrall, Andrew; Todosow, Michael

    2016-11-18

    Small modular reactors (SMRs) offer potential benefits, such as enhanced operational flexibility. However, it is vital to understand the holistic impact of SMRs on nuclear fuel cycle performance. The focus of this paper is the fuel cycle impacts of light water SMRs in a once-through fuel cycle with low-enriched uranium fuel. A key objective of this paper is to describe preliminary example reactor core physics and fuel cycle analyses conducted in support of the U.S. Department of Energy, Office of Nuclear Energy, Fuel Cycle Options Campaign. The hypothetical light water SMR example case considered in these preliminary scoping studies ismore » a cartridge type one-batch core with slightly less than 5.0% enrichment. Challenges associated with SMRs include increased neutron leakage, fewer assemblies in the core (and therefore fewer degrees of freedom in the core design), complex enrichment and burnable absorber loadings, full power operation with inserted control rods, the potential for frequent load-following operation, and shortened core height. Each of these will impact the achievable discharge burnup in the reactor and the fuel cycle performance. This paper summarizes a list of the factors relevant to SMR fuel, core, and operation that will impact fuel cycle performance. The high-level issues identified and preliminary scoping calculations in this paper are intended to inform on potential fuel cycle impacts of one-batch thermal spectrum SMRs. In particular, this paper highlights the impact of increased neutron leakage and reduced number of batches on the achievable burnup of the reactor. Fuel cycle performance metrics for a hypothetical example SMR are compared with those for a conventional three-batch light water reactor in the following areas: nuclear waste management, environmental impact, and resource utilization. The metrics performance for such an SMR is degraded for the mass of spent nuclear fuel and high-level waste disposed of, mass of depleted uranium

  3. Light-duty vehicle fuel economy improvements, 1979--1998: A consumer purchase model of corporate average fuel economy, fuel price, and income effects

    NASA Astrophysics Data System (ADS)

    Chien, David Michael

    2000-10-01

    The Energy Policy and Conservation Act of 1975, which created fuel economy standards for automobiles and light trucks, was passed by Congress in response to the rapid rise in world oil prices as a result of the 1973 oil crisis. The standards were first implemented in 1978 for automobiles and 1979 for light trucks, and began with initial standards of 18 MPG for automobiles and 17.2 MPG for light trucks. The current fuel economy standards for 1998 have been held constant at 27.5 MPG for automobiles and 20.5 MPG for light trucks since 1990--1991. While actual new automobile fuel economy has almost doubled from 14 MPG in 1974 to 27.2 MPG in 1994, it is reasonable to ask if the CAFE standards are still needed. Each year Congress attempts to pass another increase in the Corporate Average Fuel Economy (CAFE) standard and fails. Many have called for the abolition of CAFE standards citing the ineffectiveness of the standards in the past. In order to determine whether CAFE standards should be increased, held constant, or repealed, an evaluation of the effectiveness of the CAFE standards to date must be established. Because fuel prices were rising concurrently with the CAFE standards, many authors have attributed the rapid rise in new car fuel economy solely to fuel prices. The purpose of this dissertation is to re-examine the determinants of new car fuel economy via three effects: CAFE regulations, fuel price, and income effects. By measuring the marginal effects of the three fuel economy determinants upon consumers and manufacturers choices, for fuel economy, an estimate was made of the influence of each upon new fuel economy. The conclusions of this dissertation present some clear signals to policymakers: CAFE standards have been very effective in increasing fuel economy from 1979 to 1998. Furthermore, they have been the main cause of fuel economy improvement, with income being a much smaller component. Furthermore, this dissertation has suggested that fuel prices have

  4. Irradiated microstructure of U-10Mo monolithic fuel plate at very high fission density

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Gan, J.; Miller, B. D.; Keiser, D. D.

    Monolithic U-10Mo alloy fuel plates with Al-6061 cladding are being developed for use in research and test reactors as low enrichment fuel (< 20% U-235 enrichment) as a result of its high uranium loading capacity compared to that of U-7Mo dispersion fuel. These fuel plates contain a Zr diffusion barrier between the U-10Mo fuel and Al-6061 cladding that suppresses the interaction between the U-Mo fuel foil and Al alloy cladding that is known to be problematic under irradiation. This paper discusses the TEM results of the U-10Mo/Zr/Al6061 monolithic fuel plate (Plate ID: L1P09T, ~ 59% U-235 enrichment) irradiated in Advancedmore » Test Reactor at Idaho National Laboratory as part of RERTR-9B irradiation campaign with an unprecedented high local fission density of 9.8E+21 fissions/cm3. The calculated fuel foil centerline temperature at the beginning of life and the end of life is 141 and 194 C, respectively. A total of 5 TEM lamellas were prepared using focus ion beam lift-out technique. The estimated U-Mo fuel swelling, based on the fuel foil thickness change from SEM, is approximately 76%. Large bubbles (> 1 µm) are distributed evenly in U-Mo and interlink of these bubbles is evident. The average size of subdivided grains at this fission density appears similar to that at 5.2E+21 fissions/cm3. The measured average Mo and Zr content in the fuel matrix is ~ 30 at% and ~ 7 at%, respectively, in general agreement with the calculated Mo and Zr from fission density.« less

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

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Christopher Landers; Igor Bolshinsky; Ken Allen

    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 weremore » 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.« less

  6. Irradiated microstructure of U-10Mo monolithic fuel plate at very high fission density

    NASA Astrophysics Data System (ADS)

    Gan, J.; Miller, B. D.; Keiser, D. D.; Jue, J. F.; Madden, J. W.; Robinson, A. B.; Ozaltun, H.; Moore, G.; Meyer, M. K.

    2017-08-01

    Monolithic U-10Mo alloy fuel plates with Al-6061 cladding are being developed for use in research and test reactors as low enrichment fuel (<20% U-235 enrichment) as a result of its high uranium loading capacity compared to that of U-7Mo dispersion fuel. These fuel plates contain a Zr diffusion barrier between the U-10Mo fuel and Al-6061 cladding that suppresses the interaction between the U-Mo fuel foil and Al alloy cladding that is known to be problematic under irradiation. Different methods have been employed to fabricate monolithic fuel plates, including hot-rolling with no cold-rolling. L1P09T is a hot-rolled fuel plate irradiated to high fission density in the RERTR-9B experiment. This paper discusses the TEM characterization results for this U-10Mo/Zr/Al6061 monolithic fuel plate (∼59% U-235 enrichment) irradiated in Advanced Test Reactor at Idaho National Laboratory with an unprecedented high local fission density of 9.8E+21 fissions/cm3. The calculated fuel foil centerline temperature at the beginning of life and the end of life is 141 and 194 °C, respectively. TEM lamellas were prepared using focus ion beam lift-out technique. The estimated U-Mo fuel swelling, based on the fuel foil thickness change from SEM, is approximately 76%. Large bubbles (>1 μm) are distributed evenly in U-Mo and interlink of these bubbles is evident. The average size of subdivided grains at this fission density appears similar to that at 5.2E+21 fissions/cm3. The measured average Mo and Zr content in the fuel matrix is ∼30 at% and ∼7 at%, respectively, in general agreement with the calculated Mo and Zr from fission density.

  7. 76 FR 387 - Atomic Safety and Licensing Board; AREVA Enrichment Services, LLC (Eagle Rock Enrichment Facility)

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-01-04

    ... and Licensing Board; AREVA Enrichment Services, LLC (Eagle Rock Enrichment Facility) December 17, 2010... construction and operation of a gas centrifuge uranium enrichment facility--denoted as the Eagle Rock... site at http://www.nrc.gov/materials/fuel-cycle-fac/arevanc.html . These and other documents relating...

  8. Neutronics and Thermal Hydraulics Study for Using a Low-Enriched Uranium Core in the Advanced Test Reactor -- 2008 Final Report

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    G. S. Chang; M. A. Lillo; R. G. Ambrosek

    2008-06-01

    The Advanced Test Reactor (ATR) is a high power density and high neutron flux research reactor operating in the United States. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth. Because of the 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. The present work investigates the necessary modifications and evaluates the subsequent operating effects of this conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed and validated for a fuelmore » cycle burnup comparison analysis. Using the current HEU U 235 enrichment of 93.0 % as a baseline, an analysis was performed to determine the low-enriched uranium (LEU) density and U-235 enrichment required in the fuel meat to yield an equivalent K-eff versus effective full power days (EFPDs) between the HEU and the LEU cores. The MCNP ATR 1/8th core model was used to optimize the U 235 loading in the LEU core, such that the differences in K-eff and heat flux profiles between the HEU and LEU cores were minimized. The depletion methodology MCWO was used to calculate K-eff versus EFPDs in this paper. The MCWO-calculated results for the LEU demonstrated adequate excess reactivity such that the K-eff versus EFPDs plot is similar to the ATR reference HEU case study. Each HEU fuel element contains 19 fuel plates with a fuel meat thickness of 0.508 mm (20 mil). In this work, the proposed LEU (U-10Mo) core conversion case with nominal fuel meat thickness of 0.330 mm (13 mil) and U-235 enrichment of 19.7 wt% is used to optimize the radial heat flux profile by varying the fuel meat thickness from 0.191 mm (7.0 mil) to 0.330 mm (13.0 mil) at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19). A 0.8g of Boron-10, a burnable absorber, was added in the inner and outer plates to reduce the initial excess reactivity, and the peak to average ratio of

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

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Michael Tyacke; Dr. Igor Bolshinsky; Jeff Chamberlin

    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 spentmore » 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.« less

  10. Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Montierth, Leland M.

    2016-07-19

    The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element designmore » for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition, as part of a fuel meat thickness optimization effort for reactor performance, other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B, that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.« less

  11. Supply of enriched uranium for research reactors

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Mueller, H.

    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 onmore » 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.« less

  12. Reduction of gaseous pollutant emissions from gas turbine combustors using hydrogen-enriched jet fuel

    NASA Technical Reports Server (NTRS)

    Clayton, R. M.

    1976-01-01

    Recent progress in an evaluation of the applicability of the hydrogen enrichment concept to achieve ultralow gaseous pollutant emission from gas turbine combustion systems is described. The target emission indexes for the program are 1.0 for oxides of nitrogen and carbon monoxide, and 0.5 for unburned hydrocarbons. The basic concept utilizes premixed molecular hydrogen, conventional jet fuel, and air to depress the lean flammability limit of the mixed fuel. This is shown to permit very lean combustion with its low NOx production while simulataneously providing an increased flame stability margin with which to maintain low CO and HC emission. Experimental emission characteristics and selected analytical results are presented for a cylindrical research combustor designed for operation with inlet-air state conditions typical for a 30:1 compression ratio, high bypass ratio, turbofan commercial engine.

  13. Characterization of two diesel fuel degrading microbial consortia enriched from a non acclimated, complex source of microorganisms

    PubMed Central

    2010-01-01

    Background The bioremediation of soils impacted by diesel fuels is very often limited by the lack of indigenous microflora with the required broad substrate specificity. In such cases, the soil inoculation with cultures with the desired catabolic capabilities (bioaugmentation) is an essential option. The use of consortia of microorganisms obtained from rich sources of microbes (e.g., sludges, composts, manure) via enrichment (i.e., serial growth transfers) on the polluting hydrocarbons would provide bioremediation enhancements more robust and reproducible than those achieved with specialized pure cultures or tailored combinations (co-cultures) of them, together with none or minor risks of soil loading with unrelated or pathogenic allocthonous microorganisms. Results In this work, two microbial consortia, i.e., ENZ-G1 and ENZ-G2, were enriched from ENZYVEBA (a complex commercial source of microorganisms) on Diesel (G1) and HiQ Diesel (G2), respectively, and characterized in terms of microbial composition and hydrocarbon biodegradation capability and specificity. ENZ-G1 and ENZ-G2 exhibited a comparable and remarkable biodegradation capability and specificity towards n-C10 to n-C24 linear paraffins by removing about 90% of 1 g l-1 of diesel fuel applied after 10 days of aerobic shaken flask batch culture incubation at 30°C. Cultivation dependent and independent approaches evidenced that both consortia consist of bacteria belonging to the genera Chryseobacterium, Acinetobacter, Psudomonas, Stenotrophomonas, Alcaligenes and Gordonia along with the fungus Trametes gibbosa. However, only the fungus was found to grow and remarkably biodegrade G1 and G2 hydrocarbons under the same conditions. The biodegradation activity and specificity and the microbial composition of ENZ-G1 and ENZ-G2 did not significantly change after cryopreservation and storage at -20°C for several months. Conclusions ENZ-G1 and ENZ-G2 are very similar highly enriched consortia of bacteria and a

  14. Characterization of two diesel fuel degrading microbial consortia enriched from a non acclimated, complex source of microorganisms.

    PubMed

    Zanaroli, Giulio; Di Toro, Sara; Todaro, Daniela; Varese, Giovanna C; Bertolotto, Antonio; Fava, Fabio

    2010-02-16

    The bioremediation of soils impacted by diesel fuels is very often limited by the lack of indigenous microflora with the required broad substrate specificity. In such cases, the soil inoculation with cultures with the desired catabolic capabilities (bioaugmentation) is an essential option. The use of consortia of microorganisms obtained from rich sources of microbes (e.g., sludges, composts, manure) via enrichment (i.e., serial growth transfers) on the polluting hydrocarbons would provide bioremediation enhancements more robust and reproducible than those achieved with specialized pure cultures or tailored combinations (co-cultures) of them, together with none or minor risks of soil loading with unrelated or pathogenic allocthonous microorganisms. In this work, two microbial consortia, i.e., ENZ-G1 and ENZ-G2, were enriched from ENZYVEBA (a complex commercial source of microorganisms) on Diesel (G1) and HiQ Diesel (G2), respectively, and characterized in terms of microbial composition and hydrocarbon biodegradation capability and specificity. ENZ-G1 and ENZ-G2 exhibited a comparable and remarkable biodegradation capability and specificity towards n-C10 to n-C24 linear paraffins by removing about 90% of 1 g l-1 of diesel fuel applied after 10 days of aerobic shaken flask batch culture incubation at 30 degrees C. Cultivation dependent and independent approaches evidenced that both consortia consist of bacteria belonging to the genera Chryseobacterium, Acinetobacter, Psudomonas, Stenotrophomonas, Alcaligenes and Gordonia along with the fungus Trametes gibbosa. However, only the fungus was found to grow and remarkably biodegrade G1 and G2 hydrocarbons under the same conditions. The biodegradation activity and specificity and the microbial composition of ENZ-G1 and ENZ-G2 did not significantly change after cryopreservation and storage at -20 degrees C for several months. ENZ-G1 and ENZ-G2 are very similar highly enriched consortia of bacteria and a fungus capable of

  15. Determining initial enrichment, burnup, and cooling time of pressurized-water reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

    DOE PAGES

    Favalli, Andrea; Vo, D.; Grogan, Brandon R.; ...

    2016-02-26

    The purpose of the Next Generation Safeguards Initiative (NGSI)–Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuelmore » assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute 137Cs count rate and the 154Eu/ 137Cs, 134Cs/ 137Cs, 106Ru/ 137Cs, and 144Ce/ 137Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity’s behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. Furthermore, the results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.« less

  16. Determining initial enrichment, burnup, and cooling time of pressurized-water reactor spent fuel assemblies by analyzing passive gamma spectra measured at the Clab interim-fuel storage facility in Sweden

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Favalli, Andrea; Vo, D.; Grogan, Brandon R.

    The purpose of the Next Generation Safeguards Initiative (NGSI)–Spent Fuel (SF) project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins; (3) estimate the plutonium mass [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuelmore » assemblies. Since August 2013, a set of measurement campaigns has been conducted at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab), in collaboration with Swedish Nuclear Fuel and Waste Management Company (SKB). One purpose of the measurement campaigns was to acquire passive gamma spectra with high-purity germanium and lanthanum bromide scintillation detectors from Pressurized Water Reactor and Boiling Water Reactor spent fuel assemblies. The absolute 137Cs count rate and the 154Eu/ 137Cs, 134Cs/ 137Cs, 106Ru/ 137Cs, and 144Ce/ 137Cs isotopic ratios were extracted; these values were used to construct corresponding model functions (which describe each measured quantity’s behavior over various combinations of burnup, cooling time, and initial enrichment) and then were used to determine those same quantities in each measured spent fuel assembly. Furthermore, the results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.« less

  17. 41 CFR 102-34.55 - Are there fleet average fuel economy standards we must meet?

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 41 Public Contracts and Property Management 3 2010-07-01 2010-07-01 false Are there fleet average fuel economy standards we must meet? 102-34.55 Section 102-34.55 Public Contracts and Property Management Federal Property Management Regulations System (Continued) FEDERAL MANAGEMENT REGULATION PERSONAL PROPERTY 34-MOTOR VEHICLE MANAGEMENT Obtainin...

  18. Uranium Conversion & Enrichment

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Karpius, Peter Joseph

    2017-02-06

    The isotopes of uranium that are found in nature, and hence in ‘fresh’ Yellowcake’, are not in relative proportions that are suitable for power or weapons applications. The goal of conversion then is to transform the U 3O 8 yellowcake into UF 6. Conversion and enrichment of uranium is usually required to obtain material with enough 235U to be usable as fuel in a reactor or weapon. The cost, size, and complexity of practical conversion and enrichment facilities aid in nonproliferation by design.

  19. Fresh Fuel Measurements With the Differential Die-Away Self-Interrogation Instrument

    NASA Astrophysics Data System (ADS)

    Trahan, Alexis C.; Belian, Anthony P.; Swinhoe, Martyn T.; Menlove, Howard O.; Flaska, Marek; Pozzi, Sara A.

    2017-07-01

    The purpose of the Next Generation Safeguards Initiative (NGSI)-Spent Fuel (SF) Project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. The NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: 1) verify the initial enrichment, burnup, and cooling time of facility declaration; 2) detect the diversion or replacement of pins; 3) estimate the plutonium mass; 4) estimate decay heat; and 5) determine the reactivity of spent fuel assemblies. The differential die-away self-interrogation (DDSI) instrument is one instrument that was assessed for years regarding its feasibility for robust, timely verification of spent fuel assemblies. The instrument was recently built and was tested using fresh fuel assemblies in a variety of configurations, including varying enrichment, neutron absorber content, and symmetry. The early die-away method, a multiplication determination method developed in simulation space, was successfully tested on the fresh fuel assembly data and determined multiplication with a root-mean-square (RMS) error of 2.9%. The experimental results were compared with MCNP simulations of the instrument as well. Low multiplication assemblies had agreement with an average RMS error of 0.2% in the singles count rate (i.e., total neutrons detected per second) and 3.4% in the doubles count rates (i.e., neutrons detected in coincidence per second). High-multiplication assemblies had agreement with an average RMS error of 4.1% in the singles and 13.3% in the doubles count rates.

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

    NASA Astrophysics Data System (ADS)

    Myers, Astasia

    2011-06-01

    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.

  1. A cellular automaton method to simulate the microstructure and evolution of low-enriched uranium (LEU) U-Mo/Al dispersion type fuel plates

    NASA Astrophysics Data System (ADS)

    Drera, Saleem S.; Hofman, Gerard L.; Kee, Robert J.; King, Jeffrey C.

    2014-10-01

    Low-enriched uranium (LEU) fuel plates for high power materials test reactors (MTR) are composed of nominally spherical uranium-molybdenum (U-Mo) particles within an aluminum matrix. Fresh U-Mo particles typically range between 10 and 100 μm in diameter, with particle volume fractions up to 50%. As the fuel ages, reaction-diffusion processes cause the formation and growth of interaction layers that surround the fuel particles. The growth rate depends upon the temperature and radiation environment. The cellular automaton algorithm described in this paper can synthesize realistic random fuel-particle structures and simulate the growth of the intermetallic interaction layers. Examples in the present paper pack approximately 1000 particles into three-dimensional rectangular fuel structures that are approximately 1 mm on each side. The computational approach is designed to yield synthetic microstructures consistent with images from actual fuel plates and is validated by comparison with empirical data on actual fuel plates.

  2. A cellular automaton method to simulate the microstructure and evolution of low-enriched uranium (LEU) U–Mo/Al dispersion type fuel plates

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Drera, Saleem S.; Hofman, Gerard L.; Kee, Robert J.

    Low-enriched uranium (LEU) fuel plates for high power materials test reactors (MTR) are composed of nominally spherical uranium-molybdenum (U-Mo) particles within an aluminum matrix. Fresh U-Mo particles typically range between 10 and 100 mu m in diameter, with particle volume fractions up to 50%. As the fuel ages, reaction-diffusion processes cause the formation and growth of interaction layers that surround the fuel particles. The growth rate depends upon the temperature and radiation environment. The cellular automaton algorithm described in this paper can synthesize realistic random fuel-particle structures and simulate the growth of the intermetallic interaction layers. Examples in the presentmore » paper pack approximately 1000 particles into three-dimensional rectangular fuel structures that are approximately 1 mm on each side. The computational approach is designed to yield synthetic microstructures consistent with images from actual fuel plates and is validated by comparison with empirical data on actual fuel plates. (C) 2014 Elsevier B.V. All rights reserved.« less

  3. Multi-stage combustion using nitrogen-enriched air

    DOEpatents

    Fischer, Larry E.; Anderson, Brian L.

    2004-09-14

    Multi-stage combustion technology combined with nitrogen-enriched air technology for controlling the combustion temperature and products to extend the maintenance and lifetime cycles of materials in contact with combustion products and to reduce pollutants while maintaining relatively high combustion and thermal cycle efficiencies. The first stage of combustion operates fuel rich where most of the heat of combustion is released by burning it with nitrogen-enriched air. Part of the energy in the combustion gases is used to perform work or to provide heat. The cooled combustion gases are reheated by additional stages of combustion until the last stage is at or near stoichiometric conditions. Additional energy is extracted from each stage to result in relatively high thermal cycle efficiency. The air is enriched with nitrogen using air separation technologies such as diffusion, permeable membrane, absorption, and cryogenics. The combustion method is applicable to many types of combustion equipment, including: boilers, burners, turbines, internal combustion engines, and many types of fuel including hydrogen and carbon-based fuels including methane and coal.

  4. Fresh Fuel Measurements With the Differential Die-Away Self-Interrogation Instrument

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Trahan, Alexis C.; Belian, Anthony P.; Swinhoe, Martyn T.

    The purpose of the Next Generation Safeguards Initiative (NGSI)-Spent Fuel (SF) Project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. Thus the NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: 1) verify the initial enrichment, burnup, and cooling time of facility declaration; 2) detect the diversion or replacement of pins; 3) estimate the plutonium mass; 4) estimate decay heat; and 5) determine the reactivity of spent fuel assemblies. The differential die-away self-interrogation (DDSI) instrument is one instrumentmore » that was assessed for years regarding its feasibility for robust, timely verification of spent fuel assemblies. The instrument was recently built and was tested using fresh fuel assemblies in a variety of configurations, including varying enrichment, neutron absorber content, and symmetry. The early die-away method, a multiplication determination method developed in simulation space, was successfully tested on the fresh fuel assembly data and determined multiplication with a root-mean-square (RMS) error of 2.9%. The experimental results were compared with MCNP simulations of the instrument as well. Low multiplication assemblies had agreement with an average RMS error of 0.2% in the singles count rate (i.e., total neutrons detected per second) and 3.4% in the doubles count rates (i.e., neutrons detected in coincidence per second). High-multiplication assemblies had agreement with an average RMS error of 4.1% in the singles and 13.3% in the doubles count rates.« less

  5. Fresh Fuel Measurements With the Differential Die-Away Self-Interrogation Instrument

    DOE PAGES

    Trahan, Alexis C.; Belian, Anthony P.; Swinhoe, Martyn T.; ...

    2017-01-05

    The purpose of the Next Generation Safeguards Initiative (NGSI)-Spent Fuel (SF) Project is to strengthen the technical toolkit of safeguards inspectors and/or other interested parties. Thus the NGSI-SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: 1) verify the initial enrichment, burnup, and cooling time of facility declaration; 2) detect the diversion or replacement of pins; 3) estimate the plutonium mass; 4) estimate decay heat; and 5) determine the reactivity of spent fuel assemblies. The differential die-away self-interrogation (DDSI) instrument is one instrumentmore » that was assessed for years regarding its feasibility for robust, timely verification of spent fuel assemblies. The instrument was recently built and was tested using fresh fuel assemblies in a variety of configurations, including varying enrichment, neutron absorber content, and symmetry. The early die-away method, a multiplication determination method developed in simulation space, was successfully tested on the fresh fuel assembly data and determined multiplication with a root-mean-square (RMS) error of 2.9%. The experimental results were compared with MCNP simulations of the instrument as well. Low multiplication assemblies had agreement with an average RMS error of 0.2% in the singles count rate (i.e., total neutrons detected per second) and 3.4% in the doubles count rates (i.e., neutrons detected in coincidence per second). High-multiplication assemblies had agreement with an average RMS error of 4.1% in the singles and 13.3% in the doubles count rates.« less

  6. Draft environmental impact statement : corporate average fuel economy standards, passenger cars and light trucks, model years 2011-2015.

    DOT National Transportation Integrated Search

    2008-06-01

    The National Highway Traffic Safety Administration (NHTSA) has prepared this Draft Environmental Impact Statement (DEIS) to disclose and analyze the potential environmental impacts of the proposed new Corporate Average Fuel Economy (CAFE) standards a...

  7. 75 FR 62895 - Notice of Availability of Safety Evaluation Report; AREVA Enrichment Services LLC, Eagle Rock...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-10-13

    ... Evaluation Report; AREVA Enrichment Services LLC, Eagle Rock Enrichment Facility, Bonneville County, ID... report. FOR FURTHER INFORMATION CONTACT: Breeda Reilly, Senior Project Manager, Advanced Fuel Cycle, Enrichment, and Uranium Conversion, Division of Fuel Cycle Safety and Safeguards, Office of Nuclear Material...

  8. Hydrogen-enrichment-concept preliminary evaluation

    NASA Technical Reports Server (NTRS)

    Ecklund, E. E.

    1975-01-01

    A hydrogen-enriched fuels concept for automobiles is described and evaluated in terms of fuel consumption and engine exhaust emissions through multicylinder (V-8) automotive engine/hydrogen generator tests, single cylinder research engine (CFR) tests, and hydrogen-generator characterization tests. Analytical predictions are made of the fuel consumption and NO/sub x/ emissions which would result from anticipated engine improvements. The hydrogen-gas generator, which was tested to quantify its thermodynamic input-output relationships was used for integrated testing of the V-8 engine and generator.

  9. 40 CFR 600.510-12 - Calculation of average fuel economy and average carbon-related exhaust emissions.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... nearest 0.1 mpg; or (iii) For natural gas-fueled model types, the fuel economy value calculated for that... as determined in § 600.208-12(b)(5)(i). (vi) For natural gas dual fuel model types, for model years... natural gas as determined in § 600.208-12(b)(5)(ii) divided by 0.15 provided the requirements of paragraph...

  10. 40 CFR 600.510-12 - Calculation of average fuel economy and average carbon-related exhaust emissions.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... nearest 0.1 mpg; or (iii) For natural gas-fueled model types, the fuel economy value calculated for that... as determined in § 600.208-12(b)(5)(i). (vi) For natural gas dual fuel model types, for model years... natural gas as determined in § 600.208-12(b)(5)(ii) divided by 0.15 provided the requirements of paragraph...

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

    DOE Office of Scientific and Technical Information (OSTI.GOV)

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

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

  12. Precise calculation of neutron-capture reactions contribution in energy release for different types of VVER-1000 fuel assemblies

    NASA Astrophysics Data System (ADS)

    Tikhomirov, Georgy; Bahdanovich, Rynat; Pham, Phu

    2017-09-01

    Precise calculation of energy release in a nuclear reactor is necessary to obtain the correct spatial power distribution and predict characteristics of burned nuclear fuel. In this work, previously developed method for calculation neutron-capture reactions - capture component - contribution in effective energy release in a fuel core of nuclear reactor is discussed. The method was improved and implemented to the different models of VVER-1000 reactor developed for MCU 5 and MCNP 4 computer codes. Different models of equivalent cell and fuel assembly in the beginning of fuel cycle were calculated. These models differ by the geometry, fuel enrichment and presence of burnable absorbers. It is shown, that capture component depends on fuel enrichment and presence of burnable absorbers. Its value varies for different types of hot fuel assemblies from 3.35% to 3.85% of effective energy release. Average capture component contribution in effective energy release for typical serial fresh fuel of VVER-1000 is 3.5%, which is 7 MeV/fission. The method will be used in future to estimate the dependency of capture energy on fuel density, burn-up, etc.

  13. 41 CFR 102-34.60 - How do we calculate the average fuel economy for Government motor vehicles?

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 41 Public Contracts and Property Management 3 2010-07-01 2010-07-01 false How do we calculate the average fuel economy for Government motor vehicles? 102-34.60 Section 102-34.60 Public Contracts and Property Management Federal Property Management Regulations System (Continued) FEDERAL MANAGEMENT REGULATION PERSONAL PROPERTY 34-MOTOR VEHICLE...

  14. Reduced enrichment for research and test reactors: Proceedings

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    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.

  15. Microstructure of the irradiated U 3Si 2/Al silicide dispersion fuel

    NASA Astrophysics Data System (ADS)

    Gan, J.; Keiser, D. D.; Miller, B. D.; Jue, J.-F.; Robinson, A. B.; Madden, J. W.; Medvedev, P. G.; Wachs, D. M.

    2011-12-01

    The silicide dispersion fuel of U 3Si 2/Al is recognized as the best performance fuel for many nuclear research and test reactors with up to 4.8 gU/cm 3 fuel loading. An irradiated U 3Si 2/Al dispersion fuel ( 235U ˜ 75%) from the high-flux side of a fuel plate (U0R040) from the Reduced Enrichment for Research and Test Reactors (RERTR)-8 test was characterized using transmission electron microscopy (TEM). The fuel was irradiated in the Advanced Test Reactor (ATR) for 105 days. The average irradiation temperature and fission density of the U 3Si 2 fuel particles for the TEM sample are estimated to be approximately 110 °C and 5.4 × 10 27 f/m 3. The characterization was performed using a 200-kV TEM. The U/Si ratio for the fuel particle and (Si + Al)/U for the fuel-matrix-interaction layer are approximately 1.1 and 4-10, respectively. The estimated average diameter, number density and volume fraction for small bubbles (<1 μm) in the fuel particle are ˜94 nm, 1.05 × 10 20 m -3 and ˜11%, respectively. The results and their implication on the performance of the U 3Si 2/Al silicide dispersion fuel are discussed.

  16. Fabrication and testing of U-7Mo monolithic plate fuel with Zircaloy cladding

    NASA Astrophysics Data System (ADS)

    Pasqualini, E. E.; Robinson, A. B.; Porter, D. L.; Wachs, D. M.; Finlay, M. R.

    2016-10-01

    Nuclear fuel designs are being developed to replace highly enriched fuel used in research and test reactors with fuels of low enrichment. In the most challenging cases, U-(7-10 wt%)Mo monolithic plate fuels are proposed. One of the considered designs includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction during service. Zircaloy cladding, specifically Zry-4, was investigated as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry-4 clad U-7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry-4 and U-(7-10)Mo have similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch, which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly during or between roll passes. The final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction-either from fabrication or in-reactor testing-and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.7E+21 (average) fissions/cm3, 3.8E+21 (peak).

  17. Determining Reactor Fuel Type from Continuous Antineutrino Monitoring

    NASA Astrophysics Data System (ADS)

    Jaffke, Patrick; Huber, Patrick

    2017-09-01

    We investigate the ability of an antineutrino detector to determine the fuel type of a reactor. A hypothetical 5-ton antineutrino detector is placed 25 m from the core and measures the spectral shape and rate of antineutrinos emitted by fission fragments in the core for a number of 90-d periods. Our results indicate that four major fuel types can be differentiated from the variation of fission fractions over the irradiation time with a true positive probability of detection at approximately 95%. In addition, we demonstrate that antineutrinos can identify the burnup at which weapons-grade mixed-oxide (MOX) fuel would be reduced to reactor-grade MOX, on average, providing assurance that plutonium-disposition goals are met. We also investigate removal scenarios where plutonium is purposefully diverted from a mixture of MOX and low-enriched uranium fuel. Finally, we discuss how our analysis is impacted by a spectral distortion around 6 MeV observed in the antineutrino spectrum measured from commercial power reactors.

  18. 75 FR 25323 - Light-Duty Vehicle Greenhouse Gas Emission Standards and Corporate Average Fuel Economy Standards...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-05-07

    ...EPA and NHTSA are issuing this joint Final Rule to establish a National Program consisting of new standards for light-duty vehicles that will reduce greenhouse gas emissions and improve fuel economy. This joint Final Rule is consistent with the National Fuel Efficiency Policy announced by President Obama on May 19, 2009, responding to the country's critical need to address global climate change and to reduce oil consumption. EPA is finalizing greenhouse gas emissions standards under the Clean Air Act, and NHTSA is finalizing Corporate Average Fuel Economy standards under the Energy Policy and Conservation Act, as amended. These standards apply to passenger cars, light-duty trucks, and medium-duty passenger vehicles, covering model years 2012 through 2016, and represent a harmonized and consistent National Program. Under the National Program, automobile manufacturers will be able to build a single light-duty national fleet that satisfies all requirements under both programs while ensuring that consumers still have a full range of vehicle choices. NHTSA's final rule also constitutes the agency's Record of Decision for purposes of its National Environmental Policy Act (NEPA) analysis.

  19. New potentials for conventional aircraft when powered by hydrogen-enriched gasoline

    NASA Technical Reports Server (NTRS)

    Menard, W. A.; Moynihan, P. I.; Rupe, J. H.

    1976-01-01

    Hydrogen enrichment for aircraft piston engines is under study in a new NASA program. The objective of the program is to determine the feasibility of inflight injection of hydrogen in general aviation aircraft engines to reduce fuel consumption and to lower emission levels. A catalytic hydrogen generator will be incorporated as part of the air induction system of a Lycoming turbocharged engine and will generate hydrogen by breaking down small amounts of the aviation gasoline used in the normal propulsion system. This hydrogen will then be mixed with gasoline and compressed air from the turbocharger before entering the engine combustion chamber. The special properties of the hydrogen-enriched gasoline allow the engine to operate at ultralean fuel/air ratios, resulting in higher efficiencies and hence less fuel consumption. This paper summarizes the results of a systems analysis study. Calculations assuming a Beech Duke aircraft indicate that fuel savings on the order of 20% are possible. An estimate of the potential for the utilization of hydrogen enrichment to control exhaust emissions indicates that it may be possible to meet the 1979 Federal emission standards.

  20. Evaluating quantitative 3-D image analysis as a design tool for low enriched uranium fuel compacts for the transient reactor test facility: A preliminary study

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Kane, J. J.; van Rooyen, I. J.; Craft, A. E.

    In this study, 3-D image analysis when combined with a non-destructive examination technique such as X-ray computed tomography (CT) provides a highly quantitative tool for the investigation of a material’s structure. In this investigation 3-D image analysis and X-ray CT were combined to analyze the microstructure of a preliminary subsized fuel compact for the Transient Reactor Test Facility’s low enriched uranium conversion program to assess the feasibility of the combined techniques for use in the optimization of the fuel compact fabrication process. The quantitative image analysis focused on determining the size and spatial distribution of the surrogate fuel particles andmore » the size, shape, and orientation of voids within the compact. Additionally, the maximum effect of microstructural features on heat transfer through the carbonaceous matrix of the preliminary compact was estimated. The surrogate fuel particles occupied 0.8% of the compact by volume with a log-normal distribution of particle sizes with a mean diameter of 39 μm and a standard deviation of 16 μm. Roughly 39% of the particles had a diameter greater than the specified maximum particle size of 44 μm suggesting that the particles agglomerate during fabrication. The local volume fraction of particles also varies significantly within the compact although uniformities appear to be evenly dispersed throughout the analysed volume. The voids produced during fabrication were on average plate-like in nature with their major axis oriented perpendicular to the compaction direction of the compact. Finally, the microstructure, mainly the large preferentially oriented voids, may cause a small degree of anisotropy in the thermal diffusivity within the compact. α∥/α⊥, the ratio of thermal diffusivities parallel to and perpendicular to the compaction direction are expected to be no less than 0.95 with an upper bound of 1.« less

  1. Evaluating quantitative 3-D image analysis as a design tool for low enriched uranium fuel compacts for the transient reactor test facility: A preliminary study

    DOE PAGES

    Kane, J. J.; van Rooyen, I. J.; Craft, A. E.; ...

    2016-02-05

    In this study, 3-D image analysis when combined with a non-destructive examination technique such as X-ray computed tomography (CT) provides a highly quantitative tool for the investigation of a material’s structure. In this investigation 3-D image analysis and X-ray CT were combined to analyze the microstructure of a preliminary subsized fuel compact for the Transient Reactor Test Facility’s low enriched uranium conversion program to assess the feasibility of the combined techniques for use in the optimization of the fuel compact fabrication process. The quantitative image analysis focused on determining the size and spatial distribution of the surrogate fuel particles andmore » the size, shape, and orientation of voids within the compact. Additionally, the maximum effect of microstructural features on heat transfer through the carbonaceous matrix of the preliminary compact was estimated. The surrogate fuel particles occupied 0.8% of the compact by volume with a log-normal distribution of particle sizes with a mean diameter of 39 μm and a standard deviation of 16 μm. Roughly 39% of the particles had a diameter greater than the specified maximum particle size of 44 μm suggesting that the particles agglomerate during fabrication. The local volume fraction of particles also varies significantly within the compact although uniformities appear to be evenly dispersed throughout the analysed volume. The voids produced during fabrication were on average plate-like in nature with their major axis oriented perpendicular to the compaction direction of the compact. Finally, the microstructure, mainly the large preferentially oriented voids, may cause a small degree of anisotropy in the thermal diffusivity within the compact. α∥/α⊥, the ratio of thermal diffusivities parallel to and perpendicular to the compaction direction are expected to be no less than 0.95 with an upper bound of 1.« less

  2. Nuclear reactor composite fuel assembly

    DOEpatents

    Burgess, Donn M.; Marr, Duane R.; Cappiello, Michael W.; Omberg, Ronald P.

    1980-01-01

    A core and composite fuel assembly for a liquid-cooled breeder nuclear reactor including a plurality of elongated coextending driver and breeder fuel elements arranged to form a generally polygonal bundle within a thin-walled duct. The breeder elements are larger in cross section than the driver elements, and each breeder element is laterally bounded by a number of the driver elements. Each driver element further includes structure for spacing the driver elements from adjacent fuel elements and, where adjacent, the thin-walled duct. A core made up of the fuel elements can advantageously include fissile fuel of only one enrichment, while varying the effective enrichment of any given assembly or core region, merely by varying the relative number and size of the driver and breeder elements.

  3. Determination of the NPP Kr\\vsko spent fuel decay heat

    NASA Astrophysics Data System (ADS)

    Kromar, Marjan; Kurinčič, Bojan

    2017-07-01

    Nuclear fuel is designed to support fission process in a reactor core. Some of the isotopes, formed during the fission, decay and produce decay heat and radiation. Accurate knowledge of the nuclide inventory producing decay heat is important after reactor shut down, during the fuel storage and subsequent reprocessing or disposal. In this paper possibility to calculate the fuel isotopic composition and determination of the fuel decay heat with the Serpent code is investigated. Serpent is a well-known Monte Carlo code used primarily for the calculation of the neutron transport in a reactor. It has been validated for the burn-up calculations. In the calculation of the fuel decay heat different set of isotopes is important than in the neutron transport case. Comparison with the Origen code is performed to verify that the Serpent is taking into account all isotopes important to assess the fuel decay heat. After the code validation, a sensitivity study is carried out. Influence of several factors such as enrichment, fuel temperature, moderator temperature (density), soluble boron concentration, average power, burnable absorbers, and burnup is analyzed.

  4. Field test of short-notice random inspections for inventory-change verification at a low-enriched-uranium fuel-fabrication plant: Preliminary summary

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Fishbone, L.G.; Moussalli, G.; Naegele, G.

    1994-04-01

    An approach of short-notice random inspections (SNRIs) for inventory-change verification can enhance the effectiveness and efficiency of international safeguards at natural or low-enriched uranium (LEU) fuel fabrication plants. According to this approach, the plant operator declares the contents of nuclear material items before knowing if an inspection will occur to verify them. Additionally, items about which declarations are newly made should remain available for verification for an agreed time. This report details a six-month field test of the feasibility of such SNRIs which took place at the Westinghouse Electric Corporation Commercial Nuclear Fuel Division. Westinghouse personnel made daily declarations aboutmore » both feed and product items, uranium hexafluoride cylinders and finished fuel assemblies, using a custom-designed computer ``mailbox``. Safeguards inspectors from the IAEA conducted eight SNRIs to verify these declarations. Items from both strata were verified during the SNRIs by means of nondestructive assay equipment. The field test demonstrated the feasibility and practicality of key elements of the SNRI approach for a large LEU fuel fabrication plant.« less

  5. The current state of the Russian reduced enrichment research reactors program

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Aden, V.G.; Kartashov, E.F.; Lukichev, V.A.

    1997-08-01

    During the last year after the 16-th International Conference on Reducing Fuel Enrichment in Research Reactors held in October, 1993 in Oarai, Japan, the conclusive stage of the Program on reducing fuel enrichment (to 20% in U-235) in research reactors was finally made up in Russia. The Program was started late in 70th and the first stage of the Program was completed by 1986 which allowed to reduce fuel enrichment from 80-90% to 36%. The completion of the Program current stage, which is counted for 5-6 years, will exclude the use of the fuel enriched by more than 20% frommore » RF to other countries such as: Poland, Czeck Republick, Hungary, Roumania, Bulgaria, Libya, Viet-Nam, North Korea, Egypt, Latvia, Ukraine, Uzbekistan and Kazakhstan. In 1994 the Program, approved by RF Minatom authorities, has received the status of an inter-branch program since it was admitted by the RF Ministry for Science and Technical Policy. The Head of RF Minatom central administrative division N.I.Ermakov was nominated as the Head of the Russian Program, V.G.Aden, RDIPE Deputy Director, was nominated as the scientific leader. The Program was submitted to the Commission for Scientific, Technical and Economical Cooperation between USA and Russia headed by Vice-President A. Gore and Prime Minister V. Chemomyrdin and was given support also.« less

  6. Recent developments in BWR fuel design

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Congdon, S.P.; Noble, L.D.; Wood, J.E.

    1991-11-01

    Substantial increases in the cost effectiveness and performance capability of boiling water reactor (BWR) fuel designs have been implemented in the past 5 to 7 yr. This increase has been driven by (a) utility desires to lower fuel and operating costs and (b) design innovations that have lowered enrichment requirements, improved thermal-hydraulic performance, and increased discharge exposure. Higher discharge exposures reduce disposal costs for European and Asian utilities and enable US utilities to lengthen operating cycles. A typical BWR reload fuel bundle fabricated today has 25% higher {sup 235}U enrichment and a factor of 2 higher gadolinium loading than onemore » made several years ago. Today's BWR fuel bundles also contain more unheated water reduces the axial water density variation, lowers the void coefficient, and enhances the neutron efficiency of the bundle, reducing both the gadolinium poison and the enrichment requirements. In addition to these general trends, the following unique design innovations have further enhanced the fuel cost efficiency and performance characteristics of BWR fuel: ferrule spacer, part length rods, interactive channel, and bundle enhanced spectral shift. GE's fuel designs offer the flexibility for modern BWR fuel requirements and contain unique design features that enhance flexibility for modern BWR fuel requirements and contain unique design features that enhance flexibility and fuel cycle economics.« less

  7. Target-fueled nuclear reactor for medical isotope production

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Coats, Richard L.; Parma, Edward J.

    A small, low-enriched, passively safe, low-power nuclear reactor comprises a core of target and fuel pins that can be processed to produce the medical isotope .sup.99Mo and other fission product isotopes. The fuel for the reactor and the targets for the .sup.99Mo production are the same. The fuel can be low enriched uranium oxide, enriched to less than 20% .sup.235U. The reactor power level can be 1 to 2 MW. The reactor is passively safe and maintains negative reactivity coefficients. The total radionuclide inventory in the reactor core is minimized since the fuel/target pins are removed and processed after 7more » to 21 days.« less

  8. Continental-scale enrichment of atmospheric 14CO2 from the nuclear power industry: potential impact on the estimation of fossil fuel-derived CO2

    NASA Astrophysics Data System (ADS)

    Graven, H. D.; Gruber, N.

    2011-12-01

    The 14C-free fossil carbon added to atmospheric CO2 by combustion dilutes the atmospheric 14C/C ratio (Δ14C), potentially providing a means to verify fossil CO2 emissions calculated using economic inventories. However, sources of 14C from nuclear power generation and spent fuel reprocessing can counteract this dilution and may bias 14C/C-based estimates of fossil fuel-derived CO2 if these nuclear influences are not correctly accounted for. Previous studies have examined nuclear influences on local scales, but the potential for continental-scale influences on Δ14C has not yet been explored. We estimate annual 14C emissions from each nuclear site in the world and conduct an Eulerian transport modeling study to investigate the continental-scale, steady-state gradients of Δ14C caused by nuclear activities and fossil fuel combustion. Over large regions of Europe, North America and East Asia, nuclear enrichment may offset at least 20% of the fossil fuel dilution in Δ14C, corresponding to potential biases of more than -0.25 ppm in the CO2 attributed to fossil fuel emissions, larger than the bias from plant and soil respiration in some areas. Model grid cells including high 14C-release reactors or fuel reprocessing sites showed much larger nuclear enrichment, despite the coarse model resolution of 1.8°×1.8°. The recent growth of nuclear 14C emissions increased the potential nuclear bias over 1985-2005, suggesting that changing nuclear activities may complicate the use of Δ14C observations to identify trends in fossil fuel emissions. The magnitude of the potential nuclear bias is largely independent of the choice of reference station in the context of continental-scale Eulerian transport and inversion studies, but could potentially be reduced by an appropriate choice of reference station in the context of local-scale assessments.

  9. 76 FR 74853 - 2017 and Later Model Year Light-Duty Vehicle Greenhouse Gas Emissions and Corporate Average Fuel...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-12-01

    ...EPA and NHTSA, on behalf of the Department of Transportation, are issuing this joint proposal to further reduce greenhouse gas emissions and improve fuel economy for light-duty vehicles for model years 2017-2025. This proposal extends the National Program beyond the greenhouse gas and corporate average fuel economy standards set for model years 2012-2016. On May 21, 2010, President Obama issued a Presidential Memorandum requesting that NHTSA and EPA develop through notice and comment rulemaking a coordinated National Program to reduce greenhouse gas emissions of light-duty vehicles for model years 2017- 2025. This proposal, consistent with the President's request, responds to the country's critical need to address global climate change and to reduce oil consumption. NHTSA is proposing Corporate Average Fuel Economy standards under the Energy Policy and Conservation Act, as amended by the Energy Independence and Security Act, and EPA is proposing greenhouse gas emissions standards under the Clean Air Act. These standards apply to passenger cars, light-duty trucks, and medium- duty passenger vehicles, and represent a continued harmonized and consistent National Program. Under the National Program for model years 2017-2025, automobile manufacturers would be able to continue building a single light-duty national fleet that satisfies all requirements under both programs while ensuring that consumers still have a full range of vehicle choices. EPA is also proposing a minor change to the regulations applicable to MY 2012-2016, with respect to air conditioner performance and measurement of nitrous oxides.

  10. IRRADIATION PERFORMANCE OF U-Mo MONOLITHIC FUEL

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    M.K. Meyer; J. Gan; J.-F. Jue

    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.more » 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.« less

  11. Potential annealing treatments for tailoring the starting microstructure of low-enriched U-Mo dispersion fuels to optimize performance during irradiation

    NASA Astrophysics Data System (ADS)

    Keiser, Dennis D.; Jue, Jan-Fong; Woolstenhulme, Nicolas E.; Ewh, Ashley

    2011-12-01

    Low-enriched uranium-molybdenum (U-Mo) alloy particles dispersed in aluminum alloy (e.g., dispersion fuels) are being developed for application in research and test reactors. To achieve the best performance of these fuels during irradiation, optimization of the starting microstructure may be required by utilizing a heat treatment that results in the formation of uniform, Si-rich interaction layers between the U-Mo particles and Al-Si matrix. These layers behave in a stable manner under certain irradiation conditions. To identify the optimum heat treatment for producing these kinds of layers in a dispersion fuel plate, a systematic annealing study has been performed using actual dispersion fuel samples, which were fabricated at relatively low temperatures to limit the growth of any interaction layers in the samples prior to controlled heat treatment. These samples had different Al matrices with varying Si contents and were annealed between 450 and 525 °C for up to 4 h. The samples were then characterized using scanning electron microscopy (SEM) to examine the thickness, composition, and uniformity of the interaction layers. Image analysis was performed to quantify various attributes of the dispersion fuel microstructures that related to the development of the interaction layers. The most uniform layers were observed to form in fuel samples that had an Al matrix with at least 4 wt.% Si and a heat treatment temperature of at least 475 °C.

  12. Fuel cycle cost, reactor physics and fuel manufacturing considerations for Erbia-bearing PWR fuel with > 5 wt% U-235 content

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Franceschini, F.; Lahoda, E. J.; Kucukboyaci, V. N.

    2012-07-01

    The efforts to reduce fuel cycle cost have driven LWR fuel close to the licensed limit in fuel fissile content, 5.0 wt% U-235 enrichment, and the acceptable duty on current Zr-based cladding. An increase in the fuel enrichment beyond the 5 wt% limit, while certainly possible, entails costly investment in infrastructure and licensing. As a possible way to offset some of these costs, the addition of small amounts of Erbia to the UO{sub 2} powder with >5 wt% U-235 has been proposed, so that its initial reactivity is reduced to that of licensed fuel and most modifications to the existingmore » facilities and equipment could be avoided. This paper discusses the potentialities of such a fuel on the US market from a vendor's perspective. An analysis of the in-core behavior and fuel cycle performance of a typical 4-loop PWR with 18 and 24-month operating cycles has been conducted, with the aim of quantifying the potential economic advantage and other operational benefits of this concept. Subsequently, the implications on fuel manufacturing and storage are discussed. While this concept has certainly good potential, a compelling case for its short-term introduction as PWR fuel for the US market could not be determined. (authors)« less

  13. Corrosion Evaluation of RERTR Uranium Molybdenum Fuel

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    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 Fluxmore » 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

  14. ATR LEU fuel and burnable absorber neutronics performance optimization by fuel meat thickness variation

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Chang, G.S.

    2008-07-15

    The Advanced Test Reactor (ATR) is a high power density and high neutron flux research reactor operating in the United States. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth. Because of the 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. The present work investigates the necessary modifications and evaluates the subsequent operating effects of this conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed and validated for a fuelmore » cycle burnup comparison analysis. Using the current HEU U-235 enrichment of 93.0 % as a baseline, an analysis can be performed to determine the low-enriched uranium (LEU) density and U-235 enrichment required in the fuel meat to yield an equivalent K-eff between the HEU core th and the LEU core versus effective full power days (EFPD). The MCNP ATR 1/8th core model will be used to optimize the U-235 loading in the LEU core, such that the differences in K-eff and heat flux profile between the HEU and LEU core can be minimized. The depletion methodology MCWO was used to calculate K-eff versus EFPDs in this paper. The MCWO-calculated results for the LEU cases with foil (U-10Mo) types demonstrated adequate excess reactivity such that the K-eff versus EFPDs plot is similar to the reference ATR HEU case. Each HEU fuel element contains 19 fuel plates with a fuel meat thickness of 0.508 mm. In this work, the proposed LEU (U-10Mo) core conversion case with a nominal fuel meat thickness of 0.381 mm and the same U-235 enrichment (19.7 wt%) can be used to optimize the radial heat flux profile by varying the fuel meat thickness from 0.191 mm (7.5 mil) to 0.343 mm (13.5 mil) at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19). In addition, 0.8g of a burnable absorber, Boron-10, was added in the inner and outer plates to reduce the initial excess

  15. ATR LEU Fuel and Burnable Absorber Neutronics Performance Optimization by Fuel Meat Thickness Variation

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    G. S. Chang

    2007-09-01

    The Advanced Test Reactor (ATR) is a high power density and high neutron flux research reactor operating in the United States. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth. Because of the 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. The present work investigates the necessary modifications and evaluates the subsequent operating effects of this conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed and validated for a fuelmore » cycle burnup comparison analysis. Using the current HEU U 235 enrichment of 93.0 % as a baseline, an analysis can be performed to determine the low-enriched uranium (LEU) density and U-235 enrichment required in the fuel meat to yield an equivalent K-eff between the HEU core and the LEU core versus effective full power days (EFPD). The MCNP ATR 1/8th core model will be used to optimize the U-235 loading in the LEU core, such that the differences in K-eff and heat flux profile between the HEU and LEU core can be minimized. The depletion methodology MCWO was used to calculate K-eff versus EFPDs in this paper. The MCWO-calculated results for the LEU cases with foil (U-10Mo) types demonstrated adequate excess reactivity such that the K-eff versus EFPDs plot is similar to the reference ATR HEU case. Each HEU fuel element contains 19 fuel plates with a fuel meat thickness of 0.508 mm. In this work, the proposed LEU (U-10Mo) core conversion case with a nominal fuel meat thickness of 0.508 mm and the same U-235 enrichment (15.5 wt%) can be used to optimize the radial heat flux profile by varying the fuel plate thickness from 0.254 to 0.457 mm at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19). In addition, a 0.7g of burnable absorber Boron-10 was added in the inner and outer plates to reduce the initial excess reactivity, and the inner

  16. Irradiation performance of U-Mo monolithic fuel

    DOE PAGES

    Meyer, M. K.; Gan, J.; Jue, J. F.; ...

    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. U-Mo alloys represent the best known tradeoff in these properties.more » 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.« less

  17. Returning HEU Fuel from the Czech Republic to Russia

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Michael Tyacke; Dr. Igor Bolshinsky

    In December 1999, representatives from the United States, Russian Federation, and International Atomic Energy Agency began working on a program to return Russian supplied, highly enriched, uranium fuel stored at foreign research reactors to Russia. Now, under the Global Threat Reduction Initiative’s Russian Research Reactor Fuel Return Program, this effort has repatriated over 800 kg of highly enriched uranium to Russia from over 10 countries. In May 2004, the “Agreement Between the Government of the United States of America and the Government of the Russian Federation Concerning Cooperation for the Transfer of Russian Produced Research Reactor Nuclear Fuel to themore » Russian Federation” was signed. This agreement provides legal authority for the Russian Research Reactor Fuel Return Program and establishes parameters whereby eligible countries may return highly enriched uranium spent and fresh fuel assemblies and other fissile materials to Russia. On December 8, 2007, one of the largest shipments of highly 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. In February 2003, Russian Research Reactor Fuel Return Program representatives met with the Nuclear Research Institute in Rež, Czech Republic, and discussed the return of their highly enriched uranium spent nuclear fuel to the Russian Federation for reprocessing. Nearly 5 years later, the shipment was made. This article discusses the planning, preparations, coordination, and cooperation required to make this important international shipment.« less

  18. Fuel bundle design for enhanced usage of plutonium fuel

    DOEpatents

    Reese, Anthony P.; Stachowski, Russell E.

    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.

  19. A model to predict thermal conductivity of irradiated U-Mo dispersion fuel

    NASA Astrophysics Data System (ADS)

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

    2016-05-01

    Numerous global programs are focused on the continued development of existing and new research and test reactor fuels to achieve maximum attainable uranium loadings to support the conversion of a number of the world's remaining high-enriched uranium fueled reactors to low-enriched uranium fuel. Some of these programs are focused on assisting with the development and qualification of a fuel design that consists of a uranium-molybdenum (U-Mo) alloy dispersed in an aluminum matrix as one option for reactor conversion. Thermal conductivity is an important consideration in determining the operational temperature of the fuel and can be influenced by interaction layer formation between the dispersed phase and matrix and upon the concentration of the dispersed phase within the matrix. This paper extends the use of a simple model developed previously to study the influence of interaction layer formation as well as the size and volume fraction of fuel particles dispersed in the matrix, Si additions to the matrix, and Mo concentration in the fuel particles on the effective thermal conductivity of the U-Mo/Al composite during irradiation. The model has been compared to experimental measurements recently conducted on U-Mo/Al dispersion fuels at two different fission densities with acceptable agreement. Observations of the modeled results indicate that formation of an interaction layer and subsequent consumption of the matrix reveals a rather significant effect on effective thermal conductivity. The modeled interaction layer formation and subsequent consumption of the high thermal conductivity matrix was sensitive to the average dispersed fuel particle size, suggesting this parameter as one of the most effective in minimizing thermal conductivity degradation of the composite, while the influence of Si additions to the matrix in the model was highly dependent upon irradiation conditions.

  20. A model to predict thermal conductivity of irradiated U–Mo dispersion fuel

    DOE Office of Scientific and Technical Information (OSTI.GOV)

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

    The Office of Materials Management and Minimization Reactor Conversion Program continues to develop existing and new research and test reactor fuels to achieve maximum attainable uranium loadings to support the conversion of a number of the world’s remaining high-enriched uranium fueled reactors to low-enriched uranium fuel. The program is focused on assisting with the development and qualification of a fuel design that consists of a uranium-molybdenum (U-Mo) alloy dispersed in an aluminum matrix as one option for reactor conversion. Thermal conductivity is an important consideration in determining the operational temperature of the fuel and can be influenced by interaction layermore » formation between the dispersed phase and matrix and upon the concentration of the dispersed phase within the matrix. This paper extends the use of a simple model developed previously to study the influence of interaction layer formation as well as the size and volume fraction of fuel particles dispersed in the matrix, Si additions to the matrix, and Mo concentration in the fuel particles on the effective thermal conductivity of the U-Mo/Al composite during irradiation. The model has been compared to experimental measurements recently conducted on U-Mo/Al dispersion fuels at two different fission densities with acceptable agreement. Observations of the modeled results indicate that formation of an interaction layer and subsequent consumption of the matrix reveals a rather significant effect on effective thermal conductivity. The modeled interaction layer formation and subsequent consumption of the high thermal conductivity matrix was sensitive to the average dispersed fuel particle size, suggesting this parameter as one of the most effective in minimizing thermal conductivity degradation of the composite, while the influence of Si additions to the matrix in the model was highly dependent upon irradiation conditions.« less

  1. Fire extinguishment in oxygen enriched atmospheres

    NASA Technical Reports Server (NTRS)

    Robertson, A. F.; Rappaport, M. W.

    1973-01-01

    Current state-of-the-art of fire suppression and extinguishment techniques in oxygen enriched atmosphere is reviewed. Four classes of extinguishment action are considered: cooling, separation of reactants, dilution or removal of fuel, and use of chemically reactive agents. Current practice seems to show preference for very fast acting water spray applications to all interior surfaces of earth-based chambers. In space, reliance has been placed on fire prevention methods through the removal of ignition sources and use of nonflammable materials. Recommendations are made for further work related to fire suppression and extinguishment in oxygen enriched atmospheres, and an extensive bibliography is appended.

  2. Functionally Stable and Phylogenetically Diverse Microbial Enrichments from Microbial Fuel Cells during Wastewater Treatment

    PubMed Central

    Ishii, Shun'ichi; Suzuki, Shino; Norden-Krichmar, Trina M.; Nealson, Kenneth H.; Sekiguchi, Yuji; Gorby, Yuri A.; Bretschger, Orianna

    2012-01-01

    Microbial fuel cells (MFCs) are devices that exploit microorganisms as biocatalysts to recover energy from organic matter in the form of electricity. One of the goals of MFC research is to develop the technology for cost-effective wastewater treatment. However, before practical MFC applications are implemented it is important to gain fundamental knowledge about long-term system performance, reproducibility, and the formation and maintenance of functionally-stable microbial communities. Here we report findings from a MFC operated for over 300 days using only primary clarifier effluent collected from a municipal wastewater treatment plant as the microbial resource and substrate. The system was operated in a repeat-batch mode, where the reactor solution was replaced once every two weeks with new primary effluent that consisted of different microbial and chemical compositions with every batch exchange. The turbidity of the primary clarifier effluent solution notably decreased, and 97% of biological oxygen demand (BOD) was removed after an 8–13 day residence time for each batch cycle. On average, the limiting current density was 1000 mA/m2, the maximum power density was 13 mW/m2, and coulombic efficiency was 25%. Interestingly, the electrochemical performance and BOD removal rates were very reproducible throughout MFC operation regardless of the sample variability associated with each wastewater exchange. While MFC performance was very reproducible, the phylogenetic analyses of anode-associated electricity-generating biofilms showed that the microbial populations temporally fluctuated and maintained a high biodiversity throughout the year-long experiment. These results suggest that MFC communities are both self-selecting and self-optimizing, thereby able to develop and maintain functional stability regardless of fluctuations in carbon source(s) and regular introduction of microbial competitors. These results contribute significantly toward the practical application of

  3. Toward a predictive theory for environmental enrichment.

    PubMed

    Watters, Jason V

    2009-11-01

    There have been many applications of and successes with environmental enrichment for captive animals. The theoretical spine upon which much enrichment work hangs largely describes why enrichment should work. Yet, there remains no clear understanding of how enrichment should be applied to achieve the most beneficial results. This lack of understanding may stem in part from the assumptions that underlie the application of enrichment by practitioners. These assumptions are derived from an understanding that giving animals choice and control in their environment stimulates their motivation to perform behaviors that may indicate a heightened state of well-being. Learning theory provides a means to question the manner in which these constructs are routinely applied, and converting learning theory's findings to optimality predictions suggests a particularly vexing paradox-that motivation to perform appears to be maintained best when acquiring a payoff for expressing the behavior is uncertain. This effect occurs even when the actual value of the payoff is the same for all schedules of certainty of payoff acquisition. The paradox can be resolved by invoking rewards of an alternative type, such as cognitive rewards, or through an understanding of how the average payoff changes with changes in the probability of reward. This model, with measures of the average change of the payoff, suggests testable scenarios by which practitioners can measure the quality of environmental uncertainty in enrichment programs.

  4. Analysis of intergranular fission-gas bubble-size distributions in irradiated uranium-molybdenum alloy fuel

    NASA Astrophysics Data System (ADS)

    Rest, J.; Hofman, G. L.; Kim, Yeon Soo

    2009-04-01

    An analytical model for the nucleation and growth of intra and intergranular fission-gas bubbles is used to characterize fission-gas bubble development in low-enriched U-Mo alloy fuel irradiated in the advanced test reactor in Idaho as part of the Reduced Enrichment for Research and Test Reactor (RERTR) program. Fuel burnup was limited to less than ˜7.8 at.% U in order to capture the fuel-swelling stage prior to irradiation-induced recrystallization. The model couples the calculation of the time evolution of the average intergranular bubble radius and number density to the calculation of the intergranular bubble-size distribution based on differential growth rate and sputtering coalescence processes. Recent results on TEM analysis of intragranular bubbles in U-Mo were used to set the irradiation-induced diffusivity and re-solution rate in the bubble-swelling model. Using these values, good agreement was obtained for intergranular bubble distribution compared against measured post-irradiation examination (PIE) data using grain-boundary diffusion enhancement factors of 15-125, depending on the Mo concentration. This range of enhancement factors is consistent with values obtained in the literature.

  5. Fuel bundle design for enhanced usage of plutonium fuel

    DOEpatents

    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.

  6. FUEL ASSAY REACTOR

    DOEpatents

    Spinrad, B.I.; Sandmeier, H.A.; Martens, F.H.

    1962-12-25

    A reactor having maximum sensitivity to perturbations is described comprising a core consisting of a horizontally disposed, rectangular, annular fuel zone containing enriched uranium dioxide dispersed in graphite, the concentration of uranium dioxide increasing from the outside to the inside of the fuel zone, an internal reflector of graphite containing an axial test opening disposed within the fuel zone, an external graphite reflector, means for changing the neutron spectrum in the test opening, and means for measuring perturbations in the neutron flux caused by the introduction of different fuel elements into the test opening. (AEC)

  7. 77 FR 2028 - 2017 and Later Model Year Light-Duty Vehicle Greenhouse Gas Emissions and Corporate Average Fuel...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-01-13

    ...EPA and NHTSA are announcing a 14-day extension of the comment period for the joint proposed rules ``2017 and Later Model Year Light- Duty Vehicle Greenhouse Gas Emissions and Corporate Average Fuel Economy Standards,'' published in the Federal Register on December 1, 2011 (76 FR 74854). The comment period was to end on January 30, 2012 (60 days after publication of the proposals in the Federal Register). This document extends the comment period to February 13, 2012. This extension of the comment period is provided to allow the public additional time to comment on the proposed rule. The extension of the comment period does not apply to NHTSA's Draft Environmental Impact Statement (Draft EIS), available on NHTSA's Web site at www.nhtsa.gov/fuel-economy. The comment period for NHTSA's Draft EIS closes on January 31, 2012.

  8. The slightly-enriched spectral shift control reactor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Martin, W.R.; Lee, J.C.; Larsen, E.W.

    1991-11-01

    An advanced converter reactor design utilizing mechanical spectral shift control rods in a conventional pressurized water reactor configuration is under investigation. The design is based on the principle that a harder spectrum during the early part of the fuel cycle will result in large neutron captures in fertile {sup 238}U, which can then be burned in situ in a softer spectrum later in the cycle. Preliminary design calculations performed during FY 89 showed that the slightly-enriched spectral shift reactor design offers the benefit of substantially increased fuel resource utilization with the proven safety characteristics of the pressurized water reactor technologymore » retained. Optimization of the fuel design and development of fuel management strategies were carried out in FY 90, along with effort to develop and validate neutronic methodology for tight-lattice configurations with hard spectra. During FY 91, the final year of the grant, the final Slightly-Enriched Spectral Shift Reactor (SESSR) design was determined, and reference design analyses were performed for the assemblies as well as the global core configuration, both at the beginning of cycle (BOC) and with depletion. The final SESSR design results in approximately a 20% increase in the utilization of uranium resources, based on equilibrium fuel cycle analyses. Acceptable pin power peaking is obtained with the final core design, with assembly peaking factors equal to less than 1.04 for spectral shift control rods both inserted and withdrawn, and global peaking factors at BOC predicted to be 1.4. In addition, a negative Moderation Temperature Coefficient (MTC) is maintained for BOC, which is difficult to achieve with conventional advanced converter designs based on a closed fuel cycle. The SESSR design avoids the need for burnable poison absorber, although they could be added if desired to increase the cycle length while maintaining a negative MTC.« less

  9. Looking North into Lab Metallurgy Testing Area and Enrichment Motor ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    Looking North into Lab Metallurgy Testing Area and Enrichment Motor within Recycle Recovery Building - Hematite Fuel Fabrication Facility, Recycle Recovery Building, 3300 State Road P, Festus, Jefferson County, MO

  10. Microfluidic droplet enrichment for targeted sequencing

    PubMed Central

    Eastburn, Dennis J.; Huang, Yong; Pellegrino, Maurizio; Sciambi, Adam; Ptáček, Louis J.; Abate, Adam R.

    2015-01-01

    Targeted sequence enrichment enables better identification of genetic variation by providing increased sequencing coverage for genomic regions of interest. Here, we report the development of a new target enrichment technology that is highly differentiated from other approaches currently in use. Our method, MESA (Microfluidic droplet Enrichment for Sequence Analysis), isolates genomic DNA fragments in microfluidic droplets and performs TaqMan PCR reactions to identify droplets containing a desired target sequence. The TaqMan positive droplets are subsequently recovered via dielectrophoretic sorting, and the TaqMan amplicons are removed enzymatically prior to sequencing. We demonstrated the utility of this approach by generating an average 31.6-fold sequence enrichment across 250 kb of targeted genomic DNA from five unique genomic loci. Significantly, this enrichment enabled a more comprehensive identification of genetic polymorphisms within the targeted loci. MESA requires low amounts of input DNA, minimal prior locus sequence information and enriches the target region without PCR bias or artifacts. These features make it well suited for the study of genetic variation in a number of research and diagnostic applications. PMID:25873629

  11. Nuclear Fuel Reprocessing

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Harold F. McFarlane; Terry Todd

    2013-11-01

    Reprocessing is essential to closing nuclear fuel cycle. Natural uranium contains only 0.7 percent 235U, the fissile (see glossary for technical terms) isotope that produces most of the fission energy in a nuclear power plant. Prior to being used in commercial nuclear fuel, uranium is typically enriched to 3–5% in 235U. If the enrichment process discards depleted uranium at 0.2 percent 235U, it takes more than seven tonnes of uranium feed to produce one tonne of 4%-enriched uranium. Nuclear fuel discharged at the end of its economic lifetime contains less one percent 235U, but still more than the natural ore.more » Less than one percent of the uranium that enters the fuel cycle is actually used in a single pass through the reactor. The other naturally occurring isotope, 238U, directly contributes in a minor way to power generation. However, its main role is to transmute into plutoniumby neutron capture and subsequent radioactive decay of unstable uraniumand neptuniumisotopes. 239Pu and 241Pu are fissile isotopes that produce more than 40% of the fission energy in commercially deployed reactors. It is recovery of the plutonium (and to a lesser extent the uranium) for use in recycled nuclear fuel that has been the primary focus of commercial reprocessing. Uraniumtargets irradiated in special purpose reactors are also reprocessed to obtain the fission product 99Mo, the parent isotope of technetium, which is widely used inmedical procedures. Among the fission products, recovery of such expensive metals as platinum and rhodium is technically achievable, but not economically viable in current market and regulatory conditions. During the past 60 years, many different techniques for reprocessing used nuclear fuel have been proposed and tested in the laboratory. However, commercial reprocessing has been implemented along a single line of aqueous solvent extraction technology called plutonium uranium reduction extraction process (PUREX). Similarly, hundreds of types of

  12. Deuterium Enrichment in Stratospheric Molecular Hydrogen

    NASA Astrophysics Data System (ADS)

    Rahn, T.; Eiler, J.; McCarthy, M. C.; Boering, K. A.; Wennberg, P.; Atlas, E.; Donnelly, S.; Schauffler, S.

    2002-12-01

    Molecular hydrogen (H2) is the second most abundant reduced gas in the atmosphere (after methane) with a globally averaged mixing ratio of ~ 530 ppbv. Its largest source is believed to be photochemical oxidation of methane (C H4) and non-methane hydrocarbons (NMHCs); other recognized sources include biomass burning, fossil fuel burning, nitrogen fixation, and ocean degassing. As with other atmospheric trace gases, the stable isotopic content of H2 has the potential to help quantify various aspects of its production and destruction. The average deuterium content of H2 (expressed as δDH2) is enriched by ~110 ‰ relative to Vienna Standard Mean Ocean Water while CH4 in the troposphere, the precursor for photochemical H2 production, is depleted by ~ 90 ‰ relative to V-SMOW and similar values are expected for NMHCs. Both natural and anthropogenic combustion sources of H2 have been shown to be depleted in deuterium by 200 to 300 ‰ (Gerst and Quay, 2001; Rahn et al., 2002), and the ocean and N2 fixation sources are expected to be in near thermodynamic equilibrium with local H2O and should have deuterium levels of ~-700 ‰ (Rahn et al., 2002). In order to offset these deuterium depleted sources and account for the observed tropospheric δDH2, the balancing loss processes must discriminate against reaction with HD and/or the total fractionation associated with CH4 oxidation and the subsequent reactions leading to H2 must favor production of deuterated H2. We have analyzed a suite of stratospheric air samples in order to investigate the photochemical processes influencing the deuterium content of H2. While the mixing ratio of H2 is nearly constant, the deuterium content increases such that δD=440 ‰ in samples with a stratospheric mean age of ~6 years. The constant mixing ratio results from the fact that production due to CH4 oxidation and loss due to H2 oxidation are approximately equal. The observed trend in δD of stratospheric H2 can only be accounted for by an

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

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Short, Steven M.; Phillips, Jon R.; Weimar, Mark R.

    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, thatmore » 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.« less

  14. NUCLEAR REACTOR FUEL SYSTEMS

    DOEpatents

    Thamer, B.J.; Bidwell, R.M.; Hammond, R.P.

    1959-09-15

    Homogeneous reactor fuel solutions are reported which provide automatic recombination of radiolytic gases and exhibit large thermal expansion characteristics, thereby providing stability at high temperatures and enabling reactor operation without the necessity of apparatus to recombine gases formed by the radiolytic dissociation of water in the fuel and without the necessity of liquid fuel handling outside the reactor vessel except for recovery processes. The fuels consist of phosphoric acid and water solutions of enriched uranium, wherein the uranium is in either the hexavalent or tetravalent state.

  15. The effects of reduced ambient temperatures on the warm-up fuel consumption behavior of gasoline fueled automobiles

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Pucher, G.R.; Gardiner, D.P.; Mallory, R.W.

    Warm-up fuel consumption behavior as affected by ambient temperature was evaluated for five OEM gasoline fueled automobiles. Multiple EPA FTP 75 tests were performed with each vehicle at ambient test cell soak temperatures of 25 C and {minus}7 C. Fuel consumption measured during the warm-up (Bag 1, Cold Transient) test segments at these two temperature conditions was compared to the fully warmed Hot Transient (Bag 3) fuel consumption from the 25 C ambient temperature tests (the Bag 1 and Bag 3 segments involve identical speed curves). Fuel consumption increases over the 25 C Bag 3 tests for the two warm-upmore » test conditions were differentiated as those caused by increased drivetrain losses and those caused by intake charge enrichment. Results show wide variations in warm-up behavior among the five vehicles with respect to the relative increases in fuel consumption, and the proportion of the fuel consumption increases attributable to drivetrain losses and enrichment. It was discovered that the most sophisticated vehicle systems do not necessarily facilitate the least degradation in fuel consumption with respect to baseline conditions for the group of vehicles tested.« less

  16. RERTR 2009 (Reduced Enrichment for Research and Test Reactors)

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Totev, T.; Stevens, J.; Kim, Y. S.

    2010-03-01

    The U.S. Department of Energy/National Nuclear Security Administration's Office of Global Threat Reduction in cooperation with the China Atomic Energy Authority and International Atomic Energy Agency hosted the 'RERTR 2009 International Meeting on Reduced Enrichment for Research and Test Reactors.' The meeting was organized by Argonne National Laboratory, China Institute of Atomic Energy and Idaho National Laboratory and was held in Beijing, China from November 1-5, 2009. This was the 31st annual meeting in a series on the same general subject regarding the conversion of reactors within the Global Threat Reduction Initiative (GTRI). The Reduced Enrichment for Research and Testmore » Reactors (RERTR) Program develops technology necessary to enable the conversion of civilian facilities using high enriched uranium (HEU) to low enriched uranium (LEU) fuels and targets.« less

  17. Local Burn-Up Effects in the NBSR Fuel Element

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Brown N. R.; Hanson A.; Diamond, D.

    2013-01-31

    This study addresses the over-prediction of local power when the burn-up distribution in each half-element of the NBSR is assumed to be uniform. A single-element model was utilized to quantify the impact of axial and plate-wise burn-up on the power distribution within the NBSR fuel elements for both high-enriched uranium (HEU) and low-enriched uranium (LEU) fuel. To validate this approach, key parameters in the single-element model were compared to parameters from an equilibrium core model, including neutron energy spectrum, power distribution, and integral U-235 vector. The power distribution changes significantly when incorporating local burn-up effects and has lower power peakingmore » relative to the uniform burn-up case. In the uniform burn-up case, the axial relative power peaking is over-predicted by as much as 59% in the HEU single-element and 46% in the LEU single-element with uniform burn-up. In the uniform burn-up case, the plate-wise power peaking is over-predicted by as much as 23% in the HEU single-element and 18% in the LEU single-element. The degree of over-prediction increases as a function of burn-up cycle, with the greatest over-prediction at the end of Cycle 8. The thermal flux peak is always in the mid-plane gap; this causes the local cumulative burn-up near the mid-plane gap to be significantly higher than the fuel element average. Uniform burn-up distribution throughout a half-element also causes a bias in fuel element reactivity worth, due primarily to the neutronic importance of the fissile inventory in the mid-plane gap region.« less

  18. Conceptual Core Analysis of Long Life PWR Utilizing Thorium-Uranium Fuel Cycle

    NASA Astrophysics Data System (ADS)

    Rouf; Su'ud, Zaki

    2016-08-01

    Conceptual core analysis of long life PWR utilizing thorium-uranium based fuel has conducted. The purpose of this study is to evaluate neutronic behavior of reactor core using combined thorium and enriched uranium fuel. Based on this fuel composition, reactor core have higher conversion ratio rather than conventional fuel which could give longer operation length. This simulation performed using SRAC Code System based on library SRACLIB-JDL32. The calculation carried out for (Th-U)O2 and (Th-U)C fuel with uranium composition 30 - 40% and gadolinium (Gd2O3) as burnable poison 0,0125%. The fuel composition adjusted to obtain burn up length 10 - 15 years under thermal power 600 - 1000 MWt. The key properties such as uranium enrichment, fuel volume fraction, percentage of uranium are evaluated. Core calculation on this study adopted R-Z geometry divided by 3 region, each region have different uranium enrichment. The result show multiplication factor every burn up step for 15 years operation length, power distribution behavior, power peaking factor, and conversion ratio. The optimum core design achieved when thermal power 600 MWt, percentage of uranium 35%, U-235 enrichment 11 - 13%, with 14 years operation length, axial and radial power peaking factor about 1.5 and 1.2 respectively.

  19. Alternative Fuel Vehicle Adoption Increases Fleet Gasoline Consumption and Greenhouse Gas Emissions under United States Corporate Average Fuel Economy Policy and Greenhouse Gas Emissions Standards.

    PubMed

    Jenn, Alan; Azevedo, Inês M L; Michalek, Jeremy J

    2016-03-01

    The United States Corporate Average Fuel Economy (CAFE) standards and Greenhouse Gas (GHG) Emission standards are designed to reduce petroleum consumption and GHG emissions from light-duty passenger vehicles. They do so by requiring automakers to meet aggregate criteria for fleet fuel efficiency and carbon dioxide (CO2) emission rates. Several incentives for manufacturers to sell alternative fuel vehicles (AFVs) have been introduced in recent updates of CAFE/GHG policy for vehicles sold from 2012 through 2025 to help encourage a fleet technology transition. These incentives allow automakers that sell AFVs to meet less-stringent fleet efficiency targets, resulting in increased fleet-wide gasoline consumption and emissions. We derive a closed-form expression to quantify these effects. We find that each time an AFV is sold in place of a conventional vehicle, fleet emissions increase by 0 to 60 t of CO2 and gasoline consumption increases by 0 to 7000 gallons (26,000 L), depending on the AFV and year of sale. Using projections for vehicles sold from 2012 to 2025 from the Energy Information Administration, we estimate that the CAFE/GHG AFV incentives lead to a cumulative increase of 30 to 70 million metric tons of CO2 and 3 to 8 billion gallons (11 to 30 billion liters) of gasoline consumed over the vehicles' lifetimes - the largest share of which is due to legacy GHG flex-fuel vehicle credits that expire in 2016. These effects may be 30-40% larger in practice than we estimate here due to optimistic laboratory vehicle efficiency tests used in policy compliance calculations.

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

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    K. J. Allen; I. Bolshinsky; L. L. Biro

    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 Returnmore » 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.« less

  1. 76 FR 26996 - Notice of Intent To Prepare an Environmental Impact Statement for New Corporate Average Fuel...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-05-10

    ...Pursuant to the National Environmental Policy Act (NEPA), NHTSA plans to prepare an Environmental Impact Statement (EIS) to analyze the potential environmental impacts of the agency's Corporate Average Fuel Economy program for passenger automobiles (referred to herein as ``passenger cars'') and non-passenger automobiles (referred to herein as ``light trucks''). The EIS will consider the potential environmental impacts of new fuel economy standards for model years 2017-2025 passenger cars and light trucks that NHTSA will be proposing pursuant to the Energy Independence and Security Act of 2007. This notice initiates the NEPA scoping process by inviting comments from Federal, State, and local agencies, Indian tribes, and the public to help identify the environmental issues and reasonable alternatives to be examined in the EIS. This notice also provides guidance for participating in the scoping process and additional information about the alternatives NHTSA expects to consider in its NEPA analysis. In preparing this notice, NHTSA has shared the document with the Council on Environmental Quality (CEQ), the Environmental Protection Agency (EPA), and the Department of Energy (DOE).

  2. 77 FR 60482 - Regulatory Guide 5.67, Material Control and Accounting for Uranium Enrichment Facilities...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-10-03

    ... Accounting for Uranium Enrichment Facilities Authorized To Produce Special Nuclear Material of Low Strategic... Accounting for Uranium Enrichment Facilities Authorized to Produce Special Nuclear Material of Low Strategic... INFORMATION CONTACT: Glenn Tuttle, Office of Nuclear Material Safety and Safeguards, Division of Fuel Cycle...

  3. Assessment of Nuclear Fuels using Radiographic Thickness Measurement Method

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Muhammad Abir; Fahima Islam; Hyoung Koo Lee

    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 Highmore » 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.« less

  4. 40 CFR 600.510-12 - Calculation of average fuel economy and average carbon-related exhaust emissions.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... MOTOR VEHICLES Fuel Economy Regulations for Model Year 1978 Passenger Automobiles and for 1979 and Later Model Year Automobiles (Light Trucks and Passenger Automobiles)-Procedures for Determining Manufacturer... will be calculated to the nearest 0.1 mpg for the categories of automobiles identified in this section...

  5. Data Mining Techniques to Estimate Plutonium, Initial Enrichment, Burnup, and Cooling Time in Spent Fuel Assemblies

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Trellue, Holly Renee; Fugate, Michael Lynn; Tobin, Stephen Joesph

    The Next Generation Safeguards Initiative (NGSI), Office of Nonproliferation and Arms Control (NPAC), National Nuclear Security Administration (NNSA) of the U.S. Department of Energy (DOE) has sponsored a multi-laboratory, university, international partner collaboration to (1) detect replaced or missing pins from spent fuel assemblies (SFA) to confirm item integrity and deter diversion, (2) determine plutonium mass and related plutonium and uranium fissile mass parameters in SFAs, and (3) verify initial enrichment (IE), burnup (BU), and cooling time (CT) of facility declaration for SFAs. A wide variety of nondestructive assay (NDA) techniques were researched to achieve these goals [Veal, 2010 andmore » Humphrey, 2012]. In addition, the project includes two related activities with facility-specific benefits: (1) determination of heat content and (2) determination of reactivity (multiplication). In this research, a subset of 11 integrated NDA techniques was researched using data mining solutions at Los Alamos National Laboratory (LANL) for their ability to achieve the above goals.« less

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

    DOEpatents

    Zinn, Walter H.

    1976-01-01

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

  7. Enhancing the corrosion resistance of the 2205 duplex stainless steel bipolar plates in PEMFCs environment by surface enriched molybdenum

    NASA Astrophysics Data System (ADS)

    Jinlong, Lv; Zhuqing, Wang; Tongxiang, Liang; Ken, Suzuki; Hideo, Miura

    Surface molybdenum enrichment on 2205 duplex stainless steel was obtained by the ball milling technique. The electrochemical results showed molybdenum enrichment on the surface of 2205 duplex stainless steel improved its corrosion resistance in a typical proton exchange membrane fuel cell environment. This was mainly attributed to higher molybdenum content in the passive film formed on 2205 duplex stainless steel after ball milling. The decreased donor and acceptor concentrations improved significantly the corrosion resistance of surface molybdenum-enriched 2205 duplex stainless steel bipolar plates in the simulated cathodic proton exchange membrane fuel cells environment. In addition, the interfacial contact resistance of the 2205 duplex stainless steel bipolar plates slightly decreased due to surface molybdenum enrichment.

  8. Heavy Ozone Enrichments from ATMOS Infrared Solar Spectra

    NASA Technical Reports Server (NTRS)

    Irion, F. W.; Gunson, M. R.; Rinsland, C. P.; Yung, Y. L.; Abrams, M. C.; Chang, A. Y.; Goldman, A.

    1996-01-01

    Vertical enrichment profiles of stratospheric O-16O-16O-18 and O-16O-18O-16 (hereafter referred to as (668)O3 and (686)O3 respectively) have been derived from space-based solar occultation spectra recorded at 0.01 cm(exp-1) resolution by the ATMOS (Atmospheric Trace MOlecule Spectroscopy) Fourier transform infrared (FTIR) spectrometer. The observations, made during the Spacelab 3 and ATLAS-1, -2, and -3 shuttle missions, cover polar, mid-latitude and tropical regions between 26 to 2.6 mb inclusive (approximately 25 to 41 km). Average enrichments, weighted by molecular (48)O3 density, of (15 +/- 6)% were found for (668)O3 and (10 +/- 7)% for (686)O3. Defining the mixing ratio of (50)O3 as the sum of those for (668)O3 and (686)O3, an enrichment of (13 plus or minus 5)% was found for (50)O3 (1 sigma standard deviation). No latitudinal or vertical gradients were found outside this standard deviation. From a series of ground-based measurements by the ATMOS instrument at Table Mountain, California (34.4 deg N), an average total column (668)O3 enrichment of (17 +/- 4)% (1 sigma standard deviation) was determined, with no significant seasonal variation discernable. Possible biases in the spectral intensities that affect the determination of absolute enrichments are discussed.

  9. Characterization of Used Nuclear Fuel with Multivariate Analysis for Process Monitoring

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Dayman, Kenneth J.; Coble, Jamie B.; Orton, Christopher R.

    2014-01-01

    The Multi-Isotope Process (MIP) Monitor combines gamma spectroscopy and multivariate analysis to detect anomalies in various process streams in a nuclear fuel reprocessing system. Measured spectra are compared to models of nominal behavior at each measurement location to detect unexpected changes in system behavior. In order to improve the accuracy and specificity of process monitoring, fuel characterization may be used to more accurately train subsequent models in a full analysis scheme. This paper presents initial development of a reactor-type classifier that is used to select a reactor-specific partial least squares model to predict fuel burnup. Nuclide activities for prototypic usedmore » fuel samples were generated in ORIGEN-ARP and used to investigate techniques to characterize used nuclear fuel in terms of reactor type (pressurized or boiling water reactor) and burnup. A variety of reactor type classification algorithms, including k-nearest neighbors, linear and quadratic discriminant analyses, and support vector machines, were evaluated to differentiate used fuel from pressurized and boiling water reactors. Then, reactor type-specific partial least squares models were developed to predict the burnup of the fuel. Using these reactor type-specific models instead of a model trained for all light water reactors improved the accuracy of burnup predictions. The developed classification and prediction models were combined and applied to a large dataset that included eight fuel assembly designs, two of which were not used in training the models, and spanned the range of the initial 235U enrichment, cooling time, and burnup values expected of future commercial used fuel for reprocessing. Error rates were consistent across the range of considered enrichment, cooling time, and burnup values. Average absolute relative errors in burnup predictions for validation data both within and outside the training space were 0.0574% and 0.0597%, respectively. The errors seen in

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

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Fedorov, Y.S.; Bibichev, B.A.; Zilberman, B.Y.

    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)

  11. 49 CFR 531.5 - Fuel economy standards.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... ADMINISTRATION, DEPARTMENT OF TRANSPORTATION PASSENGER AUTOMOBILE AVERAGE FUEL ECONOMY STANDARDS § 531.5 Fuel... automobiles shall comply with the fleet average fuel economy standards in Table I, expressed in miles per... passenger automobile fleet shall comply with the fleet average fuel economy level calculated for that model...

  12. 49 CFR 531.5 - Fuel economy standards.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... ADMINISTRATION, DEPARTMENT OF TRANSPORTATION PASSENGER AUTOMOBILE AVERAGE FUEL ECONOMY STANDARDS § 531.5 Fuel... automobiles shall comply with the fleet average fuel economy standards in Table I, expressed in miles per... passenger automobile fleet shall comply with the fleet average fuel economy level calculated for that model...

  13. Production of Low Enriched Uranium Nitride Kernels for TRISO Particle Irradiation Testing

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    McMurray, J. W.; Silva, C. M.; Helmreich, G. W.

    2016-06-01

    A large batch of UN microspheres to be used as kernels for TRISO particle fuel was produced using carbothermic reduction and nitriding of a sol-gel feedstock bearing tailored amounts of low-enriched uranium (LEU) oxide and carbon. The process parameters, established in a previous study, produced phasepure NaCl structure UN with dissolved C on the N sublattice. The composition, calculated by refinement of the lattice parameter from X-ray diffraction, was determined to be UC 0.27N 0.73. The final accepted product weighed 197.4 g. The microspheres had an average diameter of 797±1.35 μm and a composite mean theoretical density of 89.9±0.5% formore » a solid solution of UC and UN with the same atomic ratio; both values are reported with their corresponding calculated standard error.« less

  14. Partial defect verification of spent fuel assemblies by PDET: Principle and field testing in Interim Spent fuel Storage Facility (CLAB) in Sweden

    DOE PAGES

    Ham, Y.; Kerr, P.; Sitaraman, S.; ...

    2016-05-05

    Here, the need for the development of a credible method and instrument for partial defect verification of spent fuel has been emphasized over a few decades in the safeguards communities as the diverted spent fuel pins can be the source of nuclear terrorism or devices. The need is increasingly more important and even urgent as many countries have started to transfer spent fuel to so called "difficult-to-access" areas such as dry storage casks, reprocessing or geological repositories. Partial defect verification is required by IAEA before spent fuel is placed into "difficult-to-access" areas. Earlier, Lawrence Livermore National Laboratory (LLNL) has reportedmore » the successful development of a new, credible partial defect verification method for pressurized water reactor (PWR) spent fuel assemblies without use of operator data, and further reported the validation experiments using commercial spent fuel assemblies with some missing fuel pins. The method was found to be robust as the method is relatively invariant to the characteristic variations of spent fuel assemblies such as initial fuel enrichment, cooling time, and burn-up. Since then, the PDET system has been designed and prototyped for 17×17 PWR spent fuel assemblies, complete with data acquisition software and acquisition electronics. In this paper, a summary description of the PDET development followed by results of the first successful field testing using the integrated PDET system and actual spent fuel assemblies performed in a commercial spent fuel storage site, known as Central Interim Spent fuel Storage Facility (CLAB) in Sweden will be presented. In addition to partial defect detection initial studies have determined that the tool can be used to verify the operator declared average burnup of the assembly as well as intra-assembly bunrup levels.« less

  15. Partial Defect Verification of Spent Fuel Assemblies by PDET: Principle and Field Testing in Interim Spent Fuel Storage Facility (CLAB) in Sweden

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Ham, Y.S.; Kerr, P.; Sitaraman, S.

    The need for the development of a credible method and instrument for partial defect verification of spent fuel has been emphasized over a few decades in the safeguards communities as the diverted spent fuel pins can be the source of nuclear terrorism or devices. The need is increasingly more important and even urgent as many countries have started to transfer spent fuel to so called 'difficult-to-access' areas such as dry storage casks, reprocessing or geological repositories. Partial defect verification is required by IAEA before spent fuel is placed into 'difficult-to-access' areas. Earlier, Lawrence Livermore National Laboratory (LLNL) has reported themore » successful development of a new, credible partial defect verification method for pressurized water reactor (PWR) spent fuel assemblies without use of operator data, and further reported the validation experiments using commercial spent fuel assemblies with some missing fuel pins. The method was found to be robust as the method is relatively invariant to the characteristic variations of spent fuel assemblies such as initial fuel enrichment, cooling time, and burn-up. Since then, the PDET system has been designed and prototyped for 17x17 PWR spent fuel assemblies, complete with data acquisition software and acquisition electronics. In this paper, a summary description of the PDET development followed by results of the first successful field testing using the integrated PDET system and actual spent fuel assemblies performed in a commercial spent fuel storage site, known as Central Interim Spent fuel Storage Facility (CLAB) in Sweden will be presented. In addition to partial defect detection initial studies have determined that the tool can be used to verify the operator declared average burnup of the assembly as well as intra-assembly burnup levels. (authors)« less

  16. Partial defect verification of spent fuel assemblies by PDET: Principle and field testing in Interim Spent fuel Storage Facility (CLAB) in Sweden

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Ham, Y.; Kerr, P.; Sitaraman, S.

    Here, the need for the development of a credible method and instrument for partial defect verification of spent fuel has been emphasized over a few decades in the safeguards communities as the diverted spent fuel pins can be the source of nuclear terrorism or devices. The need is increasingly more important and even urgent as many countries have started to transfer spent fuel to so called "difficult-to-access" areas such as dry storage casks, reprocessing or geological repositories. Partial defect verification is required by IAEA before spent fuel is placed into "difficult-to-access" areas. Earlier, Lawrence Livermore National Laboratory (LLNL) has reportedmore » the successful development of a new, credible partial defect verification method for pressurized water reactor (PWR) spent fuel assemblies without use of operator data, and further reported the validation experiments using commercial spent fuel assemblies with some missing fuel pins. The method was found to be robust as the method is relatively invariant to the characteristic variations of spent fuel assemblies such as initial fuel enrichment, cooling time, and burn-up. Since then, the PDET system has been designed and prototyped for 17×17 PWR spent fuel assemblies, complete with data acquisition software and acquisition electronics. In this paper, a summary description of the PDET development followed by results of the first successful field testing using the integrated PDET system and actual spent fuel assemblies performed in a commercial spent fuel storage site, known as Central Interim Spent fuel Storage Facility (CLAB) in Sweden will be presented. In addition to partial defect detection initial studies have determined that the tool can be used to verify the operator declared average burnup of the assembly as well as intra-assembly bunrup levels.« less

  17. New potentials for conventional aircraft when powered by hydrogen-enriched gasoline

    NASA Technical Reports Server (NTRS)

    Menard, W. A.; Moynihan, P. I.; Rupe, J. H.

    1976-01-01

    Hydrogen enrichment for aircraft piston engines is studied. The feasibility is examined of inflight injection of hydrogen in general aviation aircraft engines to reduce fuel consumption and to lower emission levels. Results are summarized.

  18. Irradiation performance of AGR-1 high temperature reactor fuel

    DOE PAGES

    Demkowicz, Paul A.; Hunn, John D.; Ploger, Scott A.; ...

    2015-10-23

    The AGR-1 experiment contained 72 low-enriched uranium oxide/uranium carbide TRISO coated particle fuel compacts in six capsules irradiated to burnups of 11.2 to 19.6% FIMA, with zero TRISO coating failures detected during the irradiation. The irradiation performance of the fuel including the extent of fission product release and the evolution of kernel and coating microstructures was evaluated based on detailed examination of the irradiation capsules, the fuel compacts, and individual particles. Fractional release of 110mAg from the fuel compacts was often significant, with capsule-average values ranging from 0.01 to 0.38. Analysis of silver release from individual compacts indicated that itmore » was primarily dependent on fuel temperature history. Europium and strontium were released in small amounts through intact coatings, but were found to be significantly retained in the outer pyrocarbon and compact matrix. The capsule-average fractional release from the compacts was 1 × 10 –4 to 5 × 10 –4 for 154Eu and 8 × 10 –7 to 3 × 10 –5 for 90Sr. The average 134Cs fractional release from compacts was <3 × 10 –6 when all particles maintained intact SiC. An estimated four particles out of 2.98 × 10 5 in the experiment experienced partial cesium release due to SiC failure during the irradiation, driving 134Cs fractional release in two capsules to approximately 10 –5. Identification and characterization of these particles has provided unprecedented insight into the nature and causes of SiC coating failure in high-quality TRISO fuel. In general, changes in coating morphology were found to be dominated by the behavior of the buffer and inner pyrolytic carbon (IPyC), and infrequently observed SiC layer damage was usually related to cracks in the IPyC. Palladium attack of the SiC layer was relatively minor, except for the particles that released cesium during irradiation, where SiC corrosion was found adjacent to IPyC cracks. In conclusion, palladium

  19. Measured and Predicted Vapor Liquid Equilibrium of Ethanol-Gasoline Fuels with Insight on the Influence of Azeotrope Interactions on Aromatic Species Enrichment and Particulate Matter Formation in Spark Ignition Engines

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Ratcliff, Matthew A; McCormick, Robert L; Burke, Stephen

    A relationship has been observed between increasing ethanol content in gasoline and increased particulate matter (PM) emissions from direct injection spark ignition (DISI) vehicles. The fundamental cause of this observation is not well understood. One potential explanation is that increased evaporative cooling as a result of ethanol's high HOV may slow evaporation and prevent sufficient reactant mixing resulting in the combustion of localized fuel rich regions within the cylinder. In addition, it is well known that ethanol when blended in gasoline forms positive azeotropes which can alter the liquid/vapor composition during the vaporization process. In fact, it was shown recentlymore » through a numerical study that these interactions can retain the aromatic species within the liquid phase impeding the in-cylinder mixing of these compounds, which would accentuate PM formation upon combustion. To better understand the role of the azeotrope interactions on the vapor/liquid composition evolution of the fuel, distillations were performed using the Advanced Distillation Curve apparatus on carefully selected samples consisting of gasoline blended with ethanol and heavy aromatic and oxygenated compounds with varying vapor pressures, including cumene, p-cymene, 4-tertbutyl toluene, anisole, and 4-methyl anisole. Samples collected during the distillation indicate an enrichment of the heavy aromatic or oxygenated additive with an increase in initial ethanol concentration from E0 to E30. A recently developed distillation and droplet evaporation model is used to explore the influence of dilution effects versus azeotrope interactions on the aromatic species enrichment. The results suggest that HOV-cooling effects as well as aromatic species enrichment behaviors should be considered in future development of predictive indices to forecast the PM potential of fuels containing oxygenated compounds with comparatively high HOV.« less

  20. Evaluation of biomass quality of selected woody species depending on the soil enrichment practice

    NASA Astrophysics Data System (ADS)

    Stolarski, Mariusz J.; Krzyżaniak, Michał; Załuski, Dariusz; Niksa, Dariusz

    2018-01-01

    Perennial energy crops are a source of the bio-mass used to generate energy. The aim of this study was to determine the chemical and thermophysical parameters of short rotation woody crops (black locust, poplar and willow), depending on soil enrichment practice (mineral fertilisation, lignin and mycorrhiza), in three- and four-year harvest cycles. In the study, the thermophysical properties and elemental composition of the biomass were determined. All analyses were performed in trip-licate according to the standards. The fresh black locust biomass had the lowest moisture content, which resulted in the best lower heating value (10.16 MJ kg-1, on average) in the four-year harvest cycle. The poplar biomass had the greatest higher heating value, fixed carbon, carbon and ash content, the highest concentrations of which were found in the biomass in which lignin was applied (2.00% d.m.). On the other hand, the willow biomass contained the lowest concentrations of ash and fixed carbon. Soil enrichment significantly differentiated the quality parameters of black locust, poplar and willow. This effect is of particular importance to those who grow and use biomass as a fuel.

  1. 40 CFR 63.7541 - How do I demonstrate continuous compliance under the emission averaging provision?

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... solid fuel boilers participating in the emissions averaging option as determined in § 63.7522(f) and (g... this section. (i) For each existing solid fuel boiler participating in the emissions averaging option... below the applicable limit. (ii) For each group of boilers participating in the emissions averaging...

  2. On the progressive enrichment of the oxygen isotopic composition of water along a leaf.

    PubMed

    Farquhar, G. D.; Gan, K. S.

    2003-06-01

    A model has been derived for the enrichment of heavy isotopes of water in leaves, including progressive enrichment along the leaf. In the model, lighter water is preferentially transpired leaving heavier water to diffuse back into the xylem and be carried further along the leaf. For this pattern to be pronounced, the ratio of advection to diffusion (Péclet number) has to be large in the longitudinal direction, and small in the radial direction. The progressive enrichment along the xylem is less than that occurring at the sites of evaporation in the mesophyll, depending on the isolation afforded by the radial Péclet number. There is an upper bound on enrichment, and effects of ground tissue associated with major veins are included. When transpiration rate is spatially nonuniform, averaging of enrichment occurs more naturally with transpiration weighting than with area-based weighting. This gives zero average enrichment of transpired water, the modified Craig-Gordon equation for average enrichment at the sites of evaporation and the Farquhar and Lloyd (In Stable Isotopes and Plant Carbon-Water Relations, pp. 47-70. Academic Press, New York, USA, 1993) prediction for mesophyll water. Earlier results on the isotopic composition of evolved oxygen and of retro-diffused carbon dioxide are preserved if these processes vary in parallel with transpiration rate. Parallel variation should be indicated approximately by uniform carbon isotope discrimination across the leaf.

  3. [Enrichment and toxicity effect of heavy metals in soil ecosystem].

    PubMed

    Wang, Zhenzhong; Zhang, Youmei; Deng, Jifu; Li, Zhongwu

    2006-10-01

    The study on the heavy metals-polluted soil of Qingshuitang District, Zhuzhou City showed that the main enriched heavy metals in the soil were Cd, Hg, As, Zn, Pb and Cu, among which, the former three had strong biotoxicity while the others had definite toxicity, with the average integrative pollution index being 6.40. The heavy metals enrichment in soil animals increased with increasing pollution degree. Especially for some species of Megascoiecidae, they had a high enrichment of Cd, with the enrichment index being 11.96. The species and quantity of soil animals decreased with increasing Cd concentration. Cd had an obvious biotoxic effect on Megsacoiecidae' s isozyme activity. The heavy metals enrichment in vegetables also increased with increasing pollution degree, but there existed obvious interspecies difference in Cd enrichment.

  4. Pythons metabolize prey to fuel the response to feeding.

    PubMed Central

    Starck, J. Matthias; Moser, Patrick; Werner, Roland A.; Linke, Petra

    2004-01-01

    We investigated the energy source fuelling the post-feeding metabolic upregulation (specific dynamic action, SDA) in pythons (Python regius). Our goal was to distinguish between two alternatives: (i) snakes fuel SDA by metabolizing energy depots from their tissues; or (ii) snakes fuel SDA by metabolizing their prey. To characterize the postprandial response of pythons we used transcutaneous ultrasonography to measure organ-size changes and respirometry to record oxygen consumption. To discriminate unequivocally between the two hypotheses, we enriched mice (= prey) with the stable isotope of carbon (13C). For two weeks after feeding we quantified the CO2 exhaled by pythons and determined its isotopic 13C/12C signature. Ultrasonography and respirometry showed typical postprandial responses in pythons. After feeding, the isotope ratio of the exhaled breath changed rapidly to values that characterized enriched mouse tissue, followed by a very slow change towards less enriched values over a period of two weeks after feeding. We conclude that pythons metabolize their prey to fuel SDA. The slowly declining delta13C values indicate that less enriched tissues (bone, cartilage and collagen) from the mouse become available after several days of digestion. PMID:15255044

  5. Microstructural Characterization of the U-9.1Mo Fuel/AA6061 Cladding Interface in Friction-Bonded Monolithic Fuel Plates Irradiated in the RERTR-6 Experiment

    DOE PAGES

    Keiser, Dennis D.; Jue, Jan-Fong; Miller, Brandon; ...

    2015-09-03

    Low-enrichment (U-235 < 20%) U-Mo monolithic fuel is being developed for use in research and test reactors. The earliest design for this fuel that was investigated via reactor testing was comprised of a nominally U-10Mo fuel foil encased in AA6061 (Al-6061) cladding. For a fuel design to be deemed adequate for final use in a reactor, it must maintain dimensional stability and retain fission products throughout irradiation, which means that there must be good integrity at the fuel foil/cladding interface. To investigate the nature of the fuel/cladding interface for this fuel type after irradiation, fuel plates that were tested inmore » INL's Advanced Test Reactor (ATR) were subsequently characterized using optical metallography, scanning electron microscopy, and transmission electron microscopy. Results of this characterization showed that the fuel/cladding interaction layers present at the U-Mo fuel/AA6061 cladding interface after fabrication became amorphous during irradiation. Up to two main interaction layers, based on composition, could be found at the fuel/cladding interface, depending on location. After irradiation, an Al-rich layer contained very few fission gas bubbles, but did exhibit Xe enrichment near the AA6061 cladding interface. Another layer, which contained more Si, had more observable fission gas bubbles. Adjacent to the AA6061 cladding were Mg-rich precipitates, which was in close proximity to the region where Xe is observed to be enriched. In samples produced using a focused ion beam at the interaction zone/AA6061 cladding interface were possible indications of porosity/debonding, which suggested that the interface in this location is relatively weak.« less

  6. 78 FR 63518 - Uranium Enrichment Fuel Cycle Inspection Reports Regarding Louisiana Energy Services, National...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-10-24

    ... support safe operation of Autoclave 2 of the facility have been constructed in accordance with the... Inspection Reports Regarding Louisiana Energy Services, National Enrichment Facility, Eunice, New Mexico... Louisiana Energy Services (LES), LLC, National Enrichment Facility in Eunice, New Mexico, and has authorized...

  7. Fuel weight and removal costs in fuel-break construction

    Treesearch

    James L. Murphy

    1966-01-01

    Three major fuel types were sampled during fuel-break construction on the west side of the Sierra Nevada, California. Fuel weight per acre ranged from 12.2. to 420.6 tons per acre on average. Fuel-break construction removed 27.9 to 40.5 percent of teh total fuel; costs ranged from $141.29 to $148.38 per acre and $1.16 to $4.43 per ton.

  8. Observed Changes in As-Fabricated U-10Mo Monolithic Fuel Microstructures After Irradiation in the Advanced Test Reactor

    NASA Astrophysics Data System (ADS)

    Keiser, Dennis; Jue, Jan-Fong; Miller, Brandon; Gan, Jian; Robinson, Adam; Madden, James

    2017-12-01

    A low-enriched uranium U-10Mo monolithic nuclear fuel is being developed by the Material Management and Minimization Program, earlier known as the Reduced Enrichment for Research and Test Reactors Program, for utilization in research and test reactors around the world that currently use high-enriched uranium fuels. As part of this program, reactor experiments are being performed in the Advanced Test Reactor. It must be demonstrated that this fuel type exhibits mechanical integrity, geometric stability, and predictable behavior to high powers and high fission densities in order for it to be a viable fuel for qualification. This paper provides an overview of the microstructures observed at different regions of interest in fuel plates before and after irradiation for fuel samples that have been tested. These fuel plates were fabricated using laboratory-scale fabrication methods. Observations regarding how microstructural changes during irradiation may impact fuel performance are discussed.

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

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Not Available

    1993-07-01

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

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

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    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 productmore » 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.« less

  11. Proceedings of the 1994 international meeting on reduced enrichment for research and test reactors

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    NONE

    1997-08-01

    This meeting brought together participants in the international effort to minimize and eventually eliminate the use of highly enriched uranium in civilian nuclear programs. Papers cover the following topics: National programs; fuel cycle; nuclear fuels; analyses; advanced reactors; and reactor conversions. Selected papers have been indexed separately for inclusion to the Energy Science and Technology Database.

  12. Novel Accident-Tolerant Fuel Meat and Cladding

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Robert D. Mariani; Pavel G Medvedev; Douglas L Porter

    A novel accident-tolerant fuel meat and cladding are here proposed. The fuel meat design incorporates annular fuel with inserts and discs that are fabricated from a material having high thermal conductivity, for example niobium. The inserts are rods or tubes. Discs separate the fuel pellets. Using the BISON fuel performance code it was found that the peak fuel temperature can be lowered by more than 600 degrees C for one set of conditions with niobium metal as the thermal conductor. In addition to improved safety margin, several advantages are expected from the lower temperature such as decreased fission gas releasemore » and fuel cracking. Advantages and disadvantages are discussed. An enrichment of only 7.5% fully compensates the lost reactivity of the displaced UO2. Slightly higher enrichments, such as 9%, allow uprates and increased burnups to offset the initial costs for retooling. The design has applications for fast reactors and transuranic burning, which may accelerate its development. A zirconium silicide coating is also described for accident tolerant applications. A self-limiting degradation behavior for this coating is expected to produce a glassy, self-healing layer that becomes more protective at elevated temperature, with some similarities to MoSi2 and other silicides. Both the fuel and coating may benefit from the existing technology infrastructure and the associated wide expertise for a more rapid development in comparison to other, more novel fuels and cladding.« less

  13. High-Fidelity Modelng and Simulation for a High Flux Isotope Reactor Low-Enriched Uranium Core Design

    DOE PAGES

    Betzler, Benjamin R.; Chandler, David; Davidson, Eva E.; ...

    2017-05-08

    A high-fidelity model of the High Flux Isotope Reactor (HFIR) with a low-enriched uranium (LEU) fuel design and a representative experiment loading has been developed to serve as a new reference model for LEU conversion studies. With the exception of the fuel elements, this HFIR LEU model is completely consistent with the current highly enriched uranium HFIR model. Results obtained with the new LEU model provide a baseline for analysis of alternate LEU fuel designs and further optimization studies. The newly developed HFIR LEU model has an explicit representation of the HFIR-specific involute fuel plate geometry, including the within-plate fuelmore » meat contouring, and a detailed geometry model of the fuel element side plates. Such high-fidelity models are necessary to accurately account for the self-shielding from 238U and the depletion of absorber materials present in the side plates. In addition, a method was developed to account for fuel swelling in the high-density LEU fuel plates during the depletion simulation. In conclusion, calculated time-dependent metrics for the HFIR LEU model include fission rate and cumulative fission density distributions, flux and reaction rates for relevant experiment locations, point kinetics data, and reactivity coefficients.« less

  14. High-Fidelity Modelng and Simulation for a High Flux Isotope Reactor Low-Enriched Uranium Core Design

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Betzler, Benjamin R.; Chandler, David; Davidson, Eva E.

    A high-fidelity model of the High Flux Isotope Reactor (HFIR) with a low-enriched uranium (LEU) fuel design and a representative experiment loading has been developed to serve as a new reference model for LEU conversion studies. With the exception of the fuel elements, this HFIR LEU model is completely consistent with the current highly enriched uranium HFIR model. Results obtained with the new LEU model provide a baseline for analysis of alternate LEU fuel designs and further optimization studies. The newly developed HFIR LEU model has an explicit representation of the HFIR-specific involute fuel plate geometry, including the within-plate fuelmore » meat contouring, and a detailed geometry model of the fuel element side plates. Such high-fidelity models are necessary to accurately account for the self-shielding from 238U and the depletion of absorber materials present in the side plates. In addition, a method was developed to account for fuel swelling in the high-density LEU fuel plates during the depletion simulation. In conclusion, calculated time-dependent metrics for the HFIR LEU model include fission rate and cumulative fission density distributions, flux and reaction rates for relevant experiment locations, point kinetics data, and reactivity coefficients.« less

  15. Fuel Fabrication and Nuclear Reactors

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Karpius, Peter Joseph

    2017-02-02

    The uranium from the enrichment plant is still in the form of UF 6. UF 6 is not suitable for use in a reactor due to its highly corrosive chemistry as well as its phase diagram. UF 6 is converted into UO 2 fuel pellets, which are in turn placed in fuel rods and assemblies. Reactor designs are variable in moderators, coolants, fuel, performance etc.The dream of energy ‘too-cheap to meter’ is no more, and now the nuclear power industry is pushing ahead with advanced reactor designs.

  16. Production test IP-544-A, irradiation of 1.6% enriched thick walled single tube elements in KER-1 and 2

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Kratzer, W.K.; Wise, M.J.

    1962-12-12

    The objective of this production test is to authorize the irradiation of coextruded Zr-2 jacketed thick walled 1.6% enriched tubular elements in KER loops 1 and 2 to evaluate the swelling behavior of fuel elements at high uranium temperatures Coextruded Zr-2 jacketed 1.6% enriched tubular fuel elements 1.79 inch OD, 0.97 inch ID, and 12 inches long will be irradiated KER loops 1 and 2 to exposures no greater than 2500 MWD/T.

  17. The prospect of uranium nitride (UN) and mixed nitride fuel (UN-PuN) for pressurized water reactor

    NASA Astrophysics Data System (ADS)

    Syarifah, Ratna Dewi; Suud, Zaki

    2015-09-01

    Design study of small Pressurized Water Reactors (PWRs) core loaded with uranium nitride fuel (UN) and mixed nitride fuel (UN-PuN), Pa-231 as burnable poison, and Americium has been performed. Pa-231 known as actinide material, have large capture cross section and can be converted into fissile material that can be utilized to reduce excess reactivity. Americium is one of minor actinides with long half life. The objective of adding americium is to decrease nuclear spent fuel in the world. The neutronic analysis results show that mixed nitride fuel have k-inf greater than uranium nitride fuel. It is caused by the addition of Pu-239 in mixed nitride fuel. In fuel fraction analysis, for uranium nitride fuel, the optimum volume fractions are 45% fuel fraction, 10% cladding and 45% moderator. In case of UN-PuN fuel, the optimum volume fractions are 30% fuel fraction, 10% cladding and 60% coolant/ moderator. The addition of Pa-231 as burnable poison for UN fuel, enrichment U-235 5%, with Pa-231 1.6% has k-inf more than one and excess reactivity of 14.45%. And for mixed nitride fuel, the lowest value of reactivity swing is when enrichment (U-235+Pu) 8% with Pa-231 0.4%, the excess reactivity value 13,76%. The fuel pin analyze for the addition of Americium, the excess reactivity value is lower than before, because Americium absorb the neutron. For UN fuel, enrichment U-235 8%, Pa-231 1.6% and Am 0.5%, the excess reactivity is 4.86%. And for mixed nitride fuel, when enrichment (U-235+Pu) 13%, Pa-231 0.4% and Am 0.1%, the excess reactivity is 11.94%. For core configuration, it is better to use heterogeneous than homogeneous core configuration, because the radial power distribution is better.

  18. Research and Development of Multiphysics Models in Support of the Conversion of the High Flux Isotope Reactor to Low Enriched Uranium Fuel

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Bodey, Isaac T.; Curtis, Franklin G.; Arimilli, Rao V.

    The findings presented in this report are results of a five year effort led by the RRD Division of the ORNL, which is focused on research and development toward the conversion of the High Flux Isotope Reactor (HFIR) fuel from high-enriched uranium (HEU) to low-enriched uranium (LEU). This report focuses on the tasks accomplished by the University of Tennessee Knoxville (UTK) team from the Department of Mechanical, Aerospace, and Biomedical Engineering (MABE) that provided expert support in multiphysics modeling of complex problems associated with the LEU conversion of the HFIR reactor. The COMSOL software was used as the main computationalmore » modeling tool, whereas Solidworks was also used in support of computer-aided-design (CAD) modeling of the proposed LEU fuel design. The UTK research has been governed by a statement of work (SOW), which was updated annually to clearly define the specific tasks reported herein. Ph.D. student Isaac T. Bodey has focused on heat transfer and fluid flow modeling issues and has been aided by his major professor Dr. Rao V. Arimilli. Ph.D. student Franklin G. Curtis has been focusing on modeling the fluid-structure interaction (FSI) phenomena caused by the mechanical forces acting on the fuel plates, which in turn affect the fluid flow in between the fuel plates, and ultimately the heat transfer, is also affected by the FSI changes. Franklin Curtis has been aided by his major professor Dr. Kivanc Ekici. M.Sc. student Adam R. Travis has focused two major areas of research: (1) on accurate CAD modeling of the proposed LEU plate design, and (2) reduction of the model complexity and dimensionality through interdimensional coupling of the fluid flow and heat transfer for the HFIR plate geometry. Adam Travis is also aided by his major professor, Dr. Kivanc Ekici. We must note that the UTK team, and particularly the graduate students, have been in very close collaboration with Dr. James D. Freels (ORNL technical monitor and mentor) and

  19. 49 CFR 531.5 - Fuel economy standards.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... ADMINISTRATION, DEPARTMENT OF TRANSPORTATION PASSENGER AUTOMOBILE AVERAGE FUEL ECONOMY STANDARDS § 531.5 Fuel... automobiles shall comply with the average fuel economy standards in Table I, expressed in miles per gallon, in... passenger automobile fleet shall comply with the fuel economy level calculated for that model year according...

  20. 49 CFR 531.5 - Fuel economy standards.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... ADMINISTRATION, DEPARTMENT OF TRANSPORTATION PASSENGER AUTOMOBILE AVERAGE FUEL ECONOMY STANDARDS § 531.5 Fuel... automobiles shall comply with the average fuel economy standards in Table I, expressed in miles per gallon, in... passenger automobile fleet shall comply with the fuel economy level calculated for that model year according...

  1. 40 CFR 63.7522 - Can I use emissions averaging to comply with this subpart?

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... section. You may not include new boilers or process heaters in an emissions average. (b) For a group of... subcategory of units designed to burn gas 2 (other) fuels. (iv) You may not average across the units designed to burn liquid, units designed to burn solid fuel, and units designed to burn gas 2 (other...

  2. 49 CFR 535.7 - Averaging, banking, and trading (ABT) program.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... averaging set. With the exception of FCC earned for advance technologies as further clarified below, a... advanced technology credits, may be used only within the averaging set in which they were generated. (b... (gasoline, diesel and alternative fuel) and advanced technology vehicles (hybrids with regenerative braking...

  3. Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

    NASA Astrophysics Data System (ADS)

    Mohammed, Abdul Aziz; Pauzi, Anas Muhamad; Rahman, Shaik Mohmmed Haikhal Abdul; Zin, Muhamad Rawi Muhammad; Jamro, Rafhayudi; Idris, Faridah Mohamad

    2016-01-01

    In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 (233U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintaining the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.

  4. Highly enriched multiply-labeled stable isotopic compounds as atmospheric tracers

    DOEpatents

    Goldblatt, M.; McInteer, B.B.

    1974-01-29

    Compounds multiply-labeled with stable isotopes and highly enriched in these isotopes are readily capable of detection in tracer experiments involving high dilutions. Thus, for example, /sup 13/C/sup 18/O/sub 2/ provides a useful tracer for following atmospheric pol lution produced as a result of fossil fuel burning. (Official Gazette)

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

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    James A. Smith; Barry H. Rabin; Mathieu Perton

    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.more » 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.« less

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

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Perton, M.; Levesque, D.; Monchalin, J.-P.

    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.more » 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.« less

  7. New potentials for conventional aircraft when powered by hydrogen-enriched gasoline

    NASA Technical Reports Server (NTRS)

    Menard, W. A.; Moynihan, P. I.; Rupe, J. H.

    1976-01-01

    Hydrogen enrichment for aircraft piston engines is under study in a new NASA program. The objective of the program is to determine the feasibility of inflight injection of hydrogen in general aviation aircraft engines to reduce fuel consumption and to lower emission levels. A catalytic hydrogen generator will be incorporated as part of the air induction system of a Lycoming turbocharged engine and will generate hydrogen by breaking down small amounts of the aviation gasoline used in the normal propulsion system. This hydrogen will then be mixed with gasoline and compressed air from the turbocharger before entering the engine combustion chamber. The paper summarizes the results of a systems analysis study. Calculations assuming a Beech Duke aircraft indicate that fuel savings on the order of 20% are possible. An estimate of the potential for the utilization of hydrogen enrichment to control exhaust emissions indicates that it may be possible to meet the 1979 Federal emission standards.

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

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Goodsell, Alison Victoria; Swinhoe, Martyn Thomas; Henzl, Vladimir

    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 performedmore » 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.« less

  9. Review of the TREAT Conversion Conceptual Design and Fuel Qualification Plan

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Diamond, David

    The U.S. Department of Energy (DOE) is preparing to re establish the capability to conduct transient testing of nuclear fuels at the Idaho National Laboratory (INL) Transient Reactor Test (TREAT) facility. The original TREAT core went critical in February 1959 and operated for more than 6,000 reactor startups before plant operations were suspended in 1994. DOE is now planning to restart the reactor using the plant's original high-enriched uranium (HEU) fuel. At the same time, the National Nuclear Security Administration (NNSA) Office of Material Management and Minimization Reactor Conversion Program is supporting analyses and fuel fabrication studies that will allowmore » for reactor conversion to low-enriched uranium (LEU) fuel (i.e., fuel with less than 20% by weight 235U content) after plant restart. The TREAT Conversion Program's objectives are to perform the design work necessary to generate an LEU replacement core, to restore the capability to fabricate TREAT fuel element assemblies, and to implement the physical and operational changes required to convert the TREAT facility to use LEU fuel.« less

  10. Enhanced thermal conductivity oxide nuclear fuels by co-sintering with BeO: II. Fuel performance and neutronics

    NASA Astrophysics Data System (ADS)

    McCoy, Kevin; Mays, Claude

    2008-04-01

    The fuel rod performance and neutronics of enhanced thermal conductivity oxide (ECO) nuclear fuel with BeO have been compared to those of standard UO 2 fuel. The standards of comparison were that the ECO fuel should have the same infinite neutron-multiplication factor kinf at end of life and provide the same energy extraction per fuel assembly over its lifetime. The BeO displaces some uranium, so equivalence with standard UO 2 fuel was obtained by increasing the burnup and slightly increasing the enrichment. The COPERNIC fuel rod performance code was adapted to account for the effect of BeO on thermal properties. The materials considered were standard UO 2, UO 2 with 4.0 vol.% BeO, and UO 2 with 9.6 vol.% BeO. The smaller amount of BeO was assumed to provide increases in thermal conductivity of 0, 5, or 10%, whereas the larger amount was assumed to provide an increase of 50%. A significant improvement in performance was seen, as evidenced by reduced temperatures, internal rod pressures, and fission gas release, even with modest (5-10%) increases in thermal conductivity. The benefits increased monotonically with increasing thermal conductivity. Improvements in LOCA initialization performance were also seen. A neutronic calculation considered a transition from standard UO 2 fuel to ECO fuel. The calculation indicated that only a small increase in enrichment is required to maintain the kinf at end of life. The smallness of the change was attributed to the neutron-multiplication reaction of Be with fast neutrons and the moderating effect of BeO. Adoption of ECO fuel was predicted to provide a net reduction in uranium cost. Requirements for industrial hygiene were found to be comparable to those for processing of UO 2.

  11. Device for sampling and enriching impurities in hydrogen comprising hydrogen-permeable membrane

    DOEpatents

    Ahmed, Shabbir; Papadias, Dionissios D.; Lee, Sheldon D. H.; Kumar, Romesh

    2017-01-31

    Provided herein are methods and devices to enrich trace quantities of impurities in gaseous mixtures, such as hydrogen fuel. The methods and devices rely on concentration of impurities so as to allow the detection of the impurities using commonly-available detection methods.

  12. Automobile fuel economy : potential effects of increasing the corporate average fuel economy standards

    DOT National Transportation Integrated Search

    2000-08-01

    Between 1981 and 1999, the average price of gasoline, adjusted for inflation, declined more than 60 percent. During the same period, the U.S. transportation sector's consumption of oil rose from less than 10 million to nearly 13 million barrels per d...

  13. Using NDVI to assess departure from average greenness and its relation to fire business

    Treesearch

    Robert E. Burgan; Roberta A. Hartford; Jeffery C. Eidenshink

    1996-01-01

    A new satellite-derived vegetation greenness map, departure from average, is designed to compare current-year vegetation greenness with average greenness for the same time of year. Live-fuel condition as portrayed on this map, and the calculated 1,000-hour fuel moistures, are compared to fire occurrence and area burned in Montana and Idaho during the 1993 and 1994 fire...

  14. High-temperature electrolysis of CO2-enriched mixtures by using fuel-electrode supported La0.6Sr0.4CoO3/YSZ/Ni-YSZ solid oxide cells

    NASA Astrophysics Data System (ADS)

    Kim, Si-Won; Bae, Yonggyun; Yoon, Kyung Joong; Lee, Jong-Ho; Lee, Jong-Heun; Hong, Jongsup

    2018-02-01

    To mitigate CO2 emissions, its reduction by high-temperature electrolysis using solid oxide cells is extensively investigated, for which excessive steam supply is assumed. However, such condition may degrade its feasibility due to massive energy required for generating hot steam, implying the needs for lowering steam demand. In this study, high-temperature electrolysis of CO2-enriched mixtures by using fuel-electrode supported La0.6Sr0.4CoO3/YSZ/Ni-YSZ solid oxide cells is considered to satisfy such needs. The effect of internal and external steam supply on its electrochemical performance and gas productivity is elucidated. It is shown that the steam produced in-situ inside the fuel-electrode by a reverse water gas shift reaction may decrease significantly the electrochemical resistance of dry CO2-fed operations, attributed to self-sustaining positive thermo-electrochemical reaction loop. This mechanism is conspicuous at low current density, whereas it is no longer effective at high current density in which total reactant concentrations for electrolysis is critical. To overcome such limitations, a small amount of external steam supply to the CO2-enriched feed stream may be needed, but this lowers the CO2 conversion and CO/H2 selectivity. Based on these results, it is discussed that there can be minimum steam supply sufficient for guaranteeing both low electrochemical resistance and high gas productivity.

  15. Multiple recycle of REMIX fuel at VVER-1000 operation in closed fuel cycle

    NASA Astrophysics Data System (ADS)

    Alekseev, P. N.; Bobrov, E. A.; Chibinyaev, A. V.; Teplov, P. S.; Dudnikov, A. A.

    2015-12-01

    The basic features of loading the VVER-1000 core with a new variant of REMIX fuel (REgenerated MIXture of U-Pu oxides) are considered during its multiple recycle in a closed nuclear fuel cycle. The fuel composition is produced on the basis of the uranium-plutonium regenerate extracted at processing the spent nuclear fuel (SNF) from a VVER-1000, depleted uranium, and the fissionable material: 235U as a part of highly enriched uranium (HEU) from warheads superfluous for defense purposes or 233U accumulated in thorium blankets of fusion (electronuclear) neutron sources or fast reactors. Production of such a fuel assumes no use of natural uranium in addition. When converting a part of the VVER-1000 reactors to the closed fuel cycle based on the REMIX technology, the consumption of natural uranium decreases considerably, and there is no substantial degradation of the isotopic composition of plutonium or change in the reactor-safety characteristics at the passage from recycle to recycle.

  16. Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Mohammed, Abdul Aziz, E-mail: azizM@uniten.edu.my; Rahman, Shaik Mohmmed Haikhal Abdul; Pauzi, Anas Muhamad, E-mail: anas@uniten.edu.my

    2016-01-22

    In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 ({sup 233}U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintainingmore » the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.« less

  17. Review of Membrane Oxygen Enrichment for Efficient Combustion

    NASA Astrophysics Data System (ADS)

    Ariono, Danu; Kusuma Wardani, Anita

    2017-07-01

    Oxygen enrichment from air is a simple way of increasing the efficiency of combustion process, as in oxy-combustion. Oxy-combustion has become one of the most attracting combustion technologies because of its potential to address both pollutant reduction and CO2 capture. In oxy-combustion, the fuel and recycled flue gas are combusted with oxygen enriched air (OEA). By using OEA, many benefits can be obtained, such as increasing available heat, improving ignition characteristics, flue gas reduction, increasing productivity, energy efficiency, turndown ratio, and flame stability. Membrane-based gas separation for OEA production becomes an attractive technology over the conventional technology due to the some advantages, including low capital cost, low energy consumption, compact size, and modularity. A single pass through membrane usually can enrich O2 concentration in the air up to 35% and a 50% concentration can be achieved with a double pass of membrane. The use of OEA in the combustion process eliminates the presence of nitrogen in the flue gas. Hence, the flue gas is mainly composed of CO2 and condensable water that can be easily separated. This paper gives an overview of oxy-combustion with membrane technology for oxygen enrichment process. Special attention is given to OEA production and the effect of OEA to the efficiency of combustion.

  18. Managing the Nuclear Fuel Cycle: Policy Implications of Expanding Global Access to Nuclear Power

    DTIC Science & Technology

    2008-09-03

    Spent nuclear fuel disposal has remained the most critical aspect of the nuclear fuel cycle for the United States, where longstanding nonproliferation...inalienable right and by and large, neither have U.S. government officials. However, the case of Iran raises perhaps the most critical question in...the enrichment process can take advantage of the slight difference in atomic mass between 235U and 238U. The typical enrichment process requires

  19. Electron Microscopic Examination of Irradiated TRISO Coated Particles of Compact 6-3-2 of AGR-1 Experiment

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Van Rooyen, Isabella Johanna; Demkowicz, Paul Andrew; Riesterer, Jessica Lori

    2012-12-01

    The electron microscopic examination of selected irradiated TRISO coated particles of the AGR-1 experiment of fuel compact 6-3-2 are presented in this report. Compact 6-3-2 refers to the compact in Capsule 6 at level 3 of Stack 2. The fuel used in capsule 6 compacts, are called the “baseline” fuel as it is fabricated with refined coating process conditions used to fabricate historic German fuel, because of its excellent irradiation performance with UO2 kernels. The AGR-1 fuel is however made of low-enriched uranium oxycarbide (UCO). Kernel diameters are approximately 350 µm with a U-235 enrichment of approximately 19.7%. Compact 6-3-2more » has been irradiated to 11.3% FIMA compact average burn-up with a time average, volume average temperature of 1070.2°C and with a compact average fast fluence of 2.38E21 n/cm« less

  20. Electron Microscopic Examination of Irradiated TRISO Coated Particles of Compact 6-3-2 of AGR-1 Experiment

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Van Rooyen, Isabella Johanna; Demkowicz, Paul Andrew; Riesterer, Jessica Lori

    2012-12-01

    The electron microscopic examination of selected irradiated TRISO coated particles of the AGR-1 experiment of fuel compact 6-3-2 are presented in this report. Compact 6-3-2 refers to the compact in Capsule 6 at level 3 of Stack 2. The fuel used in capsule 6 compacts, are called the “baseline” fuel as it is fabricated with refined coating process conditions used to fabricate historic German fuel, because of its excellent irradiation performance with UO 2 kernels. The AGR-1 fuel is however made of low-enriched uranium oxycarbide (UCO). Kernel diameters are approximately 350 µm with a U-235 enrichment of approximately 19.7%. Compactmore » 6-3-2 has been irradiated to 11.3% FIMA compact average burn-up with a time average, volume average temperature of 1070.2°C and with a compact average fast fluence of 2.38E21 n/cm« less

  1. 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.

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Lell, R. M.; McKnight, R. D.; Tsiboulia, A.

    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 physicsmore » 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

  2. 40 CFR 63.7522 - Can I use emissions averaging to comply with this subpart?

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    .... (vi) Dutch ovens/pile burners designed to burn biomass/bio-based solid. (vii) Fuel Cells designed to...: Industrial, Commercial, and Institutional Boilers and Process Heaters Testing, Fuel Analyses, and Initial... allowed as follows: (i) You may average among units in any of the solid fuel subcategories. (ii) You may...

  3. FY2016 Ceramic Fuels Development Annual Highlights

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Mcclellan, Kenneth James

    Key challenges for the Advanced Fuels Campaign are the development of fuel technologies to enable major increases in fuel performance (safety, reliability, power and burnup) beyond current technologies, and development of characterization methods and predictive fuel performance models to enable more efficient development and licensing of advanced fuels. Ceramic fuel development activities for fiscal year 2016 fell within the areas of 1) National and International Technical Integration, 2) Advanced Accident Tolerant Ceramic Fuel Development, 3) Advanced Techniques and Reference Materials Development, and 4) Fabrication of Enriched Ceramic Fuels. High uranium density fuels were the focus of the ceramic fuels efforts.more » Accomplishments for FY16 primarily reflect the prioritization of identification and assessment of new ceramic fuels for light water reactors which have enhanced accident tolerance while also maintaining or improving normal operation performance, and exploration of advanced post irradiation examination techniques which will support more efficient testing and qualification of new fuel systems.« less

  4. The effect of oxygen enrichment on soot formation and thermal radiation in turbulent, non-premixed methane flames

    DOE PAGES

    Shaddix, Christopher R.; Williams, Timothy C.

    2016-07-12

    Non-premixed oxy-fuel combustion of natural gas is used in industrial applications where high-intensity heat is required, such as glass manufacturing and metal forging and shaping. In these applications, the high flame temperatures achieved by oxy-fuel combustion increase radiative heat transfer to the surfaces of interest and soot formation within the flame is desired for further augmentation of radiation. However, the high cost of cryogenic air separation has limited the penetration of oxy-fuel combustion technologies. New approaches to air separation are being developed that may reduce oxygen production costs, but only for intermediate levels of oxygen enrichment of air. To determinemore » the influence of oxygen enrichment on soot formation and radiation, we developed a non-premixed coannular burner in which oxygen concentrations and oxidizer flow rates can be independently varied, to distinguish the effects of turbulent mixing intensity from oxygen enrichment on soot formation and flame radiation. Local radiation intensities, soot concentrations, and soot temperatures have been measured using a thin-film thermopile, planar laser-induced incandescence (LII), and two-color imaging pyrometry, respectively. The measurements show that soot formation increases as the oxygen concentration decreases from 100% to 50%, helping to moderate a decrease in overall flame radiation. An increase in turbulence intensity has a marked effect on flame height, soot formation and thermal radiation, leading to decreases in all of these. The soot temperature decreases with a decrease in the oxygen concentration and increases with an increase in turbulent mixing intensity. Altogether, the results suggest that properly designed oxygen-enriched burners that enhance soot formation for intermediate levels of oxygen purity may be able to achieve thermal radiation intensities as high as 85% of traditional oxy-fuel burners utilizing high-purity oxygen.« less

  5. The effect of oxygen enrichment on soot formation and thermal radiation in turbulent, non-premixed methane flames

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Shaddix, Christopher R.; Williams, Timothy C.

    Non-premixed oxy-fuel combustion of natural gas is used in industrial applications where high-intensity heat is required, such as glass manufacturing and metal forging and shaping. In these applications, the high flame temperatures achieved by oxy-fuel combustion increase radiative heat transfer to the surfaces of interest and soot formation within the flame is desired for further augmentation of radiation. However, the high cost of cryogenic air separation has limited the penetration of oxy-fuel combustion technologies. New approaches to air separation are being developed that may reduce oxygen production costs, but only for intermediate levels of oxygen enrichment of air. To determinemore » the influence of oxygen enrichment on soot formation and radiation, we developed a non-premixed coannular burner in which oxygen concentrations and oxidizer flow rates can be independently varied, to distinguish the effects of turbulent mixing intensity from oxygen enrichment on soot formation and flame radiation. Local radiation intensities, soot concentrations, and soot temperatures have been measured using a thin-film thermopile, planar laser-induced incandescence (LII), and two-color imaging pyrometry, respectively. The measurements show that soot formation increases as the oxygen concentration decreases from 100% to 50%, helping to moderate a decrease in overall flame radiation. An increase in turbulence intensity has a marked effect on flame height, soot formation and thermal radiation, leading to decreases in all of these. The soot temperature decreases with a decrease in the oxygen concentration and increases with an increase in turbulent mixing intensity. Altogether, the results suggest that properly designed oxygen-enriched burners that enhance soot formation for intermediate levels of oxygen purity may be able to achieve thermal radiation intensities as high as 85% of traditional oxy-fuel burners utilizing high-purity oxygen.« less

  6. Fuel Cycle Performance of Thermal Spectrum Small Modular Reactors

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Worrall, Andrew; Todosow, Michael

    2016-01-01

    Small modular reactors may offer potential benefits, such as enhanced operational flexibility. However, it is vital to understand the holistic impact of small modular reactors on the nuclear fuel cycle and fuel cycle performance. The focus of this paper is on the fuel cycle impacts of light water small modular reactors in a once-through fuel cycle with low-enriched uranium fuel. A key objective of this paper is to describe preliminary reactor core physics and fuel cycle analyses conducted in support of the U.S. Department of Energy Office of Nuclear Energy Fuel Cycle Options Campaign. Challenges with small modular reactors include:more » increased neutron leakage, fewer assemblies in the core (and therefore fewer degrees of freedom in the core design), complex enrichment and burnable absorber loadings, full power operation with inserted control rods, the potential for frequent load-following operation, and shortened core height. Each of these will impact the achievable discharge burn-up in the reactor and the fuel cycle performance. This paper summarizes the results of an expert elicitation focused on developing a list of the factors relevant to small modular reactor fuel, core, and operation that will impact fuel cycle performance. Preliminary scoping analyses were performed using a regulatory-grade reactor core simulator. The hypothetical light water small modular reactor considered in these preliminary scoping studies is a cartridge type one-batch core with 4.9% enrichment. Some core parameters, such as the size of the reactor and general assembly layout, are similar to an example small modular reactor concept from industry. The high-level issues identified and preliminary scoping calculations in this paper are intended to inform on potential fuel cycle impacts of one-batch thermal spectrum SMRs. In particular, this paper highlights the impact of increased neutron leakage and reduced number of batches on the achievable burn-up of the reactor. Fuel cycle

  7. Analysis of gamma ray dose for dried up pond storing low enriched UO2 fuel

    NASA Astrophysics Data System (ADS)

    Nauchi, Yasushi; Suzuki, Motomu

    2017-09-01

    Gamma ray dose is calculated for loss of coolant accident in spent fuel pond (SFP) storing irradiated fuels used in light water reactors. Influence of modelling of fuel assemblies, source distributions, and loading fraction of fuel assemblies in the fuel rack on the dose are investigated.

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

    DOE Office of Scientific and Technical Information (OSTI.GOV)

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

    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 themore » 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.« less

  9. Analysis of irradiated U-7wt%Mo dispersion fuel microstructures using automated image processing

    DOE PAGES

    Collette, R.; King, J.; Buesch, C.; ...

    2016-04-01

    The High Performance Research Reactor Fuel Development (HPPRFD) program is responsible for developing low enriched uranium (LEU) fuel substitutes for high performance reactors fueled with highly enriched uranium (HEU) that have not yet been converted to LEU. The uranium-molybdenum (U-Mo) fuel system was selected for this effort. In this study, fission gas pore segmentation was performed on U-7wt%Mo dispersion fuel samples at three separate fission densities using an automated image processing interface developed in MATLAB. Pore size distributions were attained that showed both expected and unexpected fission gas behavior. In general, it proved challenging to identify any dominant trends whenmore » comparing fission bubble data across samples from different fuel plates due to varying compositions and fabrication techniques. Here, the results exhibited fair agreement with the fission density vs. porosity correlation developed by the Russian reactor conversion program.« less

  10. Analysis of irradiated U-7wt%Mo dispersion fuel microstructures using automated image processing

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Collette, R.; King, J.; Buesch, C.

    The High Performance Research Reactor Fuel Development (HPPRFD) program is responsible for developing low enriched uranium (LEU) fuel substitutes for high performance reactors fueled with highly enriched uranium (HEU) that have not yet been converted to LEU. The uranium-molybdenum (U-Mo) fuel system was selected for this effort. In this study, fission gas pore segmentation was performed on U-7wt%Mo dispersion fuel samples at three separate fission densities using an automated image processing interface developed in MATLAB. Pore size distributions were attained that showed both expected and unexpected fission gas behavior. In general, it proved challenging to identify any dominant trends whenmore » comparing fission bubble data across samples from different fuel plates due to varying compositions and fabrication techniques. Here, the results exhibited fair agreement with the fission density vs. porosity correlation developed by the Russian reactor conversion program.« less

  11. NEUTRON REACTOR FUEL ELEMENT UTILIZING ZIRCONIUM-BASE ALLOYS

    DOEpatents

    Saller, H.A.; Keeler, J.R.; Szumachowski, E.R.

    1957-11-12

    This patent relates to clad fuel elements for use in neutronic reactors and is drawn to such a fuel element which consists of a core of fissionable material, comprised of an alloy of zirconium and U/sup 235/ enriched uranium, encased in a jacket of a binary zirconium-tin alloy in which the tin content ranges between 1 and 15% by weight.

  12. NREL Fuel Cell Bus Analysis Finds Fuel Economy to be 1.4 Times Higher than

    Science.gov Websites

    Diesel | News | NREL Fuel Cell Bus Analysis Finds Fuel Economy to be 1.4 Times Higher than Diesel NREL Fuel Cell Bus Analysis Finds Fuel Economy to be 1.4 Times Higher than Diesel December 2, 2016 NREL has published a new report showing that the average fuel economy of fuel cell electric buses from

  13. Multiple recycle of REMIX fuel at VVER-1000 operation in closed fuel cycle

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Alekseev, P. N.; Bobrov, E. A., E-mail: evgeniybobrov89@rambler.ru; Chibinyaev, A. V.

    2015-12-15

    The basic features of loading the VVER-1000 core with a new variant of REMIX fuel (REgenerated MIXture of U–Pu oxides) are considered during its multiple recycle in a closed nuclear fuel cycle. The fuel composition is produced on the basis of the uranium–plutonium regenerate extracted at processing the spent nuclear fuel (SNF) from a VVER-1000, depleted uranium, and the fissionable material: {sup 235}U as a part of highly enriched uranium (HEU) from warheads superfluous for defense purposes or {sup 233}U accumulated in thorium blankets of fusion (electronuclear) neutron sources or fast reactors. Production of such a fuel assumes no usemore » of natural uranium in addition. When converting a part of the VVER-1000 reactors to the closed fuel cycle based on the REMIX technology, the consumption of natural uranium decreases considerably, and there is no substantial degradation of the isotopic composition of plutonium or change in the reactor-safety characteristics at the passage from recycle to recycle.« less

  14. Microstructure Characterization of RERTR Fuel

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    J. Gan; B. D. Miller; D. D. Keiser

    2008-09-01

    A variety of phases have the potential to develop in the irradiated fuels for the reduced enrichment research test reactor (RERTR) program. To study the radiation stability of these potential phases, three depleted uranium alloys were cast. The phases of interest were identified including U(Si,Al)3, (U,Mo)(Si,Al)3, UMo2Al20, UAl4, and U6Mo4Al43. These alloys were irradiated with 2.6 MeV protons at 200ºC up to 3.0 dpa. The microstructure is characterized using SEM and TEM. Microstructural characterization for an archive dispersion fuel plate (U-7Mo fuel particles in Al-2%Si cladding) was also carried out. TEM sample preparation for the irradiated dispersion fuel has beenmore » developed.« less

  15. Using the Wonder of Inequalities between Averages for Mathematics Problems Solving

    ERIC Educational Resources Information Center

    Shaanan, Rachel Mogilevsky; Gordon, Moshe Stupel

    2016-01-01

    The study presents an introductory idea of using mathematical averages as a tool for enriching mathematical problem solving. Throughout students' activities, a research was conducted on their ability to solve mathematical problems, and how to cope with a variety of mathematical tasks, in a variety of ways, using the skills, tools and experiences…

  16. Influence of test fuel properties and composition on UNECE R101 CO2 and fuel economy valuation

    NASA Astrophysics Data System (ADS)

    Parker, A.

    2015-12-01

    CO2 emission and fuel consumption of passenger cars is now assessed by using a simplistic procedure measuring the emission during a test performed without any control of the fuel properties and computing the fuel consumption through an unsophisticated formula. As pump gasoline and diesel fuels are refinery products mixture of many different hydrocarbons, and in case of gasoline may also contain a significant amount of oxygenates, the fuel properties, including the density, carbon and energy content may strongly vary from one pump fuel to the other. Being the specific test fuels carefully selected by the car manufacturers and everything but randomly chosen pump fuels, the claimed CO2 emission and fuel economy figures may differ largely from the certification values. I show from the analysis of the 2014 UK government data for 2358 diesel and 2103 petrol vehicles how same volumes of only theoretically same pump fuels used during the certification test by the cars manufacturers unfortunately do not produce the same carbon dioxide emission, and very likely do not have the same energy content. The CO2 emission per liter of diesel fuel is shown to oscillate froma maximum of 3049 g to a minimum of 2125 g, with an average of 2625 g, froma +16.13% to a -19.06% of the average. TheCO2 emission per liter of petrol fuel is shown to oscillate even more from a maximum of 3735 g to a minimum of 1767 g with an average of 2327 g, from a +60.48% to a -24.05% of the average. The proposed solution is to center the assessment on the energy demand by measuring with accuracy the mass of fuel consumed and the fuel properties of the test fuel starting from the lower heating. The corrected fuel consumption and the corrected carbon dioxide emission to mention from the test are then computed by using pure hydrocarbon reference fuels for diesel and petrol having a given lower heating value and a given hydrocarbon composition. Alternatively, exactly the same test fuel should be used by all the

  17. Light-duty vehicle greenhouse gas emission standards and corporate average fuel economy standards : final rule

    DOT National Transportation Integrated Search

    2010-05-07

    Final Rule to establish a National Program consisting of new standards for light-duty vehicles that will reduce greenhouse gas emissions and improve fuel economy. This joint : Final Rule is consistent with the National Fuel Efficiency Policy announce...

  18. MSW oxy-enriched incineration technology applied in China: combustion temperature, flue gas loss and economic considerations.

    PubMed

    Fu, Zhe; Zhang, Shihong; Li, Xiangpeng; Shao, Jingai; Wang, Ke; Chen, Hanping

    2015-04-01

    To investigate the application prospect of MSW oxy-enriched incineration technology in China, the technical and economical analyses of a municipal solid waste (MSW) grate furnace with oxy-fuel incineration technology in comparison to co-incineration with coal are performed. The rated capacity of the grate furnace is 350 tonnes MSW per day. When raw MSW is burned, the amount of pure oxygen injected should be about 14.5 wt.% under 25% O2 oxy-fuel combustion conditions with the mode of oxygen supply determined by the actual situation. According to the isothermal combustion temperature (Ta), the combustion effect of 25% O2 oxy-enriched incineration (α = 1.43) is identical with that of MSW co-incineration with 20% mass ratio of coal (α = 1.91). However, the former is better than the latter in terms of plant cost, flue gas loss, and environmental impact. Despite the lower costs of MSW co-incineration with mass ratio of 5% and 10% coal (α = 1.91), 25% O2 oxy-enriched incineration (α = 1.43) is far more advantageous in combustion and pollutant control. Conventional combustion flue gas loss (q2) for co-incineration with 0% coal, 20% coal, 10% coal, 5% coal are around 17%, 13%, 14% and 15%, respectively, while that under the condition of 25% O2 oxy-enriched combustion is approximately 12% (α = 1.43). Clearly, q2 of oxy-enriched incineration is less than other methods under the same combustion conditions. High moisture content presents challenges for MSW incineration, therefore it is necessary to dry MSW prior to incineration, and making oxy-enriched incineration technology achieves higher combustion temperature and lower flue gas loss. In conclusion, based on technical and economical analysis, MSW oxy-enriched incineration retains obvious advantages and demonstrates great future prospects for MSW incineration in China. Copyright © 2015 Elsevier Ltd. All rights reserved.

  19. Microstructural Characterization of the U-9.1Mo Fuel/AA6061 Cladding Interface in Friction-Bonded Monolithic Fuel Plates Irradiated in the RERTR-6 Experiment

    NASA Astrophysics Data System (ADS)

    Keiser, Dennis D.; Jue, Jan-Fong; Miller, Brandon; Gan, Jian; Robinson, Adam; Medvedev, Pavel; Madden, James; Wachs, Dan; Clark, Curtis; Meyer, Mitch

    2015-09-01

    Low-enrichment (235U < 20 pct) U-Mo monolithic fuel is being developed for use in research and test reactors. The earliest design for this fuel that was investigated via reactor testing consisted of a nominally U-10Mo fuel foil encased in AA6061 (Al-6061) cladding. For a fuel design to be deemed adequate for final use in a reactor, it must maintain dimensional stability and retain fission products throughout irradiation, which means that there must be good integrity at the fuel foil/cladding interface. To investigate the nature of the fuel/cladding interface for this fuel type after irradiation, fuel plates were fabricated using a friction bonding process, tested in INL's advanced test reactor (ATR), and then subsequently characterized using optical metallography, scanning electron microscopy, and transmission electron microscopy. Results of this characterization showed that the fuel/cladding interaction layers present at the U-Mo fuel/AA6061 cladding interface after fabrication became amorphous during irradiation. Up to two main interaction layers, based on composition, could be found at the fuel/cladding interface, depending on location. After irradiation, an Al-rich layer contained very few fission gas bubbles, but did exhibit Xe enrichment near the AA6061 cladding interface. Another layer, which contained more Si, had more observable fission gas bubbles. In the samples produced using a focused ion beam at the interaction zone/AA6061 cladding interface, possible indications of porosity/debonding were found, which suggested that the interface in this location is relatively weak.

  20. Feasibility demonstration of a road vehicle fueled with hydrogen-enriched gasoline

    NASA Technical Reports Server (NTRS)

    Hoehn, F. W.; Dowdy, M. W.

    1974-01-01

    Evaluation of the concept of using hydrogen-enriched gasoline in a modified internal combustion engine in order to make possible the burning of ultralean mixtures. The use of such an engine in a road vehicle demonstrated that the addition of small quantities of gaseous hydrogen to gasoline resulted in significant reductions in exhaust emissions of carbon monoxide and nitrogen oxides as well as in thermal efficiency improvements of the engine performance.

  1. DESIGN CRITERIA FOR FUEL DISSOLUTION SYSTEMS AND ASSOCIATED SERVICE FACILITIES. PLANT MODIFICATIONS FOR REPROCESSING NON-PRODUCTION REACTOR FUELS. PROJECT CGC-830

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Bierman, S.R.; Graf, W.A.; Kass, M.

    1960-07-29

    Design panameters are presented for phases of the facility to reprocess low-enrichment fuels from nonproduction reactors. Included are plant flowsheets and equipment layouts for fuel element dissolution, centrifugation, solution adjustment, and waste handling. Also included are the basic design criteria for the supporting facilities which service these phases and all other facilites located in the vicinity of the selected building (Bldg. 221-U). (J.R.D.)

  2. Amazon capims (floating grassmats) - A source of C-13 enriched methane to the troposphere

    NASA Technical Reports Server (NTRS)

    Chanton, Jeffrey; Crill, Patrick; Bartlett, Karen; Martens, Christopher

    1989-01-01

    The C-13 isotopic composition of methane emitted to the troposphere from Amazon capims (floating grassmats) ranged from -36.9 to -48.0, per mil averaging -44.4 + or - 4.2 per mil. All pools of methane associated with the grassmats were enriched; methane withdrawn from plant stems ranged from -39 to -49 per mil while bubbles stirred from the root mat averaged -41.4 per mil. As the CH4 flux from these habitats makes up some 40 percent of the total flux from the Amazon floodplain, methane emissions from the region as a whole must be enriched in.

  3. Alternative Fuels Data Center: Low Rolling Resistance Tires

    Science.gov Websites

    meet their Corporate Average Fuel Economy (CAFE) standards. However, no requirements are currently Rolling Resistance Part 1: Understanding Corporate Average Fuel Economy Definitions 1015 Driving Cycle A

  4. Batch methods for enriching trace impurities in hydrogen gas for their further analysis

    DOEpatents

    Ahmed, Shabbir; Lee, Sheldon H.D.; Kumar, Romesh; Papdias, Dionissios D.

    2014-07-15

    Provided herein are batch methods and devices for enriching trace quantities of impurities in gaseous mixtures, such as hydrogen fuel. The methods and devices rely on concentrating impurities using hydrogen transport membranes wherein the time period for concentrating the sample is calculated on the basis of optimized membrane characteristics, comprising its thickness and permeance, with optimization of temperature, and wherein the enrichment of trace impurities is proportional to the pressure ratio P.sub.hi/P.sub.lo and the volume ratio V.sub.1/V.sub.2, with following detection of the impurities using commonly-available detection methods.

  5. Adhesion and enrichment of metals on human hands from contaminated soil at an Arctic urban brownfield.

    PubMed

    Siciliano, Steven D; James, K; Zhang, Guiyin; Schafer, Alexis N; Peak, J Derek

    2009-08-15

    Human exposure to contaminated soils drives clean up criteria at many urban brownfields. Current risk assessment guidelines assume that humans ingest some fraction of soil smaller than 4 mm but have no estimates of what fraction of soil is ingested by humans. Here, we evaluated soil adherence to human hands for 13 agricultural soils from Saskatchewan, Canada and 17 different soils from a brownfield located in Iqaluit, Nunavut, Canada. In addition, we estimated average particle size adhering to human hands for residents of a northern urban setting. Further, we estimated how metal concentrations differed between the adhered and bulk (< 4 mm) fraction of soil. The average particle size for adhered agricultural soils was 34 microm, adhered brownfield soils was 105 microm, and particles adhered to human residentswas 36 microm. Metals were significantly enriched in these adhered fractions with an average enrichment [(adhered-bulk)/bulk] in metal concentration of 184% (113% median) for 24 different elements. Enrichment was greater for key toxicological elements of concern such as chromium (140%), copper (140%), nickel (130%), lead (110%), and zinc (130%) and was highest for silver (810%), mercury (630%), selenium (500%), and arsenic (420%). Enrichment were positively correlated with carbonate complexation constants (but not bulk solubility products) and suggests that the dominant mechanism controlling metal enrichment in these samples is a precipitation of carbonate surfaces that subsequently adsorb metals. Our results suggest that metals of toxicological concern are selectively enriched in the fraction of soil that humans incidentally ingest. Investigators should likely process soil samples through a 45 microm sieve before estimating the risk associated with contaminated soils to humans. The chemical mechanisms resulting in metal enrichment likely differ between sites but at our site were linked to surface complexation with carbonates.

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

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Dennis D. Keiser, Jr.; Tammy L. Trowbridge; Cynthia R. Breckenridge

    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 betweenmore » 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.« less

  7. Hydrogen Generation Via Fuel Reforming

    NASA Astrophysics Data System (ADS)

    Krebs, John F.

    2003-07-01

    Reforming is the conversion of a hydrocarbon based fuel to a gas mixture that contains hydrogen. The H2 that is produced by reforming can then be used to produce electricity via fuel cells. The realization of H2-based power generation, via reforming, is facilitated by the existence of the liquid fuel and natural gas distribution infrastructures. Coupling these same infrastructures with more portable reforming technology facilitates the realization of fuel cell powered vehicles. The reformer is the first component in a fuel processor. Contaminants in the H2-enriched product stream, such as carbon monoxide (CO) and hydrogen sulfide (H2S), can significantly degrade the performance of current polymer electrolyte membrane fuel cells (PEMFC's). Removal of such contaminants requires extensive processing of the H2-rich product stream prior to utilization by the fuel cell to generate electricity. The remaining components of the fuel processor remove the contaminants in the H2 product stream. For transportation applications the entire fuel processing system must be as small and lightweight as possible to achieve desirable performance requirements. Current efforts at Argonne National Laboratory are focused on catalyst development and reactor engineering of the autothermal processing train for transportation applications.

  8. An aerosol particle containing enriched uranium encountered in the remote upper troposphere.

    PubMed

    Murphy, D M; Froyd, K D; Apel, E; Blake, D; Blake, N; Evangeliou, N; Hornbrook, R S; Peischl, J; Ray, E; Ryerson, T B; Thompson, C; Stohl, A

    2018-04-01

    We describe a submicron aerosol particle sampled at an altitude of 7 km near the Aleutian Islands that contained a small percentage of enriched uranium oxide. 235 U was 3.1 ± 0.5% of 238 U. During twenty years of aircraft sampling of millions of particles in the global atmosphere, we have rarely encountered a particle with a similarly high content of 238 U and never a particle with enriched 235 U. The bulk of the particle consisted of material consistent with combustion of heavy fuel oil. Analysis of wind trajectories and particle dispersion model results show that the particle could have originated from a variety of areas across Asia. The source of such a particle is unclear, and the particle is described here in case it indicates a novel source where enriched uranium was dispersed. Published by Elsevier Ltd.

  9. Fabrication and testing of U–7Mo monolithic plate fuel with Zircaloy cladding

    DOE PAGES

    Pasqualini, E. E.; Robinson, A. B.; Porter, D. L.; ...

    2016-07-15

    The Materials Management and Minimization program is developing fuel designs to replace highly enriched fuel with fuels of low enrichment. In the most challenging cases, U–(7–10wt%)Mo monolithic plate fuel are proposed. The chosen design includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction in service. We investigated zircaloy cladding, specifically Zry–4as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry–4 clad U–7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry–4 and U–(7–10)Mo havemore » similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly between roll passes. Our final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction, either from fabrication or in-reactor testing, and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.54E+21« less

  10. Fabrication and testing of U–7Mo monolithic plate fuel with Zircaloy cladding

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Pasqualini, E. E.; Robinson, A. B.; Porter, D. L.

    The Materials Management and Minimization program is developing fuel designs to replace highly enriched fuel with fuels of low enrichment. In the most challenging cases, U–(7–10wt%)Mo monolithic plate fuel are proposed. The chosen design includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction in service. We investigated zircaloy cladding, specifically Zry–4as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry–4 clad U–7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry–4 and U–(7–10)Mo havemore » similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly between roll passes. Our final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction, either from fabrication or in-reactor testing, and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.54E+21« less

  11. NONDESTRUCTIVE EXAMINATION OF FUEL PLATES FOR THE RERTR FUEL DEVELOPMENT EXPERIMENTS

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    N.E. Woolstenhulme; S.C. Taylor; G.A. Moore

    2012-09-01

    Nuclear fuel is the core component of reactors that is used to produce the neutron flux required for irradiation research purposes as well as commercial power generation. The development of nuclear fuels with low enrichments of uranium is a major endeavor of the RERTR program. In the development of these fuels, the RERTR program uses nondestructive examination (NDE) techniques for the purpose of determining the properties of nuclear fuel plate experiments without imparting damage or altering the fuel specimens before they are irradiated in a reactor. The vast range of properties and information about the fuel plates that can bemore » characterized using NDE makes them highly useful for quality assurance and for analyses used in modeling the behavior of the fuel while undergoing irradiation. NDE is also particularly useful for creating a control group for post-irradiation examination comparison. The two major categories of NDE discussed in this paper are X-ray radiography and ultrasonic testing (UT) inspection/evaluation. The radiographic scans are used for the characterization of fuel meat density and homogeneity as well as the determination of fuel location within the cladding. The UT scans are able to characterize indications such as voids, delaminations, inclusions, and other abnormalities in the fuel plates which are generally referred to as debonds as well as to determine the thickness of the cladding using ultrasonic acoustic microscopy methods. Additionally, the UT techniques are now also being applied to in-canal interim examination of fuel experiments undergoing irradiation and the mapping of the fuel plate surface profile to determine fuel swelling. The methods used to carry out these NDE techniques, as well as how they operate and function, are described along with a description of which properties are characterized.« less

  12. Surplus Highly Enriched Uranium Disposition Program plan

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    NONE

    1996-10-01

    The purpose of this document is to provide upper level guidance for the program that will downblend surplus highly enriched uranium for use as commercial nuclear reactor fuel or low-level radioactive waste. The intent of this document is to outline the overall mission and program objectives. The document is also intended to provide a general basis for integration of disposition efforts among all applicable sites. This plan provides background information, establishes the scope of disposition activities, provides an approach to the mission and objectives, identifies programmatic assumptions, defines major roles, provides summary level schedules and milestones, and addresses budget requirements.

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

    DOE Office of Scientific and Technical Information (OSTI.GOV)

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

    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 tomore » 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.« less

  14. Temperature sensitivity study of eddy current and digital gauge probes for nuclear fuel rod oxide measurement

    NASA Astrophysics Data System (ADS)

    Beck, Faith R.; Lind, R. Paul; Smith, James A.

    2018-04-01

    Novel fuels are part of the nationwide effort to reduce the enrichment of Uranium for energy production. Performance of such fuels is determined by irradiating their surfaces. To test irradiated samples, the instrumentation must operate remotely. The plate checker used in this experiment at Idaho National Lab (INL) performs non-destructive testing on fuel rod and plate geometries with two different types of sensors: eddy current and digital thickness gauges. The sensors measure oxide growth and total sample thickness on research fuels, respectively. Sensor measurement accuracy is crucial because even 10 microns of error is significant when determining the viability of an experimental fuel. One parameter known to affect the eddy current and thickness gauge sensors is temperature. Since both sensor accuracies depend on the ambient temperature of the system, the plate checker has been characterized for these sensitivities. The manufacturer of the digital gauge probes has noted a rather large coefficient of thermal expansion for their linear scale. It should also be noted that the accuracy of the digital gauge probes are specified at 20°C, which is approximately 7°C cooler than the average hot-cell temperature. In this work, the effect of temperature on the eddy current and digital gauge probes is studied, and thickness measurements are given as empirical functions of temperature.

  15. Mandated fuel economy standards as a strategy for improving motor vehicle fuel economy.

    DOT National Transportation Integrated Search

    1978-10-19

    The major domestic motor vehicle manufacturers have projected that their new car fleet average fuel economy will meet the federal mandated fuel economy standard for 1985, of 27.5 miles per gallon. Assuming that these projections hold true, in one dec...

  16. Effect of reactor radiation on the thermal conductivity of TREAT fuel

    NASA Astrophysics Data System (ADS)

    Mo, Kun; Miao, Yinbin; Kontogeorgakos, Dimitrios C.; Connaway, Heather M.; Wright, Arthur E.; Yacout, Abdellatif M.

    2017-04-01

    The Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory is resuming operations after more than 20 years in latency in order to produce high-neutron-flux transients for investigating transient-induced behavior of reactor fuels and their interactions with other materials and structures. A parallel program is ongoing to develop a replacement core in which the fuel, historically containing highly-enriched uranium (HEU), is replaced by low-enriched uranium (LEU). Both the HEU and prospective LEU fuels are in the form of UO2 particles dispersed in a graphite matrix, but the LEU fuel will contain a much higher volume of UO2 particles, which may create a larger area of interphase boundaries between the particles and the graphite. This may lead to a higher volume fraction of graphite exposed to the fission fragments escaping from the UO2 particles, and thus may induce a higher volume of fission-fragment damage on the fuel graphite. In this work, we analyzed the reactor-radiation induced thermal conductivity degradation of graphite-based dispersion fuel. A semi-empirical method to model the relative thermal conductivity with reactor radiation was proposed and validated based on the available experimental data. Prediction of thermal conductivity degradation of LEU TREAT fuel during a long-term operation was performed, with a focus on the effect of UO2 particle size on fission-fragment damage. The proposed method can be further adjusted to evaluate the degradation of other properties of graphite-based dispersion fuel.

  17. THE CALCULATION OF BURNABLE POISON CORRECTION FACTORS FOR PWR FRESH FUEL ACTIVE COLLAR MEASUREMENTS

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Croft, Stephen; Favalli, Andrea; Swinhoe, Martyn T.

    2012-06-19

    Verification of commercial low enriched uranium light water reactor fuel takes place at the fuel fabrication facility as part of the overall international nuclear safeguards solution to the civilian use of nuclear technology. The fissile mass per unit length is determined nondestructively by active neutron coincidence counting using a neutron collar. A collar comprises four slabs of high density polyethylene that surround the assembly. Three of the slabs contain {sup 3}He filled proportional counters to detect time correlated fission neutrons induced by an AmLi source placed in the fourth slab. Historically, the response of a particular collar design to amore » particular fuel assembly type has been established by careful cross-calibration to experimental absolute calibrations. Traceability exists to sources and materials held at Los Alamos National Laboratory for over 35 years. This simple yet powerful approach has ensured consistency of application. Since the 1980's there has been a steady improvement in fuel performance. The trend has been to higher burn up. This requires the use of both higher initial enrichment and greater concentrations of burnable poisons. The original analytical relationships to correct for varying fuel composition are consequently being challenged because the experimental basis for them made use of fuels of lower enrichment and lower poison content than is in use today and is envisioned for use in the near term. Thus a reassessment of the correction factors is needed. Experimental reassessment is expensive and time consuming given the great variation between fuel assemblies in circulation. Fortunately current modeling methods enable relative response functions to be calculated with high accuracy. Hence modeling provides a more convenient and cost effective means to derive correction factors which are fit for purpose with confidence. In this work we use the Monte Carlo code MCNPX with neutron coincidence tallies to calculate the influence

  18. THE MANUFACTURE OF FUEL ELEMENTS OF THE ARGONAUT TYPE

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Kittl, J.; Machado, R.E.; Mazza, J.A.

    1958-06-10

    The conditions required for the manufacture of the RA-1 Argonant type fuel elements are investigated. The fuel elements are in the form of a plate which is manufactured by the extrusion of a presintered mass of U/sub 3/O/sub 8/ (20% enriched) in an aluminum matrix. Steps in the investigation were obtention and specification of U/sub 3/O/sub 8/ and Al in powder form for testing, filling, and extrusion tests, finishing of the fuel elements, and computation of U/sub 3/O/sub 8/ content. (W.D.M.)

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

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    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 describedmore » as accurately as possible, given the current sources of data.« less

  20. Alternative Fuels Data Center

    Science.gov Websites

    Tax An excise tax rate of 9% of the average wholesale price on a per gallon basis applies to all . Additionally, a highway motor fuel tax of $0.02 per gallon applies to all special fuels. For taxation purposes

  1. 49 CFR 535.7 - Averaging, banking, and trading (ABT) program.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... averaging set. With the exception of FCC earned for advance technologies as further clarified below, a... transactions. Traded FCC, other than advanced technology credits, may be used by a manufacturer only within the... fleet includes conventional vehicles (gasoline, diesel and alternative fuel) and advanced technology...

  2. 49 CFR 535.7 - Averaging, banking, and trading (ABT) program.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... averaging set. With the exception of FCC earned for advance technologies as further clarified below, a... transactions. Traded FCC, other than advanced technology credits, may be used by a manufacturer only within the... fleet includes conventional vehicles (gasoline, diesel and alternative fuel) and advanced technology...

  3. Comparative analysis of LWR and FBR spent fuels for nuclear forensics evaluation

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Permana, Sidik; Suzuki, Mitsutoshi; Su'ud, Zaki

    2012-06-06

    Some interesting issues are attributed to nuclide compositions of spent fuels from thermal reactors as well as fast reactors such as a potential to reuse as recycled fuel, and a possible capability to be manage as a fuel for destructive devices. In addition, analysis on nuclear forensics which is related to spent fuel compositions becomes one of the interesting topics to evaluate the origin and the composition of spent fuels from the spent fuel foot-prints. Spent fuel compositions of different fuel types give some typical spent fuel foot prints and can be estimated the origin of source of those spentmore » fuel compositions. Some technics or methods have been developing based on some science and technological capability including experimental and modeling or theoretical aspects of analyses. Some foot-print of nuclear forensics will identify the typical information of spent fuel compositions such as enrichment information, burnup or irradiation time, reactor types as well as the cooling time which is related to the age of spent fuels. This paper intends to evaluate the typical spent fuel compositions of light water (LWR) and fast breeder reactors (FBR) from the view point of some foot prints of nuclear forensics. An established depletion code of ORIGEN is adopted to analyze LWR spent fuel (SF) for several burnup constants and decay times. For analyzing some spent fuel compositions of FBR, some coupling codes such as SLAROM code, JOINT and CITATION codes including JFS-3-J-3.2R as nuclear data library have been adopted. Enriched U-235 fuel composition of oxide type is used for fresh fuel of LWR and a mixed oxide fuel (MOX) for FBR fresh fuel. Those MOX fuels of FBR come from the spent fuels of LWR. Some typical spent fuels from both LWR and FBR will be compared to distinguish some typical foot-prints of SF based on nuclear forensic analysis.« less

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

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    J. L. Dolan; M. J. Marcath; M. Flaska

    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, amore » 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

  5. The slightly-enriched spectral shift control reactor. Final report, September 30, 1988--September 30, 1991

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Martin, W.R.; Lee, J.C.; Larsen, E.W.

    1991-11-01

    An advanced converter reactor design utilizing mechanical spectral shift control rods in a conventional pressurized water reactor configuration is under investigation. The design is based on the principle that a harder spectrum during the early part of the fuel cycle will result in large neutron captures in fertile {sup 238}U, which can then be burned in situ in a softer spectrum later in the cycle. Preliminary design calculations performed during FY 89 showed that the slightly-enriched spectral shift reactor design offers the benefit of substantially increased fuel resource utilization with the proven safety characteristics of the pressurized water reactor technologymore » retained. Optimization of the fuel design and development of fuel management strategies were carried out in FY 90, along with effort to develop and validate neutronic methodology for tight-lattice configurations with hard spectra. During FY 91, the final year of the grant, the final Slightly-Enriched Spectral Shift Reactor (SESSR) design was determined, and reference design analyses were performed for the assemblies as well as the global core configuration, both at the beginning of cycle (BOC) and with depletion. The final SESSR design results in approximately a 20% increase in the utilization of uranium resources, based on equilibrium fuel cycle analyses. Acceptable pin power peaking is obtained with the final core design, with assembly peaking factors equal to less than 1.04 for spectral shift control rods both inserted and withdrawn, and global peaking factors at BOC predicted to be 1.4. In addition, a negative Moderation Temperature Coefficient (MTC) is maintained for BOC, which is difficult to achieve with conventional advanced converter designs based on a closed fuel cycle. The SESSR design avoids the need for burnable poison absorber, although they could be added if desired to increase the cycle length while maintaining a negative MTC.« less

  6. Fuel Economy and Emissions of a Vehicle Equipped with an Aftermarket Flexible-Fuel Conversion Kit

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Thomas, John F; Huff, Shean P; West, Brian H

    2012-04-01

    The U.S. Environmental Protection Agency (EPA) grants Certificates of Conformity for alternative fuel conversion systems and also offers other forms of premarket registration of conversion kits for use in vehicles more than two model years old. Use of alternative fuels such as ethanol, natural gas, and propane are encouraged by the Energy Policy Act of 1992. Several original equipment manufacturers (OEMs) produce emissions-certified vehicles capable of using alternative fuels, and several alternative fuel conversion system manufacturers produce EPA-approved conversion systems for a variety of alternative fuels and vehicle types. To date, only one manufacturer (Flex Fuel U.S.) has received EPAmore » certifications for ethanol fuel (E85) conversion kits. This report details an independent evaluation of a vehicle with a legal installation of a Flex Fuel U.S. conversion kit. A 2006 Dodge Charger was baseline tested with ethanol-free certification gasoline (E0) and E20 (gasoline with 20 vol % ethanol), converted to flex-fuel operation via installation of a Flex Box Smart Kit from Flex Fuel U.S., and retested with E0, E20, E50, and E81. Test cycles included the Federal Test Procedure (FTP or city cycle), the highway fuel economy test (HFET), and the US06 test (aggressive driving test). Averaged test results show that the vehicle was emissions compliant on E0 in the OEM condition (before conversion) and compliant on all test fuels after conversion. Average nitrogen oxide (NOx) emissions exceeded the Tier 2/Bin 5 intermediate life NO{sub X} standard with E20 fuel in the OEM condition due to two of three test results exceeding this standard [note that E20 is not a legal fuel for non-flexible-fuel vehicles (non-FFVs)]. In addition, one E0 test result before conversion and one E20 test result after conversion exceeded the NOX standard, although the average result in these two cases was below the standard. Emissions of ethanol and acetaldehyde increased with increasing

  7. Bi-fuel System - Gasoline/LPG in A Used 4-Stroke Motorcycle - Fuel Injection Type

    NASA Astrophysics Data System (ADS)

    Suthisripok, Tongchit; Phusakol, Nachaphat; Sawetkittirut, Nuttapol

    2017-10-01

    Bi-fuel-Gasoline/LPG system has been effectively and efficiently used in gasoline vehicles with less pollutants emission. The motorcycle tested was a used Honda AirBlade i110 - fuel injection type. A 3-litre LPG storage tank, an electronic fuel control unit, a 1-mm LPG injector and a regulator were securely installed. The converted motorcycle can be started with either gasoline or LPG. The safety relief valve was set below 48 kPa and over 110 kPa. The motorcycle was tuned at the relative rich air-fuel ratio (λ) of 0.85-0.90 to attain the best power output. From dynamometer tests over the speed range of 65-100 km/h, the average power output when fuelling LPG was 5.16 hp; dropped 3.9% from the use of gasoline91. The average LPG consumption rate from the city road test at the average speed of 60 km/h was 40.1 km/l, about 17.7% more. This corresponded to lower LPG’s energy density of about 16.2%. In emission, the CO and HC concentrations were 44.4% and 26.5% lower. Once a standard gas equipment set with ECU and LPG injector were securely installed and the engine was properly tuned up to suit LPG’s characteristics, the converted bi-fuel motorcycle offers efficiently, safely and economically performance with environmental friendly emission.

  8. Trends of metals enrichment in deposited particulate matter at semi-arid area of Iran.

    PubMed

    Fouladi Fard, Reza; Naddafi, Kazem; Hassanvand, Mohammad Sadegh; Khazaei, Mohammad; Rahmani, Farah

    2018-04-30

    The presence and enrichment of heavy metals in dust depositions have been recognized as an emerging environmental health issues in the urban and industrial areas. In this study, the deposition of some metals was found in Qom, a city located in a semi-desert area in Iran that is surrounded by industrial areas. Dust deposition samples were collected using five sampling stations during a year. Dust samples were digested applying acidic condition and then, the metal content was analyzed using inductively coupled plasma technology (ICP-OES). Comparative results showed the following order, from the maximum to the minimum concentration (mg/kg dust) of elements: Ca > Al > Fe > Mg > Ti > Si > K > B > Sr > Mn > P > Ba > Cr > Zn > Ni > Sn > Pb > V > Na > Cu > Co > U > Li > Ce > Ag. The differences among the average concentrations of metals in the five stations were not significant (p value > 0.05). The average concentration of some metals increased significantly during cold seasons. In this study, the cluster analysis (CA) and princicipal component analysis (PCA) were applied, and relationships among some elements in different clusters were found. In addition, the geo-accumulation and enrichment analysis revealed that the following metals had been enriched more than the average values: boron, silver, tin, uranium, lead, zinc, cobalt, chromium, lithium, nickel, strontium, and coper. The presence of thermal power plant, pesticide manufacturing plants, publishing centers, traffic jam, and some industrial areas around the city has resulted in the enrichment of some metals (particularly in cold seasons with atmospheric stable conditions) in dust deposition.

  9. Pyroprocessing of Fast Flux Test Facility Nuclear Fuel

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    B.R. Westphal; G.L. Fredrickson; G.G. Galbreth

    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 primarymore » 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%.« less

  10. Pyroprocessing of fast flux test facility nuclear fuel

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Westphal, B.R.; Wurth, L.A.; Fredrickson, G.L.

    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 primarymore » 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)« less

  11. Development of ORIGEN Libraries for Mixed Oxide (MOX) Fuel Assembly Designs

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Mertyurek, Ugur; Gauld, Ian C.

    In this research, ORIGEN cross section libraries for reactor-grade mixed oxide (MOX) fuel assembly designs have been developed to provide fast and accurate depletion calculations to predict nuclide inventories, radiation sources and thermal decay heat information needed in safety evaluations and safeguards verification measurements of spent nuclear fuel. These ORIGEN libraries are generated using two-dimensional lattice physics assembly models that include enrichment zoning and cross section data based on ENDF/B-VII.0 evaluations. Using the SCALE depletion sequence, burnup-dependent cross sections are created for selected commercial reactor assembly designs and a representative range of reactor operating conditions, fuel enrichments, and fuel burnup.more » The burnup dependent cross sections are then interpolated to provide problem-dependent cross sections for ORIGEN, avoiding the need for time-consuming lattice physics calculations. The ORIGEN libraries for MOX assembly designs are validated against destructive radiochemical assay measurements of MOX fuel from the MALIBU international experimental program. This program included measurements of MOX fuel from a 15 × 15 pressurized water reactor assembly and a 9 × 9 boiling water reactor assembly. The ORIGEN MOX libraries are also compared against detailed assembly calculations from the Phase IV-B numerical MOX fuel burnup credit benchmark coordinated by the Nuclear Energy Agency within the Organization for Economic Cooperation and Development. Finally, the nuclide compositions calculated by ORIGEN using the MOX libraries are shown to be in good agreement with other physics codes and with experimental data.« less

  12. Development of ORIGEN Libraries for Mixed Oxide (MOX) Fuel Assembly Designs

    DOE PAGES

    Mertyurek, Ugur; Gauld, Ian C.

    2015-12-24

    In this research, ORIGEN cross section libraries for reactor-grade mixed oxide (MOX) fuel assembly designs have been developed to provide fast and accurate depletion calculations to predict nuclide inventories, radiation sources and thermal decay heat information needed in safety evaluations and safeguards verification measurements of spent nuclear fuel. These ORIGEN libraries are generated using two-dimensional lattice physics assembly models that include enrichment zoning and cross section data based on ENDF/B-VII.0 evaluations. Using the SCALE depletion sequence, burnup-dependent cross sections are created for selected commercial reactor assembly designs and a representative range of reactor operating conditions, fuel enrichments, and fuel burnup.more » The burnup dependent cross sections are then interpolated to provide problem-dependent cross sections for ORIGEN, avoiding the need for time-consuming lattice physics calculations. The ORIGEN libraries for MOX assembly designs are validated against destructive radiochemical assay measurements of MOX fuel from the MALIBU international experimental program. This program included measurements of MOX fuel from a 15 × 15 pressurized water reactor assembly and a 9 × 9 boiling water reactor assembly. The ORIGEN MOX libraries are also compared against detailed assembly calculations from the Phase IV-B numerical MOX fuel burnup credit benchmark coordinated by the Nuclear Energy Agency within the Organization for Economic Cooperation and Development. Finally, the nuclide compositions calculated by ORIGEN using the MOX libraries are shown to be in good agreement with other physics codes and with experimental data.« less

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

    NASA Astrophysics Data System (ADS)

    Kim, Yeon Soo; Hofman, G. L.

    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. Multivariate analysis of gamma spectra to characterize used nuclear fuel

    DOE PAGES

    Coble, Jamie; Orton, Christopher; Schwantes, Jon

    2017-01-17

    The Multi-Isotope Process (MIP) Monitor provides an efficient means to monitor the process conditions in used nuclear fuel reprocessing facilities to support process verification and validation. The MIP Monitor applies multivariate analysis to gamma spectroscopy of key stages in the reprocessing stream in order to detect small changes in the gamma spectrum, which may indicate changes in process conditions. This research extends the MIP Monitor by characterizing a used fuel sample after initial dissolution according to the type of reactor of origin (pressurized or boiling water reactor; PWR and BWR, respectively), initial enrichment, burn up, and cooling time. Simulated gammamore » spectra were used in this paper to develop and test three fuel characterization algorithms. The classification and estimation models employed are based on the partial least squares regression (PLS) algorithm. A PLS discriminate analysis model was developed which perfectly classified reactor type for the three PWR and three BWR reactor designs studied. Locally weighted PLS models were fitted on-the-fly to estimate the remaining fuel characteristics. For the simulated gamma spectra considered, burn up was predicted with 0.1% root mean squared percent error (RMSPE) and both cooling time and initial enrichment with approximately 2% RMSPE. Finally, this approach to automated fuel characterization can be used to independently verify operator declarations of used fuel characteristics and to inform the MIP Monitor anomaly detection routines at later stages of the fuel reprocessing stream to improve sensitivity to changes in operational parameters that may indicate issues with operational control or malicious activities.« less

  15. Multivariate analysis of gamma spectra to characterize used nuclear fuel

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Coble, Jamie; Orton, Christopher; Schwantes, Jon

    The Multi-Isotope Process (MIP) Monitor provides an efficient means to monitor the process conditions in used nuclear fuel reprocessing facilities to support process verification and validation. The MIP Monitor applies multivariate analysis to gamma spectroscopy of key stages in the reprocessing stream in order to detect small changes in the gamma spectrum, which may indicate changes in process conditions. This research extends the MIP Monitor by characterizing a used fuel sample after initial dissolution according to the type of reactor of origin (pressurized or boiling water reactor; PWR and BWR, respectively), initial enrichment, burn up, and cooling time. Simulated gammamore » spectra were used in this paper to develop and test three fuel characterization algorithms. The classification and estimation models employed are based on the partial least squares regression (PLS) algorithm. A PLS discriminate analysis model was developed which perfectly classified reactor type for the three PWR and three BWR reactor designs studied. Locally weighted PLS models were fitted on-the-fly to estimate the remaining fuel characteristics. For the simulated gamma spectra considered, burn up was predicted with 0.1% root mean squared percent error (RMSPE) and both cooling time and initial enrichment with approximately 2% RMSPE. Finally, this approach to automated fuel characterization can be used to independently verify operator declarations of used fuel characteristics and to inform the MIP Monitor anomaly detection routines at later stages of the fuel reprocessing stream to improve sensitivity to changes in operational parameters that may indicate issues with operational control or malicious activities.« less

  16. Accelerator-driven transmutation of spent fuel elements

    DOEpatents

    Venneri, Francesco; Williamson, Mark A.; Li, Ning

    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

  17. Effect of reactor radiation on the thermal conductivity of TREAT fuel

    DOE PAGES

    Mo, Kun; Miao, Yinbin; Kontogeorgakos, Dimitrios C.; ...

    2017-02-04

    The Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory is resuming operations after more than 20 years in latency in order to produce high-neutron-flux transients for investigating transient-induced behavior of reactor fuels and their interactions with other materials and structures. A parallel program is ongoing to develop a replacement core in which the fuel, historically containing highly-enriched uranium (HEU), is replaced by low-enriched uranium (LEU). Both the HEU and prospective LEU fuels are in the form of UO 2 particles dispersed in a graphite matrix, but the LEU fuel will contain a much higher volume of UO 2more » particles, which may create a larger area of interphase boundaries between the particles and the graphite. This may lead to a higher volume fraction of graphite exposed to the fission fragments escaping from the UO 2 particles, and thus may induce a higher volume of fission-fragment damage on the fuel graphite. In this work, we analyzed the reactor-radiation induced thermal conductivity degradation of graphite-based dispersion fuel. A semi-empirical method to model the relative thermal conductivity with reactor radiation was proposed and validated based on the available experimental data. Prediction of thermal conductivity degradation of LEU TREAT fuel during a long-term operation was performed, with a focus on the effect of UO 2 particle size on fission-fragment damage. Lastly, the proposed method can be further adjusted to evaluate the degradation of other properties of graphite-based dispersion fuel.« less

  18. Effect of reactor radiation on the thermal conductivity of TREAT fuel

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Mo, Kun; Miao, Yinbin; Kontogeorgakos, Dimitrios C.

    The Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory is resuming operations after more than 20 years in latency in order to produce high-neutron-flux transients for investigating transient-induced behavior of reactor fuels and their interactions with other materials and structures. A parallel program is ongoing to develop a replacement core in which the fuel, historically containing highly-enriched uranium (HEU), is replaced by low-enriched uranium (LEU). Both the HEU and prospective LEU fuels are in the form of UO 2 particles dispersed in a graphite matrix, but the LEU fuel will contain a much higher volume of UO 2more » particles, which may create a larger area of interphase boundaries between the particles and the graphite. This may lead to a higher volume fraction of graphite exposed to the fission fragments escaping from the UO 2 particles, and thus may induce a higher volume of fission-fragment damage on the fuel graphite. In this work, we analyzed the reactor-radiation induced thermal conductivity degradation of graphite-based dispersion fuel. A semi-empirical method to model the relative thermal conductivity with reactor radiation was proposed and validated based on the available experimental data. Prediction of thermal conductivity degradation of LEU TREAT fuel during a long-term operation was performed, with a focus on the effect of UO 2 particle size on fission-fragment damage. Lastly, the proposed method can be further adjusted to evaluate the degradation of other properties of graphite-based dispersion fuel.« less

  19. Juvenile psittacine environmental enrichment.

    PubMed

    Simone-Freilicher, Elisabeth; Rupley, Agnes E

    2015-05-01

    Environmental enrichment is of great import to the emotional, intellectual, and physical development of the juvenile psittacine and their success in the human home environment. Five major types of enrichment include social, occupational, physical, sensory, and nutritional. Occupational enrichment includes exercise and psychological enrichment. Physical enrichment includes the cage and accessories and the external home environment. Sensory enrichment may be visual, auditory, tactile, olfactory, or taste oriented. Nutritional enrichment includes variations in appearance, type, and frequency of diet, and treats, novelty, and foraging. Two phases of the preadult period deserve special enrichment considerations: the development of autonomy and puberty. Copyright © 2015 Elsevier Inc. All rights reserved.

  20. Managing the Nuclear Fuel Cycle: Policy Implications of Expanding Global Access to Nuclear Power

    DTIC Science & Technology

    2007-11-01

    critical aspect of the nuclear fuel cycle for the United States, where longstanding nonproliferation policy discouraged commercial nuclear fuel...perhaps the most critical question in this decade for strengthening the nuclear nonproliferation regime: how can access to sensitive fuel cycle...process can take advantage of the slight difference in atomic mass between 235U and 238U. The typical enrichment process requires about 10 lbs of uranium

  1. New potentials for conventional aircraft when powered by hydrogen-enriched gasoline

    NASA Technical Reports Server (NTRS)

    Menard, W. A.; Moynihan, P. I.; Rupe, J. H.

    1976-01-01

    Overall system efficiency and performance of a Beech Model 20 Duke aircraft was studied to provide analytical representations of an aircraft piston engine system, including all essential components required for onboard hydrogen generation. Lower emission levels and a 20% reduction in fuel consumption may be obtained by using a catalytic hydrogen generator, incorporated as part of the air induction system, to generate hydrogen by breaking down small amounts of the aviation gasoline used in the normal propulsion system. This hydrogen is then mixed with gasoline and compressed air from the turbocharger before entering the engine combustion chamber. The special properties of the hydrogen-enriched gasoline allow the engine to operate at ultra lean fuel/air ratios, resulting in higher efficiencies.

  2. U.S. Commercial Spent Nuclear Fuel Assembly Characteristics - 1968-2013

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Hu, Jianwei; Peterson, Joshua L.; Gauld, Ian C.

    2016-09-01

    Activities related to management of spent nuclear fuel (SNF) are increasing in the US and many other countries. Over 240,000 SNF assemblies have been discharged from US commercial reactors since the late 1960s. The enrichment and burnup of SNF have changed significantly over the past 40 years, and fuel assembly designs have also evolved. Understanding the general characteristics of SNF helps regulators and other stakeholders form overall strategies towards the final disposal of US SNF. This report documents a survey of all US commercial SNF assemblies in the GC-859 database and provides reference SNF source terms (e.g., nuclide inventories, decaymore » heat, and neutron/photon emission) at various cooling times up to 200 years after fuel discharge. This study reviews the distribution and evolution of fuel parameters of all SNF assemblies discharged over the past 40 years. Assemblies were categorized into three groups based on discharge year, and the median burnups and enrichments of each group were used to establish representative cases. An extended burnup case was created for boiling water reactor (BWR) fuels, and another was created for the pressurized water reactor (PWR) fuels. Two additional cases were developed to represent the eight mixed oxide (MOX) fuel assemblies in the database. Burnup calculations were performed for each representative case. Realistic parameters for fuel design and operations were used to model the SNF and to provide reference fuel characteristics representative of the current inventory. Burnup calculations were performed using the ORIGEN code, which is part of the SCALE nuclear modeling and simulation code system. Results include total activity, decay heat, photon emission, neutron flux, gamma heat, and plutonium content, as well as concentrations for 115 significant nuclides. These quantities are important in the design, regulation, and operations of SNF storage, transportation, and disposal systems.« less

  3. Effect of Fuel Temperature Profile on Eigenvalue Calculations

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Greifenkamp, Tom E; Clarno, Kevin T; Gehin, Jess C

    2008-01-01

    Use of an average fuel temperature is a current practice when modeling fuel for eigenvalue (k-inf) calculations. This is an approximation, as it is known from Heat-transfer methods that a fuel pin having linear power q', will have a temperature that varies radially and has a maximum temperature at the center line [1]. This paper describes an investigation into the effects on k-inf and isotopic concentrations of modeling a fuel pin using a single average temperature versus a radially varying fuel temperature profile. The axial variation is not discussed in this paper. A single fuel pin was modeled having 1,more » 3, 5, 8, or 10 regions of equal volumes (areas). Fig. 1 shows a model of a 10-ring fuel pin surrounded by a gap and then cladding.« less

  4. Assess How Changes in Fuel Cycle Operation Impact Safeguards

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Tobin, Stephen Joseph; Adigun, Babatunde John; Fugate, Michael Lynn

    Since the beginning of commercial nuclear power generation in the 1960s, the ability of researchers to understand and control the isotopic content of spent fuel has improved. It is therefore not surprising that both fuel assembly design and fuel assembly irradiation optimization have improved over the past 50+ years. It is anticipated that the burnup and isotopics of the spent fuel should exhibit less variation over the decades as reactor operators irradiate each assembly to the optimum amount. In contrast, older spent fuel is anticipated to vary more in burnup and resulting isotopics for a given initial enrichment. Modern fuelmore » therefore should be more uniform in composition, and thus, measured safeguards results should be easier to interpret than results from older spent fuel. With spent fuel ponds filling up, interim and long-­term storage of spent fuel will need to be addressed. Additionally after long periods of storage, spent fuel is no longer self-­protecting and, as such, the IAEA will categorize it as more attractive; in approximately 20 years many of the assemblies from early commercial cores will no longer be considered self-­protecting. This study will assess how more recent changes in the reactor operation could impact the interpretation of safeguards measurements. The status quo for spent fuel assay in the safeguards context is that the overwhelming majority of spent fuel assemblies are not measured in a quantitative way except for those assemblies about to be loaded into a difficult or impossible to access location (dry storage or, in the future, a repository). In other words, when the assembly is still accessible to a state actor, or an insider, when it is cooling in a pool, the inspectorate does not have a measurement database that could assist them in re-­verifying the integrity of that assembly. The spent fuel safeguards regime would be strengthened if spent fuel assemblies were measured from discharge to loading into a difficult or

  5. Heavy Metal Enrichment in laminated lake sediments from N-Germany and N-Poland: Geochemical background, enrichment history and land surface changes

    NASA Astrophysics Data System (ADS)

    Hoelzmann, Philipp; Brauer, Achim; Dräger, Nadine; Kienel, Ulrike; Obremska, Milena; Ott, Florian; Słowinski, Michał

    2017-04-01

    For three lake sediment records, situated in rural environments in NE-Germany (Lake Tiefer See) and N-Poland (Lake Czechowskie, Lake Głęboczek), we present a detailed heavy metal enrichment history with sub-decadal resolution for the last 200 years. We determine the local and specific geogenic background values on the base of heavy-metal analysis of pre-industrial sediments and different sediment types (e.g. calcareous gyttja, organic gyttja etc.). These results provide means to calculate and quantify anthropogenic heavy metal accumulations and enrichment factors as well as to define regional measures for a state of reference, reflecting natural conditions without human impact. All three lakes show a similar pattern of relatively low heavy metal concentrations and only Pb, Zn and Cd show a clear parallel pattern of enrichment starting around 1850. This heavy metal enrichment mainly results from atmospheric input due to increasing industrialization within the framework of the Industrial Revolution. Highest concentrations of Cd, Zn, and Pb occur around 1960 to 1980 and thereafter a clear pattern of declining anthropogenic input is registered. This data is supplemented by calculations of mass accumulation rates to determine heavy metal input to the lakes for the past 200 years. For Lake Czechowskie the heavy metal input to the lake is compared to an on average five year resolved pollen record that reflects changes in land use and vegetation.

  6. Two-Dimensional Mapping of the Calculated Fission Power for the Full-Size Fuel Plate Experiment Irradiated in the Advanced Test Reactor

    NASA Astrophysics Data System (ADS)

    Chang, G. S.; Lillo, M. A.

    2009-08-01

    The National Nuclear Security Administrations (NNSA) Reduced Enrichment for Research and Test Reactors (RERTR) program assigned to the Idaho National Laboratory (INL) the responsibility of developing and demonstrating high uranium density research reactor fuel forms to enable the use of low enriched uranium (LEU) in research and test reactors around the world. A series of full-size fuel plate experiments have been proposed for irradiation testing in the center flux trap (CFT) position of the Advanced Test Reactor (ATR). These full-size fuel plate tests are designated as the AFIP tests. The AFIP nominal fuel zone is rectangular in shape having a designed length of 21.5-in (54.61-cm), width of 1.6-in (4.064-cm), and uniform thickness of 0.014-in (0.03556-cm). This gives a nominal fuel zone volume of 0.482 in3 (7.89 cm3) per fuel plate. The AFIP test assembly has two test positions. Each test position is designed to hold 2 full-size plates, for a total of 4 full-size plates per test assembly. The AFIP test plates will be irradiated at a peak surface heat flux of about 350 W/cm2 and discharged at a peak U-235 burn-up of about 70 at.%. Based on limited irradiation testing of the monolithic (U-10Mo) fuel form, it is desirable to keep the peak fuel temperature below 250°C to achieve this, it will be necessary to keep plate heat fluxes below 500 W/cm2. Due to the heavy U-235 loading and a plate width of 1.6-in (4.064-cm), the neutron self-shielding will increase the local-to-average-ratio (L2AR) fission power near the sides of the fuel plates. To demonstrate that the AFIP experiment will meet the ATR safety requirements, a very detailed 2-dimensional (2D) Y-Z fission power profile was evaluated in order to best predict the fuel plate temperature distribution. The ability to accurately predict fuel plate power and burnup are essential to both the design of the AFIP tests as well as evaluation of the irradiated fuel performance. To support this need, a detailed MCNP Y

  7. Pre-conceptual Development and characterization of an extruded graphite composite fuel for the TREAT Reactor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Luther, Erik; Rooyen, Isabella van; Leckie, Rafael

    2015-03-01

    In an effort to explore fuel systems that are more robust under accident scenarios, the DOE-NE has identified the need to resume transient testing. The Transient Reactor Test (TREAT) facility has been identified as the preferred option for the resumption of transient testing of nuclear fuel in the United States. In parallel, NNSA’s Global Threat Reduction Initiative (GTRI) Convert program is exploring the needs to replace the existing highly enriched uranium (HEU) core with low enriched uranium (LEU) core. In order to construct a new LEU core, materials and fabrication processes similar to those used in the initial core fabricationmore » must be identified, developed and characterized. In this research, graphite matrix fuel blocks were extruded and materials properties of were measured. Initially the extrusion process followed the historic route; however, the project was expanded to explore methods to increase the graphite content of the fuel blocks and explore modern resins. Materials properties relevant to fuel performance including density, heat capacity and thermal diffusivity were measured. The relationship between process defects and materials properties will be discussed.« less

  8. 77 FR 33253 - Regulatory Guide 8.24, Revision 2, Health Physics Surveys During Enriched Uranium-235 Processing...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-06-05

    ... NUCLEAR REGULATORY COMMISSION [NRC-2010-0115] Regulatory Guide 8.24, Revision 2, Health Physics..., ``Health Physics Surveys During Enriched Uranium-235 Processing and Fuel Fabrication'' was issued with a... specifically with the following aspects of an acceptable occupational health physics program that are closely...

  9. Model Year 2013 Fuel Economy Guide: EPA Fuel Economy Estimates

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    None

    2012-12-01

    The Fuel Economy Guide is published by the U.S. Department of Energy as an aid to consumers considering the purchase of a new vehicle. The Guide lists estimates of miles per gallon (mpg) for each vehicle available for the new model year. These estimates are provided by the U.S. Environmental Protection Agency in compliance with Federal Law. By using this Guide, consumers can estimate the average yearly fuel cost for any vehicle. The Guide is intended to help consumers compare the fuel economy of similarly sized cars, light duty trucks and special purpose vehicles.

  10. Model Year 2012 Fuel Economy Guide: EPA Fuel Economy Estimates

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    None

    2011-11-01

    The Fuel Economy Guide is published by the U.S. Department of Energy as an aid to consumers considering the purchase of a new vehicle. The Guide lists estimates of miles per gallon (mpg) for each vehicle available for the new model year. These estimates are provided by the U.S. Environmental Protection Agency in compliance with Federal Law. By using this Guide, consumers can estimate the average yearly fuel cost for any vehicle. The Guide is intended to help consumers compare the fuel economy of similarly sized cars, light duty trucks and special purpose vehicles.

  11. Model Year 2011 Fuel Economy Guide: EPA Fuel Economy Estimates

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    None

    2010-11-01

    The Fuel Economy Guide is published by the U.S. Department of Energy as an aid to consumers considering the purchase of a new vehicle. The Guide lists estimates of miles per gallon (mpg) for each vehicle available for the new model year. These estimates are provided by the U.S. Environmental Protection Agency in compliance with Federal Law. By using this Guide, consumers can estimate the average yearly fuel cost for any vehicle. The Guide is intended to help consumers compare the fuel economy of similarly sized cars, light duty trucks and special purpose vehicles.

  12. Model Year 2017 Fuel Economy Guide: EPA Fuel Economy Estimates

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    None

    The Fuel Economy Guide is published by the U.S. Department of Energy as an aid to consumers considering the purchase of a new vehicle. The Guide lists estimates of miles per gallon (mpg) for each vehicle available for the new model year. These estimates are provided by the U.S. Environmental Protection Agency in compliance with Federal Law. By using this Guide, consumers can estimate the average yearly fuel cost for any vehicle. The Guide is intended to help consumers compare the fuel economy of similarly sized cars, light duty trucks and special purpose vehicles.

  13. Model Year 2018 Fuel Economy Guide: EPA Fuel Economy Estimates

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    None

    The Fuel Economy Guide is published by the U.S. Department of Energy as an aid to consumers considering the purchase of a new vehicle. The Guide lists estimates of miles per gallon (mpg) for each vehicle available for the new model year. These estimates are provided by the U.S. Environmental Protection Agency in compliance with Federal Law. By using this Guide, consumers can estimate the average yearly fuel cost for any vehicle. The Guide is intended to help consumers compare the fuel economy of similarly sized cars, light duty trucks and special purpose vehicles.

  14. Annealing tests of in-pile irradiated oxide coated U-Mo/Al-Si dispersed nuclear fuel

    NASA Astrophysics Data System (ADS)

    Zweifel, T.; Valot, Ch.; Pontillon, Y.; Lamontagne, J.; Vermersch, A.; Barrallier, L.; Blay, T.; Petry, W.; Palancher, H.

    2014-09-01

    U-Mo/Al based nuclear fuels have been worldwide considered as a promising high density fuel for the conversion of high flux research reactors from highly enriched uranium to lower enrichment. In this paper, we present the annealing test up to 1800 °C of in-pile irradiated U-Mo/Al-Si fuel plate samples. More than 70% of the fission gases (FGs) are released during two major FG release peaks around 500 °C and 670 °C. Additional characterisations of the samples by XRD, EPMA and SEM suggest that up to 500 °C FGs are released from IDL/matrix interfaces. The second peak at 670 °C representing the main release of FGs originates from the interaction between U-Mo and matrix in the vicinity of the cladding.

  15. Fuel preparation for use in the production of medical isotopes

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Policke, Timothy A.; Aase, Scott B.; Stagg, William R.

    The present invention relates generally to the field of medical isotope production by fission of uranium-235 and the fuel utilized therein (e.g., the production of suitable Low Enriched Uranium (LEU is uranium having 20 weight percent or less uranium-235) fuel for medical isotope production) and, in particular to a method for producing LEU fuel and a LEU fuel product that is suitable for use in the production of medical isotopes. In one embodiment, the LEU fuel of the present invention is designed to be utilized in an Aqueous Homogeneous Reactor (AHR) for the production of various medical isotopes including, butmore » not limited to, molybdenum-99, cesium-137, iodine-131, strontium-89, xenon-133 and yttrium-90.« less

  16. Transition from HEU to LEU fuel in Romania`s 14-MW TRIGA reactor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Bretscher, M.M.; Snelgrove, J.L.

    1991-12-31

    The 14-MW TRIGA steady state reactor (SSR) located in Pitesti, Romania, first went critical in the fall of 1979. Initially, the core configuration for full power operation used 29 fuel clusters each containing a 5 {times} 5 square array of HEU (10 wt%) -- ZrH -- Er (2.8 wt%) fuel-moderator rods (1.295 cm o.d.) clad in Incology. With a total inventory of 35 HEU fuel clusters, burnup considerations required a gradual expansion of the core from 29 to 32 and finally to 35 clusters before the reactor was shut down because of insufficient excess reactivity. At this time each ofmore » the original 29 fuel clusters had an overage {sup 235}U burnup in the range from 50 to 62%. Because of the US policy regarding the export of highly enriched uranium, fresh HEU TRIGA replacement fuel is not available. After a number of safety-related measurements, the SSR is expected to resume full power operation in the near future using a mixed core containing five LEU TRIGA clusters of the same geometry as the original fuel but with fuel-moderator rods containing 45 wt% U (19.7% {sup 235}U enrichment) and 1.1 wt% Er. Rods for 14 additional LEU fuel clusters will be fabricated by General Atomics. In support of the SSR mixed core operation numerous neutronic calculations have been performed. This paper presents some of the results of those calculations.« less

  17. Demonstration of optimum fuel-to-moderator ratio in a PWR unit fuel cell

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Feltus, M.A.; Pozsgai, C.

    1992-01-01

    Nuclear engineering students at The Pennsylvania State University develop scaled-down [[approx]350 MW(thermal)] pressurized water reactors (PWRs) using actual plants as references. The design criteria include maintaining the clad temperature below 2200[degree]F, fuel temperature below melting point, sufficient departure from nucleate boiling ratio (DNBR) margin, a beginning-of-life boron concentration that yields a negative moderator temperature coefficient, an adequate cycle power production (330 effective full-power days), and a batch loading scheme that is economical. The design project allows for many degrees of freedom (e.g., assembly number, pitch and height and batch enrichments) so that each student's result is unique. The iterative naturemore » of the design process is stressed in the course. The LEOPARD code is used for the unit cell depletion, critical boron, and equilibrium xenon calculations. Radial two-group diffusion equations are solved with the TWIDDLE-DEE code. The steady-state ZEBRA thermal-hydraulics program is used for calculating DNBR. The unit fuel cell pin radius and pitch (fuel-to-moerator ratio) for the scaled-down design, however, was set equal to the already optimized ratio for the reference PWR. This paper describes an honors project that shows how the optimum fuel-to-moderator ratio is found for a unit fuel cell shown in terms of neutron economics. This exercise illustrates the impact of fuel-to-moderator variations on fuel utilization factor and the effect of assuming space and energy separability.« less

  18. Method for processing coal-enrichment waste with solid and volatile fuel inclusions

    NASA Astrophysics Data System (ADS)

    Khasanova, A. V.; Zhirgalova, T. B.; Osintsev, K. V.

    2017-10-01

    The method relates to the field of industrial heat and power engineering. It can be used in coal preparation plants for processing coal waste. This new way is realized to produce a loose ash residue directed to the production of silicate products and fuel gas in rotary kilns. The proposed method is associated with industrial processing of brown coal beneficiation waste. Waste is obtained by flotation separation of rock particles up to 13 mm in size from coal particles. They have in their composition both solid and volatile fuel inclusions (components). Due to the high humidity and significant rock content, low heat of combustion, these wastes are not used on energy boilers, they are stored in dumps polluting the environment.

  19. Status of reduced enrichment programs for research reactors in Japan

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Kanda, Keiji; Nishihara, Hedeaki; Shirai, Eiji

    1997-08-01

    The reduced enrichment programs for the JRR-2, JRR-3, JRR-4 and JMTR of Japan Atomic Energy Research Institute (JAERI), and the KUR of Kyoto University Research Reactor Institute (KURRI) have been partially completed and are mostly still in progress under the Joint Study Programs with Argonne National Laboratory (ANL). The JMTR and JRR-2 have been already converted to use MEU aluminide fuels in 1986 and 1987, respectively. The operation of the upgraded JRR-3(JRR-3M) has started in March 1990 with the LEU aluminide fuels. Since May 1992, the two elements have been inserted in the KUR. The safety review application for themore » full core conversion to use LEU silicide in the JMTR was approved in February 1992 and the conversion has been done in January 1994. The Japanese Government approved a cancellation of the KUHFR Project in February 1991, and in April 1994 the U.S. Government gave an approval to utilize HEU in the KUR instead of the KUHFR. Therefore, the KUR will be operated with HEU fuel until 2001. Since March 1994, Kyoto University is continuing negotiation with UKAEA Dounreay on spent fuel reprocessing and blending down of recovered uranium, in addition to that with USDOE.« less

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

    DOE Office of Scientific and Technical Information (OSTI.GOV)

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

    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 processmore » was three tables containing the radionuclide inventories for LEU/MOX in elemental form. Masses, thermal powers, and activities were reported for each category.« less

  1. Preservation of microbial communities enriched on lignocellulose under thermophilic and high-solid conditions.

    PubMed

    Yu, Chaowei; Reddy, Amitha P; Simmons, Christopher W; Simmons, Blake A; Singer, Steven W; VanderGheynst, Jean S

    2015-01-01

    Microbial communities enriched from diverse environments have shown considerable promise for the targeted discovery of microorganisms and enzymes for bioconversion of lignocellulose to liquid fuels. While preservation of microbial communities is important for commercialization and research, few studies have examined storage conditions ideal for preservation. The goal of this study was to evaluate the impact of preservation method on composition of microbial communities enriched on switchgrass before and after storage. The enrichments were completed in a high-solid and aerobic environment at 55 °C. Community composition was examined for each enrichment to determine when a stable community was achieved. Preservation methods included cryopreservation with the cryoprotective agents DMSO and glycerol, and cryopreservation without cryoprotective agents. Revived communities were examined for their ability to decompose switchgrass under high-solid and thermophilic conditions. High-throughput 16S rRNA gene sequencing of DNA extracted from enrichment samples showed that the majority of the shift in composition of the switchgrass-degrading community occurred during the initial three 2-week enrichments. Shifts in community structure upon storage occurred in all cryopreserved samples. Storage in liquid nitrogen in the absence of cryoprotectant resulted in variable preservation of dominant microorganisms in enriched samples. Cryopreservation with either DMSO or glycerol provided consistent and equivalent preservation of dominant organisms. A stable switchgrass-degrading microbial community was achieved after three 2-week enrichments. Dominant microorganisms were preserved equally well with DMSO and glycerol. DMSO-preserved communities required more incubation time upon revival to achieve pre-storage activity levels during high-solid thermophilic cultivation on switchgrass. Despite shifts in the community with storage, the samples were active upon revival under thermophilic and

  2. Preservation of microbial communities enriched on lignocellulose under thermophilic and high-solid conditions

    DOE PAGES

    Yu, Chaowei; Reddy, Amitha P.; Simmons, Christopher W.; ...

    2015-12-02

    Microbial communities enriched from diverse environments have shown considerable promise for the targeted discovery of microorganisms and enzymes for bioconversion of lignocellulose to liquid fuels. While preservation of microbial communities is important for commercialization and research, few studies have examined storage conditions ideal for preservation. The goal of this study was to evaluate the impact of preservation method on composition of microbial communities enriched on switchgrass before and after storage. The enrichments were completed in a high-solid and aerobic environment at 55 °C. Community composition was examined for each enrichment to determine when a stable community was achieved. Preservation methodsmore » included cryopreservation with the cryoprotective agents DMSO and glycerol, and cryopreservation without cryoprotective agents. Revived communities were examined for their ability to decompose switchgrass under high-solid and thermophilic conditions. High-throughput 16S rRNA gene sequencing of DNA extracted from enrichment samples showed that the majority of the shift in composition of the switchgrass-degrading community occurred during the initial three 2-week enrichments. Shifts in community structure upon storage occurred in all cryopreserved samples. Storage in liquid nitrogen in the absence of cryoprotectant resulted in variable preservation of dominant microorganisms in enriched samples. Cryopreservation with either DMSO or glycerol provided consistent and equivalent preservation of dominant organisms. In conclusion, a stable switchgrass-degrading microbial community was achieved after three 2-week enrichments. Dominant microorganisms were preserved equally well with DMSO and glycerol. DMSO-preserved communities required more incubation time upon revival to achieve pre-storage activity levels during high-solid thermophilic cultivation on switchgrass. Despite shifts in the community with storage, the samples were active upon revival

  3. Stationary Fuel Cell System Composite Data Products | Hydrogen and Fuel

    Science.gov Websites

    Capacity by Equipment Type CDP STAT 14, 10/21/15 Average Eligible Cost by Equipment Type, including Other Distributed Generation CDP STAT 15, 10/21/15 Average Eligible Cost for Biogas Sources CDP STAT 16, 10/21/15 Capacity and Eligible Cost (CHP Fuel Cells) CDP STAT 22, 10/21/15 Distribution of Eligible Cost with and

  4. Analysis of Accidents at the Pakistan Research Reactor-1 Using Proposed Mixed-Fuel (HEU and LEU) Core

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Bokhari, Ishtiaq H.

    2004-12-15

    The Pakistan Research Reactor-1 (PARR-1) was converted from highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel in 1991. The reactor is running successfully, with an upgraded power level of 10 MW. To save money on the purchase of costly fresh LEU fuel elements, the use of less burnt HEU spent fuel elements along with the present LEU fuel elements is being considered. The proposal calls for the HEU fuel elements to be placed near the thermal column to gain the required excess reactivity. In the present study the safety analysis of a proposed mixed-fuel core has been carried outmore » at a calculated steady-state power level of 9.8 MW. Standard computer codes and correlations were employed to compute various parameters. Initiating events in reactivity-induced accidents involve various modes of reactivity insertion, namely, start-up accident, accidental drop of a fuel element on the core, flooding of a beam tube with water, and removal of an in-pile experiment during reactor operation. For each of these transients, time histories of reactor power, energy released, temperature, and reactivity were determined.« less

  5. Microstructural Characteristics of HIP-bonded Monolithic Nuclear Fuels with a Diffusion Barrier

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Jan-Fong Jue; Dennis D. Keiser, Jr.; Cynthia R. Breckenridge

    Due to the limitation of maximum uranium load achievable by dispersion fuel type, the Global Threat Reduction Initiative (GTRI) is developing an advanced monolithic fuel to convert US high performance research reactors to low-enriched uranium. Hot-isostatic-press bonding was the single process down-selected to bond monolithic U-Mo fuel meat to aluminum alloy cladding. A diffusion barrier was applied to the U–Mo fuel meat by roll-bonding process to prevent extensive interaction between fuel meat and aluminum-alloy cladding. Microstructural characterization was performed on fresh fuel plates fabricated at Idaho National Laboratory. Interfaces between fuel meat, cladding, and diffusion barrier, as well as U–10Momore » fuel meat and Al–6061 cladding were characterized by scanning electron microscopy. Preliminary results indicate that the interfaces contain many different phases while decomposition, second phases, and chemical banding were also observed in the fuel meat. The important attributes of the HIP-bonded monolithic fuel are • A typical Zr diffusion barrier of thickness 25 µm • Transverse cross section that exhibits relatively equiaxed grains with an average grain diameter of 10 µm • Chemical banding, in some areas more than 100 µm in length, that is very pronounced in longitudinal (i.e., rolling) direction with Mo concentration varying from 7–13 wt% • Decomposed areas containing plate-shaped low-Mo phase • A typical Zr/cladding interaction layer of thickness 1-2 µm • A visible UZr2 bearing layer of thickness 1-2 µm • Mo-rich precipitates (mainly Mo2Zr, forming a layer in some areas) followed by a Mo-depleted sub-layer between the visible UZr2-bearing layer and the U–Mo matrix • No excessive interaction between cladding and the uncoated fuel edge • Cladding-to-cladding bonding that exhibits no cracks or porosity with second phases high in Mg, Si, and O decorating the bond line. • Some of these attributes might be critical to

  6. Microstructural characteristics of HIP-bonded monolithic nuclear fuels with a diffusion barrier

    NASA Astrophysics Data System (ADS)

    Jue, Jan-Fong; Keiser, Dennis D.; Breckenridge, Cynthia R.; Moore, Glenn A.; Meyer, Mitchell K.

    2014-05-01

    Due to the limitation of maximum uranium load achievable by dispersion fuel type, the Global Threat Reduction Initiative is developing an advanced monolithic fuel to convert US high-performance research reactors to low-enriched uranium. Hot-isostatic-press (HIP) bonding was the single process down-selected to bond monolithic U-Mo fuel meat to aluminum alloy cladding. A diffusion barrier was applied to the U-Mo fuel meat by roll-bonding process to prevent extensive interaction between fuel meat and aluminum-alloy cladding. Microstructural characterization was performed on fresh fuel plates fabricated at Idaho National Laboratory. Interfaces between the fuel meat, the cladding, and the diffusion barrier, as well as between the U-10Mo fuel meat and the Al-6061 cladding, were characterized by scanning electron microscopy. Preliminary results indicate that the interfaces contain many different phases while decomposition, second phases, and chemical banding were also observed in the fuel meat. The important attributes of the HIP-bonded monolithic fuel are: average grain diameter of 10 μm. Chemical banding, in some areas more than 100 μm in length, that is very pronounced in longitudinal (i.e., rolling) direction with Mo concentration varying from 7-13 wt.%. Decomposed areas containing plate-shaped low-Mo phase. A typical Zr/cladding interaction layer with a thickness of 1-2 μm. A visible UZr2 bearing layer with a thickness of 1-2 μm. Mo-rich precipitates (mainly Mo2Zr, forming a layer in some areas) followed by a Mo-depleted sub-layer between the visible UZr2-bearing layer and the U-Mo matrix. No excessive interaction between cladding and the uncoated fuel edge. Cladding-to-cladding bonding that exhibits no cracks or porosity with second phases high in Mg, Si, and O decorating the bond line. Some of these attributes might be

  7. Neutronics Benchmarks for the Utilization of Mixed-Oxide Fuel: Joint U.S./Russian Progress Report for Fiscal Year 1997

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Akkurt, H

    2001-01-11

    In 1967, a series of critical experiments were conducted at the Westinghouse Reactor Evaluation Center (WREC) using mixed-oxide (MOX) PuO{sub 2}-UO{sub 2} and/or UO{sub 2} fuels in various lattices and configurations . These experiments were performed under the joint sponsorship of the Empire State Atomic Development Associates (ESADA) plutonium program and Westinghouse . The purpose of these experiments was to develop experimental data to validate analytical methods used in the design of a plutonium-bearing replacement fuel for water reactors. Three different fuels were used during the experimental program: two MOX fuels and a low-enriched UO{sub 2} fuel. The MOX fuelsmore » were distinguished by their {sup 240}Pu content: 8 wt% {sup 240}Pu and 24 wt% {sup 240}Pu. Both MOX fuels contained 2.0 wt % PuO{sub 2} in natural UO{sub 2} . The UO{sub 2} fuel with 2.72 wt % enrichment was used for comparison with the plutonium data and for use in multiregion experiments.« less

  8. Hybrid Reynolds-Averaged/Large Eddy Simulation of the Flow in a Model SCRamjet Cavity Flameholder

    NASA Technical Reports Server (NTRS)

    Baurle, R. A.

    2016-01-01

    Steady-state and scale-resolving simulations have been performed for flow in and around a model scramjet combustor flameholder. Experimental data available for this configuration include velocity statistics obtained from particle image velocimetry. Several turbulence models were used for the steady-state Reynolds-averaged simulations which included both linear and non-linear eddy viscosity models. The scale-resolving simulations used a hybrid Reynolds-averaged/large eddy simulation strategy that is designed to be a large eddy simulation everywhere except in the inner portion (log layer and below) of the boundary layer. Hence, this formulation can be regarded as a wall-modeled large eddy simulation. This e ort was undertaken to not only assess the performance of the hybrid Reynolds-averaged / large eddy simulation modeling approach in a flowfield of interest to the scramjet research community, but to also begin to understand how this capability can best be used to augment standard Reynolds-averaged simulations. The numerical errors were quantified for the steady-state simulations, and at least qualitatively assessed for the scale-resolving simulations prior to making any claims of predictive accuracy relative to the measurements. The steady-state Reynolds-averaged results displayed a high degree of variability when comparing the flameholder fuel distributions obtained from each turbulence model. This prompted the consideration of applying the higher-fidelity scale-resolving simulations as a surrogate "truth" model to calibrate the Reynolds-averaged closures in a non-reacting setting prior to their use for the combusting simulations. In general, the Reynolds-averaged velocity profile predictions at the lowest fueling level matched the particle imaging measurements almost as well as was observed for the non-reacting condition. However, the velocity field predictions proved to be more sensitive to the flameholder fueling rate than was indicated in the measurements.

  9. Mixed sulfate-reducing bacteria-enriched microbial fuel cells for the treatment of wastewater containing copper.

    PubMed

    Miran, Waheed; Jang, Jiseon; Nawaz, Mohsin; Shahzad, Asif; Jeong, Sang Eun; Jeon, Che Ok; Lee, Dae Sung

    2017-12-01

    Microbial fuel cells (MFCs) have been widely investigated for organic-based waste/substrate conversion to electricity. However, toxic compounds such as heavy metals are ubiquitous in organic waste and wastewater. In this work, a sulfate reducing bacteria (SRB)-enriched anode is used to study the impact of Cu 2+ on MFC performance. This study demonstrates that MFC performance is slightly enhanced at concentrations of up to 20 mg/L of Cu 2+ , owing to the stimulating effect of metals on biological reactions. Cu 2+ removal involves the precipitation of metalloids out of the solution, as metal sulfide, after they react with the sulfide produced by SRB. Simultaneous power generation of 224.1 mW/m 2 at lactate COD/SO 4 2- mass ratio of 2.0 and Cu 2+ of 20 mg/L, and high Cu 2+ removal efficiency, at >98%, are demonstrated in the anodic chamber of a dual-chamber MFC. Consistent MFC performance at 20 mg/L of Cu 2+ for ten successive cycles shows the excellent reproducibility of this system. In addition, total organic content and sulfate removal efficiencies greater than 85% and 70%, respectively, are achieved up to 20 mg/L of Cu 2+ in 48 h batches. However, higher metal concentration and very low pH at <4.0 inhibit the SRB MFC system. Microbial community analysis reveals that Desulfovibrio is the most abundant SRB in anode biofilm at the genus level, at 38.1%. The experimental results demonstrate that biological treatment of low-concentration metal-containing wastewater with SRB in MFCs can be an attractive technique for the bioremediation of this type of medium with simultaneous energy generation. Copyright © 2017 Elsevier Ltd. All rights reserved.

  10. Characterization of Delayed-Particle Emission Signatures for Pyroprocessing. Part 1: ABTR Fuel Assembly.

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Durkee, Jr., Joe W.

    A three-part study is conducted using the MCNP6 Monte Carlo radiation-transport code to calculate delayed-neutron (DN) and delayed-gamma (DG) emission signatures for nondestructive assay (NDA) metal-fuel pyroprocessing. In Part 1, MCNP6 is used to produce irradiation-induced used nuclear fuel (UNF) isotopic inventories for an Argonne National Laboratory (ANL) Advanced Burner Test Reactor (ABTR) preconceptual design fuel assembly (FA) model. The initial fuel inventory consists of uranium mixed with light-water-reactor transuranic (TRU) waste and 10 wt% zirconium (U-LWR-SFTRU-10%Zr). To facilitate understanding, parametric evaluation is done using models for 3% and 5% initial 235U a% enrichments, burnups of 5, 10, 15, 20,more » 30, …, 120 GWd/MTIHM, and 3-, 5-, 10-, 20-, and 30- year cooling times. Detailed delayed-particle radioisotope source terms for the irradiate FA are created using BAMF-DRT and SOURCES3A. Using simulation tallies, DG activity ratios (DGARs) are developed for 134Cs/ 137Cs 134Cs/ 154Eu, and 154Eu/ 137Cs markers as a function of (1) burnup and (2) actinide mass, including elemental uranium, neptunium, plutonium, americium, and curium. Spectral-integrated DN emission is also tallied. The study reveals a rich assortment of DGAR behavior as a function of DGAR type, enrichment, burnup, and cooling time. Similarly, DN emission plots show variation as a function of burnup and of actinide mass. Sensitivity of DGAR and DN signatures to initial 235U enrichment, burnup, and cooling time is evident. Comparisons of the ABTR radiation signatures and radiation signatures previously reported for a generic Westinghouse oxide-fuel assembly indicate that there are pronounced differences in the ABTR and Westinghouse oxide-fuel DN and DG signatures. These differences are largely attributable to the initial TRU inventory in the ABTR fuel. The actinide and nonactinide inventories for the FA models serve as source materials for the pre- and postelectrorefining

  11. Post-irradiation-examination of irradiated fuel outside the hot cell

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Dawn E. Janney; Adam B. Robinson; Thomas P. O'Holleran

    Because of their high radioactivity, irradiated fuels are commonly examined in a hot cell. However, the Idaho National Laboratory (INL) has recently investigated irradiated U-Mo-Al metallic fuel from the Reduced Enrichment for Research and Test Reactors (RERTR) project using a conventional unshielded scanning electron microscope outside a hot cell. This examination was possible because of a two-step sample-preparation approach in which a small volume of fuel was isolated in a hot cell and shielding was introduced during later stages of sample preparation. The resulting sample contained numerous sample-preparation artifacts but allowed analysis of microstructures from selected areas.

  12. 49 CFR 531.5 - Fuel economy standards.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 49 Transportation 6 2010-10-01 2010-10-01 false Fuel economy standards. 531.5 Section 531.5 Transportation Other Regulations Relating to Transportation (Continued) NATIONAL HIGHWAY TRAFFIC SAFETY ADMINISTRATION, DEPARTMENT OF TRANSPORTATION PASSENGER AUTOMOBILE AVERAGE FUEL ECONOMY STANDARDS § 531.5 Fuel economy standards. (a) Except as provided...

  13. Measurement of the 235U Induced Fission Gamma-ray Spectrum as an Active Non-destructive Assay of Fresh Nucleear Fuel

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Sarnoski, Sarah E.; Fast, James E.; Fulsom, Bryan G.

    2017-07-17

    Non-destructive assay is a powerful tool the International Atomic Energy Agency (IAEA) employs to verify adherence to safeguards agreements. Current IAEA veri- cation techniques for fresh nuclear fuel include passive gamma-ray spectroscopy to determine fuel enrichment. This technique suers from self-shielding and lakes the percision to detect diversion of central fuel rods. The aim of this research is to develop a new, more capable non-destructive analysis technique using active neutron interroga- tion of fuel assemblies and determining the yields of short-lived ssion products from high-resolution gamma-ray spectroscopy using high-purity germanium (HPGe). This paper reports results from irradiation of a onemore » meter tall mock fresh fuel assembly with low enriched uranium (LEU) or depleted uranium (DU) rods using a down-scattered deuterium-tritium (D-T) neutron source. Both prompt and delayed gamma-ray spec- tra were collected as time-stamped list-mode data in a coax detector and without list mode data in a planar strip detector. No dierentiating signatures were observed in the prompt spectra in either detector; however, both detectors observed several short-lived ssion product signatures in LEU and not DU fuel, indicating that this technique has potential for determination of enrichment of fresh fuel assemblies. There were eight unique ssion products observed in the LEU spectra with the coax detector spectra, and three ssion products were observed in the LEU spectra with the strip detector.« less

  14. Fabrication of Monolithic RERTR Fuels by Hot Isostatic Pressing

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Jan-Fong Jue; Blair H. Park; Curtis R. Clark

    2010-11-01

    The RERTR (Reduced Enrichment for Research and Test Reactors) Program is developing advanced nuclear fuels for high-power test reactors. Monolithic fuel design provides higher uranium loading than that of the traditional dispersion fuel design. Hot isostatic pressing is a promising process for low-cost batch fabrication of monolithic RERTR fuel plates for these high-power reactors. Bonding U Mo fuel foil and 6061 Al cladding by hot isostatic press bonding was successfully developed at Idaho National Laboratory. Due to the relatively high processing temperature, the interaction between fuel meat and aluminum cladding is a concern. Two different methods were employed to mitigatemore » this effect: (1) a diffusion barrier and (2) a doping addition to the interface. Both types of fuel plates have been fabricated by hot isostatic press bonding. Preliminary results show that the direct fuel/cladding interaction during the bonding process was eliminated by introducing a thin zirconium diffusion barrier layer between the fuel and the cladding. Fuel plates were also produced and characterized with a silicon-rich interlayer between fuel and cladding. This paper reports the recent progress of this developmental effort and identifies the areas that need further attention.« less

  15. Application of polyimide membranes for biogas purification and enrichment.

    PubMed

    Harasimowicz, M; Orluk, P; Zakrzewska-Trznadel, G; Chmielewski, A G

    2007-06-18

    Biogas is a clean environment friendly fuel that is produced by bacterial conversion of organic matter under anaerobic (oxygen-free) conditions. Raw biogas contains about 55-65% methane (CH(4)), 30-45% carbon dioxide (CO(2)), traces of hydrogen sulphide (H(2)S) and fractions of water vapour. Pure methane has a calorific value of 9100 kcal/m(3) at 15.5 degrees C and 1 atm; the calorific value of biogas varies from 4800 to 6900 kcal/m(3). To achieve the standard composition of the biogas and calorific value of 5500 kcal/m(3) the treatment techniques like absorption or membrane separation should be applied. In the paper the results of the tests of the CH(4) enrichment in simulated biogas mixture consisted of methane, carbon dioxide and hydrogen sulphide were presented. It was showed that using the capillary module with polyimide membranes it was possible to achieve the enrichment of CH(4) from the concentrations of 55-85% up to 91-94.4%. The membrane material was resistant to the small concentrations of sour gases and assured the reduction of H(2)S and water vapour concentrations, as well. The required enrichment was achieved in the single module, however to prevent CH(4) losses the multistage or hybrid systems should be used to improve process efficiency.

  16. Alternative Fuels Data Center: Transportation System Efficiency

    Science.gov Websites

    energy use. Transportation planners and corporate decision makers can implement combinations of these corporate decision makers can help employees telework to conserve fuel. Maps & Data Average Annual Fuel

  17. Spatial variation of deuterium enrichment in bulk water of snowgum leaves.

    PubMed

    Santrucek, Jirí; Kveton, Jirí; Setlík, Jirí; Bulícková, Lenka

    2007-01-01

    Deuterium enrichment of bulk water was measured and modeled in snowgum (Eucalyptus pauciflora Sieber ex Sprengel) leaves grown under contrasting air and soil humidity in arid and wet conditions in a glasshouse. A map of the enrichment was constructed with a resolution of 4 mm by using a newly designed cryodistillation method. There was progressively increasing enrichment in both longitudinal (along the leaf midrib) and transversal (perpendicular to the midrib) directions, most pronounced in the arid-grown leaf. The whole-leaf average of the enrichment was well below the value estimated by the Craig-Gordon model. The discrepancy between model and measurements persisted when the estimates were carried out separately for the leaf base and tip, which differed in temperature and stomatal conductance. The discrepancy was proportional to the transpiration rate, indicating the significance of diffusion-advection interplay (Péclet effect) of deuterium-containing water molecules in small veins close to the evaporating sites in the leaf. Combined Craig-Gordon and desert-river models, with or without the Péclet number, P, were used for predicting the leaf longitudinal enrichment. The predictions without P overestimated the measured values of deltadeuterium. Fixed P value partially improved the coincidence. We suggest that P should vary along the leaf length l to reconcile the modeled data with observations of longitudinal enrichment. Local values of P, P(l), integrating the upstream fraction of water used or the leaf area, substantially improved the model predictions.

  18. A model for recovery of scrap monolithic uranium molybdenum fuel by electrorefining

    NASA Astrophysics Data System (ADS)

    Van Kleeck, Melissa A.

    The goal of the Reduced Enrichment for Research and Test Reactors program (RERTR) is toreduce enrichment at research and test reactors, thereby decreasing proliferation risk at these facilities. A new fuel to accomplish this goal is being manufactured experimentally at the Y12 National Security Complex. This new fuel will require its own waste management procedure,namely for the recovery of scrap from its manufacture. The new fuel is a monolithic uraniummolybdenum alloy clad in zirconium. Feasibility tests were conducted in the Planar Electrode Electrorefiner using scrap U-8Mo fuel alloy. These tests proved that a uranium product could be recovered free of molybdenum from this scrap fuel by electrorefining. Tests were also conducted using U-10Mo Zr clad fuel, which confirmed that product could be recovered from a clad version of this scrap fuel at an engineering scale, though analytical results are pending for the behavior of Zr in the electrorefiner. A model was constructed for the simulation of electrorefining the scrap material produced in the manufacture of this fuel. The model was implemented on two platforms, Microsoft Excel and MatLab. Correlations, used in the model, were developed experimentally, describing area specific resistance behavior at each electrode. Experiments validating the model were conducted using scrap of U-10Mo Zr clad fuel in the Planar Electrode Electrorefiner. The results of model simulations on both platforms were compared to experimental results for the same fuel, salt and electrorefiner compositions and dimensions for two trials. In general, the model demonstrated behavior similar to experimental data but additional refinements are needed to improve its accuracy. These refinements consist of a function for surface area at anode and cathode based on charge passed. Several approximations were made in the model concerning areas of electrodes which should be replaced by a more accurate function describing these areas.

  19. Surface engineering of low enriched uranium-molybdenum

    NASA Astrophysics Data System (ADS)

    Leenaers, A.; Van den Berghe, S.; Detavernier, C.

    2013-09-01

    Recent attempts to qualify the LEU(Mo) dispersion plate fuel with Si addition to the Al matrix up to high power and burn-up have not yet been successful due to unacceptable fuel plate swelling at a local burn-up above 60% 235U. The root cause of the failures is clearly related directly to the formation of the U(Mo)-Al(Si) interaction layer. Excessive formation of these layers around the fuel kernels severely weakens the local mechanical integrity and eventually leads to pillowing of the plate. In 2008, SCK·CEN has launched the SELENIUM U(Mo) dispersion fuel development project in an attempt to find an alternative way to reduce the interaction between U(Mo) fuel kernels and the Al matrix to a significantly low level: by applying a coating on the U(Mo) kernels. Two fuel plates containing 8gU/cc U(Mo) coated with respectively 600 nm Si and 1000 nm ZrN in a pure Al matrix were manufactured. These plates were irradiated in the BR2 reactor up to a maximum heat flux of 470 W/cm2 until a maximum local burn-up of approximately 70% 235U (˜50% plate average) was reached. Awaiting the PIE results, the advantages of applying a coating are discussed in this paper through annealing experiments and TRIM (the Transport of Ions in Matter) calculations.

  20. Comparison of flexible fuel vehicle and life-cycle fuel consumption and emissions of selected pollutants and greenhouse gases for ethanol 85 versus gasoline.

    PubMed

    Zhai, Haibo; Frey, H Christopher; Rouphail, Nagui M; Gonçalves, Gonçalo A; Farias, Tiago L

    2009-08-01

    The objective of this research is to evaluate differences in fuel consumption and tailpipe emissions of flexible fuel vehicles (FFVs) operated on ethanol 85 (E85) versus gasoline. Theoretical ratios of fuel consumption and carbon dioxide (CO2) emissions for both fuels are estimated based on the same amount of energy released. Second-by-second fuel consumption and emissions from one FFV Ford Focus fueled with E85 and gasoline were measured under real-world traffic conditions in Lisbon, Portugal, using a portable emissions measurement system (PEMS). Cycle average dynamometer fuel consumption and emission test results for FFVs are available from the U.S. Department of Energy, and emissions certification test results for ethanol-fueled vehicles are available from the U.S. Environmental Protection Agency. On the basis of the PEMS data, vehicle-specific power (VSP)-based modal average fuel and emission rates for both fuels are estimated. For E85 versus gasoline, empirical ratios of fuel consumption and CO2 emissions agree within a margin of error to the theoretical expectations. Carbon monoxide (CO) emissions were found to be typically lower. From the PEMS data, nitric oxide (NO) emissions associated with some higher VSP modes are higher for E85. From the dynamometer and certification data, average hydrocarbon (HC) and nitrogen oxides (NOx) emission differences vary depending on the vehicle. The differences of average E85 versus gasoline emission rates for all vehicle models are -22% for CO, 12% for HC, and -8% for NOx emissions, which imply that replacing gasoline with E85 reduces CO emissions, may moderately decrease NOx tailpipe emissions, and may increase HC tailpipe emissions. On a fuel life cycle basis for corn-based ethanol versus gasoline, CO emissions are estimated to decrease by 18%. Life-cycle total and fossil CO2 emissions are estimated to decrease by 25 and 50%, respectively; however, life-cycle HC and NOx emissions are estimated to increase by 18 and 82

  1. Music for the birds: effects of auditory enrichment on captive bird species.

    PubMed

    Robbins, Lindsey; Margulis, Susan W

    2016-01-01

    With the increase of mixed species exhibits in zoos, targeting enrichment for individual species may be problematic. Often, mammals may be the primary targets of enrichment, yet other species that share their environment (such as birds) will unavoidably be exposed to the enrichment as well. The purpose of this study was to determine if (1) auditory stimuli designed for enrichment of primates influenced the behavior of captive birds in the zoo setting, and (2) if the specific type of auditory enrichment impacted bird behavior. Three different African bird species were observed at the Buffalo Zoo during exposure to natural sounds, classical music and rock music. The results revealed that the average frequency of flying in all three bird species increased with naturalistic sounds and decreased with rock music (F = 7.63, df = 3,6, P = 0.018); vocalizations for two of the three species (Superb Starlings and Mousebirds) increased (F = 18.61, df = 2,6, P = 0.0027) in response to all auditory stimuli, however one species (Lady Ross's Turacos) increased frequency of duetting only in response to rock music (X(2) = 18.5, df = 2, P < 0.0001). Auditory enrichment implemented for large mammals may influence behavior in non-target species as well, in this case leading to increased activity by birds. © 2016 Wiley Periodicals, Inc.

  2. Bioavailability and possible benefits of wheat intake naturally enriched with selenium and its products.

    PubMed

    Djujić, I S; Jozanov-Stankov, O N; Milovac, M; Janković, V; Djermanović, V

    2000-12-01

    Bioavailability and possible benefits of wheat intake naturally enriched with selenium and its products was tested. Wheat obtained by application of an original combination and procedure for foliar supplementation of plants with Se was characterized on the average by five times higher content of Se, the main form being L-(+)-selenomethionine (SeMet). Substitution of Se-deficient wheat by wheat naturally enriched with Se and its products contributed to the increase of daily intake on the average by 18 microg (12-35 microg) in volunteers, which is more than 50% of the average daily intake. Six weeks after the beginning of its application, increased daily intake of Se brought about the increase of its concentration in the plasma of the examined persons by 53%, in their erythrocytes by 37%, in their hair by 44%, and in their urine by 54%. This result was comparable to the effect obtained in the course of an 8-wk daily intake of supplements with 100 microg Se in the form of enriched bakery yeast. Analysis of glutathione peroxidase (GSH-Px) activity in blood, thiobarbituric acid reactive substances (TBARS) in plasma, lipid parameters (total cholesterol, high-density lipoprotein cholesterol, low-density lipoprotein cholesterol, and triglycerides), and glucose in serum of volunteers showed that the increased Se intake induced increased GSH-Px activity in blood and decreased concentrations of TBARS, lipid parameters, and glucose in blood. Using only one crop (wheat enriched with Se), the existing deficiency of Se in our population can be alleviated. In this way, one-fourth of our population with lower Se intake than 21 microg/d will satisfy basal requirements, whereas one-half will become moderately deficient in Se instead of distinctly deficient in Se.

  3. Assessment for advanced fuel cycle options in CANDU

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Morreale, A.C.; Luxat, J.C.; Friedlander, Y.

    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 drivermore » 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.« less

  4. Accident Analysis for the NIST Research Reactor Before and After Fuel Conversion

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Baek J.; Diamond D.; Cuadra, A.

    Postulated accidents have been analyzed for the 20 MW D2O-moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analysis has been carried out for the present core, which contains high enriched uranium (HEU) fuel and for a proposed equilibrium core with low enriched uranium (LEU) fuel. The analyses employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron transport calculations were performed with the MCNPX code to determine homogenized fuel compositions in the lower and upper halves of each fuel element and to determine the resulting neutronic properties of the core. The accident analysis employed a modelmore » of the primary loop with the RELAP5 code. The model includes the primary pumps, shutdown pumps outlet valves, heat exchanger, fuel elements, and flow channels for both the six inner and twenty-four outer fuel elements. Evaluations were performed for the following accidents: (1) control rod withdrawal startup accident, (2) maximum reactivity insertion accident, (3) loss-of-flow accident resulting from loss of electrical power with an assumption of failure of shutdown cooling pumps, (4) loss-of-flow accident resulting from a primary pump seizure, and (5) loss-of-flow accident resulting from inadvertent throttling of a flow control valve. In addition, natural circulation cooling at low power operation was analyzed. The analysis shows that the conversion will not lead to significant changes in the safety analysis and the calculated minimum critical heat flux ratio and maximum clad temperature assure that there is adequate margin to fuel failure.« less

  5. Estimation of average burnup of damaged fuels loaded in Fukushima Dai-ichi reactors by using the {sup 134}Cs/{sup 137}Cs ratio method

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Endo, T.; Sato, S.; Yamamoto, A.

    2012-07-01

    Average burnup of damaged fuels loaded in Fukushima Dai-ichi reactors is estimated, using the {sup 134}Cs/{sup 137}Cs ratio method for measured radioactivities of {sup 134}Cs and {sup 137}Cs in contaminated soils within the range of 100 km from the Fukushima Dai-ichi nuclear power plants. As a result, the measured {sup 134}Cs/{sup 137}Cs ratio from the contaminated soil is 0.996{+-}0.07 as of March 11, 2011. Based on the {sup 134}Cs/{sup 137}Cs ratio method, the estimated burnup of damaged fuels is approximately 17.2{+-}1.5 [GWd/tHM]. It is noted that the numerical results of various calculation codes (SRAC2006/PIJ, SCALE6.0/TRITON, and MVP-BURN) are almost themore » same evaluation values of {sup 134}Cs/ {sup 137}Cs ratio with same evaluated nuclear data library (ENDF-B/VII.0). The void fraction effect in depletion calculation has a major impact on {sup 134}Cs/{sup 137}Cs ratio compared with the differences between JENDL-4.0 and ENDF-B/VII.0. (authors)« less

  6. Management of thermal peaking factors in CONFU-B PWR assemblies using neutron poisons and tailored enrichment

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Visosky, M.; Hejzlar, P.; Kazimi, M.

    2006-07-01

    CONFU-B assemblies are PWR assemblies containing standard Uranium fuel rods and TRU bearing inert material fuel rods and are designed to achieve net TRU destruction over a 4.5-year irradiation. These highly heterogeneous assemblies tend to exhibit large intra-assembly power peaking factors (IAPPF). Neutronic strategies to reduce IAPPF are developed. The IAPPF are calculated at the assembly level using CASMO4, and these are used to calculate the most restrictive thermal margin (the Minimum Departure from Nucleate Boiling Ratio, MDNBR) using a whole-core VIPRE-01 model. This paper examines two strategies to manage the thermal margin of a CONFU-B assembly while retaining themore » TRU destruction performance: use of neutron poisons and tailored enrichment schemes. Burnable poisons can be used to suppress BOL reactivity of fresh CONFU-B assemblies with only minor impact on MDNBR and TRU destruction performance. Tailored enrichment, along with the use of soluble boron, can achieve significant improvements in MDNBR, but at some cost to TRU destruction performance. (authors)« less

  7. Nuclear energy in Europe: uranium flow modeling and fuel cycle scenario trade-offs from a sustainability perspective.

    PubMed

    Tendall, Danielle M; Binder, Claudia R

    2011-03-15

    The European nuclear fuel cycle (covering the EU-27, Switzerland and Ukraine) was modeled using material flow analysis (MFA).The analysis was based on publicly available data from nuclear energy agencies and industries, national trade offices, and nongovernmental organizations. Military uranium was not considered due to lack of accessible data. Nuclear fuel cycle scenarios varying spent fuel reprocessing, depleted uranium re-enrichment, enrichment assays, and use of fast neutron reactors, were established. They were then assessed according to environmental, economic and social criteria such as resource depletion, waste production, chemical and radiation emissions, costs, and proliferation risks. The most preferable scenario in the short term is a combination of reduced tails assay and enrichment grade, allowing a 17.9% reduction of uranium demand without significantly increasing environmental, economic, or social risks. In the long term, fast reactors could theoretically achieve a 99.4% decrease in uranium demand and nuclear waste production. However, this involves important costs and proliferation risks. Increasing material efficiency is not systematically correlated with the reduction of other risks. This suggests that an overall optimization of the nuclear fuel cycle is difficult to obtain. Therefore, criteria must be weighted according to stakeholder interests in order to determine the most sustainable solution. This paper models the flows of uranium and associated materials in Europe, and provides a decision support tool for identifying the trade-offs of the alternative nuclear fuel cycles considered.

  8. Controlling factors of soil selenium distribution in a watershed in Se-enriched and longevity region of South China.

    PubMed

    Shao, Ya; Cai, Chongfa; Zhang, Haitao; Fu, Wei; Zhong, Xuemei; Tang, Shen

    2018-05-10

    Selenium (Se) is an essential nutritional element for human beings. Many studies have been conducted on concentration and distribution patterns of soil Se in low Se, Se-enriched, and selenosis areas; however, soil Se has not been systematically studied in a watershed, especially in Se-enriched longevity region and karst area in South China. This study is carried out to explore the controlling factors of Se-enriched soils in Baishou river tributary watershed, where soils are Se-enriched, and local people have the phenomenon of longevity. The area-weighted average rock Se concentration in the watershed is 0.054 mg/kg, and there are no significant differences in rock Se concentration between different strata and between different lithological rocks. The area-weighted average concentration of Se in soils (0-20 cm) is 0.80 mg/kg, and the soil Se concentration is of high level in the watershed. Soil Se concentration decreases from upstream to downstream in the watershed, and significantly correlated with elevation. Climate is the main factor causing high content of soil Se in the watershed which lacks black rock series. The difference of clastic and carbonate parent materials in soil forming process and the physical and chemical properties (pH, OM, etc.) are the main reasons for the spatial variation of Se distribution in the watershed. The research will be beneficial to the development and utilization of Se-enriched soil in Se-enriched area.

  9. PWR and BWR spent fuel assembly gamma spectra measurements

    NASA Astrophysics Data System (ADS)

    Vaccaro, S.; Tobin, S. J.; Favalli, A.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Hu, J.; Schwalbach, P.; Sjöland, A.; Trellue, H.; Vo, D.

    2016-10-01

    A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative-Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

  10. PWR and BWR spent fuel assembly gamma spectra measurements

    DOE PAGES

    Vaccaro, S.; Tobin, Stephen J.; Favalli, Andrea; ...

    2016-07-17

    A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detectmore » the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. As a result, to compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.« less

  11. PWR and BWR spent fuel assembly gamma spectra measurements

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Vaccaro, S.; Tobin, Stephen J.; Favalli, Andrea

    A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detectmore » the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of 137Cs, 154Eu, and 134Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. As a result, to compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.« less

  12. Nuclear Forensics Attributing the Source of Spent Fuel Used in an RDD Event

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Scott, Mark Robert

    2005-05-01

    An RDD attack against the U.S. is something America needs to prepare against. If such an event occurs the ability to quickly identify the source of the radiological material used in an RDD would aid investigators in identifying the perpetrators. Spent fuel is one of the most dangerous possible radiological sources for an RDD. In this work, a forensics methodology was developed and implemented to attribute spent fuel to a source reactor. The specific attributes determined are the spent fuel burnup, age from discharge, reactor type, and initial fuel enrichment. It is shown that by analyzing the post-event material, thesemore » attributes can be determined with enough accuracy to be useful for investigators. The burnup can be found within a 5% accuracy, enrichment with a 2% accuracy, and age with a 10% accuracy. Reactor type can be determined if specific nuclides are measured. The methodology developed was implemented into a code call NEMASYS. NEMASYS is easy to use and it takes a minimum amount of time to learn its basic functions. It will process data within a few minutes and provide detailed information about the results and conclusions.« less

  13. Characteristics and Dose Levels for Spent Reactor Fuels

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Coates, Cameron W

    2007-01-01

    Current guidance considers highly radioactive special nuclear materials to be those materials that, unshielded, emit a radiation dose [rate] measured at 1 m which exceeds 100 rem/h. Smaller, less massive fuel assemblies from research reactors can present a challenge from the point of view of self protection because of their size (lower dose, easier to handle) and the desirability of higher enrichments; however, a follow-on study to cross-compare dose trends of research reactors and power reactors was deemed useful to confirm/verify these trends. This paper summarizes the characteristics and dose levels of spent reactor fuels for both research reactors andmore » power reactors and extends previous studies aimed at quantifying expected dose rates from research reactor fuels worldwide.« less

  14. Supplemental Reactor Physics Calculations and Analysis of ELF Mk 1A Fuel

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Pope, Michael A.

    2014-10-01

    These calculations supplement previous the reactor physics work evaluating the Enhanced Low Enriched Uranium (LEU) Fuel (ELF) Mk 1A element. This includes various additional comparisons between the current Highly Enriched Uranium (HEU) and LEU along with further characterization of the performance of the ELF fuel. The excess reactivity to be held down at BOC for ELF Mk 1A fuel is estimated to be approximately $2.75 greater than with HEU for a typical cycle. This is a combined effect of the absence of burnable poison in the ELF fuel and the reduced neck shim worth in LEU fuel compared to HEU.more » Burnable poison rods were conceptualized for use in the small B positions containing Gd2O3 absorber. These were shown to provide $2.37 of negative reactivity at BOC and to burn out in less than half of a cycle. The worth of OSCCs is approximately the same between HEU and ELF Mk 1A (LEU) fuels in the representative loading evaluated. This was evaluated by rotating all banks simultaneously. The safety rod worth is relatively unchanged between HEU and ELF Mk 1A (LEU) fuels in the representative loading evaluated. However, this should be reevaluated with different loadings. Neutron flux, both total and fast (>1 MeV), is either the same or reduced upon changing from HEU to ELF Mk 1A (LEU) fuels in the representative loading evaluated. This is consistent with the well-established trend of lower neutron fluxes for a given power in LEU than HEU.The IPT loop void reactivity is approximately the same or less positive with ELF Mk 1A (LEU) fuel than HEU in the representative loading evaluated.« less

  15. Nuclear Nonproliferation: Concerns With U.S. Delays in Accepting Foreign Research Reactors’ Spent Fuel

    DTIC Science & Technology

    1994-03-01

    transport or storage plans. The return of some of the spent fuel will also depend on the readiness of dry storage . One expert told us that...enriched uranium fuel (HEU), a material that can be used to make nuclear bombs, in civilian nuclear programs worldwide. Research reactors are of...address the environmental impact of transporting the fuel and storing it in both existing and new storage units, possibly by June 1995. Under the

  16. Alternative Fuels Data Center: Maps and Data

    Science.gov Websites

    View Graph Graph Download Data Generated_thumb20140811-21276-p5mcbz Average Fuel Economy at Different Road Grades Generated_thumb20140811-21276-p5mcbz Trend of Fuel Economy and Consumption at different at Various Driving Speeds Generated_thumb20170119-11720-1o8tesk Trend of fuel efficiency at different

  17. VISION User Guide - VISION (Verifiable Fuel Cycle Simulation) Model

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Jacob J. Jacobson; Robert F. Jeffers; Gretchen E. Matthern

    2009-08-01

    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; 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 for U.S. nuclear power. The model is not intendedmore » 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 of 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. Note that recovered uranium is itself often partitioned: some RU flows with recycled transuranic elements, some flows with wastes, and the rest is designated RU. RU comes out of storage if needed to correct the U/TRU ratio in new recycled fuel. Neither RU nor DU are designated as wastes. VISION is comprised of

  18. Virginia ridesharing statistics : methodologies for determining carpooler and vanpool average life bases and the average fuel economy of commuter vehicles.

    DOT National Transportation Integrated Search

    1985-01-01

    The objective of this research was to investigate methods of computing average life values for carpoolers and vanpools in Virginia. These statistics are to be used by the Rail and Public Transportation Division in evaluating the efficiency and cost-e...

  19. LIFE Materials: Overview of Fuels and Structural Materials Issues Volume 1

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Farmer, J

    2008-09-08

    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 spentmore » 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

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

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Jacob J. Jacobson; Robert F. Jeffers; Gretchen E. Matthern

    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 amore » 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

  1. Analysis of the Reactor Physics of Low-Enrichment Fuel for the INL Advanced Test Reactor in support of RERTR

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Mark DeHart; William Skerjanc; Sean Morrell

    2012-06-01

    Analysis of the performance of the ATR with a LEU fuel design shows promise in terms of a core design that will yield the same neutron sources in target locations. A proposed integral cladding burnable absorber design appears to meet power profile requirements that will satisfy power distributions for safety limits. Performance of this fuel design is ongoing; the current work is the initial evaluation of the core performance of this fuel design with increasing burnup. Results show that LEU fuel may have a longer lifetime that HEU fuel however, such limits may be set by mechanical performance of themore » fuel rather that available reactivity. Changes seen in the radial fuel power distribution with burnup in LEU fuel will require further study to ascertain the impact on neutron fluxes in target locations. Source terms for discharged fuel have also been studied. By its very nature, LEU fuel produces much more plutonium than is present in HEU fuel at discharge. However, the effect of the plutonium inventory appears to have little affect on radiotoxicity or decay heat in the fuel.« less

  2. Succinic acid production on xylose-enriched biorefinery streams by Actinobacillus succinogenes in batch fermentation

    DOE PAGES

    Salvachua, Davinia; Mohagheghi, Ali; Smith, Holly; ...

    2016-02-02

    Co-production of chemicals from lignocellulosic biomass alongside fuels holds promise for improving the economic outlook of integrated biorefineries. In current biochemical conversion processes that use thermochemical pretreatment and enzymatic hydrolysis, fractionation of hemicellulose-derived and cellulose-derived sugar streams is possible using hydrothermal or dilute acid pretreatment (DAP), which then offers a route to parallel trains for fuel and chemical production from xylose- and glucose-enriched streams. Succinic acid (SA) is a co-product of particular interest in biorefineries because it could potentially displace petroleum-derived chemicals and polymer precursors for myriad applications. Furthermore, SA production from biomass-derived hydrolysates has not yet been fully exploredmore » or developed.« less

  3. Carbon monoxide exposure from aircraft fueling vehicles.

    PubMed

    McCammon, C S; Halperin, W F; Lemen, R A

    1981-01-01

    Investigators from the National Institute for Occupational Safety and Health observed deficiencies in maintenance of fueling trucks at an international airport. The exhaust system is vented under the front bumper, a standard design on fueling trucks which is intended to minimize the proximity of the exhaust system to the jet fuel in the vehicles. Carbon monoxide levels were measured in the cabs of 17 fueling trucks with windows closed, heaters on, and in different positions relative to the wind. One truck had an average CO level of 300 ppm, two exceeded 100 ppm, five others exceeded 50 ppm, while levels in the other nine averaged less than or equal to 500 ppm. Levels of CO depended on the mechanical condition of the vehicle and the vehicle's orientation to the wind. Stringent maintenance is required as the exhaust design is not fail-safe.

  4. Construction and Characterization of a Cellulolytic Consortium Enriched from the Hindgut of Holotrichia parallela Larvae.

    PubMed

    Sheng, Ping; Huang, Jiangli; Zhang, Zhihong; Wang, Dongsheng; Tian, Xiaojuan; Ding, Jiannan

    2016-09-30

    Degradation of rice straw by cooperative microbial activities is at present the most attractive alternative to fuels and provides a basis for biomass conversion. The use of microbial consortia in the biodegradation of lignocelluloses could reduce problems such as incomplete synergistic enzymes, end-product inhibition, and so on. In this study, a cellulolytic microbial consortium was enriched from the hindgut of Holotrichia parallela larvae via continuous subcultivation (20 subcultures in total) under static conditions. The degradation ratio for rice straw was about 83.1% after three days of cultivation, indicating its strong cellulolytic activity. The diversity analysis results showed that the bacterial diversity and richness decreased during the consortium enrichment process, and the consortium enrichment process could lead to a significant enrichment of phyla Proteobacteria and Spirochaetes, classes Clostridia, Epsilonproteobacteria, and Betaproteobacteria, and genera Arcobacter , Treponema , Comamonas , and Clostridium . Some of these are well known as typical cellulolytic and hemicellulolytic microorganisms. Our results revealed that the microbial consortium identified herein is a potential candidate for use in the degradation of waste lignocellulosic biomass and further highlights the hindgut of the larvae as a reservoir of extensive and specific cellulolytic and hemicellulolytic microbes.

  5. Electrochemical and genomic analysis of novel electroactive isolates obtained via potentiostatic enrichment from tropical sediment

    NASA Astrophysics Data System (ADS)

    Doyle, Lucinda E.; Yung, Pui Yi; Mitra, Sumitra D.; Wuertz, Stefan; Williams, Rohan B. H.; Lauro, Federico M.; Marsili, Enrico

    2017-07-01

    Enrichment of electrochemically-active microorganisms (EAM) to date has mostly relied on microbial fuel cells fed with wastewater. This study aims to enrich novel EAM by exposing tropical sediment, not frequently reported in the literature, to sustained anodic potentials. Voltamperometric techniques and electrochemical impedance spectroscopy, performed over a wide range of potentials, characterise extracellular electron transfer (EET) over time. Applied potential is found to affect biofilm electrochemical signature. Geobacter metallireducens is heavily enriched on the electrodes, as determined by metagenomic and metatranscriptomic analysis, in the first report of the species in a lactate-fed system. Two novel isolates are grown in pure culture from the enrichment, identified by 16S rRNA gene sequencing as Aeromonas and Enterobacter, respectively. The names proposed are Aeromonas sp. CL-1 and Enterobacter sp. EA-1. Both isolates are capable of EET on carbon felt and screen-printed carbon electrodes without the addition of exogenous redox mediators. Enterobacter sp. EA-1 can also perform mediated electron transfer using the soluble redox mediator 2-hydroxy-1,4-naphthoquinone (HNQ). Both isolates are able to use acetate and lactate as electron donors. This work outlines a comprehensive methodology for characterising novel EAM from unconventional inocula.

  6. Coupling fuel cycles with repositories: how repository institutional choices may impact fuel cycle design

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Forsberg, C.; Miller, W.F.

    2013-07-01

    The historical repository siting strategy in the United States has been a top-down approach driven by federal government decision making but it has been a failure. This policy has led to dispatching fuel cycle facilities in different states. The U.S. government is now considering an alternative repository siting strategy based on voluntary agreements with state governments. If that occurs, state governments become key decision makers. They have different priorities. Those priorities may change the characteristics of the repository and the fuel cycle. State government priorities, when considering hosting a repository, are safety, financial incentives and jobs. It follows that statesmore » will demand that a repository be the center of the back end of the fuel cycle as a condition of hosting it. For example, states will push for collocation of transportation services, safeguards training, and navy/private SNF (Spent Nuclear Fuel) inspection at the repository site. Such activities would more than double local employment relative to what was planned for the Yucca Mountain-type repository. States may demand (1) the right to take future title of the SNF so if recycle became economic the reprocessing plant would be built at the repository site and (2) the right of a certain fraction of the repository capacity for foreign SNF. That would open the future option of leasing of fuel to foreign utilities with disposal of the SNF in the repository but with the state-government condition that the front-end fuel-cycle enrichment and fuel fabrication facilities be located in that state.« less

  7. Utilization of TRISO Fuel with LWR Spent Fuel in Fusion-Fission Hybrid Reactor System

    NASA Astrophysics Data System (ADS)

    Acır, Adem; Altunok, Taner

    2010-10-01

    HTRs use a high performance particulate TRISO fuel with ceramic multi-layer coatings due to the high burn up capability and very neutronic performance. TRISO fuel because of capable of high burn up and very neutronic performance is conducted in a D-T fusion driven hybrid reactor. In this study, TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 68%. The neutronic effect of TRISO coated LWR spent fuel in the fuel rod used hybrid reactor on the fuel performance has been investigated for Flibe, Flinabe and Li20Sn80 coolants. The reactor operation time with the different first neutron wall loads is 24 months. Neutron transport calculations are evaluated by using XSDRNPM/SCALE 5 codes with 238 group cross section library. The effect of TRISO coated LWR spent fuel in the fuel rod used hybrid reactor on tritium breeding (TBR), energy multiplication (M), fissile fuel breeding, average burn up values are comparatively investigated. It is shown that the high burn up can be achieved with TRISO fuel in the hybrid reactor.

  8. Sooting Limits Of Diffusion Flames With Oxygen-Enriched Air And Diluted Fuel

    NASA Technical Reports Server (NTRS)

    Sunderland, P. B.; Urban, D. L.; Stocker, D. P.; Chao, B. H.; Axelbaum, R. L.

    2003-01-01

    Oxygen-enhanced combustion permits certain benefits and flexibility that are not otherwise available in the design of practical combustors, as discussed by Baukal. The cost of pure and enriched oxygen has declined to the point that oxygen-enhanced combustion is preferable to combustion in air for many applications. Carbon sequestration is greatly facilitated by oxygen enrichment because nitrogen can be eliminated from the product stream. For example, when natural gas (or natural gas diluted with CO2) is burned in pure oxygen, the only significant products are water and CO2. Oxygen-enhanced combustion also has important implications for soot formation, as explored in this work. We propose that soot inception in nonpremixed flames requires a region where C/O ratio, temperature, and residence time are above certain critical values. Soot does not form at low temperatures, with the threshold in nonpremixed flames ranging from about 1250-1650 K, a temperature referred to here as the critical temperature for soot inception, Tc. Soot inception also can be suppressed when residence time is short (equivalently, when the strain rate in counterflow flames is high). Soot induction times of 0.8-15 ms were reported by Tesner and Shurupov for acetylene/nitrogen mixtures at 1473 K. Burner stabilized spherical microgravity flames are employed in this work for two main reasons. First, this configuration offers unrestricted control over convection direction. Second, in steady state these flames are strain-free and thus can yield intrinsic sooting limits in diffusion flames, similar to the way past work in premixed flames has provided intrinsic values of C/O ratio associated with soot inception limits.

  9. Significance of and prospects for fuel recycle in Japan

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Otsuka, K.; Ikeda, K.

    Japan's nuclear power plant capacity ranks fourth in the world at around 20 GW. But nuclear fuel cycle industries (enrichment, reprocessing and radioactive waste management) are still in their infancy compared with the size and stage of the power plants. Thus it is a matter of urgency to establish a nuclear fuel cycle in Japan which can promote nuclear energy as a quasi-indigenous energy source. Some moves toward establishing a nuclear fuel cycle have been observed recently. As a case in point, in July 1984, the Federation of Electric Power Companies has formally requested Aomori Prefecture to locate nuclear fuelmore » cycle facilities in the Shimokita Peninsula region. Plutonium recovered from spent fuel will be utilized in LWR, ATR, and FBR. Research and development activities on these technologies are in progress.« less

  10. Positive responses of belowground C dynamics to nitrogen enrichment in China.

    PubMed

    Deng, Lei; Peng, Changhui; Zhu, Guangyu; Chen, Lei; Liu, Yulin; Shangguan, Zhouping

    2018-03-01

    Determining how nitrogen (N) impacts ecosystem carbon (C) cycling is critical to using C sequestration to offset anthropogenic CO 2 emissions. The N deposition rate in China is higher than the global average; however, many results of N enrichment experiments in China have not been included in global syntheses. In this study, we assembled a large dataset that comprised 124 published studies concerning N addition experiments, including 570 observations at 127 sites across China, to quantify the responses of belowground C dynamics to N enrichment in terrestrial ecosystems in China by a meta-analysis. The results showed that overall soil organic C, dissolved organic C (DOC) and soil microbial biomass C (MBC) increased by 1.8, 7.4, and 8.8%, respectively (P<0.05), in response to N enrichment; belowground biomass and litter increased by 14.6 and 24.4%, respectively (P<0.05); and soil respiration increased by 6.1% (P<0.05). N enrichment promoted C inputs into the soil mainly by increasing litter and belowground biomass inputs. Additionally, N enrichment increased C output by increasing soil respiration. Land use type and N addition level had different impacts on the soil C pool and on soil respiration. DOC, MBC, and litter exhibited more positive responses to N deposition in cooler and more arid regions than in other regions. The meta-analysis indicated that N enrichment had a positive impact on belowground C cycles in China. Climate played a greater role than did N deposition level in affecting processes of ecosystem C cycling. Moreover, belowground C cycle processes are determined by complicated interactions among land use type, N enrichment, and climate. Copyright © 2017 Elsevier B.V. All rights reserved.

  11. Development of monolithic nuclear fuels for RERTR by hot isostatic pressing

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Jue, J.-F.; Park, Blair; Chapple, Michael

    2008-07-15

    The RERTR Program (Reduced Enrichment for Research and Test Reactors) is developing advanced nuclear fuels for high power test reactors. Monolithic fuel design provides a higher uranium loading than that of the traditional dispersion fuel design. In order to bond monolithic fuel meat to aluminum cladding, several bonding methods such as roll bonding, friction stir bonding and hot isostatic pressing, have been explored. Hot isostatic pressing is a promising process for low cost, batch fabrication of monolithic RERTR fuel plates. The progress on the development of this process at the Idaho National Laboratory will be presented. Due to the relativelymore » high processing temperature used, the reaction between fuel meat and aluminum cladding to form brittle intermetallic phases may be a concern. The effect of processing temperature and time on the fuel/cladding reaction will be addressed. The influence of chemical composition on the reaction will also be discussed. (author)« less

  12. Enrichment of intergalactic matter.

    NASA Technical Reports Server (NTRS)

    Silk, J.; Siluk, R. S.

    1972-01-01

    The primordial gas out of which the Galaxy condensed may have been significantly enriched in heavy elements. A specific mechanism of enrichment is described, in which quasi-stellar sources eject enriched matter into the intergalactic medium. This matter is recycled through successive generations of these sources, and is progressively enriched. The enriched intergalactic matter is accreted by the protogalaxy and we find, for rates of mass ejection by quasi-stellar sources equal to about one solar mass per year in heavy elements, that this mechanism can account for the heavy-element abundances in the oldest Population II stars. Expressions are given for the degree of enrichment of the intergalactic gas as a function of redshift, and we show that our hypothesis implies that the present density of intergalactic gas must be at least a factor 3 larger than the mean density in galaxies at the present epoch.

  13. Thermochemical properties of flame gases from fine wildland fuels

    Treesearch

    Frank A. Albini

    1979-01-01

    Describes a theoretical model for calculating thermochemical properties of the gaseous fuel that burns in the free flame at the edge of a spreading fire in fine forest fuels. Predicted properties are the heat of combustion, stoichiometric air/fuel mass ratio, mass-averaged temperature, and mass fraction of unburned fuel in the gas mixture emitted from the flame-...

  14. Dry Storage of Research Reactor Spent Nuclear Fuel - 13321

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Adams, T.M.; Dunsmuir, M.D.; Leduc, D.R.

    2013-07-01

    Spent fuel from domestic and foreign research reactors is received and stored at the Savannah River Site's L Area Material Storage (L Basin) Facility. This DOE-owned fuel consists primarily of highly enriched uranium in metal, oxide or silicide form with aluminum cladding. Upon receipt, the fuel is unloaded and transferred to basin storage awaiting final disposition. Disposition alternatives include processing via the site's H Canyon facility for uranium recovery, or packaging and shipment of the spent fuel to a waste repository. A program has been developed to provide a phased approach for dry storage of the L Basin fuel. Themore » initial phase of the dry storage program will demonstrate loading, drying, and storage of fuel in twelve instrumented canisters to assess fuel performance. After closure, the loaded canisters are transferred to pad-mounted concrete overpacks, similar to those used for dry storage of commercial fuel. Unlike commercial spent fuel, however, the DOE fuel has high enrichment, very low to high burnup, and low decay heat. The aluminum cladding presents unique challenges due to the presence of an oxide layer that forms on the cladding surface, and corrosion degradation resulting from prolonged wet storage. The removal of free and bound water is essential to the prevention of fuel corrosion and radiolytic generation of hydrogen. The demonstration will validate models predicting pressure, temperature, gas generation, and corrosion performance, provide an engineering scale demonstration of fuel handling, drying, leak testing, and canister backfill operations, and establish 'road-ready' storage of fuel that is suitable for offsite repository shipment or retrievable for onsite processing. Implementation of the Phase I demonstration can be completed within three years. Phases II and III, leading to the de-inventory of L Basin, would require an additional 750 canisters and 6-12 years to complete. Transfer of the fuel from basin storage to dry

  15. Assessment of Possible Cycle Lengths for Fully-Ceramic Micro-Encapsulated Fuel-Based Light Water Reactor Concepts

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    R. Sonat Sen; Michael A. Pope; Abderrafi M. Ougouag

    2012-04-01

    The tri-isotropic (TRISO) fuel developed for High Temperature reactors is known for its extraordinary fission product retention capabilities [1]. Recently, the possibility of extending the use of TRISO particle fuel to Light Water Reactor (LWR) technology, and perhaps other reactor concepts, has received significant attention [2]. The Deep Burn project [3] currently focuses on once-through burning of transuranic fissile and fissionable isotopes (TRU) in LWRs. The fuel form for this purpose is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the TRISO fuel particle design from high temperature reactor technology, but uses SiC as a matrix material rather thanmore » graphite. In addition, FCM fuel may also use a cladding made of a variety of possible material, again including SiC as an admissible choice. The FCM fuel used in the Deep Burn (DB) project showed promising results in terms of fission product retention at high burnup values and during high-temperature transients. In the case of DB applications, the fuel loading within a TRISO particle is constituted entirely of fissile or fissionable isotopes. Consequently, the fuel was shown to be capable of achieving reasonable burnup levels and cycle lengths, especially in the case of mixed cores (with coexisting DB and regular LWR UO2 fuels). In contrast, as shown below, the use of UO2-only FCM fuel in a LWR results in considerably shorter cycle length when compared to current-generation ordinary LWR designs. Indeed, the constraint of limited space availability for heavy metal loading within the TRISO particles of FCM fuel and the constraint of low (i.e., below 20 w/0) 235U enrichment combine to result in shorter cycle lengths compared to ordinary LWRs if typical LWR power densities are also assumed and if typical TRISO particle dimensions and UO2 kernels are specified. The primary focus of this summary is on using TRISO particles with up to 20 w/0 enriched uranium kernels loaded in

  16. Model Year 2014 Fuel Economy Guide: EPA Fuel Economy Estimates

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    None

    2013-12-01

    The Fuel Economy Guide is published by the U.S. Department of Energy as an aid to consumers considering the purchase of a new vehicle. The Guide lists estimates of miles per gallon (mpg) for each vehicle available for the new model year. These estimates are provided by the U.S. Environmental Protection Agency in compliance with Federal Law. By using this Guide, consumers can estimate the average yearly fuel cost for any vehicle. The Guide is intended to help consumers compare the fuel economy of similarly sized cars, light duty trucks and special purpose vehicles. The vehicles listed have been dividedmore » into three classes of cars, three classes of light duty trucks, and three classes of special purpose vehicles.« less

  17. Model Year 2015 Fuel Economy Guide: EPA Fuel Economy Estimates

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    None

    2014-12-01

    The Fuel Economy Guide is published by the U.S. Department of Energy as an aid to consumers considering the purchase of a new vehicle. The Guide lists estimates of miles per gallon (mpg) for each vehicle available for the new model year. These estimates are provided by the U.S. Environmental Protection Agency in compliance with Federal Law. By using this Guide, consumers can estimate the average yearly fuel cost for any vehicle. The Guide is intended to help consumers compare the fuel economy of similarly sized cars, light duty trucks and special purpose vehicles. The vehicles listed have been dividedmore » into three classes of cars, three classes of light duty trucks, and three classes of special purpose vehicles.« less

  18. Model Year 2016 Fuel Economy Guide: EPA Fuel Economy Estimates

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    None

    The Fuel Economy Guide is published by the U.S. Department of Energy as an aid to consumers considering the purchase of a new vehicle. The Guide lists estimates of miles per gallon (mpg) for each vehicle available for the new model year. These estimates are provided by the U.S. Environmental Protection Agency in compliance with Federal Law. By using this Guide, consumers can estimate the average yearly fuel cost for any vehicle. The Guide is intended to help consumers compare the fuel economy of similarly sized cars, light duty trucks and special purpose vehicles. The vehicles listed have been dividedmore » into three classes of cars, three classes of light duty trucks, and three classes of special purpose vehicles.« less

  19. Delgado Community College/Sears-Roebuck Keeping America Working. Math, Science and Technology Summer Youth Enrichment Program. Final Report.

    ERIC Educational Resources Information Center

    Delgado Community Coll., New Orleans, LA.

    Recognizing the need for better preparation of high school students in mathematics, science, and technology, Delgado Community College and the Orleans Parish School System entered into an agreement for the provision of a summer enrichment program for minority students in grades 7 through 9 who had exhibited average or above average abilities in…

  20. Three-dimensional neutronics optimization of helium-cooled blanket for multi-functional experimental fusion-fission hybrid reactor (FDS-MFX)

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Jiang, J.; Yuan, B.; Jin, M.

    2012-07-01

    Three-dimensional neutronics optimization calculations were performed to analyse the parameters of Tritium Breeding Ratio (TBR) and maximum average Power Density (PDmax) in a helium-cooled multi-functional experimental fusion-fission hybrid reactor named FDS (Fusion-Driven hybrid System)-MFX (Multi-Functional experimental) blanket. Three-stage tests will be carried out successively, in which the tritium breeding blanket, uranium-fueled blanket and spent-fuel-fueled blanket will be utilized respectively. In this contribution, the most significant and main goal of the FDS-MFX blanket is to achieve the PDmax of about 100 MW/m3 with self-sustaining tritium (TBR {>=} 1.05) based on the second-stage test with uranium-fueled blanket to check and validate themore » demonstrator reactor blanket relevant technologies based on the viable fusion and fission technologies. Four different enriched uranium materials were taken into account to evaluate PDmax in subcritical blanket: (i) natural uranium, (ii) 3.2% enriched uranium, (iii) 19.75% enriched uranium, and (iv) 64.4% enriched uranium carbide. These calculations and analyses were performed using a home-developed code VisualBUS and Hybrid Evaluated Nuclear Data Library (HENDL). The results showed that the performance of the blanket loaded with 64.4% enriched uranium was the most attractive and it could be promising to effectively obtain tritium self-sufficiency (TBR-1.05) and a high maximum average power density ({approx}100 MW/m{sup 3}) when the blanket was loaded with the mass of {sup 235}U about 1 ton. (authors)« less

  1. Hyb-Seq: Combining target enrichment and genome skimming for plant phylogenomics1

    PubMed Central

    Weitemier, Kevin; Straub, Shannon C. K.; Cronn, Richard C.; Fishbein, Mark; Schmickl, Roswitha; McDonnell, Angela; Liston, Aaron

    2014-01-01

    • Premise of the study: Hyb-Seq, the combination of target enrichment and genome skimming, allows simultaneous data collection for low-copy nuclear genes and high-copy genomic targets for plant systematics and evolution studies. • Methods and Results: Genome and transcriptome assemblies for milkweed (Asclepias syriaca) were used to design enrichment probes for 3385 exons from 768 genes (>1.6 Mbp) followed by Illumina sequencing of enriched libraries. Hyb-Seq of 12 individuals (10 Asclepias species and two related genera) resulted in at least partial assembly of 92.6% of exons and 99.7% of genes and an average assembly length >2 Mbp. Importantly, complete plastomes and nuclear ribosomal DNA cistrons were assembled using off-target reads. Phylogenomic analyses demonstrated signal conflict between genomes. • Conclusions: The Hyb-Seq approach enables targeted sequencing of thousands of low-copy nuclear exons and flanking regions, as well as genome skimming of high-copy repeats and organellar genomes, to efficiently produce genome-scale data sets for phylogenomics. PMID:25225629

  2. Differences in Pop Levels between Conventional and Omega-3 Fatty Acid-Enriched Milk and Dairy Products

    PubMed Central

    Guerranti, Cristiana; Focardi, Silvano Ettore

    2011-01-01

    Conventional and omega-3 fatty acid-enriched milk and cheese were analyzed for persistent organic pollutants (POPs). Omega-3-enriched products are usually supplemented with fish oil which is potentially contaminated. All classes of the considered POPs (PCBs, DDT, HCB, PBDEs, and PCDD/Fs) were found in the samples, with average concentrations higher in omega-3-enriched products than in conventional ones. For PCBs, DDT, and HCB, differences were statistically significant and, therefore, cannot be ascribed to normal variability. With regard to all classes of compounds, the highest levels in individual samples were always found in omega-3 products, in line with the hypothesis that these foods are potentially more contaminated than conventional ones. PMID:23724286

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

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Heidet, F.; Kim, T.; Grandy, C.

    2012-07-30

    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.« less

  4. Managing the Nuclear Fuel Cycle: Policy Implications of Expanding Global Access to Nuclear Power

    DTIC Science & Technology

    2010-03-05

    However, the case of Iran raises perhaps the most critical question in this decade for strengthening the nuclear nonproliferation regime: How can...enrichment process can take advantage of the slight difference in atomic mass between 235U and 238U. The typical enrichment process requires about 10 lbs of...neutrons but can induce fission in all actinides , including all plutonium isotopes. Therefore, nuclear fuel for a fast reactor must have a higher

  5. Simulated Real-World Energy Impacts of a Thermally Sensitive Powertrain Considering Viscous Losses and Enrichment (Presentation)

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Wood, E.; Gonder, J.; Lopp, S.

    It is widely understood that cold-temperature engine operation negatively impacts vehicle fuel use due to a combination of increased friction (high-viscosity engine oil) and temporary enrichment (accelerated catalyst heating). However, relatively little effort has been dedicated to thoroughly quantifying these impacts across a large number of driving cycles and ambient conditions. This work leverages high-quality dynamometer data collected at various ambient conditions to develop a modeling framework for quantifying engine cold-start fuel penalties over a wide array of real-world usage profiles. Additionally, mitigation strategies including energy retention and exhaust heat recovery are explored with benefits quantified for each approach.

  6. Effects of Fuel Distribution on Detonation Tube Performance

    NASA Technical Reports Server (NTRS)

    Perkins, H. Douglas; Sung, Chih-Jen

    2003-01-01

    A pulse detonation engine uses a series of high frequency intermittent detonation tubes to generate thrust. The process of filling the detonation tube with fuel and air for each cycle may yield non-uniform mixtures. Uniform mixing is commonly assumed when calculating detonation tube thrust performance. In this study, detonation cycles featuring idealized non-uniform Hz/air mixtures were analyzed using a two-dimensional Navier-Stokes computational fluid dynamics code with detailed chemistry. Mixture non-uniformities examined included axial equivalence ratio gradients, transverse equivalence ratio gradients, and partially fueled tubes. Three different average test section equivalence ratios were studied; one stoichiometric, one fuel lean, and one fuel rich. All mixtures were detonable throughout the detonation tube. Various mixtures representing the same average test section equivalence ratio were shown to have specific impulses within 1% of each other, indicating that good fuel/air mixing is not a prerequisite for optimal detonation tube performance under conditions investigated.

  7. Sensitivity analysis of FeCrAl cladding and U3Si2 fuel under accident conditions

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Gamble, Kyle Allan Lawrence; Hales, Jason Dean

    2016-08-01

    The purpose of this milestone report is to highlight the results of sensitivity analyses performed on two accident tol- erant fuel concepts: U3Si2 fuel and FeCrAl cladding. The BISON fuel performance code under development at Idaho National Laboratory was coupled to Sandia National Laboratories’ DAKOTA software to perform the sensitivity analyses. Both Loss of Coolant (LOCA) and Station blackout (SBO) scenarios were analyzed using main effects studies. The results indicate that for FeCrAl cladding the input parameters with greatest influence on the output metrics of interest (fuel centerline temperature and cladding hoop strain) during the LOCA were the isotropic swellingmore » and fuel enrichment. For U3Si2 the important inputs were found to be the intergranular diffusion coefficient, specific heat, and fuel thermal conductivity. For the SBO scenario, Young’s modulus was found to be influential in FeCrAl in addition to the isotropic swelling and fuel enrichment. Contrarily to the LOCA case, the specific heat of U3Si2 was found to have no effect during the SBO. The intergranular diffusion coefficient and fuel thermal conductivity were still found to be of importance. The results of the sensitivity analyses have identified areas where further research is required including fission gas behavior in U3Si2 and irradiation swelling in FeCrAl. Moreover, the results highlight the need to perform the sensitivity analyses on full length fuel rods for SBO scenarios.« less

  8. 76 FR 34103 - In the Matter of Areva Enrichment Services, LLC (Eagle Rock Enrichment Facility); Notice of...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-06-10

    .... 10-899-02-ML-BD01] In the Matter of Areva Enrichment Services, LLC (Eagle Rock Enrichment Facility... gas centrifuge uranium enrichment facility--denoted as the Eagle Rock Enrichment Facility (EREF)--in... Information for Contention Preparation; In the Matter of Areva Enrichment Services, LLC (Eagle Rock Enrichment...

  9. Effect of fuel-air-ratio nonuniformity on emissions of nitrogen oxides

    NASA Technical Reports Server (NTRS)

    Lyons, V. J.

    1981-01-01

    The inlet fuel-air ratio nonuniformity is studied to deterine how nitrogen oxide (NOx) emissions are affected. An increase in NOx emissions with increased fuel-air ratio nonuniformity for average equivalence ratios less than 0.7 and a decrease in NOx emissions for average equivalence ratios near stoichiometric is predicted. The degree of uniformityy of fuel-air ratio profiles that is necessary to achieve NOx emissions goals for actual engines that use lean, premixed, prevaporized combustion systems is determined.

  10. Virtual Averaging Making Nonframe-Averaged Optical Coherence Tomography Images Comparable to Frame-Averaged Images.

    PubMed

    Chen, Chieh-Li; Ishikawa, Hiroshi; Wollstein, Gadi; Bilonick, Richard A; Kagemann, Larry; Schuman, Joel S

    2016-01-01

    Developing a novel image enhancement method so that nonframe-averaged optical coherence tomography (OCT) images become comparable to active eye-tracking frame-averaged OCT images. Twenty-one eyes of 21 healthy volunteers were scanned with noneye-tracking nonframe-averaged OCT device and active eye-tracking frame-averaged OCT device. Virtual averaging was applied to nonframe-averaged images with voxel resampling and adding amplitude deviation with 15-time repetitions. Signal-to-noise (SNR), contrast-to-noise ratios (CNR), and the distance between the end of visible nasal retinal nerve fiber layer (RNFL) and the foveola were assessed to evaluate the image enhancement effect and retinal layer visibility. Retinal thicknesses before and after processing were also measured. All virtual-averaged nonframe-averaged images showed notable improvement and clear resemblance to active eye-tracking frame-averaged images. Signal-to-noise and CNR were significantly improved (SNR: 30.5 vs. 47.6 dB, CNR: 4.4 vs. 6.4 dB, original versus processed, P < 0.0001, paired t -test). The distance between the end of visible nasal RNFL and the foveola was significantly different before (681.4 vs. 446.5 μm, Cirrus versus Spectralis, P < 0.0001) but not after processing (442.9 vs. 446.5 μm, P = 0.76). Sectoral macular total retinal and circumpapillary RNFL thicknesses showed systematic differences between Cirrus and Spectralis that became not significant after processing. The virtual averaging method successfully improved nontracking nonframe-averaged OCT image quality and made the images comparable to active eye-tracking frame-averaged OCT images. Virtual averaging may enable detailed retinal structure studies on images acquired using a mixture of nonframe-averaged and frame-averaged OCT devices without concerning about systematic differences in both qualitative and quantitative aspects.

  11. Virtual Averaging Making Nonframe-Averaged Optical Coherence Tomography Images Comparable to Frame-Averaged Images

    PubMed Central

    Chen, Chieh-Li; Ishikawa, Hiroshi; Wollstein, Gadi; Bilonick, Richard A.; Kagemann, Larry; Schuman, Joel S.

    2016-01-01

    Purpose Developing a novel image enhancement method so that nonframe-averaged optical coherence tomography (OCT) images become comparable to active eye-tracking frame-averaged OCT images. Methods Twenty-one eyes of 21 healthy volunteers were scanned with noneye-tracking nonframe-averaged OCT device and active eye-tracking frame-averaged OCT device. Virtual averaging was applied to nonframe-averaged images with voxel resampling and adding amplitude deviation with 15-time repetitions. Signal-to-noise (SNR), contrast-to-noise ratios (CNR), and the distance between the end of visible nasal retinal nerve fiber layer (RNFL) and the foveola were assessed to evaluate the image enhancement effect and retinal layer visibility. Retinal thicknesses before and after processing were also measured. Results All virtual-averaged nonframe-averaged images showed notable improvement and clear resemblance to active eye-tracking frame-averaged images. Signal-to-noise and CNR were significantly improved (SNR: 30.5 vs. 47.6 dB, CNR: 4.4 vs. 6.4 dB, original versus processed, P < 0.0001, paired t-test). The distance between the end of visible nasal RNFL and the foveola was significantly different before (681.4 vs. 446.5 μm, Cirrus versus Spectralis, P < 0.0001) but not after processing (442.9 vs. 446.5 μm, P = 0.76). Sectoral macular total retinal and circumpapillary RNFL thicknesses showed systematic differences between Cirrus and Spectralis that became not significant after processing. Conclusion The virtual averaging method successfully improved nontracking nonframe-averaged OCT image quality and made the images comparable to active eye-tracking frame-averaged OCT images. Translational Relevance Virtual averaging may enable detailed retinal structure studies on images acquired using a mixture of nonframe-averaged and frame-averaged OCT devices without concerning about systematic differences in both qualitative and quantitative aspects. PMID:26835180

  12. Determination of the plutonium content in a spent fuel assembly by passive and active interrogation using a differential die-away instrument

    NASA Astrophysics Data System (ADS)

    Henzl, V.; Croft, S.; Richard, J.; Swinhoe, M. T.; Tobin, S. J.

    2013-06-01

    In this paper, we present a novel approach to estimating the total plutonium content in a spent fuel assembly (SFA) that is based on combining information from a passive measurement of the total neutron count rate (PN) of the assayed SFA and a measure of its multiplication. While PN can be measured essentially with any non-destructive assay (NDA) technique capable of neutron detection, the measure of multiplication is, in our approach, determined by means of active interrogation using an instrument based on the Differential Die-Away technique (DDA). The DDA is a NDA technique developed within the U.S. Department of Energy's Next Generation Safeguards Initiative (NGSI) project focused on the utilization of NDA techniques to determine the elemental plutonium content in commercial nuclear SFA's [1]. This approach was adopted since DDA also allows determination of other SFA characteristics, such as burnup, initial enrichment, and cooling time, and also allows for detection of certain types of diversion of nuclear material. The quantification of total plutonium is obtained using an analytical correlation function in terms of the observed PN and active multiplication. Although somewhat similar approaches relating Pu content with PN have been adopted in the past, we demonstrate by extensive simulation of the fuel irradiation and NDA process that our analytical method is independent of explicit knowledge of the initial enrichment, burnup, and an absolute value of the SFA's reactivity (i.e. multiplication factor). We show that when tested with MCNPX™ simulations comprising the 64 SFA NGSI Spent Fuel Library-1 we were able to determine elemental plutonium content, using just a few calibration parameters, with an average variation in the prediction of around 1-2% across the wide dynamic range of irradiation history parameters used, namely initial enrichment (IE=2-5%), burnup (BU=15-60 GWd/tU) and cooling time (CT=1-80 y). In this paper we describe the basic approach and the

  13. A study of the stabilities, microstructures and fuel characteristics of tri-fuel (diesel-biodiesel-ethanol) using various fuel preparation methods

    NASA Astrophysics Data System (ADS)

    Lee, K. H.; Mukhtar, N. A. M.; Yohaness Hagos, Ftwi; Noor, M. M.

    2017-10-01

    In this study, the work was carried out to investigate the effects of ethanol proportions on the stabilities and physicochemical characteristics of tri-fuel (Diesel-Biodiesel-Ethanol). For the first time, tri-fuel emulsions and blended were compared side by side. The experiment was done with composition having 5%, 10%, 15%, 20% and 25 % of ethanol with fixed 10% of biodiesel from palm oil origin on a volume basis into diesel. The results indicated that the phase stabilities of the emulsified fuels were higher compared to the blended fuels. In addition, tri-fuel composition with higher proportion of ethanol were found unstable with high tendency to form layer separation. It was found that tri-fuel emulsion with 5% ethanol content (D85B10E5) was of the best in stability with little separation. Furthermore, tri-fuel with lowest ethanol proportion indicated convincing physicochemical characteristics compared to others. Physicochemical characteristics of tri-fuel blending yield almost similar results to tri-fuel emulsion but degrading as more proportion ethanol content added. Emulsion category had cloudy look but on temporarily basis. Under the microscope, tri-fuel emulsion and blending droplet were similar for its active moving about micro-bubble but distinct in term of detection of collision, average disperse micro-bubble size, the spread and organization of the microstructure.

  14. Are leaf physiological traits related to leaf water isotopic enrichment in restinga woody species?

    PubMed

    Rosado, Bruno H P; De Mattos, Eduardo A; Sternberg, Leonel Da S L

    2013-09-01

    During plant-transpiration, water molecules having the lighter stable isotopes of oxygen and hydrogen evaporate and diffuse at a faster rate through the stomata than molecules having the heavier isotopes, which cause isotopic enrichment of leaf water. Although previous models have assumed that leaf water is well-mixed and isotopically uniform, non-uniform stomatal closure, promoting different enrichments between cells, and different pools of water within leaves, due to morpho-physiological traits, might lead to inaccuracies in isotopic models predicting leaf water enrichment. We evaluate the role of leaf morpho-physiological traits on leaf water isotopic enrichment in woody species occurring in a coastal vegetation of Brazil known as restinga. Hydrogen and oxygen stable isotope values of soil, plant stem and leaf water and leaf traits were measured in six species from restinga vegetation during a drought and a wet period. Leaf water isotopic enrichment relative to stem water was more homogeneous among species during the drought in contrast to the wet period suggesting convergent responses to deal to temporal heterogeneity in water availability. Average leaf water isotopic enrichment relative to stem water during the drought period was highly correlated with relative apoplastic water content. We discuss this observation in the context of current models of leaf water isotopic enrichment as a function of the Péclet effect. We suggest that future studies should include relative apoplastic water content in isotopic models.

  15. The STAT7 Code for Statistical Propagation of Uncertainties In Steady-State Thermal Hydraulics Analysis of Plate-Fueled Reactors

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Dunn, Floyd E.; Hu, Lin-wen; Wilson, Erik

    The STAT code was written to automate many of the steady-state thermal hydraulic safety calculations for the MIT research reactor, both for conversion of the reactor from high enrichment uranium fuel to low enrichment uranium fuel and for future fuel re-loads after the conversion. A Monte-Carlo statistical propagation approach is used to treat uncertainties in important parameters in the analysis. These safety calculations are ultimately intended to protect against high fuel plate temperatures due to critical heat flux or departure from nucleate boiling or onset of flow instability; but additional margin is obtained by basing the limiting safety settings onmore » avoiding onset of nucleate boiling. STAT7 can simultaneously analyze all of the axial nodes of all of the fuel plates and all of the coolant channels for one stripe of a fuel element. The stripes run the length of the fuel, from the bottom to the top. Power splits are calculated for each axial node of each plate to determine how much of the power goes out each face of the plate. By running STAT7 multiple times, full core analysis has been performed by analyzing the margin to ONB for each axial node of each stripe of each plate of each element in the core.« less

  16. 77 FR 18272 - Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services LLC...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-03-27

    ... that cascades number 1.5, 1.6, 1.7, 1.8, 2.1, and 2.4 as well as autoclave one of the facility have... 2.4 as well as autoclave one of the facility have been constructed in accordance with the... Facility Inspection Reports Regarding Louisiana Energy Services LLC, National Enrichment Facility, Eunice...

  17. Predictors of Success in Accelerated and Enrichment Summer Mathematics Courses for Academically Talented Adolescents

    ERIC Educational Resources Information Center

    Young, Adena E.; Worrell, Frank C.; Gabelko, Nina H.

    2011-01-01

    In this study, we used logistic regression to examine how well student background and prior achievement variables predicted success among students attending accelerated and enrichment mathematics courses at a summer program (N = 459). Socioeconomic status, grade point average (GPA), and mathematics diagnostic test scores significantly predicted…

  18. Simulated Real-World Energy Impacts of a Thermally Sensitive Powertrain Considering Viscous Losses and Enrichment: Preprint

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Wood, E.; Gonder, J.; Lopp, S.

    It is widely understood that cold ambient temperatures increase vehicle fuel consumption due to heat transfer losses, increased friction (increased viscosity lubricants), and enrichment strategies (accelerated catalyst heating). However, relatively little effort has been dedicated to thoroughly quantifying these impacts across a large set of real world drive cycle data and ambient conditions. This work leverages experimental dynamometer vehicle data collected under various drive cycles and ambient conditions to develop a simplified modeling framework for quantifying thermal effects on vehicle energy consumption. These models are applied over a wide array of real-world usage profiles and typical meteorological data to developmore » estimates of in-use fuel economy. The paper concludes with a discussion of how this integrated testing/modeling approach may be applied to quantify real-world, off-cycle fuel economy benefits of various technologies.« less

  19. Further evaluations of the toxicity of irradiated advanced heavy water reactor fuels.

    PubMed

    Edwards, Geoffrey W R; Priest, Nicholas D

    2014-11-01

    The neutron economy and online refueling capability of heavy water moderated reactors enable them to use many different fuel types, such as low enriched uranium, plutonium mixed with uranium, or plutonium and/or U mixed with thorium, in addition to their traditional natural uranium fuel. However, the toxicity and radiological protection methods for fuels other than natural uranium are not well established. A previous paper by the current authors compared the composition and toxicity of irradiated natural uranium to that of three potential advanced heavy water fuels not containing plutonium, and this work uses the same method to compare irradiated natural uranium to three other fuels that do contain plutonium in their initial composition. All three of the new fuels are assumed to incorporate plutonium isotopes characteristic of those that would be recovered from light water reactor fuel via reprocessing. The first fuel investigated is a homogeneous thorium-plutonium fuel designed for a once-through fuel cycle without reprocessing. The second fuel is a heterogeneous thorium-plutonium-U bundle, with graded enrichments of U in different parts of a single fuel assembly. This fuel is assumed to be part of a recycling scenario in which U from previously irradiated fuel is recovered. The third fuel is one in which plutonium and Am are mixed with natural uranium. Each of these fuels, because of the presence of plutonium in the initial composition, is determined to be considerably more radiotoxic than is standard natural uranium. Canadian nuclear safety regulations require that techniques be available for the measurement of 1 mSv of committed effective dose after exposure to irradiated fuel. For natural uranium fuel, the isotope Pu is a significant contributor to the committed effective dose after exposure, and thermal ionization mass spectrometry is sensitive enough that the amount of Pu excreted in urine is sufficient to estimate internal doses, from all isotopes, as low

  20. Predicting Large Deflections of Multiplate Fuel Elements Using a Monolithic FSI Approach

    DOE PAGES

    Curtis, Franklin G.; Freels, James D.; Ekici, Kivanc

    2017-10-26

    As part of the Global Threat Reduction Initiative, the Oak Ridge National Laboratory is evaluating conversion of fuel for the High Flux Isotope Reactor (HFIR) from high-enriched uranium to low-enriched uranium. Currently, multiphysics simulations that model fluid-structure interaction phenomena are being performed to ensure the safety of the reactor with the new fuel type. A monolithic solver that fully couples fluid and structural dynamics is used to model deflections in the new design. A classical experiment is chosen to validate the capabilities of the current solver and the method. Here, a single-plate simulation with various boundary conditions as well asmore » a five-plate simulation are presented. Finally, use of the monolithic solver provides stable solutions for the large deflections and the tight coupling of the fluid and structure and the maximum deflections are captured accurately.« less

  1. Need for higher fuel burnup at the Hatch Plant

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Beckhman, J.T.

    1996-03-01

    Hatch is a BWR 4 and has been in operation for some time. The first unit became commercial about 1975. Obtaining higher burnups, or higher average discharge exposures, is nothing new at Hatch. Since we have started, the discharge exposure of the plant has increased. Now, of course, we are not approaching the numbers currently being discussed but, the average discharge exposure has increased from around 20,000 MWD/MTU in the early to mid-1980s to 34,000 MWD/MTU in 1994, I am talking about batch average values. There are also peak bundle and peak rod values. You will have to make themore » conversions if you think in one way or the other because I am talking in batch averages. During Hatch`s operating history we have had some problems with fuel failure. Higher burnup fuel raises a concern about how much fuel failure you are going to have. Fuel failure is, of course, an economic issue with us. Back in the early 1980s, we had a problem with crud-induced localized corrosion, known as CILC. We have gotten over that, but we had some times when it was up around 27 fuel failures a year. That is not a pleasant time to live through because it is not what you want from an economic viewpoint or any other. We have gotten that down. We have had some fuel failures recently, but they have not been related to fuel burnup or to corrosion. In fact, the number of failures has decreased from the early 1980s to the 90s even though burnup increased during that time. The fuel failures are more debris-related-type failures. In addition to increasing burnups, utilities are actively evaluating or have already incorporated power uprate and longer fuel cycles (e.g., 2-year cycles). The goal is to balance out the higher power density, longer cycles, higher burnup, and to have no leakers. Why do we as an industry want to have higher burnup fuel? That is what I want to tell you a little bit about.« less

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

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    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 spentmore » 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

  3. Mammalian DNA enriched for replication origins is enriched for snap-back sequences.

    PubMed

    Zannis-Hadjopoulos, M; Kaufmann, G; Martin, R G

    1984-11-15

    Using the instability of replication loops as a method for the isolation of double-stranded nascent DNA, extruded DNA enriched for replication origins was obtained and denatured. Snap-back DNA, single-stranded DNA with inverted repeats (palindromic sequences), reassociates rapidly into stem-loop structures with zero-order kinetics when conditions are changed from denaturing to renaturing, and can be assayed by chromatography on hydroxyapatite. Origin-enriched nascent DNA strands from mouse, rat and monkey cells growing either synchronously or asynchronously were purified and assayed for the presence of snap-back sequences. The results show that origin-enriched DNA is also enriched for snap-back sequences, implying that some origins for mammalian DNA replication contain or lie near palindromic sequences.

  4. Predicting vehicle fuel consumption patterns using floating vehicle data.

    PubMed

    Du, Yiman; Wu, Jianping; Yang, Senyan; Zhou, Liutong

    2017-09-01

    The status of energy consumption and air pollution in China is serious. It is important to analyze and predict the different fuel consumption of various types of vehicles under different influence factors. In order to fully describe the relationship between fuel consumption and the impact factors, massive amounts of floating vehicle data were used. The fuel consumption pattern and congestion pattern based on large samples of historical floating vehicle data were explored, drivers' information and vehicles' parameters from different group classification were probed, and the average velocity and average fuel consumption in the temporal dimension and spatial dimension were analyzed respectively. The fuel consumption forecasting model was established by using a Back Propagation Neural Network. Part of the sample set was used to train the forecasting model and the remaining part of the sample set was used as input to the forecasting model. Copyright © 2017. Published by Elsevier B.V.

  5. Evaluation of HFIR LEU Fuel Using the COMSOL Multiphysics Platform

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Primm, Trent; Ruggles, Arthur; Freels, James D

    2009-03-01

    A finite element computational approach to simulation of the High Flux Isotope Reactor (HFIR) Core Thermal-Fluid behavior is developed. These models were developed to facilitate design of a low enriched core for the HFIR, which will have different axial and radial flux profiles from the current HEU core and thus will require fuel and poison load optimization. This report outlines a stepwise implementation of this modeling approach using the commercial finite element code, COMSOL, with initial assessment of fuel, poison and clad conduction modeling capability, followed by assessment of mating of the fuel conduction models to a one dimensional fluidmore » model typical of legacy simulation techniques for the HFIR core. The model is then extended to fully couple 2-dimensional conduction in the fuel to a 2-dimensional thermo-fluid model of the coolant for a HFIR core cooling sub-channel with additional assessment of simulation outcomes. Finally, 3-dimensional simulations of a fuel plate and cooling channel are presented.« less

  6. 16. VIEW OF THE ENRICHED URANIUM RECOVERY SYSTEM. ENRICHED URANIUM ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    16. VIEW OF THE ENRICHED URANIUM RECOVERY SYSTEM. ENRICHED URANIUM RECOVERY PROCESSED RELATIVELY PURE MATERIALS AND SOLUTIONS AND SOLID RESIDUES WITH RELATIVELY LOW URANIUM CONTENT. URANIUM RECOVERY INVOLVED BOTH SLOW AND FAST PROCESSES. (4/4/66) - Rocky Flats Plant, General Manufacturing, Support, Records-Central Computing, Southern portion of Plant, Golden, Jefferson County, CO

  7. Laser and gas centrifuge enrichment

    NASA Astrophysics Data System (ADS)

    Heinonen, Olli

    2014-05-01

    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.

  8. The influence of cladding on fission gas release from irradiated U-Mo monolithic fuel

    NASA Astrophysics Data System (ADS)

    Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.

    2017-04-01

    The monolithic uranium-molybdenum (U-Mo) alloy has been proposed as a fuel design capable of converting the world's highest power research reactors from use of high enriched uranium to low enriched uranium. However, a zirconium (Zr) diffusion barrier must be used to eliminate interactions that form between the U-Mo monolith and aluminum alloy 6061 (AA6061) cladding during fabrication and are enhanced during irradiation. One aspect of fuel development and qualification is to demonstrate an appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An exothermic reaction has previously been observed between the AA6061 cladding and Zr diffusion layer. In this paper, two fuel segments with different irradiation history were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Samples from each segment were tested with cladding and without cladding to investigate the effect, if any, that the exothermic reaction has on fission gas release mechanisms. Measurements revealed there is an instantaneous effect of the cladding/Zr exothermic reaction, but not necessarily a cumulative effect above approximately 973 K (700 °C). The mechanisms responsible for fission gas release events are discussed.

  9. The influence of cladding on fission gas release from irradiated U-Mo monolithic fuel

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.

    2017-04-01

    The monolithic uranium-molybdenum (U-Mo) alloy has been proposed as a fuel design capable of converting the world’s highest power research reactors from use of high enriched uranium to low enriched uranium. However, a zirconium (Zr) diffusion barrier must be used to eliminate interactions that form during fabrication and are enhanced during irradiation between the U-Mo monolith and aluminum alloy 6061 (AA6061) cladding. One aspect of fuel development and qualification is to demonstrate appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An exothermic reaction has previously been observed between the AA6061 cladding andmore » Zr diffusion layer. In this paper, two fuel segments with different irradiation history were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Samples from each segment were tested with cladding and without cladding to investigate the effect, if any, that the exothermic reaction has on fission gas release mechanisms. Measurements revealed there is an instantaneous effect of the cladding/Zr exothermic reaction, but not necessarily a cumulative effect above approximately 973 K (700 oC). The mechanisms responsible for fission gas release events are discussed.« less

  10. Comparison of fresh fuel experimental measurements to MCNPX calculations using self-interrogation neutron resonance densitometry

    NASA Astrophysics Data System (ADS)

    LaFleur, Adrienne M.; Charlton, William S.; Menlove, Howard O.; Swinhoe, Martyn T.

    2012-07-01

    A new non-destructive assay technique called Self-Interrogation Neutron Resonance Densitometry (SINRD) is currently being developed at Los Alamos National Laboratory (LANL) to improve existing nuclear safeguards measurements for Light Water Reactor (LWR) fuel assemblies. SINRD consists of four 235U fission chambers (FCs): bare FC, boron carbide shielded FC, Gd covered FC, and Cd covered FC. Ratios of different FCs are used to determine the amount of resonance absorption from 235U in the fuel assembly. The sensitivity of this technique is based on using the same fissile materials in the FCs as are present in the fuel because the effect of resonance absorption lines in the transmitted flux is amplified by the corresponding (n,f) reaction peaks in the fission chamber. In this work, experimental measurements were performed in air with SINRD using a reference Pressurized Water Reactor (PWR) 15×15 low enriched uranium (LEU) fresh fuel assembly at LANL. The purpose of this experiment was to assess the following capabilities of SINRD: (1) ability to measure the effective 235U enrichment of the PWR fresh LEU fuel assembly and (2) sensitivity and penetrability to the removal of fuel pins from an assembly. These measurements were compared to Monte Carlo N-Particle eXtended transport code (MCNPX) simulations to verify the accuracy of the MCNPX model of SINRD. The reproducibility of experimental measurements via MCNPX simulations is essential to validating the results and conclusions obtained from the simulations of SINRD for LWR spent fuel assemblies.

  11. Development and Validation of Capabilities to Measure Thermal Properties of Layered Monolithic U-Mo Alloy Plate-Type Fuel

    NASA Astrophysics Data System (ADS)

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

    2014-07-01

    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 to low enriched uranium. One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the thermal-conductivity behavior of the fuel system as a function of temperature and expected irradiation conditions. The purpose of this paper is to verify functionality of equipment installed in hot cells for eventual measurements on irradiated uranium-molybdenum (U-Mo) monolithic fuel specimens, refine procedures to operate the equipment, and validate 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 Zr 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.

  12. Alternative Fuels Data Center

    Science.gov Websites

    gross vehicle weight limits by a weight equal to the difference between the average weight of the diesel tank and fueling system. The NGV maximum gross weight may not exceed 82,000 pounds. (Reference

  13. Image fusion of Secondary Ion Mass Spectrometry and Energy-dispersive X-Ray Spectroscopy data for the characterization of uranium-molybdenum fuel foils

    NASA Astrophysics Data System (ADS)

    Willingham, David; Naes, Benjamin E.; Tarolli, Jay G.; Schemer-Kohrn, Alan; Rhodes, Mark; Dahl, Michael; Guzman, Anthony; Burkes, Douglas E.

    2018-01-01

    Uranium-molybdenum (U-Mo) monolithic fuels represent one option for converting civilian research and test reactors operating with high enriched uranium (HEU) to low enriched uranium (LEU), effectively reducing the threat of nuclear proliferation world-wide. However, processes associated with fabrication of U-Mo monolithic fuels result in regions of elemental heterogeneity, observed as bands traversing the cross-section of representative samples. Isotopic variations (e.g., 235U and 238U) could also be introduced because of associated processing steps, particularly since HEU feedstock is melted with natural or depleted uranium diluent to produce LEU. This study demonstrates the utility of correlative analysis of Energy-Dispersive X-ray Spectroscopy (EDS) and Secondary Ion Mass Spectrometry (SIMS) with their image data streams using image fusion, resulting in a comprehensive microanalytical characterization toolbox. Elemental and isotopic measurements were made on a sample from the Advanced Test Reactor (ATR) Full-sized plate In-center flux trap Position (AFIP)-7 experiment and compared to previous optical and electron microscopy results. The image fusion results are characteristic of SIMS isotopic maps, but with the spatial resolution of EDS images and, therefore, can be used to increase the effective spatial resolution of the SIMS imaging results to better understand homogeneity or heterogeneity that persists because of processing selections. Visual inspection using the image fusion methodology indicated slight variations in the 235U/238U ratio and quantitative analysis using the image intensities across several FoVs revealed an average 235U atom percent value of 17.9 ± 2.4%, which was indicative of a non-uniform U isotopic distribution in the area sampled. Further development of this capability is useful for understanding the connections between the properties of LEU fuel alternatives and the ability to predict performance under irradiation.

  14. Yersinia enterocolitica in slaughter pig tonsils: enumeration and detection by enrichment versus direct plating culture.

    PubMed

    Van Damme, Inge; Habib, Ihab; De Zutter, Lieven

    2010-02-01

    Tonsil samples from 139 slaughter pigs were examined for the presence of pathogenic Yersinia enterocolitica by enrichment procedures based on the standard method ISO 10273:2003. In addition, samples were tested by direct plating method to evaluate its efficiency compared to the enrichment culture methods and to quantify the level of contamination in porcine tonsils. In total, 52 samples (37.4%) were positive for pathogenic Y. enterocolitica, all belonging to bioserotype 4/O:3. Fifty out of the 52 positive samples (96.2%) were detected by direct plating. Enumeration showed an average concentration of 4.5 log(10) CFU g(-1) and 4.4 log(10) CFU g(-1) tonsil on Salmonella-Shigella-desoxycholate-calcium chloride (SSDC) and cefsulodin-irgasan-novobiocin (CIN) agar plates, respectively. The enrichment procedures recommended by the ISO 10273:2003 method were not optimal for the isolation of pathogenic Y. enterocolitica from pig tonsils: two days enrichment in irgasan-ticarcillin-potassium chlorate (ITC) broth resulted in an isolation rate of 84.6%, while 5 days enrichment in peptone-sorbitol-bile (PSB) broth recovered only 59.6% of positive samples. Reducing the enrichment time in PSB from 5 to 2 days resulted in a significantly higher recovery rate (94.2%) and might serve as an appropriate enrichment protocol for the isolation of pathogenic Y. enterocolitica from pig tonsils. Compared to enrichment culture methods, results based on direct plating can be obtained in a shorter time course and provide quantitative data that might be needed for further risk assessment studies.

  15. Low-temperature irradiation behavior of uranium-molybdenum alloy dispersion fuel

    NASA Astrophysics Data System (ADS)

    Meyer, M. K.; Hofman, G. L.; Hayes, S. L.; Clark, C. R.; Wiencek, T. C.; Snelgrove, J. L.; Strain, R. V.; Kim, K.-H.

    2002-08-01

    Irradiation tests have been conducted to evaluate the performance of a series of high-density uranium-molybdenum (U-Mo) alloy, aluminum matrix dispersion fuels. Fuel plates incorporating alloys with molybdenum content in the range of 4-10 wt% were tested. Two irradiation test vehicles were used to irradiate low-enrichment fuels to approximately 40 and 70 at.% 235U burnup in the advanced test reactor at fuel temperatures of approximately 65 °C. The fuel particles used to fabricate dispersion specimens for most of the test were produced by generating filings from a cast rod. In general, fuels with molybdenum contents of 6 wt% or more showed stable in-reactor fission gas behavior, exhibiting a distribution of small, stable gas bubbles. Fuel particle swelling was moderate and decreased with increasing alloy content. Fuel particles with a molybdenum content of 4 wt% performed poorly, exhibiting extensive fuel-matrix interaction and the growth of relatively large fission gas bubbles. Fuel particles with 4 or 6 wt% molybdenum reacted more rapidly with the aluminum matrix than those with higher-alloy content. Fuel particles produced by an atomization process were also included in the test to determine the effect of fuel particle morphology and microstructure on fuel performance for the U-10Mo composition. Both of the U-10Mo fuel particle types exhibited good irradiation performance, but showed visible differences in fission gas bubble nucleation and growth behavior.

  16. Managing the Nuclear Fuel Cycle: Policy Implications of Expanding Global Access to Nuclear Power

    DTIC Science & Technology

    2008-01-20

    critical aspect of the nuclear fuel cycle for the United States, where longstanding nonproliferation policy discouraged commercial nuclear fuel...have U.S. government officials. However, the case of Iran raises perhaps the most critical question in this decade for strengthening the nuclear...slight difference in atomic mass between 235U and 238U. The typical enrichment process requires about 10 lbs of uranium U3O8 to produce 1 lb of low

  17. Trends of jet fuel demand and properties

    NASA Technical Reports Server (NTRS)

    Friedman, R.

    1984-01-01

    Petroleum industry forecasts predict an increasing demand for jet fuels, a decrease in the gasoline-to-distillate (heavier fuel) demand ratio, and a greater influx of poorer quality petroleum in the next two to three decades. These projections are important for refinery product analyses. The forecasts have not been accurate, however, in predicting the recent, short term fluctuations in jet fuel and competing product demand. Changes in petroleum quality can be assessed, in part, by a review of jet fuel property inspections. Surveys covering the last 10 years show that average jet fuel freezing points, aromatic contents, and smoke points have trends toward their specification limits.

  18. Experimental Investigations Of The Influence Of Pressure On Critical Extinction Conditions Of Laminar Nonpremixed Flames Burning Condensed Hydrocarbon Fuels, Jet Fuels, And Surrogates

    DTIC Science & Technology

    2014-10-25

    cycloalkanes, aromatics, and alkenes [3,23–26]. On average, the composition by volume is approximately 60% alkanes, 20% cycloalkanes, 18% aromatics, and 2... alkenes [26]. Efforts to develop chemical-kinetic models that accurately describe the combustion of these practical fuels are critical to the development... method also allows for measurements of fuel mass burning rates by recording the required fuel feed rate to maintain the fuel cup level under steady

  19. Preparation of the Second Shipment of Spent Nuclear Fuel from the Ustav Jaderneho Vyzkumu Rez (UJV Rez), a.s., Czech Republic to the Russian Federation for Reprocessing - 13478

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Trtilek, Radek; Podlaha, Josef

    After more than 50 years of operation of the LVR-15 research reactor operated by the UJV Rez, a. s. (formerly Nuclear Research Institute - NRI), a large amount of the spent nuclear fuel (SNF) of Russian origin has been accumulated. In 2005 UJV Rez, a. s. jointed the Russian Research Reactor Fuel Return (RRRFR) program under the United States (US) - Russian Global Threat Reduction Initiative (GTRI) and started the process of SNF shipment from the LVR-15 research reactor back to the Russian Federation (RF). In 2007 the first shipment of SNF was realized. In 2011, preparation of the secondmore » shipment of spent fuel from the Czech Republic started. The experience obtained from the first shipment will be widely used, but some differences must be taken into the account. The second shipment will be realized in 2013 and will conclude the return transport of all, both fresh and spent, high-enriched nuclear fuel from the Czech Republic to the Russian Federation. After the shipment is completed, there will be only low-enriched nuclear fuel on the territory of the Czech Republic, containing maximum of 20% of U-235, which is the conventionally recognized limit between the low- and high-enriched nuclear materials. The experience (technical, organizational, administrative, logistic) obtained from the each SNF shipment as from the Czech Republic as from other countries using the Russian type research reactors are evaluated and projected onto preparation of next shipment of high enriched nuclear fuel back to the Russian Federation. The results shown all shipments provided by the UJV Rez, a. s. in the frame of the GTRI Program have been performed successfully and safely. It is expected the experience and results will be applied to preparation and completing of the Chinese Miniature Neutron Source Reactors (MNSR) Spent Nuclear Fuel Repatriation in the near future. (authors)« less

  20. Effects of bulk precipitation pH and growth period on cation enrichment in precipitation beneath the canopy of a beech (Fagus moesiaca) forest stand.

    PubMed

    Michopoulos, P; Baloutsos, G; Nakos, G; Economou, A

    2001-12-17

    The effects of bulk precipitation pH and growth period (growing and dormant) on cation enrichment beneath foliage were examined in a beech (Fagus moesiaca) forest stand during a 48-month period. The bulk precipitation pH values ranged from 4.2 to 7.2. The lowest values were observed in winter due to fossil fuel combustion in a nearby big city. The ratio of monthly ion fluxes of throughfall plus stemflow over monthly ion fluxes of bulk precipitation was chosen as an index of cation enrichment and, therefore, as the dependent variable. Bulk precipitation pH and growth period were chosen as independent factors. Precipitation interception (%) by tree canopies was also taken into account. It was found that the pH factor was significant only for H+ ion enrichment suggesting neutralization of H+ ions in the beech canopy, whereas Mg2+ and K+ enrichment were greater in the growing period, probably as a result of leaching. Crown interception was negatively significant for NH4+-N enrichment.

  1. Benchmarking criticality analysis of TRIGA fuel storage racks.

    PubMed

    Robinson, Matthew Loren; DeBey, Timothy M; Higginbotham, Jack F

    2017-01-01

    A criticality analysis was benchmarked to sub-criticality measurements of the hexagonal fuel storage racks at the United States Geological Survey TRIGA MARK I reactor in Denver. These racks, which hold up to 19 fuel elements each, are arranged at 0.61m (2 feet) spacings around the outer edge of the reactor. A 3-dimensional model was created of the racks using MCNP5, and the model was verified experimentally by comparison to measured subcritical multiplication data collected in an approach to critical loading of two of the racks. The validated model was then used to show that in the extreme condition where the entire circumference of the pool was lined with racks loaded with used fuel the storage array is subcritical with a k value of about 0.71; well below the regulatory limit of 0.8. A model was also constructed of the rectangular 2×10 fuel storage array used in many other TRIGA reactors to validate the technique against the original TRIGA licensing sub-critical analysis performed in 1966. The fuel used in this study was standard 20% enriched (LEU) aluminum or stainless steel clad TRIGA fuel. Copyright © 2016. Published by Elsevier Ltd.

  2. REACTOR PHYSICS MODELING OF SPENT RESEARCH REACTOR FUEL FOR TECHNICAL NUCLEAR FORENSICS

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Nichols, T.; Beals, D.; Sternat, M.

    2011-07-18

    Technical nuclear forensics (TNF) refers to the collection, analysis and evaluation of pre- and post-detonation radiological or nuclear materials, devices, and/or debris. TNF is an integral component, complementing traditional forensics and investigative work, to help enable the attribution of discovered radiological or nuclear material. Research is needed to improve the capabilities of TNF. One research area of interest is determining the isotopic signatures of research reactors. Research reactors are a potential source of both radiological and nuclear material. Research reactors are often the least safeguarded type of reactor; they vary greatly in size, fuel type, enrichment, power, and burn-up. Manymore » research reactors are fueled with highly-enriched uranium (HEU), up to {approx}93% {sup 235}U, which could potentially be used as weapons material. All of them have significant amounts of radiological material with which a radioactive dispersal device (RDD) could be built. Therefore, the ability to attribute if material originated from or was produced in a specific research reactor is an important tool in providing for the security of the United States. Currently there are approximately 237 operating research reactors worldwide, another 12 are in temporary shutdown and 224 research reactors are reported as shut down. Little is currently known about the isotopic signatures of spent research reactor fuel. An effort is underway at Savannah River National Laboratory (SRNL) to analyze spent research reactor fuel to determine these signatures. Computer models, using reactor physics codes, are being compared to the measured analytes in the spent fuel. This allows for improving the reactor physics codes in modeling research reactors for the purpose of nuclear forensics. Currently the Oak Ridge Research reactor (ORR) is being modeled and fuel samples are being analyzed for comparison. Samples of an ORR spent fuel assembly were taken by SRNL for analytical and

  3. Enrichment and characterization of sulfate reducing, naphthalene degrading microorganisms

    NASA Astrophysics Data System (ADS)

    Steffen, Kümmel; Florian-Alexander, Herbst; Márcia, Duarte; Dietmar, Pieper; Jana, Seifert; Bergen Martin, von; Hans-Hermann, Richnow; Carsten, Vogt

    2014-05-01

    Polycyclic aromatic hydrocarbons (PAH) are pollutants of great concern due to their potential toxicity, mutagenicity and carcinogenicity. PAH are widely distributed in the environment by accidental discharges during the transport, use and disposal of petroleum products, and during forest and grass fires. Caused by their hydrophobic nature, PAH basically accumulate in sediments from where they are slowly released into the groundwater. Although generally limited by the low water solubility of PAH, microbial degradation is one of the major mechanisms leading to the complete clean-up of PAH-contaminated sites. Whereas organisms and biochemical pathways responsible for the aerobic breakdown of PAH are well known, anaerobic PAH biodegradation is less understood; only a few anaerobic PAH degrading cultures have been described. We studied the anaerobic PAH degradation in a microcosm approach to enrich anaerobic PAH degraders. Anoxic groundwater and sediment samples were used as inoculum. Groundwater samples were purchased from the erstwhile gas works facility and a former wood impregnation site. In contrast, sources of sediment samples were a former coal refining area and an old fuel depot. Samples were incubated in anoxic mineral salt medium with naphthalene as sole carbon source and sulfate as terminal electron acceptor. Grown cultures were characterized by feeding with 13C-labeled naphthalene, 16S rRNA gene sequencing using an Illumina® approach, and functional proteome analyses. Finally, six enrichment cultures able to degrade naphthalene under anoxic conditions were established. First results point to a dominance of identified sequences affiliated to the freshwater sulfate-reducing strain N47, which is a known anaerobic naphthalene degrader, in four out of the six enrichments. In those enrichments, peptides related to the pathway of anoxic naphthalene degradation in N47 were abundant. Overall the data underlines the importance of Desulfobacteria for natural

  4. Air-surface exchange measurements of gaseous elemental mercury over naturally enriched and background terrestrial landscapes in Australia

    NASA Astrophysics Data System (ADS)

    Edwards, G. C.; Howard, D. A.

    2013-05-01

    This paper presents the first gaseous elemental mercury (GEM) air-surface exchange measurements obtained over naturally enriched and background (<0.1 μg g-1 Hg) terrestrial landscapes in Australia. Two pilot field studies were carried out during the Australian autumn and winter periods at a copper-gold-cobalt-arsenic-mercury mineral field near Pulganbar, NSW. GEM fluxes using a dynamic flux chamber approach were measured, along with controlling environmental parameters over three naturally enriched and three background substrates. The enriched sites results showed net emission to the atmosphere and a strong correlation between flux and substrate Hg concentration, with average fluxes ranging from 14 ± 1 ng m-2 h-1 to 113 ± 6 ng m-2 h-1. Measurements at background sites showed both emission and deposition. The average Hg flux from all background sites showed an overall net emission of 0.36 ± 0.06 ng m-2 h-1. Fluxes show strong relationships with temperature, radiation, and substrate parameters. A compensation point of 2.48, representative of bare soils was determined. For periods of deposition, dry deposition velocities ranged from 0.00025 cm s-1 to 0.0083 cm s-1 with an average of 0.0041 ± 0.00018 cm s-1, representing bare soil, nighttime conditions. Comparison of the Australian data to North American data suggests the need for Australian-specific mercury air-surface exchange data representative of Australia's unique climatic conditions, vegetation types, land use patterns and soils.

  5. Cooking rice in excess water reduces both arsenic and enriched vitamins in the cooked grain.

    PubMed

    Gray, Patrick J; Conklin, Sean D; Todorov, Todor I; Kasko, Sasha M

    2016-01-01

    This paper reports the effects of rinsing rice and cooking it in variable amounts of water on total arsenic, inorganic arsenic, iron, cadmium, manganese, folate, thiamin and niacin in the cooked grain. We prepared multiple rice varietals both rinsed and unrinsed and with varying amounts of cooking water. Rinsing rice before cooking has a minimal effect on the arsenic (As) content of the cooked grain, but washes enriched iron, folate, thiamin and niacin from polished and parboiled rice. Cooking rice in excess water efficiently reduces the amount of As in the cooked grain. Excess water cooking reduces average inorganic As by 40% from long grain polished, 60% from parboiled and 50% from brown rice. Iron, folate, niacin and thiamin are reduced by 50-70% for enriched polished and parboiled rice, but significantly less so for brown rice, which is not enriched.

  6. Neutron Characterization of Encapsulated ATF-1/LANL-1 Mockup Fuel Capsules

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Vogel, Sven C.; Borges, Nicholas Paul; Losko, Adrian Simon

    Twenty pellets of mock-up accident tolerant fuels UN-U3Si5 were produced at LANL and loaded in two rodlet/capsule assemblies. Tomographic imaging and diffraction measurements were performed to characterize these samples at the Flight-Path 5 and HIPPO beam lines at LANSCE/LANL between November 2016 and January 2017 as well as in August 2017. The entire ~10 cm long, ~1 cm diameter fuel volume could be characterized, however due to time constraints only 2 mm slices in 4mm increments were characterized with neutron diffraction and a 28mm subset of the entire sample was characterized with energy-resolved neutron imaging. The double encapsulation of themore » fuel into two steel containers does not pose a problem for the neutron analysis and the methods could be applied to enriched as well irradiated fuels.« less

  7. The Development of Environmentally Friendly Technologies of Using Coals and Products of Their Enrichment in the Form of Coal Water Slurries

    NASA Astrophysics Data System (ADS)

    Murko, Vasily; Hamalainen, Veniamin

    2017-11-01

    The article presents the current state of the technology for production and combustion of fuel coal water slurries in Russia and foreign countries. Experimental and industrial facilities show the technological and economic efficiency of using this technology for disposal of wastes resulting after coal processing and enrichment. The feasibility studies of use of the technology at large Kuzbass thermal power stations are presented. The possibility of solving a serious environmental problem of reducing storage of the most toxic waste of coal enrichment in the location areas of coal washing plants and coal mining enterprises is demonstrated.

  8. Production, purification and utilization of biogas as fuel for internal combustion engine

    NASA Astrophysics Data System (ADS)

    Hernandez, Noel M.; Villanueva, Eliseo P.

    2018-03-01

    This study attempts to modify a 4-cylinder gasoline engine to run with a purified compressed biogas as substitute for fossil fuels. Water scrubbing method was used as the easiest purification technique to remove CO2 and iron filing for H2S. The pressurized raw biogas was fed in a low cost made portable floating type gas holder with volume capacity of 0.74 m3. The purified biogas was compressed using a reciprocating compressor through a two stage series of enrichment and moisture removal process using activated alumina into the steel cylinder to improve the quality of the methane content. The enriched biogas was filled in the LPG tank for 20 minutes at 10 bars at an average of 73.67% CH4 with no traces of H2S as storage for engine utilization. The modification involved the installation and mounting of LPG conversion kit. A comparative analysis of the performance and combustion characteristics of the engine was evaluated separately with gasoline and purified compressed biogas using electro-dynamometer as variable loads. The findings show that power output deterioration in compressed biogas was mainly due to high percentage of CO2 and other gases impurities. It also shows that because of the calorific value of biogas, the thermal efficiency is lesser than that of gasoline. It implies that the overall engine performance can be improved by removing undesirable gases in the mixture.

  9. Methodology and Software for Gross Defect Detection of Spent Nuclear Fuel at the Atucha-I Reactor [Novel Methodology and Software for Spent Fuel Gross Defect Detection at the Atucha-I Reactor

    DOE PAGES

    Sitaraman, Shivakumar; Ham, Young S.; Gharibyan, Narek; ...

    2017-03-27

    Here, fuel assemblies in the spent fuel pool are stored by suspending them in two vertically stacked layers at the Atucha Unit 1 nuclear power plant (Atucha-I). This introduces the unique problem of verifying the presence of fuel in either layer without physically moving the fuel assemblies. Given that the facility uses both natural uranium and slightly enriched uranium at 0.85 wt% 235U and has been in operation since 1974, a wide range of burnups and cooling times can exist in any given pool. A gross defect detection tool, the spent fuel neutron counter (SFNC), has been used at themore » site to verify the presence of fuel up to burnups of 8000 MWd/t. At higher discharge burnups, the existing signal processing software of the tool was found to fail due to nonlinearity of the source term with burnup.« less

  10. Methodology and Software for Gross Defect Detection of Spent Nuclear Fuel at the Atucha-I Reactor [Novel Methodology and Software for Spent Fuel Gross Defect Detection at the Atucha-I Reactor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Sitaraman, Shivakumar; Ham, Young S.; Gharibyan, Narek

    Here, fuel assemblies in the spent fuel pool are stored by suspending them in two vertically stacked layers at the Atucha Unit 1 nuclear power plant (Atucha-I). This introduces the unique problem of verifying the presence of fuel in either layer without physically moving the fuel assemblies. Given that the facility uses both natural uranium and slightly enriched uranium at 0.85 wt% 235U and has been in operation since 1974, a wide range of burnups and cooling times can exist in any given pool. A gross defect detection tool, the spent fuel neutron counter (SFNC), has been used at themore » site to verify the presence of fuel up to burnups of 8000 MWd/t. At higher discharge burnups, the existing signal processing software of the tool was found to fail due to nonlinearity of the source term with burnup.« less

  11. STUDIES OF FAST REACTOR FUEL ELEMENT BEHAVIOR UNDER TRANSIENT HEATING TO FAILURE. I. INITIAL EXPERIMENTS ON METALLIC SAMPLES IN THE ABSENCE OF COOLANT

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Dickerman, C. E.; Sowa, E. S.; Okrent, D.

    1961-08-01

    Meltdown tests on single metallic unirradiated fuel elements in TREAT are described. The fuel elements (EBRII Mark I fuel pins, EBR-II fuel pins with retractory Nb or Ta cladding, and Fermi-I fuel pins) are tested in an inert atmosphere, with no coolant. The fuel elements are exposed to reactor power bursts of 200 msec to 25 sec duration, under conditions simulating fast reactor operations. For these tests, the type of power burst, the integrated power, the fuel enrichment, the maximum cladding temperature, and the effects of the test on the fuel element are recorded. ( T.F.H.)

  12. Reporting Codes and Fuel Pathways for the EPA Moderated Transaction System (EMTS)

    EPA Pesticide Factsheets

    Users should reference this document for a complete list of all reporting codes and all possible fuel pathways for Renewable Fuel Standard (RFS) and Fuels Averaging, Banking and Trading (ABT) users of the EPA Moderated Transaction System (EMTS).

  13. Neutron diffraction measurement of residual stresses, dislocation density and texture in Zr-bonded U-10Mo “mini” fuel foils and plates

    DOE PAGES

    Brown, Donald William; Okuniewski, Maria A.; Sisneros, Thomas A.; ...

    2016-12-01

    Here, Al clad U-10Mo fuel plates are being considered for conversion of several research reactors from high-enriched to low-enriched U fuel. Neutron diffraction measurements of the textures, residual phase stresses, and dislocation densities in the individual phases of the mini-foils throughout several processing steps and following hot-isostatic pressing to the Al cladding, have been completed. Recovery and recrystallization of the bare U-10Mo fuel foil, as indicated by the dislocation density and texture, are observed depending on the state of the material prior to annealing and the duration and temperature of the annealing process. In general, the cladding procedure significantly reducesmore » the dislocation density, but the final state of the clad plate, both texture and dislocation density, depends strongly on the final processing step of the fuel foil. In contrast, the residual stress state of the final plate is dominated by the thermal expansion mismatch of the constituent materials.« less

  14. Neutron diffraction measurement of residual stresses, dislocation density and texture in Zr-bonded U-10Mo “mini” fuel foils and plates

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Brown, Donald William; Okuniewski, Maria A.; Sisneros, Thomas A.

    Here, Al clad U-10Mo fuel plates are being considered for conversion of several research reactors from high-enriched to low-enriched U fuel. Neutron diffraction measurements of the textures, residual phase stresses, and dislocation densities in the individual phases of the mini-foils throughout several processing steps and following hot-isostatic pressing to the Al cladding, have been completed. Recovery and recrystallization of the bare U-10Mo fuel foil, as indicated by the dislocation density and texture, are observed depending on the state of the material prior to annealing and the duration and temperature of the annealing process. In general, the cladding procedure significantly reducesmore » the dislocation density, but the final state of the clad plate, both texture and dislocation density, depends strongly on the final processing step of the fuel foil. In contrast, the residual stress state of the final plate is dominated by the thermal expansion mismatch of the constituent materials.« less

  15. A compact tritium enrichment unit for large sample volumes with automated re-filling and higher enrichment factor.

    PubMed

    Kumar, B; Han, L-F; Wassenaar, L I; Klaus, P M; Kainz, G G; Hillegonds, D; Brummer, D; Ahmad, M; Belachew, D L; Araguás, L; Aggarwal, P

    2016-12-01

    Tritium ( 3 H) in natural waters is a powerful tracer of hydrological processes, but its low concentrations require electrolytic enrichment before precise measurements can be made with a liquid scintillation counter. Here, we describe a newly developed, compact tritium enrichment unit which can be used to enrich up to 2L of a water sample. This allows a high enrichment factor (>100) for measuring low 3 H contents of <0.05TU. The TEU uses a small cell (250mL) with automated re-filling and a CO 2 bubbling technique to neutralize the high alkalinity of enriched samples. The enriched residual sample is retrieved from the cell under vacuum by cryogenic distillation at -20°C and the tritium enrichment factor for each sample is accurately determined by measuring pre- and post- enrichment 2 H concentrations with laser spectrometry. Copyright © 2016. Published by Elsevier Ltd.

  16. 40 CFR 600.311-12 - Determination of values for fuel economy labels.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... value from paragraph (a) of this section, in miles per kW-hour. (4) For hydrogen fuel cell vehicles... for hydrogen fuel cell vehicles. MPG = The combined fuel economy value from paragraph (a) of this... hydrogen fuel cell vehicles. Average Annual Miles = The estimated annual mileage figure provided by EPA, in...

  17. 40 CFR 600.311-12 - Determination of values for fuel economy labels.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... value from paragraph (a) of this section, in miles per kW-hour. (4) For hydrogen fuel cell vehicles... for hydrogen fuel cell vehicles. MPG = The combined fuel economy value from paragraph (a) of this... hydrogen fuel cell vehicles. Average Annual Miles = The estimated annual mileage figure provided by EPA, in...

  18. Quantification of histone modification ChIP-seq enrichment for data mining and machine learning applications

    PubMed Central

    2011-01-01

    Background The advent of ChIP-seq technology has made the investigation of epigenetic regulatory networks a computationally tractable problem. Several groups have applied statistical computing methods to ChIP-seq datasets to gain insight into the epigenetic regulation of transcription. However, methods for estimating enrichment levels in ChIP-seq data for these computational studies are understudied and variable. Since the conclusions drawn from these data mining and machine learning applications strongly depend on the enrichment level inputs, a comparison of estimation methods with respect to the performance of statistical models should be made. Results Various methods were used to estimate the gene-wise ChIP-seq enrichment levels for 20 histone methylations and the histone variant H2A.Z. The Multivariate Adaptive Regression Splines (MARS) algorithm was applied for each estimation method using the estimation of enrichment levels as predictors and gene expression levels as responses. The methods used to estimate enrichment levels included tag counting and model-based methods that were applied to whole genes and specific gene regions. These methods were also applied to various sizes of estimation windows. The MARS model performance was assessed with the Generalized Cross-Validation Score (GCV). We determined that model-based methods of enrichment estimation that spatially weight enrichment based on average patterns provided an improvement over tag counting methods. Also, methods that included information across the entire gene body provided improvement over methods that focus on a specific sub-region of the gene (e.g., the 5' or 3' region). Conclusion The performance of data mining and machine learning methods when applied to histone modification ChIP-seq data can be improved by using data across the entire gene body, and incorporating the spatial distribution of enrichment. Refinement of enrichment estimation ultimately improved accuracy of model predictions. PMID

  19. Occupational safety data and casualty rates for the uranium fuel cycle. [Glossaries

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    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/supmore » 12/ Btu of energy output, and per other appropriate units of output.« less

  20. 31 CFR 540.316 - Uranium enrichment.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 31 Money and Finance:Treasury 3 2013-07-01 2013-07-01 false Uranium enrichment. 540.316 Section... FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.316 Uranium enrichment. The term uranium enrichment means the process of...

  1. 31 CFR 540.316 - Uranium enrichment.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 31 Money and Finance:Treasury 3 2014-07-01 2014-07-01 false Uranium enrichment. 540.316 Section... FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.316 Uranium enrichment. The term uranium enrichment means the process of...

  2. 31 CFR 540.316 - Uranium enrichment.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 31 Money and Finance:Treasury 3 2011-07-01 2011-07-01 false Uranium enrichment. 540.316 Section... FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.316 Uranium enrichment. The term uranium enrichment means the process of...

  3. 31 CFR 540.316 - Uranium enrichment.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 31 Money and Finance:Treasury 3 2012-07-01 2012-07-01 false Uranium enrichment. 540.316 Section... FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.316 Uranium enrichment. The term uranium enrichment means the process of...

  4. An Optically Stimulated Luminescence Uranium Enrichment Monitor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Miller, Steven D.; Tanner, Jennifer E.; Simmons, Kevin L.

    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 themore » 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

  5. Partially-reflected water-moderated square-piteched U(6.90)O 2 fuel rod lattices with 0.67 fuel to water volume ratio (0.800 CM Pitch)

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Harms, Gary A.

    The US Department of Energy (DOE) Nuclear Energy Research Initiative funded the design and construction of the Seven Percent Critical Experiment (7uPCX) at Sandia National Laboratories. The start-up of the experiment facility and the execution of the experiments described here were funded by the DOE Nuclear Criticality Safety Program. The 7uPCX is designed to investigate critical systems with fuel for light water reactors in the enrichment range above 5% 235U. The 7uPCX assembly is a water-moderated and -reflected array of aluminum-clad square-pitched U(6.90%)O 2 fuel rods.

  6. Fuel/air nonuniformity - Effect on nitric oxide emissions

    NASA Technical Reports Server (NTRS)

    Lyons, V. J.

    1981-01-01

    An analytical and experimental study was performed to determine the effect of inlet fuel/air profile nonuniformity on NO(x) emissions. The theoretical NO(x) levels were verified in a flame-tube rig at inlet air temperatures of 600, 700, and 800 K, 0.3 MPa rig pressure, 25 m/sec reference velocity, overall equivalence ratio of 0.6 and residence time near 0.002 sec. The theory predicts an increase in NO(x) emissions for increased fuel/air nonuniformity for average equivalence ratios less than 0.7, while for average equivalence ratios near stoichiometric, increasing the nonuniformity will decrease NO(x) emissions. The results can be used to predict the degree of uniformity of fuel/air profiles necessary to achieve NO(x) emissions goals for actual engines that use lean premixed, prevaporized combustion systems.

  7. Space reactor fuel element testing in upgraded TREAT

    NASA Astrophysics Data System (ADS)

    Todosow, M.; Bezler, P.; Ludewig, H.; Kato, W. Y.

    The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc.; a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR); NERVA-derivative; and other concepts are discussed. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggest that full-scale PBR elements could be tested at an average energy deposition of approximately 60-80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of approximately 100 MW/L may be achievable.

  8. Space reactor fuel element testing in upgraded TREAT

    NASA Astrophysics Data System (ADS)

    Todosow, Michael; Bezler, Paul; Ludewig, Hans; Kato, Walter Y.

    1993-01-01

    The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., is a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggests that full-scale PBR elements could be tested at an average energy deposition of ˜60-80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperture limit, average energy deposition of ˜100 MW/L may be achievable.

  9. Average biomass of four Northwest shrubs by fuel size class and crown cover.

    Treesearch

    Robert E. Martin; David W. Frewing; James L. McClanahan

    1981-01-01

    The average biomass of big sagebrush (Artemisia tridentata Nutt.), antelope bitterbrush (Purshia tridentata (Pursh) DC.), snowbrush ceanothus (Ceanothus velutinus Dougl. ex Hook.), and greenleaf manzanita (Arctostaphylos patula Greene) was 6.1, 5.1, 10.7, and 16.2 tons per acre (13.9,...

  10. Recent trends in aviation turbine fuel properties

    NASA Technical Reports Server (NTRS)

    Friedman, R.

    1982-01-01

    Plots and tables, compiled from Department of Energy (and predecessor agency) inspection reports from 1969 to 1980, present ranges, averages, extremes, and trends for most of the 22 properties of Jet A aviation turbine fuel. In recent years, average values of aromatics content, mercaptan sulfur content, distillation temperature of 10 percent recovered, smoke point, and freezing point show small but recognizable trends toward their specification limits. About 80 percent of the fuel samples had at least one property near specification, defined as within a standard band about the specification limit. By far the most common near-specification properties were aromatics content, smoke point, and freezing point.

  11. Neutronics Investigations for the Lower Part of a Westinghouse SVEA-96+ Assembly

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Murphy, M.F.; Luethi, A.; Seiler, R.

    2002-05-15

    Accurate critical experiments have been performed for the validation of total fission (F{sub tot}) and {sup 238}U-capture (C{sub 8}) reaction rate distributions obtained with CASMO-4, HELIOS, BOXER, and MCNP4B for the lower axial region of a real Westinghouse SVEA-96+ fuel assembly. The assembly comprised fresh fuel with an average {sup 235}U enrichment of 4.02 wt%, a maximum enrichment of 4.74 wt%, 14 burnable-absorber fuel pins, and full-density water moderation. The experimental configuration investigated was core 1A of the LWR-PROTEUS Phase I project, where 61 different fuel pins, representing {approx}64% of the assembly, were gamma-scanned individually. Calculated (C) and measured (E)more » values have been compared in terms of C/E distributions. For F{sub tot}, the standard deviations are 1.2% for HELIOS, 0.9% for CASMO-4, 0.8% for MCNP4B, and 1.7% for BOXER. Standard deviations of 1.1% for HELIOS, CASMO-4, and MCNP4B and 1.2% for BOXER were obtained in the case of C{sub 8}. Despite the high degree of accuracy observed on the average, it was found that the five burnable-absorber fuel pins investigated showed a noticeable underprediction of F{sub tot}, quite systematically, for the deterministic codes evaluated (average C/E for the burnable-absorber fuel pins in the range 0.974 to 0.988, depending on the code)« less

  12. Comprehensive viral enrichment enables sensitive respiratory virus genomic identification and analysis by next generation sequencing.

    PubMed

    O'Flaherty, Brigid M; Li, Yan; Tao, Ying; Paden, Clinton R; Queen, Krista; Zhang, Jing; Dinwiddie, Darrell L; Gross, Stephen M; Schroth, Gary P; Tong, Suxiang

    2018-06-01

    Next generation sequencing (NGS) technologies have revolutionized the genomics field and are becoming more commonplace for identification of human infectious diseases. However, due to the low abundance of viral nucleic acids (NAs) in relation to host, viral identification using direct NGS technologies often lacks sufficient sensitivity. Here, we describe an approach based on two complementary enrichment strategies that significantly improves the sensitivity of NGS-based virus identification. To start, we developed two sets of DNA probes to enrich virus NAs associated with respiratory diseases. The first set of probes spans the genomes, allowing for identification of known viruses and full genome sequencing, while the second set targets regions conserved among viral families or genera, providing the ability to detect both known and potentially novel members of those virus groups. Efficiency of enrichment was assessed by NGS testing reference virus and clinical samples with known infection. We show significant improvement in viral identification using enriched NGS compared to unenriched NGS. Without enrichment, we observed an average of 0.3% targeted viral reads per sample. However, after enrichment, 50%-99% of the reads per sample were the targeted viral reads for both the reference isolates and clinical specimens using both probe sets. Importantly, dramatic improvements on genome coverage were also observed following virus-specific probe enrichment. The methods described here provide improved sensitivity for virus identification by NGS, allowing for a more comprehensive analysis of disease etiology. © 2018 O'Flaherty et al.; Published by Cold Spring Harbor Laboratory Press.

  13. Effects of heat treatment on U–Mo fuel foils with a zirconium diffusion barrier

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Jue, Jan-Fong; Trowbridge, Tammy L.; Breckenridge, Cynthia R.

    A monolith fuel design based on U–Mo alloy has been selected as the fuel type for conversion of the United States’ high performance research reactors (HPRRs) from highly enriched uranium (HEU) to low-enriched uranium (LEU). In this fuel design, a thin layer of zirconium is used to eliminate the direct interaction between the U–Mo fuel meat and the aluminum-alloy cladding during irradiation. The co-rolling process used to bond the Zr barrier layer to the U–Mo foil during fabrication alters the microstructure of both the U–10Mo fuel meat and the U–Mo/Zr interface. This work studied the effects of post-rolling annealing treatmentmore » on the microstructure of the co-rolled U–Mo fuel meat and the U–Mo/Zr interaction layer. Microscopic characterization shows that the grain size of U–Mo fuel meat increases with the annealing temperature, as expected. The grain sizes were ~9, ~13, and ~20 μm for annealing temperature of 650, 750, and 850 °C, respectively. No abnormal grain growth was observed. The U–Mo/Zr interaction-layer thickness increased with the annealing temperature with an Arrhenius constant for growth of 184 kJ/mole, consistent with a previous diffusion-couple study. The interaction layer thickness was 3.2 ± 0.5 μm, 11.1 ± 2.1 μm, 27.1 ± 0.9 μm for annealing temperature of 650, 750, to 850 °C, respectively. The homogeneity of Mo improves with post rolling annealing temperature and with U–Mo coupon homogenization. The phases in the Zr/U–Mo interaction layer produced by co-rolling, however, differ from those reported in the previous diffusion couple studies.« less

  14. Effects of heat treatment on U-Mo fuel foils with a zirconium diffusion barrier

    NASA Astrophysics Data System (ADS)

    Jue, Jan-Fong; Trowbridge, Tammy L.; Breckenridge, Cynthia R.; Moore, Glenn A.; Meyer, Mitchell K.; Keiser, Dennis D.

    2015-05-01

    A monolith fuel design based on U-Mo alloy has been selected as the fuel type for conversion of the United States' high performance research reactors (HPRRs) from highly enriched uranium (HEU) to low-enriched uranium (LEU). In this fuel design, a thin layer of zirconium is used to eliminate the direct interaction between the U-Mo fuel meat and the aluminum-alloy cladding during irradiation. The co-rolling process used to bond the Zr barrier layer to the U-Mo foil during fabrication alters the microstructure of both the U-10Mo fuel meat and the U-Mo/Zr interface. This work studied the effects of post-rolling annealing treatment on the microstructure of the co-rolled U-Mo fuel meat and the U-Mo/Zr interaction layer. Microscopic characterization shows that the grain size of U-Mo fuel meat increases with the annealing temperature, as expected. The grain sizes were ∼9, ∼13, and ∼20 μm for annealing temperature of 650, 750, and 850 °C, respectively. No abnormal grain growth was observed. The U-Mo/Zr interaction-layer thickness increased with the annealing temperature with an Arrhenius constant for growth of 184 kJ/mole, consistent with a previous diffusion-couple study. The interaction layer thickness was 3.2 ± 0.5 μm, 11.1 ± 2.1 μm, 27.1 ± 0.9 μm for annealing temperature of 650, 750, to 850 °C, respectively. The homogeneity of Mo improves with post rolling annealing temperature and with U-Mo coupon homogenization. The phases in the Zr/U-Mo interaction layer produced by co-rolling, however, differ from those reported in the previous diffusion couple studies.

  15. Irradiation of TZM: Uranium dioxide fuel pin at 1700 K

    NASA Technical Reports Server (NTRS)

    Mcdonald, G. E.

    1973-01-01

    A fuel pin clad with TZM and containing solid pellets of uranium dioxide was fission heated in a static helium-cooled capsule at a maximum surface temperature of 1700 K for approximately 1000 hr and to a total burnup of 2.0 percent of the uranium-235. The results of the postirradiation examination indicated: (1) A transverse, intergranular failure of the fuel pin occurred when the fuel pin reached 2.0-percent burnup. This corresponds to 1330 kW-hr/cu cm, where the volume is the sum of the fuel, clad, and void volumes in the fuel region. (2) The maximum swelling of the fuel pin was less than 1.5 percent on the fuel-pin diameter. (3) There was no visible interaction between the TZM clad and the UO2. (4) Irradiation at 1700 K produced a course-grained structure, with an average grain diameter of 0.02 centimeter and with some of the grains extending one-half of the thickness of the clad. (5) Below approximately 1500 K, the irradiation of the clad produced a moderately fine-grained structure, with an average grain diameter of 0.004 centimeter.

  16. Afternoon Remedial and Enrichment Program, Buffalo, New York. Elementary Program in Compensatory Education, 2.

    ERIC Educational Resources Information Center

    American Institutes for Research in the Behavioral Sciences, Palo Alto, CA.

    This afternoon remedial and enrichment program was offered to inner-city low income children (grades 3-8). About 75 percent of the children were black, 20 percent white and 5 percent Puerto Rican. Remedial instruction was offered in reading and mathematics. Average class size was six pupils; these small groups allowed for better diagnosis of needs…

  17. Decoupling of DAMO archaea from DAMO bacteria in a methane-driven microbial fuel cell.

    PubMed

    Ding, Jing; Lu, Yong-Ze; Fu, Liang; Ding, Zhao-Wei; Mu, Yang; Cheng, Shuk H; Zeng, Raymond J

    2017-03-01

    Anaerobic oxidation of methane (AOM) contributes significantly to the global methane sink. Previously, studies of anaerobic methanotrophic (ANME) archaea have been limited as they have not been separable from their bacterial partners during the AOM process because of their dependence on the bacteria. A microbial fuel cell (MFC) is a device capable of directly transforming chemical energy to electrical energy via electrochemical reactions involving biochemical pathways. In this study, decoupling of denitrifying anaerobic methane oxidation (DAMO) archaea and DAMO bacteria was investigated in an microbial fuel cell (MFC) using methane as the fuel. The DAMO fuel cell worked successfully but demonstrated weak electrogenic capability with around 25 mV production. After 45 days' enrichment, the sequencing and fluorescence in situ hybridization results showed the DAMO archaea percentage had increased from 26.96% (inoculum) to 65.77% (electrode biofilm), while the DAMO bacteria percentage decreased from 24.39% to 2.07%. Moreover, the amount of ANME-2d had doubled in the electrode biofilm compared with the inoculum. The sequencing results also showed substantial enrichment of the Ignavibacterium and Geobacter genera. The roles of Ignavibacterium and Geobacter in the MFC system need to be further investigated. Nevertheless, these results illustrate that an MFC device may provide a possible approach to separate DAMO archaea from DAMO bacteria. Copyright © 2016 Elsevier Ltd. All rights reserved.

  18. 21 CFR 137.165 - Enriched flour.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ... 21 Food and Drugs 2 2011-04-01 2011-04-01 false Enriched flour. 137.165 Section 137.165 Food and... Related Products § 137.165 Enriched flour. Enriched flour conforms to the definition and standard of.... Enriched flour may be acidified with monocalcium phosphate within the limits prescribed by § 137.175 for...

  19. 21 CFR 137.165 - Enriched flour.

    Code of Federal Regulations, 2012 CFR

    2012-04-01

    ... 21 Food and Drugs 2 2012-04-01 2012-04-01 false Enriched flour. 137.165 Section 137.165 Food and... Related Products § 137.165 Enriched flour. Enriched flour conforms to the definition and standard of.... Enriched flour may be acidified with monocalcium phosphate within the limits prescribed by § 137.175 for...

  20. 21 CFR 137.165 - Enriched flour.

    Code of Federal Regulations, 2014 CFR

    2014-04-01

    ... 21 Food and Drugs 2 2014-04-01 2014-04-01 false Enriched flour. 137.165 Section 137.165 Food and... Related Products § 137.165 Enriched flour. Enriched flour conforms to the definition and standard of.... Enriched flour may be acidified with monocalcium phosphate within the limits prescribed by § 137.175 for...

  1. 21 CFR 137.165 - Enriched flour.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    ... 21 Food and Drugs 2 2013-04-01 2013-04-01 false Enriched flour. 137.165 Section 137.165 Food and... Related Products § 137.165 Enriched flour. Enriched flour conforms to the definition and standard of.... Enriched flour may be acidified with monocalcium phosphate within the limits prescribed by § 137.175 for...

  2. Selection of Nuclear Fuel for TREAT: UO 2 vs U 3O 8

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Glazoff, Michael Vasily; Van Rooyen, Isabella Johanna; Coryell, Benjamin David

    The Transient Reactor Test (TREAT) that resides at the Materials and Fuels Complex (MFC) at Idaho National Laboratory (INL), first achieved criticality in 1959, and successfully performed many transient tests on nuclear fuel until 1994 when its operations were suspended. Resumption of operations at TREAT was approved in February 2014 to meet the U.S. Department of Energy (DOE) Office of Nuclear Energy’s objectives in transient testing of nuclear fuels. The National Nuclear Security Administration’s is converting TREAT from its existing highly enriched uranium (HEU) core to a new core containing low enriched uranium (LEU) (i.e., U-235< 20% by weight). Themore » TREAT Conversion project is currently progressing with conceptual design phase activities. Dimensional stability of the fuel element assemblies, predictable fuel can oxidation and sufficient heat conductivity by the fuel blocks are some of the critical performance requirements of the new LEU fuel. Furthermore, to enable the design team to design fuel block and can specifications, it is amongst the objectives to evaluate TREAT LEU fuel and cladding material’s chemical interaction. This information is important to understand the viability of Zr-based alloys and fuel characteristics for the fabrication of the TREAT LEU fuel and cladding. Also, it is very important to make the right decision on what type of nuclear fuel will be used at TREAT. In particular, one has to consider different oxides of uranium, and most importantly, UO 2 vs U 3O 8. In this report, the results are documented pertaining to the choice mentioned above (UO 2 vs U 3O 8). The conclusion in favor of using UO 2 was made based on the analysis of historical data, up-to-date literature, and self-consistent calculations of phase equilibria and thermodynamic properties in the U-O and U-O-C systems. The report is organized as follows. First, the criteria that were used to make the choice are analyzed. Secondly, existing historical data and

  3. Toxicity of irradiated advanced heavy water reactor fuels.

    PubMed

    Priest, N D; Richardson, R B; Edwards, G W R

    2013-02-01

    The good neutron economy and online refueling capability of the CANDU® heavy water moderated reactor (HWR) enable it to use many different fuels such as low enriched uranium (LEU), plutonium, or thorium, in addition to its traditional natural uranium (NU) fuel. The toxicity and radiological protection methods for these proposed fuels, unlike those for NU, are not well established. This study uses software to compare the fuel composition and toxicity of irradiated NU fuel against those of two irradiated advanced HWR fuel bundles as a function of post-irradiation time. The first bundle investigated is a CANFLEX® low void reactor fuel (LVRF), of which only the dysprosium-poisoned central element, and not the outer 42 LEU elements, is specifically analyzed. The second bundle investigated is a heterogeneous high-burnup (LEU,Th)O(2) fuelled bundle, whose two components (LEU in the outer 35 elements and thorium in the central eight elements) are analyzed separately. The LVRF central element was estimated to have a much lower toxicity than that of NU at all times after shutdown. Both the high burnup LEU and the thorium fuel had similar toxicity to NU at shutdown, but due to the creation of such inhalation hazards as (238)Pu, (240)Pu, (242)Am, (242)Cm, and (244)Cm (in high burnup LEU), and (232)U and (228)Th (in irradiated thorium), the toxicity of these fuels was almost double that of irradiated NU after 2,700 d of cooling. New urine bioassay methods for higher actinoids and the analysis of thorium in fecal samples are recommended to assess the internal dose from these two fuels.

  4. 21 CFR 139.115 - Enriched macaroni products.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ... defatted wheat germ, enriched farina, or enriched flour, or through the direct additions of any of the... 21 Food and Drugs 2 2011-04-01 2011-04-01 false Enriched macaroni products. 139.115 Section 139... and Noodle Products § 139.115 Enriched macaroni products. (a) Description. Enriched macaroni products...

  5. 21 CFR 139.115 - Enriched macaroni products.

    Code of Federal Regulations, 2012 CFR

    2012-04-01

    ... defatted wheat germ, enriched farina, or enriched flour, or through the direct additions of any of the... 21 Food and Drugs 2 2012-04-01 2012-04-01 false Enriched macaroni products. 139.115 Section 139... and Noodle Products § 139.115 Enriched macaroni products. (a) Description. Enriched macaroni products...

  6. 21 CFR 139.115 - Enriched macaroni products.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    ... defatted wheat germ, enriched farina, or enriched flour, or through the direct additions of any of the... 21 Food and Drugs 2 2013-04-01 2013-04-01 false Enriched macaroni products. 139.115 Section 139... and Noodle Products § 139.115 Enriched macaroni products. (a) Description. Enriched macaroni products...

  7. 21 CFR 139.115 - Enriched macaroni products.

    Code of Federal Regulations, 2014 CFR

    2014-04-01

    ... defatted wheat germ, enriched farina, or enriched flour, or through the direct additions of any of the... 21 Food and Drugs 2 2014-04-01 2014-04-01 false Enriched macaroni products. 139.115 Section 139... and Noodle Products § 139.115 Enriched macaroni products. (a) Description. Enriched macaroni products...

  8. Local finite element enrichment strategies for 2D contact computations and a corresponding post-processing scheme

    NASA Astrophysics Data System (ADS)

    Sauer, Roger A.

    2013-08-01

    Recently an enriched contact finite element formulation has been developed that substantially increases the accuracy of contact computations while keeping the additional numerical effort at a minimum reported by Sauer (Int J Numer Meth Eng, 87: 593-616, 2011). Two enrich-ment strategies were proposed, one based on local p-refinement using Lagrange interpolation and one based on Hermite interpolation that produces C 1-smoothness on the contact surface. Both classes, which were initially considered for the frictionless Signorini problem, are extended here to friction and contact between deformable bodies. For this, a symmetric contact formulation is used that allows the unbiased treatment of both contact partners. This paper also proposes a post-processing scheme for contact quantities like the contact pressure. The scheme, which provides a more accurate representation than the raw data, is based on an averaging procedure that is inspired by mortar formulations. The properties of the enrichment strategies and the corresponding post-processing scheme are illustrated by several numerical examples considering sliding and peeling contact in the presence of large deformations.

  9. Local geology determines responses of stream producers and fungal decomposers to nutrient enrichment: A field experiment.

    PubMed

    Mykrä, Heikki; Sarremejane, Romain; Laamanen, Tiina; Karjalainen, Satu Maaria; Markkola, Annamari; Lehtinen, Sirkku; Lehosmaa, Kaisa; Muotka, Timo

    2018-04-16

    We examined how short-term (19 days) nutrient enrichment influences stream fungal and diatom communities, and rates of leaf decomposition and algal biomass accrual. We conducted a field experiment using slow-releasing nutrient pellets to increase nitrate (NO 3 -N) and phosphate (PO 4 -P) concentrations in a riffle section of six naturally acidic (naturally low pH due to catchment geology) and six circumneutral streams. Nutrient enrichment increased microbial decomposition rate on average by 14%, but the effect was significant only in naturally acidic streams. Nutrient enrichment also decreased richness and increased compositional variability of fungal communities in naturally acidic streams. Algal biomass increased in both stream types, but algal growth was overall very low. Diatom richness increased in response to nutrient addition by, but only in circumneutral streams. Our results suggest that primary producers and decomposers are differentially affected by nutrient enrichment and that their responses to excess nutrients are context dependent, with a potentially stronger response of detrital processes and fungal communities in naturally acidic streams than in less selective environments.

  10. Air-surface exchange measurements of gaseous elemental mercury over naturally enriched and background terrestrial landscapes in Australia

    NASA Astrophysics Data System (ADS)

    Edwards, G. C.; Howard, D. A.

    2012-10-01

    This paper presents the first gaseous elemental mercury (GEM) air-surface exchange measurements obtained over naturally enriched and background (< 0.1 μg g-1 Hg) terrestrial landscapes in Australia. Two pilot field studies were carried out during the Australian autumn and winter periods at a copper-gold-cobalt-arsenic-mercury mineral field near Pulganbar, NSW. GEM fluxes using a dynamic flux chamber approach were measured, along with controlling environmental parameters over three naturally enriched and three background substrates. The enriched sites results showed net emission to the atmosphere and a strong correlation between flux and substrate Hg concentration, with average fluxes ranging from 14 ± 1 ng m-2 h-1 to 113 ± 6 ng m-2 h-1. Measurements at background sites showed both emission and deposition. The average Hg flux from all background sites showed an overall net emission of 0.36 ± 0.06 ng m-2 h-1. Fluxes show strong relationships with temperature, radiation, and substrate parameters. A compensation point of 2.48, representative of bare soils was determined. Comparison of the Australian data to North American data confirmed the need for Australian specific mercury air-surface exchange data representative of Australia's unique climatic conditions, vegetation types, land use patterns, and soils.

  11. The performance of a boron-loaded gel-fuel ramjet

    NASA Astrophysics Data System (ADS)

    Haddad, A.; Natan, B.; Arieli, R.

    2011-10-01

    The present work focuses on the possibility of combining the advantages of ramjet propulsion with the high energetic potential of boron. However, the use of boron poses two major challenges. The first, common to all solid additives to liquid fuels is particle sedimentation and poor dispersion. This problem is solved through the use of a gel fuel. The second obstacle, specific to boron-enriched fuels, is the difficulty in realizing the full energetic potential of boron. This could be overcome by means of an aft-combustion chamber, where fuel-rich combustion products are mixed with cold bypass air. Cooling causes the gaseous boron oxide to condense and, as a consequence, the heat of evaporation trapped in the gaseous oxide is released. The merits of such a combination are assessed through its ability to power an air-to-surface missile of relatively small size, capable of delivering a large payload to over a distance of about 1000 km in short time. The paper presents a preliminary design of a ramjet missile using a gel fuel loaded with boron. The thermochemical aspects of the two-stage combustion of the fuel are considered. A comparison with a solid rocket motor (SRM) missile launched under the same conditions as the ramjet missile is made. The boron-loaded gel-fuel ramjet is found superior for this mission.

  12. In-use activity, fuel use, and emissions of heavy-duty diesel roll-off refuse trucks.

    PubMed

    Sandhu, Gurdas S; Frey, H Christopher; Bartelt-Hunt, Shannon; Jones, Elizabeth

    2015-03-01

    The objectives of this study were to quantify real-world activity, fuel use, and emissions for heavy duty diesel roll-off refuse trucks; evaluate the contribution of duty cycles and emissions controls to variability in cycle average fuel use and emission rates; quantify the effect of vehicle weight on fuel use and emission rates; and compare empirical cycle average emission rates with the U.S. Environmental Protection Agency's MOVES emission factor model predictions. Measurements were made at 1 Hz on six trucks of model years 2005 to 2012, using onboard systems. The trucks traveled 870 miles, had an average speed of 16 mph, and collected 165 tons of trash. The average fuel economy was 4.4 mpg, which is approximately twice previously reported values for residential trash collection trucks. On average, 50% of time is spent idling and about 58% of emissions occur in urban areas. Newer trucks with selective catalytic reduction and diesel particulate filter had NOx and PM cycle average emission rates that were 80% lower and 95% lower, respectively, compared to older trucks without. On average, the combined can and trash weight was about 55% of chassis weight. The marginal effect of vehicle weight on fuel use and emissions is highest at low loads and decreases as load increases. Among 36 cycle average rates (6 trucks×6 cycles), MOVES-predicted values and estimates based on real-world data have similar relative trends. MOVES-predicted CO2 emissions are similar to those of the real world, while NOx and PM emissions are, on average, 43% lower and 300% higher, respectively. The real-world data presented here can be used to estimate benefits of replacing old trucks with new trucks. Further, the data can be used to improve emission inventories and model predictions. In-use measurements of the real-world activity, fuel use, and emissions of heavy-duty diesel roll-off refuse trucks can be used to improve the accuracy of predictive models, such as MOVES, and emissions

  13. Quaternion Averaging

    NASA Technical Reports Server (NTRS)

    Markley, F. Landis; Cheng, Yang; Crassidis, John L.; Oshman, Yaakov

    2007-01-01

    Many applications require an algorithm that averages quaternions in an optimal manner. For example, when combining the quaternion outputs of multiple star trackers having this output capability, it is desirable to properly average the quaternions without recomputing the attitude from the the raw star tracker data. Other applications requiring some sort of optimal quaternion averaging include particle filtering and multiple-model adaptive estimation, where weighted quaternions are used to determine the quaternion estimate. For spacecraft attitude estimation applications, derives an optimal averaging scheme to compute the average of a set of weighted attitude matrices using the singular value decomposition method. Focusing on a 4-dimensional quaternion Gaussian distribution on the unit hypersphere, provides an approach to computing the average quaternion by minimizing a quaternion cost function that is equivalent to the attitude matrix cost function Motivated by and extending its results, this Note derives an algorithm that deterniines an optimal average quaternion from a set of scalar- or matrix-weighted quaternions. Rirthermore, a sufficient condition for the uniqueness of the average quaternion, and the equivalence of the mininiization problem, stated herein, to maximum likelihood estimation, are shown.

  14. Space reactor fuel element testing in upgraded TREAT

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Todosow, M.; Bezler, P.; Ludewig, H.

    1993-01-14

    The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. initial results suggest that full-scale PBR, elements could be tested at an average energy deposition of {approximately}60--80 MW-s/L in the current TREAT reactor. Ifmore » the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of {approximately}100 MW/L may be achievable.« less

  15. Space reactor fuel element testing in upgraded TREAT

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Todosow, M.; Bezler, P.; Ludewig, H.

    1993-05-01

    The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. initial results suggest that full-scale PBR, elements could be tested at an average energy deposition of {approximately}60--80 MW-s/L in the current TREAT reactor. Ifmore » the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of {approximately}100 MW/L may be achievable.« less

  16. Multirecycling of Plutonium from LMFBR Blanket in Standard PWRs Loaded with MOX Fuel

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Sonat Sen; Gilles Youinou

    2013-02-01

    It is now well-known that, from a physics standpoint, Pu, or even TRU (i.e. Pu+M.A.), originating from LEU fuel irradiated in PWRs can be multirecycled also in PWRs using MOX fuel. However, the degradation of the isotopic composition during irradiation necessitates using enriched U in conjunction with the MOX fuel either homogeneously or heterogeneously to maintain the Pu (or TRU) content at a level allowing safe operation of the reactor, i.e. below about 10%. The study is related to another possible utilization of the excess Pu produced in the blanket of a LMFBR, namely in a PWR(MOX). In this casemore » the more Pu is bred in the LMFBR, the more PWR(MOX) it can sustain. The important difference between the Pu coming from the blanket of a LMFBR and that coming from a PWR(LEU) is its isotopic composition. The first one contains about 95% of fissile isotopes whereas the second one contains only about 65% of fissile isotopes. As it will be shown later, this difference allows the PWR fed by Pu from the LMFBR blanket to operate with natural U instead of enriched U when it is fed by Pu from PWR(LEU)« less

  17. Final Report on Two-Stage Fast Spectrum Fuel Cycle Options

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Yang, Won Sik; Lin, C. S.; Hader, J. S.

    2016-01-30

    This report presents the performance characteristics of two “two-stage” fast spectrum fuel cycle options proposed to enhance uranium resource utilization and to reduce nuclear waste generation. One is a two-stage fast spectrum fuel cycle option of continuous recycle of plutonium (Pu) in a fast reactor (FR) and subsequent burning of minor actinides (MAs) in an accelerator-driven system (ADS). The first stage is a sodium-cooled FR fuel cycle starting with low-enriched uranium (LEU) fuel; at the equilibrium cycle, the FR is operated using the recovered Pu and natural uranium without supporting LEU. Pu and uranium (U) are co-extracted from the dischargedmore » fuel and recycled in the first stage, and the recovered MAs are sent to the second stage. The second stage is a sodium-cooled ADS in which MAs are burned in an inert matrix fuel form. The discharged fuel of ADS is reprocessed, and all the recovered heavy metals (HMs) are recycled into the ADS. The other is a two-stage FR/ADS fuel cycle option with MA targets loaded in the FR. The recovered MAs are not directly sent to ADS, but partially incinerated in the FR in order to reduce the amount of MAs to be sent to the ADS. This is a heterogeneous recycling option of transuranic (TRU) elements« less

  18. Enrichment of clinical study populations.

    PubMed

    Temple, R

    2010-12-01

    Those who conduct clinical trials "enrich" study populations in a variety of ways in order to identify a population of patients in whom a drug effect, if present, is more likely to be demonstrable. The principal ways to do this are as follows: (i) practical enrichment, i.e., generally seeking to reduce noise (variability of measurement) and heterogeneity (by avoiding the enrollment of patients with other diseases and individuals in whom the disease disappears spontaneously), (ii) prognostic enrichment, i.e., finding patients who are likely to have the event of interest when enrolling for risk-reduction studies, and (iii) predictive enrichment, i.e., finding the individuals who are more likely to respond. Enrichment fits well into the growing interest in "individualization" of therapy but creates some tension with another trend, namely, the desire for "real-world studies" with less restrictive entry criteria and other requirements.

  19. Irradiation performance of (Th,Pu)O2 fuel under Pressurized Water Reactor conditions

    NASA Astrophysics Data System (ADS)

    Boer, B.; Lemehov, S.; Wéber, M.; Parthoens, Y.; Gysemans, M.; McGinley, J.; Somers, J.; Verwerft, M.

    2016-04-01

    This paper examines the in-pile safety performance of (Th,Pu)O2 fuel pins under simulated Pressurized Water Reactor (PWR) conditions. Both sol-gel and SOLMAS produced (Th,Pu)O2 fuels at enrichments of 7.9% and 12.8% in Pu/HM have been irradiated at SCK·CEN. The irradiation has been performed under PWR conditions (155 bar, 300 °C) in a dedicated loop of the BR-2 reactor. The loop is instrumented with flow and temperature monitors at inlet and outlet, which allow for an accurate measurement of the deposited enthalpy.

  20. Performance, Cost, and Market Share of Conventional Vehicle Efficiency Technologies? Retrospective Comparison of Regulatory Document Projections for Corporate Average Fuel Economy and Greenhouse Gas Standards

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Xie, Fei; Lin, Zhenhong; Nealer, Rachael

    This paper conducted an analysis of regulatory documents on current energy- and greenhouse gas–relevant conventional vehicle efficiency technologies in the corporate average fuel economy standards (2017 to 2025) and greenhouse gas rulemaking context by NHTSA and EPA. The focus was on identifying what technologies today—as estimated now (2015 to 2016)—receive higher or lower expectations with regard to effectiveness, cost, and consumer adoption than what experts projected during the 2010 to 2011 rulemaking period. A broad range of conventional vehicle efficiency technologies, including gasoline engine and diesel engine, transmission, accessory, hybrid, and vehicle body technologies, was investigated in this analysis. Finally,more » most assessed technologies were found to have had better competitiveness than expected with regard to effectiveness or costs, or both, with costs and market penetration more difficult to predict than technology effectiveness.« less

  1. Performance, Cost, and Market Share of Conventional Vehicle Efficiency Technologies? Retrospective Comparison of Regulatory Document Projections for Corporate Average Fuel Economy and Greenhouse Gas Standards

    DOE PAGES

    Xie, Fei; Lin, Zhenhong; Nealer, Rachael

    2017-09-30

    This paper conducted an analysis of regulatory documents on current energy- and greenhouse gas–relevant conventional vehicle efficiency technologies in the corporate average fuel economy standards (2017 to 2025) and greenhouse gas rulemaking context by NHTSA and EPA. The focus was on identifying what technologies today—as estimated now (2015 to 2016)—receive higher or lower expectations with regard to effectiveness, cost, and consumer adoption than what experts projected during the 2010 to 2011 rulemaking period. A broad range of conventional vehicle efficiency technologies, including gasoline engine and diesel engine, transmission, accessory, hybrid, and vehicle body technologies, was investigated in this analysis. Finally,more » most assessed technologies were found to have had better competitiveness than expected with regard to effectiveness or costs, or both, with costs and market penetration more difficult to predict than technology effectiveness.« less

  2. Sonication-based isolation and enrichment of Chlorella protothecoides chloroplasts for illumina genome sequencing

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Angelova, Angelina; Park, Sang-Hycuk; Kyndt, John

    2013-09-01

    With the increasing world demand for biofuel, a number of oleaginous algal species are being considered as renewable sources of oil. Chlorella protothecoides Krüger synthesizes triacylglycerols (TAGs) as storage compounds that can be converted into renewable fuel utilizing an anabolic pathway that is poorly understood. The paucity of algal chloroplast genome sequences has been an important constraint to chloroplast transformation and for studying gene expression in TAGs pathways. In this study, the intact chloroplasts were released from algal cells using sonication followed by sucrose gradient centrifugation, resulting in a 2.36-fold enrichment of chloroplasts from C. protothecoides, based on qPCR analysis.more » The C. protothecoides chloroplast genome (cpDNA) was determined using the Illumina HiSeq 2000 sequencing platform and found to be 84,576 Kb in size (8.57 Kb) in size, with a GC content of 30.8 %. This is the first report of an optimized protocol that uses a sonication step, followed by sucrose gradient centrifugation, to release and enrich intact chloroplasts from a microalga (C. prototheocoides) of sufficient quality to permit chloroplast genome sequencing with high coverage, while minimizing nuclear genome contamination. The approach is expected to guide chloroplast isolation from other oleaginous algal species for a variety of uses that benefit from enrichment of chloroplasts, ranging from biochemical analysis to genomics studies.« less

  3. The enrichment behavior of natural radionuclides in pulverized oil shale-fired power plants.

    PubMed

    Vaasma, Taavi; Kiisk, Madis; Meriste, Tõnis; Tkaczyk, Alan Henry

    2014-12-01

    The oil shale industry is the largest producer of NORM (Naturally Occurring Radioactive Material) waste in Estonia. Approximately 11-12 million tons of oil shale containing various amounts of natural radionuclides is burned annually in the Narva oil shale-fired power plants, which accounts for approximately 90% of Estonian electricity production. The radionuclide behavior characteristics change during the fuel combustion process, which redistributes the radionuclides between different ash fractions. Out of 24 operational boilers in the power plants, four use circulating fluidized bed (CFB) technology and twenty use pulverized fuel (PF) technology. Over the past decade, the PF boilers have been renovated, with the main objective to increase the efficiency of the filter systems. Between 2009 and 2012, electrostatic precipitators (ESP) in four PF energy blocks were replaced with novel integrated desulphurization technology (NID) for the efficient removal of fly ash and SO2 from flue gases. Using gamma spectrometry, activity concentrations and enrichment factors for the (238)U ((238)U, (226)Ra, (210)Pb) and (232)Th ((232)Th, (228)Ra) family radionuclides as well as (40)K were measured and analyzed in different PF boiler ash fractions. The radionuclide activity concentrations in the ash samples increased from the furnace toward the back end of the flue gas duct. The highest values in different PF boiler ash fractions were in the last field of the ESP and in the NID ash, where radionuclide enrichment factors were up to 4.2 and 3.3, respectively. The acquired and analyzed data on radionuclide activity concentrations in different PF boiler ashes (operating with an ESP and a NID system) compared to CFB boiler ashes provides an indication that changes in the fuel (oil shale) composition and boiler working parameters, as well as technological enhancements in Estonian oil shale fired power plants, have had a combined effect on the distribution patterns of natural radionuclides

  4. Loss-of-Flow and Loss-of-Pressure Simulations of the BR2 Research Reactor with HEU and LEU Fuel

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Licht, J.; Bergeron, A.; Dionne, B.

    2016-01-01

    Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The reactor core of BR2 is located inside a pressure vessel that contains 79 channels in a hyperboloid configuration. The core configuration is highly variable as each channel can contain a fuel assembly, a control or regulating rod, an experimentalmore » device, or a beryllium or aluminum plug. Because of this variability, a representative core configuration, based on current reactor use, has been defined for the fuel conversion analyses. The code RELAP5/Mod 3.3 was used to perform the transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. The input model has been modernized relative to that historically used at BR2 taking into account the best modeling practices developed by Argonne National Laboratory (ANL) and BR2 engineers.« less

  5. Effects of Enrichment and Litter Parity on Reproductive Performance and Behavior in BALB/c and 129/Sv Mice.

    PubMed

    Whitaker, Julia W; Moy, Sheryl S; Pritchett-Corning, Kathleen R; Fletcher, Craig A

    2016-01-01

    We examined the effect of adding species-appropriate environmental enrichment items to breeding cages of BALB/cAnNCrl and 129S2/SvPasCrl mice. The 3 enrichment conditions were: 1) cotton nesting material; 2) nesting material plus a paper shelter and rolled paper bedding; and 3) an igloo dome with an exercise wheel in addition to the shelter-group enrichments. We measured litter size, litter survival to weaning age, average pup weight at 21 d, and the interlitter interval to evaluate reproductive performance. A random subset of the first- or second-litter offspring from each enrichment condition and strain was assessed in multiple behavioral tests. Enrichment significantly affected anxiety-like behavior and sociability, with the direction of change dependent on strain and sex. Litter parity had greater effects on some reproductive parameters than did the enrichment condition, and this effect was not solely due to a difference between the first compared with subsequent litters. The significant effects of litter parity on the number of pups born and weaned, female pup weight, and interlitter interval were dependent on the enrichment condition in BALB/c but not 129/Sv mice. Offspring from the first or second litter were included in a generational component to investigate whether enrichment effects on reproduction persist in adult offspring after transfer to a different facility for breeding. Natal cage enrichment had no effect on any reproductive parameter in the transferred mice. Overall, additional enrichment beyond nesting material had a beneficial effect on the interlitter interval in BALB/c mice and on the number of pups weaned in 129/Sv mice.

  6. Effects of Enrichment and Litter Parity on Reproductive Performance and Behavior in BALB/c and 129/Sv Mice

    PubMed Central

    Whitaker, Julia W; Moy, Sheryl S; Pritchett-Corning, Kathleen R; Fletcher, Craig A

    2016-01-01

    We examined the effect of adding species-appropriate environmental enrichment items to breeding cages of BALB/cAnNCrl and 129S2/SvPasCrl mice. The 3 enrichment conditions were: 1) cotton nesting material; 2) nesting material plus a paper shelter and rolled paper bedding; and 3) an igloo dome with an exercise wheel in addition to the shelter-group enrichments. We measured litter size, litter survival to weaning age, average pup weight at 21 d, and the interlitter interval to evaluate reproductive performance. A random subset of the first- or second-litter offspring from each enrichment condition and strain was assessed in multiple behavioral tests. Enrichment significantly affected anxiety-like behavior and sociability, with the direction of change dependent on strain and sex. Litter parity had greater effects on some reproductive parameters than did the enrichment condition, and this effect was not solely due to a difference between the first compared with subsequent litters. The significant effects of litter parity on the number of pups born and weaned, female pup weight, and interlitter interval were dependent on the enrichment condition in BALB/c but not 129/Sv mice. Offspring from the first or second litter were included in a generational component to investigate whether enrichment effects on reproduction persist in adult offspring after transfer to a different facility for breeding. Natal cage enrichment had no effect on any reproductive parameter in the transferred mice. Overall, additional enrichment beyond nesting material had a beneficial effect on the interlitter interval in BALB/c mice and on the number of pups weaned in 129/Sv mice. PMID:27423144

  7. Inoculation Stress Hypothesis of Environmental Enrichment

    PubMed Central

    Crofton, Elizabeth J.; Zhang, Yafang; Green, Thomas A.

    2014-01-01

    One hallmark of psychiatric conditions is the vast continuum of individual differences in susceptibility vs. resilience resulting from the interaction of genetic and environmental factors. The environmental enrichment paradigm is an animal model that is useful for studying a range of psychiatric conditions, including protective phenotypes in addiction and depression models. The major question is how environmental enrichment, a non-drug and non-surgical manipulation, can produce such robust individual differences in such a wide range of behaviors. This paper draws from a variety of published sources to outline a coherent hypothesis of inoculation stress as a factor producing the protective enrichment phenotypes. The basic tenet suggests that chronic mild stress from living in a complex environment and interacting non-aggressively with conspecifics can inoculate enriched rats against subsequent stressors and/or drugs of abuse. This paper reviews the enrichment phenotypes, mulls the fundamental nature of environmental enrichment vs. isolation, discusses the most appropriate control for environmental enrichment, and challenges the idea that cortisol/corticosterone equals stress. The intent of the inoculation stress hypothesis of environmental enrichment is to provide a scaffold with which to build testable hypotheses for the elucidation of the molecular mechanisms underlying these protective phenotypes and thus provide new therapeutic targets to treat psychiatric/neurological conditions. PMID:25449533

  8. Alternative Fuels Data Center: Mississippi Transportation Data for

    Science.gov Websites

    with other local stakeholders. Gasoline Diesel Natural Gas Transportation Fuel Consumption Source Renewable Power Plants 0 Renewable Power Plant Capacity (nameplate, MW) 0 Source: BioFuels Atlas from the $2.19/GGE $2.50/gallon $2.50/GGE Diesel $2.61/gallon $2.35/GGE $2.96/gallon $2.66/GGE Source: Average

  9. 10 CFR Appendix II to Part 504 - Fuel Price Computation

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 504—Fuel Price Computation (a) Introduction. This appendix provides the equations and parameters... inflation indices must follow standard statistical procedures and must be fully documented within the... the weighted average fuel price must follow standard statistical procedures and be fully documented...

  10. Microstructural characterization of an irradiated RERTR-6 U-7Mo/AA4043 alloy dispersion fuel plate specimen blister-tested to a final temperature of 500°C

    DOE PAGES

    Keiser, Jr., Dennis D.; Jue, Jan -Fong; Gan, Jian; ...

    2017-02-27

    The Material Management and Minimization (M3) Reactor Conversion Program, in the past called the Reduced Enrichment for Research and Test Reactor (RERTR) Program, is developing low-enriched uranium (LEU) fuels for application in research reactors. U–Mo alloy dispersion fuel is one type being developed. Blister testing has been performed on different fuel plate samples to determine the margin to failure for fuel plates irradiated to different fission densities. Microstructural characterization was performed using scanning electron microscopy and transmission electron microscopy on a sample taken from a U-7Mo/AA4043 matrix dispersion fuel plate irradiated in the RERTR-6 experiment that was blister-tested up tomore » a final temperature of 500°C. The results indicated that two types of grain/cell boundaries were observed in the U- 7Mo fuel particles, one with a relatively low Mo content and fission gas bubbles and a second type enriched in Si, due to interdiffusion from the Si-containing matrix, with little evidence of fission gas bubbles. With respect to the behavior of the major fission gas Xe, a significant amount of the Xe was still observed within the U-7Mo fuel particle, along with microns into the AA4043 matrix. For the fuel/matrix interaction layers that form during fabrication and then grow during irradiation, they change from the as-irradiated amorphous structure to one that is crystalline after blister testing. In the AA4043 matrix, the original Si-rich precipitates, which are typically observed in as-irradiated U-Mo dispersion fuel, get consumed due to interdiffusion with the U-7Mo fuel particles during the blister test. Lastly, the fission gas bubbles that were originally around 2 nm in diameter and resided on a fission gas superlattice in the intragranular regions of as-irradiated U-7Mo fuel grew in size (up to ~20 nm diameter) during blister testing.« less

  11. Microstructural characterization of an irradiated RERTR-6 U-7Mo/AA4043 alloy dispersion fuel plate specimen blister-tested to a final temperature of 500°C

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Keiser, Jr., Dennis D.; Jue, Jan -Fong; Gan, Jian

    The Material Management and Minimization (M3) Reactor Conversion Program, in the past called the Reduced Enrichment for Research and Test Reactor (RERTR) Program, is developing low-enriched uranium (LEU) fuels for application in research reactors. U–Mo alloy dispersion fuel is one type being developed. Blister testing has been performed on different fuel plate samples to determine the margin to failure for fuel plates irradiated to different fission densities. Microstructural characterization was performed using scanning electron microscopy and transmission electron microscopy on a sample taken from a U-7Mo/AA4043 matrix dispersion fuel plate irradiated in the RERTR-6 experiment that was blister-tested up tomore » a final temperature of 500°C. The results indicated that two types of grain/cell boundaries were observed in the U- 7Mo fuel particles, one with a relatively low Mo content and fission gas bubbles and a second type enriched in Si, due to interdiffusion from the Si-containing matrix, with little evidence of fission gas bubbles. With respect to the behavior of the major fission gas Xe, a significant amount of the Xe was still observed within the U-7Mo fuel particle, along with microns into the AA4043 matrix. For the fuel/matrix interaction layers that form during fabrication and then grow during irradiation, they change from the as-irradiated amorphous structure to one that is crystalline after blister testing. In the AA4043 matrix, the original Si-rich precipitates, which are typically observed in as-irradiated U-Mo dispersion fuel, get consumed due to interdiffusion with the U-7Mo fuel particles during the blister test. Lastly, the fission gas bubbles that were originally around 2 nm in diameter and resided on a fission gas superlattice in the intragranular regions of as-irradiated U-7Mo fuel grew in size (up to ~20 nm diameter) during blister testing.« less

  12. Microstructural characterization of an irradiated RERTR-6 U-7Mo/AA4043 alloy dispersion fuel plate specimen blister-tested to a final temperature of 500 °C

    NASA Astrophysics Data System (ADS)

    Keiser, Dennis D.; Jue, Jan-Fong; Gan, Jian; Miller, Brandon D.; Robinson, Adam B.; Madden, James W.; Ross Finlay, M.; Moore, Glenn; Medvedev, Pavel; Meyer, Mitch

    2017-05-01

    The Material Management and Minimization (M3) Reactor Conversion Program, in the past called the Reduced Enrichment for Research and Test Reactor (RERTR) Program, is developing low-enriched uranium (LEU) fuels for application in research and test reactors. U-Mo alloy dispersion fuel is one type being developed. Blister testing has been performed on different fuel plate samples to determine the margin to failure for fuel plates irradiated to different fission densities. Microstructural characterization was performed using scanning electron microscopy and transmission electron microscopy on a sample taken from a U-7Mo/AA4043 matrix dispersion fuel plate irradiated in the RERTR-6 experiment that was blister-tested up to a final temperature of 500 °C. The results indicated that two types of grain/cell boundaries were observed in the U-7Mo fuel particles, one with a relatively low Mo content and fission gas bubbles and a second type enriched in Si, due to interdiffusion from the Si-containing matrix, with little evidence of fission gas bubbles. With respect to the behavior of the major fission gas Xe, a significant amount of the Xe was still observed within the U-7Mo fuel particle, along with microns into the AA4043 matrix. For the fuel/matrix interaction layers that form during fabrication and then grow during irradiation, they change from the as-irradiated amorphous structure to one that is crystalline after blister testing. In the AA4043 matrix, the original Si-rich precipitates, which are typically observed in as-irradiated U-Mo dispersion fuel, get consumed due to interdiffusion with the U-7Mo fuel particles during the blister test. Finally, the fission gas bubbles that were originally around 3 nm in diameter and resided on a fission gas superlattice (FGS) in the intragranular regions of as-irradiated U-7Mo fuel grew in size (up to ∼20 nm diameter) during blister testing and, in many areas, are no longer organized as a superlattice.

  13. Microstructural analysis of as-processed U-10 wt.%Mo monolithic fuel plate in AA6061 matrix with Zr diffusion barrier

    NASA Astrophysics Data System (ADS)

    Perez, E.; Yao, B.; Keiser, D. D., Jr.; Sohn, Y. H.

    2010-07-01

    For higher U-loading in low-enriched U-10 wt.%Mo fuels, monolithic fuel plate clad in AA6061 is being developed as a part of Reduced Enrichment for Research and Test Reactor (RERTR) program. This paper reports the first characterization results from a monolithic U-10 wt.%Mo fuel plate with a Zr diffusion barrier that was fabricated as part of a plate fabrication campaign for irradiation testing in the Advanced Test Reactor (ATR). Both scanning and transmission electron microscopy (SEM and TEM) were employed for analysis. At the interface between the Zr barrier and U-10 wt.%Mo, going from Zr to U(Mo), UZr 2, γ-UZr, Zr solid-solution and Mo 2Zr phases were observed. The interface between AA6061 cladding and Zr barrier plate consisted of four layers, going from Al to Zr, (Al, Si) 2Zr, (Al, Si)Zr 3 (Al, Si) 3Zr, and AlSi 4Zr 5. Irradiation behavior of these intermetallic phases is discussed based on their constituents. Characterization of as-fabricated phase constituents and microstructure would help understand the irradiation behavior of these fuel plates, interpret post-irradiation examination, and optimize the processing parameters of monolithic fuel system.

  14. Alternative Fuels Data Center

    Science.gov Websites

    . (Reference West Virginia Code 11-14C-2, 11-14C-5, 11-14C-6a, 11-15A-13a, and 11-15-18b equivalent, with a variable component equal to at least 5% of the average wholesale price of the fuel

  15. ToNER: A tool for identifying nucleotide enrichment signals in feature-enriched RNA-seq data.

    PubMed

    Promworn, Yuttachon; Kaewprommal, Pavita; Shaw, Philip J; Intarapanich, Apichart; Tongsima, Sissades; Piriyapongsa, Jittima

    2017-01-01

    Biochemical methods are available for enriching 5' ends of RNAs in prokaryotes, which are employed in the differential RNA-seq (dRNA-seq) and the more recent Cappable-seq protocols. Computational methods are needed to locate RNA 5' ends from these data by statistical analysis of the enrichment. Although statistical-based analysis methods have been developed for dRNA-seq, they may not be suitable for Cappable-seq data. The more efficient enrichment method employed in Cappable-seq compared with dRNA-seq could affect data distribution and thus algorithm performance. We present Transformation of Nucleotide Enrichment Ratios (ToNER), a tool for statistical modeling of enrichment from RNA-seq data obtained from enriched and unenriched libraries. The tool calculates nucleotide enrichment scores and determines the global transformation for fitting to the normal distribution using the Box-Cox procedure. From the transformed distribution, sites of significant enrichment are identified. To increase power of detection, meta-analysis across experimental replicates is offered. We tested the tool on Cappable-seq and dRNA-seq data for identifying Escherichia coli transcript 5' ends and compared the results with those from the TSSAR tool, which is designed for analyzing dRNA-seq data. When combining results across Cappable-seq replicates, ToNER detects more known transcript 5' ends than TSSAR. In general, the transcript 5' ends detected by ToNER but not TSSAR occur in regions which cannot be locally modeled by TSSAR. ToNER uses a simple yet robust statistical modeling approach, which can be used for detecting RNA 5'ends from Cappable-seq data, in particular when combining information from experimental replicates. The ToNER tool could potentially be applied for analyzing other RNA-seq datasets in which enrichment for other structural features of RNA is employed. The program is freely available for download at ToNER webpage (http://www4a.biotec.or.th/GI/tools/toner) and Git

  16. ToNER: A tool for identifying nucleotide enrichment signals in feature-enriched RNA-seq data

    PubMed Central

    Promworn, Yuttachon; Kaewprommal, Pavita; Shaw, Philip J.; Intarapanich, Apichart; Tongsima, Sissades

    2017-01-01

    Background Biochemical methods are available for enriching 5′ ends of RNAs in prokaryotes, which are employed in the differential RNA-seq (dRNA-seq) and the more recent Cappable-seq protocols. Computational methods are needed to locate RNA 5′ ends from these data by statistical analysis of the enrichment. Although statistical-based analysis methods have been developed for dRNA-seq, they may not be suitable for Cappable-seq data. The more efficient enrichment method employed in Cappable-seq compared with dRNA-seq could affect data distribution and thus algorithm performance. Results We present Transformation of Nucleotide Enrichment Ratios (ToNER), a tool for statistical modeling of enrichment from RNA-seq data obtained from enriched and unenriched libraries. The tool calculates nucleotide enrichment scores and determines the global transformation for fitting to the normal distribution using the Box-Cox procedure. From the transformed distribution, sites of significant enrichment are identified. To increase power of detection, meta-analysis across experimental replicates is offered. We tested the tool on Cappable-seq and dRNA-seq data for identifying Escherichia coli transcript 5′ ends and compared the results with those from the TSSAR tool, which is designed for analyzing dRNA-seq data. When combining results across Cappable-seq replicates, ToNER detects more known transcript 5′ ends than TSSAR. In general, the transcript 5′ ends detected by ToNER but not TSSAR occur in regions which cannot be locally modeled by TSSAR. Conclusion ToNER uses a simple yet robust statistical modeling approach, which can be used for detecting RNA 5′ends from Cappable-seq data, in particular when combining information from experimental replicates. The ToNER tool could potentially be applied for analyzing other RNA-seq datasets in which enrichment for other structural features of RNA is employed. The program is freely available for download at ToNER webpage (http://www4a

  17. Improved modelling of ship SO 2 emissions—a fuel-based approach

    NASA Astrophysics Data System (ADS)

    Endresen, Øyvind; Bakke, Joachim; Sørgård, Eirik; Flatlandsmo Berglen, Tore; Holmvang, Per

    Significant variations are apparent between the various reported regional and global ship SO 2 emission inventories. Important parameters for SO 2 emission modelling are sulphur contents and marine fuel consumption. Since 1993, the global average sulphur content for heavy fuel has shown an overall downward trend, while the bunker sale has increased. We present an improved bottom up approach to estimate marine sulphur emissions from ship transportation, including the geographical distribution. More than 53,000 individual bunker samples are used to establish regionally and globally (volume) weighted average sulphur contents for heavy and distillate marine fuels. We find that the year 2002 sulphur content in heavy fuels varies regionally from 1.90% (South America) to 3.07% (Asia), with a globally weighted average of 2.68% sulphur. The calculated globally weighted average content for heavy fuels is found to be 5% higher than the average (arithmetic mean) sulphur content commonly used. The reason for this is likely that larger bunker stems are mainly of high-viscosity heavy fuel, which tends to have higher sulphur values compared to lower viscosity fuels. The uncertainties in SO 2 inventories are significantly reduced using our updated SO 2 emission factors (volume-weighted sulphur content). Regional marine bunker sales figures are combined with volume-weighted sulphur contents for each region to give a global SO 2 emission estimate in the range of 5.9-7.2 Tg (SO 2) for international marine transportation. Also taking into account the domestic sales, the total emissions from all ocean-going transportation is estimated to be 7.0-8.5 Tg (SO 2). Our estimate is significantly lower than recent global estimate reported by Corbett and Koehler [2003. Journal of Geophysical Research: Atmospheres 108] (6.49 Tg S or about 13.0 Tg SO 2). Endresen et al. [2004. Journal of Geophysical Research 109, D23302] claim that uncertainties in input data for the activity-based method will

  18. Alternative Fuels Data Center: Nevada Transportation Data for Alternative

    Science.gov Websites

    . Gasoline Diesel Natural Gas Electricity Transportation Fuel Consumption Source: State Energy Data System Renewable Power Plant Capacity (nameplate, MW) 1,684 Source: BioFuels Atlas from the National Renewable Source: Average prices per gasoline gallon equivalent (GGE) for the West Coast PADD from the Alternative

  19. Alternative Fuels Data Center: Delaware Transportation Data for Alternative

    Science.gov Websites

    local stakeholders. Gasoline Diesel Natural Gas Transportation Fuel Consumption Source: State Energy Plants 1 Renewable Power Plant Capacity (nameplate, MW) 2 Source: BioFuels Atlas from the National /gallon $2.66/GGE Source: Average prices per gasoline gallon equivalent (GGE) for the Central Atlantic

  20. Carbon-13 and proton nuclear magnetic resonance analysis of shale-derived refinery products and jet fuels and of experimental referee broadened-specification jet fuels

    NASA Technical Reports Server (NTRS)

    Dalling, D. K.; Bailey, B. K.; Pugmire, R. J.

    1984-01-01

    A proton and carbon-13 nuclear magnetic resonance (NMR) study was conducted of Ashland shale oil refinery products, experimental referee broadened-specification jet fuels, and of related isoprenoid model compounds. Supercritical fluid chromatography techniques using carbon dioxide were developed on a preparative scale, so that samples could be quantitatively separated into saturates and aromatic fractions for study by NMR. An optimized average parameter treatment was developed, and the NMR results were analyzed in terms of the resulting average parameters; formulation of model mixtures was demonstrated. Application of novel spectroscopic techniques to fuel samples was investigated.

  1. Advanced Fuel Cycle Cost Basis – 2017 Edition

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Dixon, B. W.; Ganda, F.; Williams, K. A.

    This report, commissioned by the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the DOE Nuclear Technology Research and Development (NTRD) Program (previously the Fuel Cycle Research and Development (FCRD) and the Advanced Fuel Cycle Initiative (AFCI)). The report describes the NTRD cost basis development process, reference information on NTRD 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 reportmore » contains reference cost data for numerous fuel cycle cost modules (modules A-O) as well as cost modules for a number of reactor types (R modules). The fuel cycle cost modules were developed in the areas of natural uranium mining and milling, thorium 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, managed decay storage, recycled product 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. Since its inception, this report has been periodically updated. The last such internal document was published in August 2015 while the last external edition was published in December of 2009 as INL/EXT-07-12107 and is available on the Web at URL: www.inl.gov/technicalpublications/Documents/4536700.pdf. This current report (Sept 2017) is planned to be reviewed for external release, at which time it will replace the 2009 report as an external publication. This information is used in the ongoing evaluation of nuclear fuel cycles by the NE NTRD program.« less

  2. Sampling Enrichment toward Target Structures Using Hybrid Molecular Dynamics-Monte Carlo Simulations

    PubMed Central

    Yang, Kecheng; Różycki, Bartosz; Cui, Fengchao; Shi, Ce; Chen, Wenduo; Li, Yunqi

    2016-01-01

    Sampling enrichment toward a target state, an analogue of the improvement of sampling efficiency (SE), is critical in both the refinement of protein structures and the generation of near-native structure ensembles for the exploration of structure-function relationships. We developed a hybrid molecular dynamics (MD)-Monte Carlo (MC) approach to enrich the sampling toward the target structures. In this approach, the higher SE is achieved by perturbing the conventional MD simulations with a MC structure-acceptance judgment, which is based on the coincidence degree of small angle x-ray scattering (SAXS) intensity profiles between the simulation structures and the target structure. We found that the hybrid simulations could significantly improve SE by making the top-ranked models much closer to the target structures both in the secondary and tertiary structures. Specifically, for the 20 mono-residue peptides, when the initial structures had the root-mean-squared deviation (RMSD) from the target structure smaller than 7 Å, the hybrid MD-MC simulations afforded, on average, 0.83 Å and 1.73 Å in RMSD closer to the target than the parallel MD simulations at 310K and 370K, respectively. Meanwhile, the average SE values are also increased by 13.2% and 15.7%. The enrichment of sampling becomes more significant when the target states are gradually detectable in the MD-MC simulations in comparison with the parallel MD simulations, and provide >200% improvement in SE. We also performed a test of the hybrid MD-MC approach in the real protein system, the results showed that the SE for 3 out of 5 real proteins are improved. Overall, this work presents an efficient way of utilizing solution SAXS to improve protein structure prediction and refinement, as well as the generation of near native structures for function annotation. PMID:27227775

  3. Sampling Enrichment toward Target Structures Using Hybrid Molecular Dynamics-Monte Carlo Simulations.

    PubMed

    Yang, Kecheng; Różycki, Bartosz; Cui, Fengchao; Shi, Ce; Chen, Wenduo; Li, Yunqi

    2016-01-01

    Sampling enrichment toward a target state, an analogue of the improvement of sampling efficiency (SE), is critical in both the refinement of protein structures and the generation of near-native structure ensembles for the exploration of structure-function relationships. We developed a hybrid molecular dynamics (MD)-Monte Carlo (MC) approach to enrich the sampling toward the target structures. In this approach, the higher SE is achieved by perturbing the conventional MD simulations with a MC structure-acceptance judgment, which is based on the coincidence degree of small angle x-ray scattering (SAXS) intensity profiles between the simulation structures and the target structure. We found that the hybrid simulations could significantly improve SE by making the top-ranked models much closer to the target structures both in the secondary and tertiary structures. Specifically, for the 20 mono-residue peptides, when the initial structures had the root-mean-squared deviation (RMSD) from the target structure smaller than 7 Å, the hybrid MD-MC simulations afforded, on average, 0.83 Å and 1.73 Å in RMSD closer to the target than the parallel MD simulations at 310K and 370K, respectively. Meanwhile, the average SE values are also increased by 13.2% and 15.7%. The enrichment of sampling becomes more significant when the target states are gradually detectable in the MD-MC simulations in comparison with the parallel MD simulations, and provide >200% improvement in SE. We also performed a test of the hybrid MD-MC approach in the real protein system, the results showed that the SE for 3 out of 5 real proteins are improved. Overall, this work presents an efficient way of utilizing solution SAXS to improve protein structure prediction and refinement, as well as the generation of near native structures for function annotation.

  4. The Learning Enrichment Service: A Triad-Based Secondary School Model for Enrichment Programming.

    ERIC Educational Resources Information Center

    Smyth, Elizabeth; And Others

    1983-01-01

    Three secondary teachers describe a school-wide support system for meeting the needs of gifted students in and beyond the regular classroom. A management team coordinates enrichment within the school and community while a computerized data bank of enrichment resources is accessible to all learners. (CL)

  5. Early experience and results of bone graft enriched with autologous platelet gel for recalcitrant nonunions of lower extremity.

    PubMed

    Chiang, Chao-Ching; Su, Chen-Yao; Huang, Ching-Kuei; Chen, Wei-Ming; Chen, Tain-Hsiung; Tzeng, Yun-Hsuan

    2007-09-01

    Refractory nonunions of the tibia or femur are physically and mentally devastating conditions for the patients, and the treatment is challenging for orthopedic surgeons. The goal of this study was to assess the feasibility and outcome of surgical treatment in recalcitrant nonunions of a lower extremity with bone graft enriched with autologous platelet gel (APG). Twelve patients with four femoral and eight tibial atrophic nonunions after multiple prior procedures were included. All of them were treated with the bone grafting procedures with autograft complex enriched with APG. They were evaluated with radiographs, bone mineral density for bony healing process, and the Short-Form 36 Health Survey for functional outcome. Of the 12 patients, 11 healed at an average of 19.7 weeks after the first attempt and 1 healed after the second attempt at 21 weeks. The bone mineral density continued to increase steadily from early healing to the remodeling phase. Functional status was greatly improved at an average follow-up of 32.4 months. The results of this preliminary study implied the possible potential of bone graft enriched with APG in the treatment of recalcitrant nonunions of the lower extremity. More research is necessary to clarify its role in augmentation of bone graft to enhance healing of nonunion.

  6. Inoculation stress hypothesis of environmental enrichment.

    PubMed

    Crofton, Elizabeth J; Zhang, Yafang; Green, Thomas A

    2015-02-01

    One hallmark of psychiatric conditions is the vast continuum of individual differences in susceptibility vs. resilience resulting from the interaction of genetic and environmental factors. The environmental enrichment paradigm is an animal model that is useful for studying a range of psychiatric conditions, including protective phenotypes in addiction and depression models. The major question is how environmental enrichment, a non-drug and non-surgical manipulation, can produce such robust individual differences in such a wide range of behaviors. This paper draws from a variety of published sources to outline a coherent hypothesis of inoculation stress as a factor producing the protective enrichment phenotypes. The basic tenet suggests that chronic mild stress from living in a complex environment and interacting non-aggressively with conspecifics can inoculate enriched rats against subsequent stressors and/or drugs of abuse. This paper reviews the enrichment phenotypes, mulls the fundamental nature of environmental enrichment vs. isolation, discusses the most appropriate control for environmental enrichment, and challenges the idea that cortisol/corticosterone equals stress. The intent of the inoculation stress hypothesis of environmental enrichment is to provide a scaffold with which to build testable hypotheses for the elucidation of the molecular mechanisms underlying these protective phenotypes and thus provide new therapeutic targets to treat psychiatric/neurological conditions. Copyright © 2014 Elsevier Ltd. All rights reserved.

  7. Modeling transit bus fuel consumption on the basis of cycle properties.

    PubMed

    Delgado, Oscar F; Clark, Nigel N; Thompson, Gregory J

    2011-04-01

    A method exists to predict heavy-duty vehicle fuel economy and emissions over an "unseen" cycle or during unseen on-road activity on the basis of fuel consumption and emissions data from measured chassis dynamometer test cycles and properties (statistical parameters) of those cycles. No regression is required for the method, which relies solely on the linear association of vehicle performance with cycle properties. This method has been advanced and examined using previously published heavy-duty truck data gathered using the West Virginia University heavy-duty chassis dynamometer with the trucks exercised over limited test cycles. In this study, data were available from a Washington Metropolitan Area Transit Authority emission testing program conducted in 2006. Chassis dynamometer data from two conventional diesel buses, two compressed natural gas buses, and one hybrid diesel bus were evaluated using an expanded driving cycle set of 16 or 17 different driving cycles. Cycle properties and vehicle fuel consumption measurements from three baseline cycles were selected to generate a linear model and then to predict unseen fuel consumption over the remaining 13 or 14 cycles. Average velocity, average positive acceleration, and number of stops per distance were found to be the desired cycle properties for use in the model. The methodology allowed for the prediction of fuel consumption with an average error of 8.5% from vehicles operating on a diverse set of chassis dynamometer cycles on the basis of relatively few experimental measurements. It was found that the data used for prediction should be acquired from a set that must include an idle cycle along with a relatively slow transient cycle and a relatively high speed cycle. The method was also applied to oxides of nitrogen prediction and was found to have less predictive capability than for fuel consumption with an average error of 20.4%.

  8. Alternative Fuels Data Center: Oklahoma Transportation Data for Alternative

    Science.gov Websites

    Fuel Consumption Source: State Energy Data System based on beta data converted to gasoline gallon ) 2,573 Source: BioFuels Atlas from the National Renewable Energy Laboratory Case Studies Video thumbnail Source: Average prices per gasoline gallon equivalent (GGE) for the Midwest PADD from the Alternative

  9. Closed fuel cycle with increased fuel burn-up and economy applying of thorium resources

    NASA Astrophysics Data System (ADS)

    Kulikov, G. G.; Apse, V. A.

    2017-01-01

    The possible role of existing thorium reserves in the Russian Federation on engaging thorium in being currently closed (U-Pu)-fuel cycle of nuclear power of the country is considered. The application efficiency of thermonuclear neutron sources with thorium blanket for the economical use of existing thorium reserves is demonstrated. The aim of the work is to find solutions of such major tasks as the reduction of both front-end and back-end of nuclear fuel cycle and an enhancing its protection against the uncontrolled proliferation of fissile materials by means of the smallest changes in the fuel cycle. During implementation of the work we analyzed the results obtained earlier by the authors, brought new information on the number of thorium available in the Russian Federation and made further assessments. On the basis of proposal on the inclusion of hybrid reactors with Th-blanket into the future nuclear power for the production of light uranium fraction 232+233+234U, and 231Pa, we obtained the following results: 1. The fuel cycle will shift from fissile 235U to 233U which is more attractive for thermal power reactors. 2. The light uranium fraction is the most "protected" in the uranium component of fuel and mixed with regenerated uranium will in addition become a low enriched uranium fuel, that will weaken the problem of uncontrolled proliferation of fissile materials. 3. 231Pa doping into the fuel stabilizes its multiplying properties that will allow us to implement long-term fuel residence time and eventually to increase the export potential of all nuclear power technologies. 4. The thorium reserves being near city Krasnoufimsk (Russia) are large enough for operation of large-scale nuclear power of the Russian Federation of 70 GWe capacity during more than a quarter century under assumption that thorium is loaded into blankets of hybrid TNS only. The general conclusion: the inclusion of a small number of hybrid reactors with Th-blanket into the future nuclear

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

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    E. Schneider; B. Carlsen; E. Tavrides

    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 wellmore » 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

  11. Recent GE BWR fuel experience and design evolution

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Wood, J.E.; Potts, G.A.; Proebstle, R.A.

    1992-01-01

    Reliable fuel operation is essential to the safe, reliable, and economic power production by today's commercial nuclear reactors. GE Nuclear Energy is committed to maximize fuel reliability through the progressive development of improved fuel design features and dedication to provide the maximum quality of the design features and dedication to provide the maximum quality of the design, fabrication, and operation of GE BWR fuel. Over the last 35 years, GE has designed, fabricated, and placed in operation over 82,000 BWR fuel bundles containing over 5 million fuel rods. This experience includes successful commercial reactor operation of fuel assemblies to greatermore » than 45000 MWd/MTU bundle average exposure. This paper reports that this extensive experience base has enabled clear identification and characterization of the active failure mechanisms. With this failure mechanism characterization, mitigating actions have been developed and implemented by GE to provide the highest reliability BWR fuel bundles possible.« less

  12. Refuse-Derived Fuels in U.S. Air Force Heating and Power Systems.

    DTIC Science & Technology

    1986-01-01

    pellet degradation . Postpellet - cooling may not prevent disintegration into more fines. The Argonne W National Laboratory is sponsoring a research...of fuel,% by using the fuel Orsat (volumetric) analysis and both the carbon and sulfur in the fuel. This is known as a carbon balance. 00 The test... sulfur in the fuel per pound of fuel C ’ = average specific heat of the dry flue gas PG T = flue gas temperature leaving the economizer (or leaving

  13. 31 CFR 540.316 - Uranium enrichment.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false Uranium enrichment. 540.316 Section 540.316 Money and Finance: Treasury Regulations Relating to Money and Finance (Continued) OFFICE OF... REGULATIONS General Definitions § 540.316 Uranium enrichment. The term uranium enrichment means the process of...

  14. Analysis of Fuel Vaporization, Fuel-Air Mixing, and Combustion in Integrated Mixer-Flame Holders

    NASA Technical Reports Server (NTRS)

    Deur, J. M.; Cline, M. C.

    2004-01-01

    Requirements to limit pollutant emissions from the gas turbine engines for the future High-Speed Civil Transport (HSCT) have led to consideration of various low-emission combustor concepts. One such concept is the Integrated Mixer-Flame Holder (IMFH). This report describes a series of IMFH analyses performed with KIVA-II, a multi-dimensional CFD code for problems involving sprays, turbulence, and combustion. To meet the needs of this study, KIVA-II's boundary condition and chemistry treatments are modified. The study itself examines the relationships between fuel vaporization, fuel-air mixing, and combustion. Parameters being considered include: mixer tube diameter, mixer tube length, mixer tube geometry (converging-diverging versus straight walls), air inlet velocity, air inlet swirl angle, secondary air injection (dilution holes), fuel injection velocity, fuel injection angle, number of fuel injection ports, fuel spray cone angle, and fuel droplet size. Cases are run with and without combustion to examine the variations in fuel-air mixing and potential for flashback due to the above parameters. The degree of fuel-air mixing is judged by comparing average, minimum, and maximum fuel/air ratios at the exit of the mixer tube, while flame stability is monitored by following the location of the flame front as the solution progresses from ignition to steady state. Results indicate that fuel-air mixing can be enhanced by a variety of means, the best being a combination of air inlet swirl and a converging-diverging mixer tube geometry. With the IMFH configuration utilized in the present study, flashback becomes more common as the mixer tube diameter is increased and is instigated by disturbances associated with the dilution hole flow.

  15. Final report on accident tolerant fuel performance analysis of APMT-Steel Clad/UO₂ fuel and APMT-Steel Clad/UN-U₃Si₅ fuel concepts

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Unal, Cetin; Galloway, Jack D.

    2014-09-12

    In FY2014 our group completed and documented analysis of new Accident Tolerant Fuel (ATF) concepts using BISON. We have modeled the viability of moving from Zircaloy to stainless steel cladding in traditional light water reactors (LWRs). We have explored the reactivity penalty of this change using the MCNP-based burnup code Monteburns, while attempting to minimize this penalty by increasing the fuel pellet radius and decreasing the cladding thickness. Fuel performance simulations using BISON have also been performed to quantify changes to structural integrity resulting from thinner stainless steel claddings. We account for thermal and irradiation creep, fission gas swelling, thermalmore » swelling and fuel relocation in the models for both Zircaloy and stainless steel claddings. Additional models that account for the lower oxidation stainless steel APMT are also invoked where available. Irradiation data for HT9 is used as a fallback in the absence of appropriate models. In this study the isotopic vectors within each natural element are varied to assess potential reactivity gains if advanced enrichment capabilities were levied towards cladding technologies. Recommendations on cladding thicknesses for a robust cladding as well as the constitutive components of a less penalizing composition are provided. In the first section (section 1-3), we present results accepted for publication in the 2014 TOPFUEL conference regarding the APMT/UO₂ ATF concept (J. Galloway & C. Unal, Accident Tolerant and Neutronically Favorable LWR Cladding, Proceedings of WRFPM 2014, Sendai, Japan, Paper No.1000050). Next we discuss our preliminary findings from the thermo-mechanical analysis of UN-U₃Si₅ fuel with APMT clad. In this analysis we used models developed from limited data that need to be updated when the irradiation data from ATF-1 test is available. Initial results indicate a swelling rate less than 1.5% is needed to prevent excessive clad stress.« less

  16. Fuel-rich catalytic combustion: A soot-free technique for in situ hydrogen-like enrichment

    NASA Technical Reports Server (NTRS)

    Brabbs, T. A.; Olson, S. L.

    1985-01-01

    An experimental program on the catalytic oxidation of iso-octane demonstrated the feasibility of the two-stage combustion system for reducing particulate emissions. With a fuel-rich (phi = 4.8 to 7.8) catalytic combustion preburner as the first stage the combustion process was soot free at reactor outlet temperatures of 1200 K or less. Although soot was not measured directly, its absence was indicated. Reaction products collected at two positions downstream of the catalyst bed were analyzed on a gas chromatograph. Comparison of these products indicated that pyrolysis of the larger molecules continued along the drift tube and that benzene formation was a gas-phase reaction. The effective hydrogen-carbon ratio calculated from the reaction products increased by 20 to 68 percent over the range of equivalence ratios tested. The catalytic partial oxidation process also yielded a large number of smaller-containing molecules. The fraction of fuel carbon in compounds having two or fewer carbon atoms ranged from 30 percent at 1100 K to 80 percent at 1200 K.

  17. Dissociating Averageness and Attractiveness: Attractive Faces Are Not Always Average

    ERIC Educational Resources Information Center

    DeBruine, Lisa M.; Jones, Benedict C.; Unger, Layla; Little, Anthony C.; Feinberg, David R.

    2007-01-01

    Although the averageness hypothesis of facial attractiveness proposes that the attractiveness of faces is mostly a consequence of their averageness, 1 study has shown that caricaturing highly attractive faces makes them mathematically less average but more attractive. Here the authors systematically test the averageness hypothesis in 5 experiments…

  18. Evaluation of design parameters for TRISO-coated fuel particles to establish manufacturing critical limits using PARFUME

    DOE PAGES

    Skerjanc, William F.; Maki, John T.; Collin, Blaise P.; ...

    2015-12-02

    The success of modular high temperature gas-cooled reactors is highly dependent on the performance of the tristructural-isotopic (TRISO) coated fuel particle and the quality to which it can be manufactured. During irradiation, TRISO-coated fuel particles act as a pressure vessel to contain fission gas and mitigate the diffusion of fission products to the coolant boundary. The fuel specifications place limits on key attributes to minimize fuel particle failure under irradiation and postulated accident conditions. PARFUME (an integrated mechanistic coated particle fuel performance code developed at the Idaho National Laboratory) was used to calculate fuel particle failure probabilities. By systematically varyingmore » key TRISO-coated particle attributes, failure probability functions were developed to understand how each attribute contributes to fuel particle failure. Critical manufacturing limits were calculated for the key attributes of a low enriched TRISO-coated nuclear fuel particle with a kernel diameter of 425 μm. As a result, these critical manufacturing limits identify ranges beyond where an increase in fuel particle failure probability is expected to occur.« less

  19. 21 CFR 139.155 - Enriched noodle products.

    Code of Federal Regulations, 2012 CFR

    2012-04-01

    ... farina, or enriched flour, or through the direct additions of any of the substances prescribed in... 21 Food and Drugs 2 2012-04-01 2012-04-01 false Enriched noodle products. 139.155 Section 139.155... Noodle Products § 139.155 Enriched noodle products. (a) Enriched noodle products are the class of food...

  20. 21 CFR 139.155 - Enriched noodle products.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ... farina, or enriched flour, or through the direct additions of any of the substances prescribed in... 21 Food and Drugs 2 2011-04-01 2011-04-01 false Enriched noodle products. 139.155 Section 139.155... Noodle Products § 139.155 Enriched noodle products. (a) Enriched noodle products are the class of food...

  1. 21 CFR 139.155 - Enriched noodle products.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    ... farina, or enriched flour, or through the direct additions of any of the substances prescribed in... 21 Food and Drugs 2 2013-04-01 2013-04-01 false Enriched noodle products. 139.155 Section 139.155... Noodle Products § 139.155 Enriched noodle products. (a) Enriched noodle products are the class of food...

  2. 21 CFR 139.155 - Enriched noodle products.

    Code of Federal Regulations, 2014 CFR

    2014-04-01

    ... farina, or enriched flour, or through the direct additions of any of the substances prescribed in... 21 Food and Drugs 2 2014-04-01 2014-04-01 false Enriched noodle products. 139.155 Section 139.155... Noodle Products § 139.155 Enriched noodle products. (a) Enriched noodle products are the class of food...

  3. Swelling of U-7Mo/Al-Si dispersion fuel plates under irradiation - Non-destructive analysis of the AFIP-1 fuel plates

    NASA Astrophysics Data System (ADS)

    Wachs, D. M.; Robinson, A. B.; Rice, F. J.; Kraft, N. C.; Taylor, S. C.; Lillo, M.; Woolstenhulme, N.; Roth, G. A.

    2016-08-01

    Extensive fuel-matrix interactions leading to plate pillowing have proven to be a significant impediment to the development of a suitable high density low-enriched uranium molybdenum alloy (U-Mo) based dispersion fuel for high power applications in research reactors. The addition of silicon to the aluminum matrix was previously demonstrated to reduce interaction layer growth in mini-plate experiments. The AFIP-1 project involved the irradiation, in-canal examination, and post-irradiation examination of two fuel plates. The irradiation of two distinct full size, flat fuel plates (one using an Al-2wt%Si matrix and the other an Al-4043 (∼4.8 wt% Si) matrix) was performed in the INL ATR reactor in 2008-2009. The irradiation conditions were: ∼250 W/cm2 peak Beginning Of Life (BOL) power, with a ∼3.5e21 f/cm3 peak burnup. The plates were successfully irradiated and did not show any pillowing at the end of the irradiation. This paper reports the results and interpretation of the in-canal and post-irradiation non-destructive examinations that were performed on these fuel plates. It further compares additional PIE results obtained on fuel plates irradiated in contemporary campaigns in order to allow a complete comparison with all results obtained under similar conditions. Except for a brief indication of accelerated swelling early in the irradiation of the Al-2Si plate, the fuel swelling is shown to evolve linearly with the fission density through the maximum burnup.

  4. 21 CFR 137.305 - Enriched farina.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ... 21 Food and Drugs 2 2011-04-01 2011-04-01 false Enriched farina. 137.305 Section 137.305 Food and... CONSUMPTION CEREAL FLOURS AND RELATED PRODUCTS Requirements for Specific Standardized Cereal Flours and Related Products § 137.305 Enriched farina. (a) Enriched farina conforms to the definition and standard of...

  5. 21 CFR 137.305 - Enriched farina.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    ... 21 Food and Drugs 2 2013-04-01 2013-04-01 false Enriched farina. 137.305 Section 137.305 Food and... CONSUMPTION CEREAL FLOURS AND RELATED PRODUCTS Requirements for Specific Standardized Cereal Flours and Related Products § 137.305 Enriched farina. (a) Enriched farina conforms to the definition and standard of...

  6. 21 CFR 137.305 - Enriched farina.

    Code of Federal Regulations, 2014 CFR

    2014-04-01

    ... 21 Food and Drugs 2 2014-04-01 2014-04-01 false Enriched farina. 137.305 Section 137.305 Food and... CONSUMPTION CEREAL FLOURS AND RELATED PRODUCTS Requirements for Specific Standardized Cereal Flours and Related Products § 137.305 Enriched farina. (a) Enriched farina conforms to the definition and standard of...

  7. 21 CFR 137.305 - Enriched farina.

    Code of Federal Regulations, 2012 CFR

    2012-04-01

    ... 21 Food and Drugs 2 2012-04-01 2012-04-01 false Enriched farina. 137.305 Section 137.305 Food and... CONSUMPTION CEREAL FLOURS AND RELATED PRODUCTS Requirements for Specific Standardized Cereal Flours and Related Products § 137.305 Enriched farina. (a) Enriched farina conforms to the definition and standard of...

  8. Report of a parent survey of cannabidiol-enriched cannabis use in pediatric treatment-resistant epilepsy.

    PubMed

    Porter, Brenda E; Jacobson, Catherine

    2013-12-01

    Severe childhood epilepsies are characterized by frequent seizures, neurodevelopmental delays, and impaired quality of life. In these treatment-resistant epilepsies, families often seek alternative treatments. This survey explored the use of cannabidiol-enriched cannabis in children with treatment-resistant epilepsy. The survey was presented to parents belonging to a Facebook group dedicated to sharing information about the use of cannabidiol-enriched cannabis to treat their child's seizures. Nineteen responses met the following inclusion criteria for the study: a diagnosis of epilepsy and current use of cannabidiol-enriched cannabis. Thirteen children had Dravet syndrome, four had Doose syndrome, and one each had Lennox-Gastaut syndrome and idiopathic epilepsy. The average number of antiepileptic drugs (AEDs) tried before using cannabidiol-enriched cannabis was 12. Sixteen (84%) of the 19 parents reported a reduction in their child's seizure frequency while taking cannabidiol-enriched cannabis. Of these, two (11%) reported complete seizure freedom, eight (42%) reported a greater than 80% reduction in seizure frequency, and six (32%) reported a 25-60% seizure reduction. Other beneficial effects included increased alertness, better mood, and improved sleep. Side effects included drowsiness and fatigue. Our survey shows that parents are using cannabidiol-enriched cannabis as a treatment for their children with treatment-resistant epilepsy. Because of the increasing number of states that allow access to medical cannabis, its use will likely be a growing concern for the epilepsy community. Safety and tolerability data for cannabidiol-enriched cannabis use among children are not available. Objective measurements of a standardized preparation of pure cannabidiol are needed to determine whether it is safe, well tolerated, and efficacious at controlling seizures in this pediatric population with difficult-to-treat seizures. © 2013.

  9. CERCA LEU fuel assemblies testing in Maria Reactor - safety analysis summary and testing program scope.

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Pytel, K.; Mieleszczenko, W.; Lechniak, J.

    2010-03-01

    The presented paper contains neutronic and thermal-hydraulic (for steady and unsteady states) calculation results prepared to support annex to Safety Analysis Report for MARIA reactor in order to obtain approval for program of testing low-enriched uranium (LEU) lead test fuel assemblies (LTFA) manufactured by CERCA. This includes presentation of the limits and operational constraints to be in effect during the fuel testing investigations. Also, the scope of testing program (which began in August 2009), including additional measurements and monitoring procedures, is described.

  10. Comparison of hecter fuel with export aviation gasoline

    NASA Technical Reports Server (NTRS)

    Dickinson, H C; Gage, V R; Sparrow, S W

    1921-01-01

    Among the fuels which will operate at compression ratios up to at least 8.0 without preignition or "pinking" is hecter fuel, whence a careful determination of its performance is of importance. For the test data presented in this report the hecter fuel used was a mixture of 30 per cent benzol and 70 per cent cyclohexane, having a low freezing point, and distilling from first drop to 90 per cent at nearly a constant temperature, about 20 degrees c. below the average distillation temperature ("mean volatility") of the x gasoline (export grade). The results of these experiments show that the power developed by hecter fuel is the same as that developed by export aviation gasoline at about 1,800 r.p.m. at all altitudes. At lower speeds differences in the power developed by the fuels become evident. Comparisons at ground level were omitted to avoid any possibility of damaging the engine by operating with open throttle on gasoline at so high a compression. The fuel consumption per unit power based on weight, not volume, averaged more than 10 per cent greater with hecter than with x gasoline. The thermal efficiency of the engine when using hecter is less than when using gasoline, particularly at higher speeds. A generalization of the difference for all altitudes and speeds being 8 per cent. A general deduction from these facts is that more hecter is exhausted unburnt. Hecter can withstand high compression pressures and temperature without preignition. (author)

  11. Differential effects of enriched environment at work on cognitive decline in old age.

    PubMed

    Then, Francisca S; Luck, Tobias; Luppa, Melanie; König, Hans-Helmut; Angermeyer, Matthias C; Riedel-Heller, Steffi G

    2015-05-26

    The aim of the present study was to investigate how different mentally demanding work conditions during the professional life-i.e., enriched environments at work-might influence the rate of cognitive decline in old age. Individuals (n = 1,054) of the Leipzig Longitudinal Study of the Aged, a representative population-based cohort study of individuals aged 75 years and older, underwent cognitive testing via the Mini-Mental State Examination (MMSE) in up to 6 measurement waves. Type and level of mentally demanding work conditions in the participants' former professional life were classified based on the O*NET job descriptor database. In multivariate mixed-model analyses (controlling for sociodemographic and health-related factors), a high level of mentally demanding work tasks stimulating verbal intelligence was significantly associated with a better cognitive functioning at baseline (on average 5 MMSE points higher) as well as a lower rate of cognitive decline (on average 2 MMSE points less) over the 8-year follow-up period compared with a low level. The rate of cognitive decline in old age was also significantly lower (on average 3 MMSE points less) in individuals who had a high level of mentally demanding work tasks stimulating executive functions than those who had a low level. The results suggest that a professional life enriched with work tasks stimulating verbal intelligence and executive functions may help to sustain a good cognitive functioning in old age (75+ years). The findings thus emphasize that today's challenging work conditions may also promote positive health effects. © 2015 American Academy of Neurology.

  12. 21 CFR 137.160 - Enriched bromated flour.

    Code of Federal Regulations, 2013 CFR

    2013-04-01

    ... 21 Food and Drugs 2 2013-04-01 2013-04-01 false Enriched bromated flour. 137.160 Section 137.160... Flours and Related Products § 137.160 Enriched bromated flour. Enriched bromated flour conforms to the..., prescribed for enriched flour by § 137.165, except that potassium bromate is added in a quantity not...

  13. 21 CFR 137.160 - Enriched bromated flour.

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... 21 Food and Drugs 2 2010-04-01 2010-04-01 false Enriched bromated flour. 137.160 Section 137.160... Flours and Related Products § 137.160 Enriched bromated flour. Enriched bromated flour conforms to the..., prescribed for enriched flour by § 137.165, except that potassium bromate is added in a quantity not...

  14. 21 CFR 137.160 - Enriched bromated flour.

    Code of Federal Regulations, 2012 CFR

    2012-04-01

    ... 21 Food and Drugs 2 2012-04-01 2012-04-01 false Enriched bromated flour. 137.160 Section 137.160... Flours and Related Products § 137.160 Enriched bromated flour. Enriched bromated flour conforms to the..., prescribed for enriched flour by § 137.165, except that potassium bromate is added in a quantity not...

  15. 21 CFR 137.160 - Enriched bromated flour.

    Code of Federal Regulations, 2014 CFR

    2014-04-01

    ... 21 Food and Drugs 2 2014-04-01 2014-04-01 false Enriched bromated flour. 137.160 Section 137.160... Flours and Related Products § 137.160 Enriched bromated flour. Enriched bromated flour conforms to the..., prescribed for enriched flour by § 137.165, except that potassium bromate is added in a quantity not...

  16. 21 CFR 137.160 - Enriched bromated flour.

    Code of Federal Regulations, 2011 CFR

    2011-04-01

    ... 21 Food and Drugs 2 2011-04-01 2011-04-01 false Enriched bromated flour. 137.160 Section 137.160... Flours and Related Products § 137.160 Enriched bromated flour. Enriched bromated flour conforms to the..., prescribed for enriched flour by § 137.165, except that potassium bromate is added in a quantity not...

  17. Alternative Fuels Data Center: Maine Transportation Data for Alternative

    Science.gov Websites

    connect with other local stakeholders. Gasoline Diesel Natural Gas Transportation Fuel Consumption Source Renewable Power Plants 58 Renewable Power Plant Capacity (nameplate, MW) 984 Source: BioFuels Atlas from the $2.96/gallon $2.66/GGE Source: Average prices per gasoline gallon equivalent (GGE) for the New England

  18. Alternative Fuels Data Center: Hawaii Transportation Data for Alternative

    Science.gov Websites

    Diesel Natural Gas Transportation Fuel Consumption Source: State Energy Data System based on beta data Plant Capacity (nameplate, MW) 145 Source: BioFuels Atlas from the National Renewable Energy Laboratory $2.96/gallon $2.66/GGE Source: Average prices per gasoline gallon equivalent (GGE) for the West Coast

  19. Alternative Fuels Data Center: Montana Transportation Data for Alternative

    Science.gov Websites

    . Gasoline Diesel Natural Gas Transportation Fuel Consumption Source: State Energy Data System based on beta Renewable Power Plant Capacity (nameplate, MW) 2,955 Source: BioFuels Atlas from the National Renewable /gallon $2.66/GGE Source: Average prices per gasoline gallon equivalent (GGE) for the Rocky Mountain PADD

  20. Method for modeling driving cycles, fuel use, and emissions for over snow vehicles.

    PubMed

    Hu, Jiangchuan; Frey, H Christopher; Sandhu, Gurdas S; Graver, Brandon M; Bishop, Gary A; Schuchmann, Brent G; Ray, John D

    2014-07-15

    As input to a winter use plan, activity, fuel use, and tailpipe exhaust emissions of over snow vehicles (OSV), including five snow coaches and one snowmobile, were measured on a designated route in Yellowstone National Park (YNP). Engine load was quantified in terms of vehicle specific power (VSP), which is a function of speed, acceleration, and road grade. Compared to highway vehicles, VSP for OSVs is more sensitive to rolling resistance and less sensitive to aerodynamic drag. Fuel use rates increased linearly (R2>0.96) with VSP. For gasoline-fueled OSVs, fuel-based emission rates of carbon monoxide (CO) and nitrogen oxides (NOx) typically increased with increasing fuel use rate, with some cases of very high CO emissions. For the diesel OSVs, which had selective catalytic reduction and diesel particulate filters, fuel-based NOx and particulate matter (PM) emission rates were not sensitive to fuel flow rate, and the emission controls were effective. Inter vehicle variability in cycle average fuel use and emissions rates for CO and NOx was substantial. However, there was relatively little inter-cycle variation in cycle average fuel use and emission rates when comparing driving cycles. Recommendations are made regarding how real-world OSV activity, fuel use, and emissions data can be improved.