Science.gov

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. Hydrogen-enriched fuels

    SciTech Connect

    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 conventional 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.

  3. Enrichment Monitor for 235U Fuel Tubes

    SciTech Connect

    Winn, W.G.

    2001-08-22

    This report describes the performance of this prototype y-monitor of 235 Uranium enrichment. In this proposed method y-rates associated with 235U and 232U are correlated with enrichment. Instrumentation for appraising fuel tubes with this method has been assembled and tested.

  4. Overview of reduced enrichment fuels: Development, testing, and specification

    SciTech Connect

    Snelgrove, J.L.

    1987-01-01

    The US Reduced Enrichment Research and Test Reactor (RERTR) Program was established in 1978 to provide the technical means to operate research and test reactors with low enrichment uranium (LEU) fuels without significant penalty in experiment performance, operation costs, component modifications, or safety characteristics. This paper discusses relevant developments in fuel developments. 9 refs., 1 tab.

  5. Qualification status of LEU (low-enriched uranium) fuels

    SciTech Connect

    Snelgrove, J.L.

    1987-01-01

    Sufficient data has been obtained from tests of high-density, low-enriched fuels for research and test reactors to declare them qualified for use. These fuels include UZrH/sub x/ (TRIGA fuel) and UO/sub 2/ (SPERT fuel) for rod-type reactors and UA1/sub x/, U/sub 3/O/sub 8/, U/sub 3/Si/sub 2/, and U/sub 3/Si dispersed in aluminum for plate-type reactors. Except for U/sub 3/Si, the allowable fission density for LEU applications is limited only by the available /sup 235/U. Several reactors are now using these fuels, and additional conversions are in progress. The basic performance characteristics and limits, if any, of the qualified low-enriched (and medium-enriched) fuels are discussed. Continuing and planned work to qualify additional fuels is also discussed.

  6. Delays hit conversion to low enriched uranium fuel

    NASA Astrophysics Data System (ADS)

    Allen, Michael

    2016-03-01

    Eliminating highly enriched uranium (HEU) fuel from civilian research reactors around the world will take a lot longer than anticipated, according to a new study by the US National Academies of Sciences, Engineering and Medicine.

  7. Low enrichment fuel conversion for Iowa State University

    SciTech Connect

    Rohach, A.F.

    1992-08-01

    This report discusses the UTR-10 reactor at Iowa State University which went critical on low enriched uranium (LEU) fuel on August 14, 1991. However, subsequent to the criticality experiments the fuel plates started to discolor. In addition, roll pins used to lift the fuel assemblies were discovered to be cracked. It was determined that these problems were due to chemical agents in the primary coolant water. The roll pins were replaced by solid stainless steel pins. The primary coolant was replaced and the reactor is currently in operation. Surveillance specimens will be used to monitor any possible future discoloration. The high enriched fuel (HEU) is being prepared for eventual shipment to a high enriched fuel receiving facility.

  8. Fuel optimum low-thrust elliptic transfer using numerical averaging

    NASA Astrophysics Data System (ADS)

    Tarzi, Zahi; Speyer, Jason; Wirz, Richard

    2013-05-01

    Low-thrust electric propulsion is increasingly being used for spacecraft missions primarily due to its high propellant efficiency. As a result, a simple and fast method for low-thrust trajectory optimization is of great value for preliminary mission planning. However, few low-thrust trajectory tools are appropriate for preliminary mission design studies. The method presented in this paper provides quick and accurate solutions for a wide range of transfers by using numerical orbital averaging to improve solution convergence and include orbital perturbations. Thus, preliminary trajectories can be obtained for transfers which involve many revolutions about the primary body. This method considers minimum fuel transfers using first-order averaging to obtain the fuel optimum rates of change of the equinoctial orbital elements in terms of each other and the Lagrange multipliers. Constraints on thrust and power, as well as minimum periapsis, are implemented and the equations are averaged numerically using a Gausian quadrature. The use of numerical averaging allows for more complex orbital perturbations to be added in the future without great difficulty. The effects of zonal gravity harmonics, solar radiation pressure, and thrust limitations due to shadowing are included in this study. The solution to a transfer which minimizes the square of the thrust magnitude is used as a preliminary guess for the minimum fuel problem, thus allowing for faster convergence to a wider range of problems. Results from this model are shown to provide a reduction in propellant mass required over previous minimum fuel solutions.

  9. Diesel cogeneration plant using oxygen enriched air and emulsified fuels

    SciTech Connect

    Marciniak, T.J.; Cole, R.L.; Sekar, R.R.; Stodolsky, F. ); Eustis, J.N. )

    1990-01-01

    The investigation of oxygen-enriched combustion of alternative fuels in diesel engines at Argonne National Laboratory (ANL) is based on information gathered from two previous Department of Energy programs. The first was the slow-speed diesel engine program which used fuels such as coal-water slurry and coal derived liquid fuels in a slow speed diesel engine. The second was the development of membrane oxygen separation equipment. The results of these programs indicated that using the new membrane oxygen enrichment technology with medium- and high-speed diesel engines would do two things. First, oxygen enrichment could reduce some emissions from stationary diesel engines, particularly smoke, particulates and hydrocarbons while significantly increasing power output. The second, was that it might be possible to use less expensive liquid fuels such as No. 4, No. 6 and residual oil emulsified with water in medium- to high-speed diesel engines. The water would (1) help to eliminate the undesirable increase in nitrogen oxide production when enriched oxygen is used, and (2) by reducing the viscosity of the heavier liquid fuels, make them easier to use in smaller industrial cogeneration applications. This program consists of four steps: preliminary feasibility study, exploratory experiments, system development, and demonstration and commercialization of an industrial cogeneration system. 3 refs., 13 figs.

  10. Evaluation of oxygen-enrichment system for alternative fuel vehicles

    SciTech Connect

    Poola, R.B.; Sekar, R.R.; Ng, H.K.

    1995-12-01

    This report presents results on the reduction in exhaust emissions achieved by using oxygen-enriched intake air on a flexible fuel vehicle (FFV) that used Indolene and M85 as test fuels. The standard federal test procedure (FTP) and the US Environmental Protection Agency`s (EPA`s) off-cycle (REP05) test were followed. The report also provides a review of literature on the oxygen membrane device and design considerations. It presents information on the sources and contributions of cold-phase emissions to the overall exhaust emissions from light-duty vehicles (LDVs) and on the various emission standards and present-day control technologies under consideration. The effects of oxygen-enriched intake air on FTP and off-cycle emissions are discussed on the basis of test results. Conclusions are drawn from the results and discussion, and different approaches for the practical application of this technology in LDVs are recommended.

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

    SciTech Connect

    Bzandes, S.; Lonhert, G.

    1982-07-01

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

  12. Effect of reduced enrichment on the fuel cycle for research reactors

    SciTech Connect

    Travelli, A.

    1982-01-01

    The new fuels developed by the RERTR Program and by other international programs for application in research reactors with reduced uranium enrichment (<20% EU) are discussed. It is shown that these fuels, combined with proper fuel-element design and fuel-management strategies, can provide at least the same core residence time as high-enrichment fuels in current use, and can frequently significantly extend it. The effect of enrichment reduction on other components of the research reactor fuel cycle, such as uranium and enrichment requirements, fuel fabrication, fuel shipment, and reprocessing are also briefly discussed with their economic implications. From a systematic comparison of HEU and LEU cores for the same reference research reactor, it is concluded that the new fuels have a potential for reducing the research reactor fuel cycle costs while reducing, at the same time, the uranium enrichment of the fuel.

  13. 49 CFR 525.11 - Termination of exemption; amendment of alternative average fuel economy standard.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 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...

  14. 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...

  15. 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...

  16. 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...

  17. 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...

  18. 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

    ... COMMISSION Notice of Availability of Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding...: Ty Naquin, Project Manager, Uranium Enrichment Branch, Division of Fuel Cycle Safety and Safeguards... amended. The introduction of uranium hexafluoride into any module of the National Enrichment Facility...

  19. ZPR-3 Assembly 6F : A spherical assembly of highly enriched uranium, depleted uranium, aluminum and steel with an average {sup 235}U enrichment of 47 atom %.

    SciTech Connect

    Lell, R. M.; McKnight, R. D; Schaefer, R. W.; Nuclear Engineering Division

    2010-09-30

    Over a period of 30 years, more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited for nuclear data validation and to form the basis for criticality safety benchmarks. A number of the Argonne ZPR/ZPPR critical assemblies have been evaluated as ICSBEP and IRPhEP benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. ZPR-3 Assembly 6 consisted of six phases, A through F. In each phase a critical configuration was constructed to simulate a very simple shape such as a slab, cylinder or sphere that could be analyzed with the limited analytical tools available in the 1950s. In each case the configuration consisted of a core region of metal plates surrounded by a thick depleted uranium metal reflector. The average compositions of the core configurations were essentially identical in phases A - F. ZPR-3

  20. 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... PROPERTY 34-MOTOR VEHICLE MANAGEMENT Obtaining Fuel Efficient Motor Vehicles § 102-34.55 Are there fleet... that each executive agency meet the fleet average fuel economy standards in place as of January 1...

  1. 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...

  2. 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

    ... COMMISSION Notice of Availability of Uranium Enrichment Fuel Cycle Facility's Inspection Reports Regarding... CONTACT: Gregory Chapman, Project Manager, Uranium Enrichment Branch, Division of Fuel Cycle Safety and... Energy Act of 1954, as amended. The introduction of uranium hexafluoride into any module of the...

  3. Conversion of TRIGA research reactors from high-enriched- to low-enriched-uranium fuels: owner/operator view

    SciTech Connect

    Feltz, D.E.

    1986-01-01

    In June 1985, the US Nuclear Regulatory Commission (NRC) commissioners issued a four-point directive to the NRC staff concerning the conversion of research reactors from the use of high-enriched-uranium (HEU) to low-enriched-uranium (LEU) fuels. As a result of this directive, the earlier concerns of the research reactor community that were presented to the NRC during the comment period of the 1984 proposed rule on HEU-LEU conversion must be dealt with now. This paper discusses the items of most concern to HEU TRIGA owner/operators for conversion to LEU fuel.

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

    SciTech Connect

    Primm, Trent; Guida, Tracey

    2010-02-01

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

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

    SciTech Connect

    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 storage sites for spent fuel in the USA by disposing of the spent fuel in deep sub-seabed sediments after storing the spent fuel at U.S. Navy Nuclear Shipyards that have the space and all of the necessary equipment and security already in place. The non-proliferating fuel cycle also reduces transportation risks by utilizing barges for the collection of spent fuel and transport to the Navy shipyards and specially designed ships to take the spent fuel to designated disposal sites at sea and to dispose of them there in deep sub-seabed sediments. Disposal in the sub-seabed sediments practically eliminates human intrusion. Potential disposal sites include Great Meteor East and Southern Nares Abyssal Plain. Such sites then could easily become international disposal sites since they occur in the open ocean. It also reduces the level of human exposure in case of failure because of the large physical and chemical dilution and the elimination of a major pathway to man-seawater is not potable. Of course, the recovery of uranium from sea water and the disposal of spent fuel in sub-seabed sediments must be proven on an industrial scale. All other technologies are already operating on an industrial scale. If externalities, such as reduced terrorist threats, environmental damage (including embedded emissions), long

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

    SciTech Connect

    Bradley, E.; Adelfang, P.; Goldman, I.N.

    2008-07-15

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

  7. Low enrichment fuel conversion for Iowa State University

    SciTech Connect

    Rohach, A.F.; Hendrickson, R.A.

    1990-08-01

    Work during the reported period was centered primarily in preparation for receiving the LEU fuel and the shipping of the HEU fuel. The LEU fuel has not been received. The HEU fuel assemblies for the UTR-10 reactor will not fit into any current research reactor shipping containers; therefore, the fuel assemblies must be disassembled and the fuel shipped as fuel plates. Procedures and practices have been developed so that the fuel assemblies will be disassembled in a shielded environment.

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

    SciTech Connect

    Ilas, Germina; Primm, Trent

    2009-11-01

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

  9. ZPR-3 Assembly 12 : A cylindrical assembly of highly enriched uranium, depleted uranium and graphite with an average {sup 235}U enrichment of 21 atom %.

    SciTech Connect

    Lell, R. M.; McKnight, R. D.; Perel, R. L.; Wagschal, J. J.; Nuclear Engineering Division; Racah Inst. of Physics

    2010-09-30

    Over a period of 30 years, more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited for nuclear data validation and to form the basis for criticality safety benchmarks. A number of the Argonne ZPR/ZPPR critical assemblies have been evaluated as ICSBEP and IRPhEP benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. ZPR-3 Assembly 12 (ZPR-3/12) was designed as a fast reactor physics benchmark experiment with an average core {sup 235}U enrichment of approximately 21 at.%. Approximately 68.9% of the total fissions in this assembly occur above 100 keV, approximately 31.1% occur below 100 keV, and essentially none below 0.625 eV - thus the classification as a 'fast' assembly. This assembly is Fast Reactor Benchmark No. 9 in the Cross Section Evaluation Working Group (CSEWG) Benchmark Specifications

  10. Final qualification testing results for TRIGA low-enriched uranium fuel

    SciTech Connect

    West, G.B.; Simnad, M.T.; Copeland, G.L.

    1987-01-01

    Following the adoption of policies by the US government and other fuel supplier nations to limit, with few exceptions, export of fuels to those enriched to < 20% in /sup 235/U, GA Technologies undertook, starting in 1976, a rigorous program to develop uranium zirconium-hydride fuels with high uranium concentrations up to 3.7 g U/cm/sup 3/, while limiting the enrichment to < 20% in /sup 235/U. By increasing the uranium concentration from 8.5 to 45 wt% and by including erbium as a burnable poison, the long reactivity lifetime of the high-enriched uranium cores has been preserved. The achieved purpose of these low-enriched uranium (LEU) fuels was also to preserve all the unique safety features of the TRIGA fuel system - prompt negative temperature coefficient of reactivity, high fission product retentivity, chemical stability when quenched from high temperatures in water, and dimensional stability over a broad range of operating temperatures.

  11. 41 CFR 102-34.55 - Are there fleet average fuel economy standards we must meet?

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 41 Public Contracts and Property Management 3 2011-01-01 2011-01-01 false Are there fleet average fuel economy standards we must meet? 102-34.55 Section 102-34.55 Public Contracts and Property... PROPERTY 34-MOTOR VEHICLE MANAGEMENT Obtaining Fuel Efficient Motor Vehicles § 102-34.55 Are there...

  12. 41 CFR 102-34.55 - Are there fleet average fuel economy standards we must meet?

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 41 Public Contracts and Property Management 3 2013-07-01 2013-07-01 false Are there fleet average fuel economy standards we must meet? 102-34.55 Section 102-34.55 Public Contracts and Property... PROPERTY 34-MOTOR VEHICLE MANAGEMENT Obtaining Fuel Efficient Motor Vehicles § 102-34.55 Are there...

  13. 41 CFR 102-34.55 - Are there fleet average fuel economy standards we must meet?

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 41 Public Contracts and Property Management 3 2014-01-01 2014-01-01 false Are there fleet average fuel economy standards we must meet? 102-34.55 Section 102-34.55 Public Contracts and Property... PROPERTY 34-MOTOR VEHICLE MANAGEMENT Obtaining Fuel Efficient Motor Vehicles § 102-34.55 Are there...

  14. Development of Enriched Borated Aluminum Alloy for Basket Material of Cask for Spent Nuclear Fuel

    SciTech Connect

    Katsura Kajihara; Yasuhiro Aruga; Jun Shimojo; Hiroaki Taniuchi; Tsutomu Takeda; Masatosi Sasaki

    2002-07-01

    New enriched borated aluminum alloys manufactured by melting process are developed, which resulted in supplying structural basket materials for spent nuclear fuel packagings. In this process, the borated aluminum alloys were melted in a vacuum induction furnace at elevated temperature than that of ordinary aluminum melting processes. Boron dissolves into the matrix at the temperature of 1273 K or more, and fine aluminum diboride is precipitated and uniformly dispersed upon cooling rapidity. It is confirmed that boron is homogeneously dispersed with the fine particles of approximate 5 in average size in the product. Tensile strength and creep property at elevated temperature in 1 mass-%B 6061-T651 plate and 1 mass-%B 3004 extruded rectangular pipe as structural materials are examined. It is confirmed that the both of borated aluminum alloys have stable strength and creep properties that are similar to those of ordinary aluminum alloys. (authors)

  15. 78 FR 23312 - Uranium Enrichment Fuel Cycle Inspection Reports Regarding Louisiana Energy Services, National...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-04-18

    ... COMMISSION Uranium Enrichment Fuel Cycle Inspection Reports Regarding Louisiana Energy Services, National... shell operations such as: movement of cylinders; fire walls; transient combustible inspections; cylinder movers; and worker evacuation. The NRC staff has prepared inspection reports documenting its findings...

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

    SciTech Connect

    Pesic, Milan P

    2003-10-15

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

  17. Low enrichment fuel conversion for Iowa State University. Progress report, August 1, 1991--July 31, 1992

    SciTech Connect

    Rohach, A.F.

    1992-08-01

    This report discusses the UTR-10 reactor at Iowa State University which went critical on low enriched uranium (LEU) fuel on August 14, 1991. However, subsequent to the criticality experiments the fuel plates started to discolor. In addition, roll pins used to lift the fuel assemblies were discovered to be cracked. It was determined that these problems were due to chemical agents in the primary coolant water. The roll pins were replaced by solid stainless steel pins. The primary coolant was replaced and the reactor is currently in operation. Surveillance specimens will be used to monitor any possible future discoloration. The high enriched fuel (HEU) is being prepared for eventual shipment to a high enriched fuel receiving facility.

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

    SciTech Connect

    SCHWINKENDORF, K.N.

    2006-05-12

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

  19. 49 CFR 537.9 - Determination of fuel economy values and average fuel economy.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... under paragraph (c) of this section and has been determined and approved under 40 CFR part 600, the...)(1) of this section for which a fuel economy value approved under 40 CFR part 600, does not exist... section for which a fuel economy value has been neither determined nor approved under 40 CFR part 600,...

  20. 49 CFR 537.9 - Determination of fuel economy values and average fuel economy.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... under paragraph (c) of this section and has been determined and approved under 40 CFR part 600, the...)(1) of this section for which a fuel economy value approved under 40 CFR part 600, does not exist... section for which a fuel economy value has been neither determined nor approved under 40 CFR part 600,...

  1. 49 CFR 537.9 - Determination of fuel economy values and average fuel economy.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... under paragraph (c) of this section and has been determined and approved under 40 CFR part 600, the...)(1) of this section for which a fuel economy value approved under 40 CFR part 600, does not exist... section for which a fuel economy value has been neither determined nor approved under 40 CFR part 600,...

  2. 49 CFR 537.9 - Determination of fuel economy values and average fuel economy.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... under paragraph (c) of this section and has been determined and approved under 40 CFR part 600, the...)(1) of this section for which a fuel economy value approved under 40 CFR part 600, does not exist... section for which a fuel economy value has been neither determined nor approved under 40 CFR part 600,...

  3. 49 CFR 537.9 - Determination of fuel economy values and average fuel economy.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... under paragraph (c) of this section and has been determined and approved under 40 CFR part 600, the...)(1) of this section for which a fuel economy value approved under 40 CFR part 600, does not exist... section for which a fuel economy value has been neither determined nor approved under 40 CFR part 600,...

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

    SciTech Connect

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

    2010-10-01

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

  5. Fuel development activities of the US RERTR Program. [Reduced Enrichment Research and Test Reactor

    SciTech Connect

    Snelgrove, J.L.; Domagala, R.F.; Wiencek, T.C.; Copeland, G.L.

    1983-01-01

    Progress in the development and irradiation testing of high-density fuels for use with low-enriched uranium in research and test reactors is reported. Swelling and blister-threshold temperature data obtained from the examination of miniature fuel plates containing UAl/sub x/, U/sub 3/O/sub 8/, U/sub 3/Si/sub 2/, or U/sub 3/Si dispersed in an aluminum matrix are presented. Combined with the results of metallurgical examinations, these data show that these four fuel types will perform adequately to full burnup of the /sup 235/U contained in the low-enriched fuel. The exothermic reaction of the uranium-silicide fuels with aluminum has been found to occur at about the same temperature as the melting of the aluminum matrix and cladding and to be essentially quenched by the melting endotherm. A new series of miniature fuel plate irradiations is also discussed.

  6. Irradiation testing of full-sized, reduced-enrichment fuel elements

    SciTech Connect

    Snelgrove, J.L.; Copeland, G.L.

    1983-01-01

    The current status of the irradiation testing of full-sized, reduced-enrichment fuel elements and fuel rods under the US Reduced Enrichment Research and Test Reactor Program is reported. Being tested are UAl/sub x/-Al, U/sub 3/O/sub 8/-Al, U/sub 3/Si/sub 2/-Al, and U/sub 3/Si-Al dispersion fuels and UZrH/sub x/ (TRIGA) fuel at uranium densities in the fuel meat ranging from 1.7 to 6.0 Mg/m/sup 3/. Generally good performance has been experienced to date. Some preliminary results of postirradiation examinations are also included. A whole-core demonstration in the Oak Ridge Research Reactor is planned. Some details of this demonstration are provided.

  7. 41 CFR 102-34.55 - Are there fleet average fuel economy standards we must meet?

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 41 Public Contracts and Property Management 3 2012-01-01 2012-01-01 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...

  8. 40 CFR 600.510-86 - Calculation of average fuel economy.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 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 § 600... 0.1 mpg for the classes of automobiles identified herein, and the results of such calculations...

  9. 40 CFR 600.510-93 - Calculation of average fuel economy.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... meet the minimum driving range requirements established by the Secretary of Transportation (49 CFR part... 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...

  10. 40 CFR 600.510-93 - Calculation of average fuel economy.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... meet the minimum driving range requirements established by the Secretary of Transportation (49 CFR part... 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 §...

  11. Low-Enriched Fuel Design Concept for the Prismatic Very High Temperature Reactor Core

    SciTech Connect

    Sterbentz, James W

    2007-05-01

    A new non-TRISO fuel and clad design concept is proposed for the prismatic, heliumcooled Very High Temperature Reactor core. The new concept could substantially reduce the current 10-20 wt% TRISO uranium enrichments down to 4-6 wt% for both initial and reload cores. The proposed fuel form would be a high-temperature, high-density uranium ceramic, for example UO2, configured into very small diameter cylindrical rods. The small diameter fuel rods significantly increase core reactivity through improved neutron moderation and fuel lumping. Although a high-temperature clad system for the concept remains to be developed, recent success in tube fabrication and preliminary irradiation testing of silicon carbide (SiC) cladding for light water reactor applications offers good potential for this application, and for future development of other carbide clad designs. A high-temperature ceramic fuel, together with a high-temperature clad material, could also lead to higher thermal safety margins during both normal and transient reactor conditions relative to TRISO fuel. The calculated neutronic results show that the lowenrichment, small diameter fuel rods and low thermal neutron absorbing clad retain the strong negative Doppler fuel temperature coefficient of reactivity that ensures inherent safe operation of the VHTR, and depletion studies demonstrate that an 18-month power cycle can be achieved with the lower enrichment fuel.

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

    SciTech Connect

    TOFFER, H.

    2006-07-18

    With the shutdown of the Hanford PUREX (Plutonium-Uranium Extraction Plant) reprocessing plant in the 1970s, adequate storage capacity for spent Hanford N Reactor fuel elements in the K and N Reactor pools became a concern. To maximize space utilization in the pools, accounting for fuel burnup was considered. Fuel that had experienced a neutron environment in a reactor is known as spent, exposed, or irradiated fuel. In contrast fuel that has not yet been placed in a reactor is known as green, unexposed, or unirradiated fuel. Calculations indicated that at typical fuel exposures for N Reactor, the spent-fuel critical mass would be twice the critical mass for green fuel. A decision was reached to test the calculational result with a definitive experiment. If the results proved positive, storage capacity could be increased and N Reactor operation could be prolonged. An experiment to be conducted in the N Reactor spent-fuel storage pool was designed and assembled (References 1 and 2) and the services of the Battelle Northwest Laboratories (BNWL) (now Pacific Northwest National Laboratory [PNNL]) critical mass laboratory were procured for the measurements (Reference 3). The experiments were performed in April 1975 in the Hanford N Reactor fuel storage pool. The fuel elements were MKIA fuel assemblies, comprised of two concentric tubes of low-enriched metallic uranium. Two separate sets of measurements were performed: one with unirradiated fuel and one with irradiated fuel. Both the unirradiated and irradiated fuel, were measured in the same geometry. The spent-fuel MKIA assemblies had an average burnup of 2865 MWd (megawatt days)/t. A constraint was imposed restricting the measurements to a subcritical limit of k{sub eff} = 0.97. Subcritical count rate data was obtained with pulsed-neutron and approach-to-critical measurements. Ten (10) configurations with green fuel and nine (9) configurations with spent fuel are described and evaluated. Of these, three (3) green fuel

  13. High-density reduced-enrichment fuels for Research and Test Reactors

    SciTech Connect

    Snelgrove, J.L.; Hofman, G.L.; Copeland, G.L.

    1983-01-01

    Development and irradiation testing of high-density fuels have been conducted by the US RERTR Program in order to provide the technical means to reduce the enrichment of fuels for research and test reactors. The traditional aluminum dispersion fuel technology has been extended to include the highest practical loadings of uranium-aluminide (UAl/sub x/, 2.3 MgU/m/sup 3/), uranium-oxide (U/sub 3/O/sub 8/, 3.2 MgU/m/sup 3/), and uranium-silicide (U/sub 3/Si/sub 2/, 5.5 MgU/m/sup 3/; U/sub 3/Si, 7.0 MgU/m/sup 3/) fuels. A third uranium-silicide alloy, U/sub 3/SiAl (U + 3.5 wt % Si + 1.5 wt % Al) has been found to perform poorly at high burnup. Testing of miniature fuel plates and full-sized fuel elements is at an advanced stage for the highest loadings of the aluminide and oxide fuels and intermediate loadings of the silicide fuels, and good results have been obtained for low-enriched uranium. The data obtained to date are discussed. 1 reference, 3 figures, 1 table.

  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

    ... COMMISSION Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services LLC... Act of 1954, as amended. The introduction of uranium hexafluoride into any module of the National... documents. The ADAMS accession numbers for the documents related to this document are: Inspection...

  15. 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

    ... well as autoclave one of the facility have been constructed in accordance with the requirements of the... number 1.5, 1.6, 1.7, 1.8, 2.1, and 2.4 as well as autoclave one of the facility have been constructed in... COMMISSION Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services...

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

    SciTech Connect

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

    2008-02-01

    The National Nuclear Security Agency Global Threat Reduction Initiative (GTRI) is tasked with minimizing the use of high-enriched uranium (HEU) worldwide. A key component of that effort is the conversion of research reactors from HEU to low-enriched uranium (LEU) fuels. The GTRI Convert Fuel Development program, previously known as the Reduced Enrichment for Research and Test Reactors program was initiated in 1978 by the United States Department of Energy to develop the nuclear fuels necessary to enable these conversions. The program cooperates with the research reactors’ operators to achieve this goal of HEU to LEU conversion without reduction in reactor performance. The programmatic mandate is to complete the conversion of all civilian domestic research reactors by 2014. These reactors include the five domestic high-performance research reactors (HPRR), namely: the High Flux Isotope Reactor at the Oak Ridge National Laboratory, the Advanced Test Reactor at the Idaho National Laboratory, the National Bureau of Standards Reactor at the National Institute of Standards and Technology, the Missouri University Research Reactor at the University of Missouri–Columbia, and the MIT Reactor-II at the Massachusetts Institute of Technology. Characteristics for each of the HPRRs are given in Appendix A. The GTRI Convert Fuel Development program is currently engaged in the development of a novel nuclear fuel that will enable these conversions. The fuel design is based on a monolithic fuel meat (made from a uranium-molybdenum alloy) clad in Al-6061 that has shown excellent performance in irradiation testing. The unique aspects of the fuel design, however, necessitate the development and implementation of new fabrication techniques and, thus, establishment of the infrastructure to ensure adequate fuel fabrication capability. A conceptual fabrication process description and rough estimates of the total facility throughput are described in this document as a basis for

  17. Low temperature combustion using nitrogen enrichment to mitigate NOx from large bore natural gas fueled engines.

    SciTech Connect

    Biruduganti, M.; Gupta, S.; Sekar, R.; Energy Systems

    2010-01-01

    Low temperature combustion is identified as one of the pathways to meet the mandatory ultra low NO{sub x} emissions levels set by the regulatory agencies. Exhaust gas recirculation (EGR) is a well known technique to realize low NO{sub x} emissions. However, EGR has many built-in adverse ramifications that negate its advantages in the long term. This paper discusses nitrogen enrichment of intake air using air separation membranes as a better alternative to the mature EGR technique. This investigation was undertaken to determine the maximum acceptable level of nitrogen enrichment of air for a single-cylinder spark-ignited natural gas engine. NO{sub x} reduction as high as 70% was realized with a modest 2% nitrogen enrichment while maintaining power density and simultaneously improving fuel conversion efficiency (FCE). Any enrichment beyond this level degraded engine performance in terms of power density, FCE, and unburned hydrocarbon emissions. The effect of ignition timing was also studied with and without N{sub 2} enrichment. Finally, lean burn versus stoichiometric operation utilizing nitrogen enrichment was compared. Analysis showed that lean burn operation along with nitrogen enrichment is one of the effective pathways for realizing better FCE and lower NO{sub x} emissions.

  18. Conversion and standardization of university reactor fuels using low-enrichment uranium: Plans and schedules

    SciTech Connect

    Young, H.H.; Brown, K.R.; Matos, J.E.

    1986-01-01

    The highly-enriched uranium (HEU) fuel used in twenty United States university reactors can be viewed as contributing to the risk of theft or diversion of weapons-useable material. To minimize this risk, the US Nuclear Regulatory Commission issued its final rule on ''Limiting the Use of Highly Enriched Uranium in Domestically Licensed Research and Test Reactors,'' in February 1986. This paper describes the plans and schedules developed by the US Department of Energy to coordinate an orderly transition from HEU to LEU fuel in most of these reactors. An important element in the planning process has been the desire to standardize the LEU fuels used in US university reactors and to enhance the performance and utilization of a number of these reactors. The program is estimated to cost about $10 million and to last about five years.

  19. 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

    ... must meet the minimum driving range requirements established by the Secretary of Transportation (49 CFR... 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...

  20. 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

    ... must meet the minimum driving range requirements established by the Secretary of Transportation (49 CFR... 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...

  1. Conversion and Evaluation of the University of Massachusetts Lowell Research Reactor From High-Enriched To Low-Enriched Uranium Fuel

    SciTech Connect

    Leo M. Bobek

    2003-11-19

    The process for converting the University of Massachusetts Lowell Research Reactor (UMLRR) from high-enrichment uranium (HEU) fuel to low-enrichment uranium (LEU) fuel began in 1988. Several years of design reviews, computational modeling, and thermal hydraulic analyses resulted in a preliminary reference core design and configuration based on 20 standard, MTR-type, flat-plate, 19.75% enriched, uranium silicide (u3Si2) fuel elements. A final safety analysis for the fuel conversion was submitted to the Nuclear Regulatory Commission (NRC) in 1993. The NRC made two additional requests for additional information and supplements were submitted in 1994 and 1997. The new UMLRR Reactor Supervisor initiated an effort to change the LEU reference core configuration to eliminate a complicated control rod modification needed for the smaller core.

  2. 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.

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

    SciTech Connect

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

    2014-10-30

    Engineering design studies of the feasibility of conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL) as part of an effort sponsored by the U.S. Department of Energy’s Global Threat Reduction Initiative (GTRI)/Reduced Enrichment for Research and Test Reactors (RERTR) program. The fuel type selected by the program for the conversion of the five high-power research reactors in the U.S. that still use HEU fuel is a new U-Mo monolithic fuel. Studies by ORNL have previously indicated that HFIR can be successfully converted using the new fuel provided (1) the reactor power can be increased from 85 MW to 100 MW and (2) the fuel can be fabricated to a specific reference design. Fabrication techniques for the new fuel are under development by the program but are still immature, especially for the “complex” aspects of the HFIR fuel design. In FY 2012, the program underwent a major shift in focus to emphasize developing and qualifying processes for the fabrication of reliable and affordable LEU fuel. In support of this new focus and in an effort to ensure that the HFIR fuel design is as suitable for reliable fabrication as possible, ORNL undertook the present study to propose and evaluate several alternative design features. These features include (1) eliminating the fuel zone axial contouring in the previous reference design by substituting a permanent neutron absorber in the lower unfueled region of all of the fuel plates, (2) relocating the burnable neutron absorber from the fuel plates of the inner fuel element to the side plates of the inner fuel element (the fuel plates of the outer fuel element do not contain a burnable absorber), (3) relocating the fuel zone inside the fuel plate to be centered on the centerline of the depth of the plate, and (4) reshaping the radial contour of the relocated fuel zone to be symmetric about this centerline. The

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

    SciTech Connect

    Renfro, David G; Chandler, David; Cook, David Howard; Ilas, Germina; Jain, Prashant K; Valentine, Jennifer R

    2014-11-01

    Engineering design studies of the feasibility of conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL) as part of an effort sponsored by the U.S. Department of Energy s Global Threat Reduction Initiative (GTRI)/Reduced Enrichment for Research and Test Reactors (RERTR) program. The fuel type selected by the program for the conversion of the five high-power research reactors in the U.S. that still use HEU fuel is a new U-Mo monolithic fuel. Studies by ORNL have previously indicated that HFIR can be successfully converted using the new fuel provided (1) the reactor power can be increased from 85 MW to 100 MW and (2) the fuel can be fabricated to a specific reference design. Fabrication techniques for the new fuel are under development by the program but are still immature, especially for the complex aspects of the HFIR fuel design. In FY 2012, the program underwent a major shift in focus to emphasize developing and qualifying processes for the fabrication of reliable and affordable LEU fuel. In support of this new focus and in an effort to ensure that the HFIR fuel design is as suitable for reliable fabrication as possible, ORNL undertook the present study to propose and evaluate several alternative design features. These features include (1) eliminating the fuel zone axial contouring in the previous reference design by substituting a permanent neutron absorber in the lower unfueled region of all of the fuel plates, (2) relocating the burnable neutron absorber from the fuel plates of the inner fuel element to the side plates of the inner fuel element (the fuel plates of the outer fuel element do not contain a burnable absorber), (3) relocating the fuel zone inside the fuel plate to be centered on the centerline of the depth of the plate, and (4) reshaping the radial contour of the relocated fuel zone to be symmetric about this centerline. The present

  5. 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

    ... excluded from both the calculation of the fleet average standard for a manufacturer under 40 CFR 86.1818-12... megagrams, from 40 CFR 86.1868-12(c), and rounded to the nearest whole number; VLM = vehicle lifetime miles... technology credits for the applicable vehicle category, in megagrams, from 40 CFR 86.1869-12(e), and...

  6. 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

    ... excluded from both the calculation of the fleet average standard for a manufacturer under 40 CFR 86.1818-12... conditioning efficiency credits for the applicable vehicle category, in megagrams, from 40 CFR 86.1868-12(c... vehicle category, in megagrams, from 40 CFR 86.1869-12(e), and rounded to the nearest whole number;...

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

    SciTech Connect

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

    1985-01-01

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

  8. Effect of changes in DOE pricing policies for enrichment and reprocessing on research reactor fuel cycle costs

    SciTech Connect

    Matos, J.E.; Freese, K.E.

    1986-11-03

    Fuel cycle costs with HEU and LEU fuels for the IAEA generic 10 MW reactor are updated to reflect the change in DOE pricing policy for enrichment services as of October 1985 and the published charges for LEU reprocessing services as of February 1986. The net effects are essentially no change in HEU fuel cycle costs and a reduction of about 8 to 10% in the fuel cycle costs for LEU silicide fuel.

  9. Process for producing dry, sulfur-free, CH[sub 4]-enriched synthesis or fuel gas

    SciTech Connect

    Child, E.T.; Lafferty, W.L. Jr.; Suggitt, R.M.; Jahnke, F.C.

    1993-08-03

    A process is described for the production of a dry, sulfur-free methane enriched synthesis gas or fuel gas stream comprising: (1) cooling a particulate-free raw synthesis or fuel gas feed stream comprising H[sub 2], CO, CO[sub 2], H[sub 2]O, N[sub 2], H[sub 2]S, COS and with or without methane to a temperature in the range of about 60 F to 130 F and separating out at least a portion of water condensate; (2) mixing together said cooled raw synthesis or fuel gas from (1) and a portion of cryogenic liquefied natural gas (LNG) thereby further cooling the new synthesis or fuel gas to a temperature in the range of about [minus]75 F to 60 F; (3) directly contacting the mixture from (2) in an acid-gas removal zone with liquid acid-gas absorbent solvent thereby absorbing sulfur-containing compounds, water, and at least a portion of the CO[sub 2], and thereby producing acid-gas rich liquid absorbent solvent containing dissolved water and a dry stream of methane enriched synthesis or fuel gas; (4) separating said acid-gas rich liquid absorbent from said dry stream of methane enriched synthesis or fuel gas comprising H[sub 2], CO, CH[sub 4], and substantially no sulfur-containing gas or moisture; (5) regenerating the separated acid-gas rich liquid absorbent solvent to remove the sulfur-containing gas and the dissolved water; and (6) introducing regenerated liquid acid-gas absorbent solvent into said acid gas removal zone.

  10. Determining Spent Nuclear Fuel's Plutonium Content, Initial Enrichment, Burnup, and Cooling Time

    SciTech Connect

    Cheatham, Jesse R; Francis, Matthew W

    2011-01-01

    The Next Generation of Safeguards Initiative is examining nondestructive assay techniques to determine the total plutonium content in spent nuclear fuel. The goal of this research was to develop new techniques that can independently verify the plutonium content in a spent fuel assembly without relying on an operator's declarations. Fundamentally this analysis sought to answer the following questions: (1) do spent fuel assemblies contain unique, identifiable isotopic characteristics as a function of their burnup, cooling time, and initial enrichment; (2) how much variation can be seen in spent fuel isotopics from similar and dissimilar reactor power operations; and (3) what isotopes (if any) could be used to determine burnup, cooling time, and initial enrichment? To answer these questions, 96,000 ORIGEN cases were run that simulated typical two-cycle operations with burnups ranging from 21,900 to 72,000 MWd/MTU, cooling times from 5 to 25 years, and initial enrichments between 3.5 and 5.0 weight percent. A relative error coefficient was determined to show how numerically close a reference solution has to be to another solution for the two results to be indistinguishable. By looking at the indistinguishable solutions, it can be shown how a precise measurement of spent fuel isotopics can be inconclusive when used in the absence of an operator's declarations. Using this Method of Indistinguishable Solutions (MIS), we evaluated a prominent method of nondestructive analysis - gamma spectroscopy. From this analysis, a new approach is proposed that demonstrates great independent forensic examination potential for spent nuclear fuel by examining both the neutron emissions of Cm-244 and the gamma emissions of Cs-134 and Eu-154.

  11. Reference (Axially Graded) Low Enriched Uranium Fuel Design for the High Flux Isotope Reactor (HFIR)

    SciTech Connect

    Ilas, Germina; Primm, Trent

    2010-01-01

    During the past five years, staff at the Oak Ridge National Laboratory (ORNL) have studied the issue of whether the HFIR could be converted to low enriched uranium (LEU) fuel without degrading the performance of the reactor. Using state-of-the-art reactor physics methods and behind-the-state-of-the-art thermal hydraulics methods, the staff have developed fuel plate designs (HFIR uses two types of fuel plates) that are believed to meet physics and thermal hydraulic criteria provided the reactor power is increased from 85 to 100 MW. The paper will present a defense of the results by explaining the design and validation process. A discussion of the requirements for showing applicability of analyses to approval for loading the fuel to HFIR lead test core irradiation currently scheduled for 2016 will be provided. Finally, the potential benefits of upgrading thermal hydraulics methods will be discussed.

  12. Functionally stable and phylogenetically diverse microbial enrichments from microbial fuel cells during wastewater treatment.

    PubMed

    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/m(2), the maximum power density was 13 mW/m(2), 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

  13. 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

  14. The Nuclear Fuel Services, Inc. program to support disposition of enriched uranium-bearing materials

    SciTech Connect

    Schutt, Stephen M.; Jacob, Norman P.

    2007-07-01

    The disposition of surplus nuclear materials has become one of the most pressing issues of our time. Numerous agencies have invoked programs with the purpose of removing such materials from various international venues and disposing these materials in a manner that achieves non-proliferability. This paper describes the Nuclear Fuel Services, Inc (NFS) Nuclear Material Disposition Program, which to date has focused on a variety of Special Nuclear Material (SNM), in particular uranium of various enrichments. The major components of this program are discussed, with emphasis on recycle and return of material to the nuclear fuel cycle. (authors)

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

    SciTech Connect

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

    1990-01-01

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

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

    SciTech Connect

    Pope, M. A.; DeHart, M. D.; Morrell, S. R.; Jamison, R. K.; Nef, E. C.; Nigg, D. W.

    2015-03-01

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

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

    SciTech Connect

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

    2010-07-01

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

  18. 77 FR 68070 - 2017 and Later Model Year Light-Duty Vehicle Greenhouse Gas Emissions and Corporate Average Fuel...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-11-15

    ... Light-Duty Vehicle Greenhouse Gas Emissions and Corporate Average Fuel Economy Standards Correction In rule document 2012-21972 appearing on pages 62623-63200 in the issue of Monday, October 15, 2012,...

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

    SciTech Connect

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

    2009-07-01

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

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

    SciTech Connect

    Christopher Landers; Igor Bolshinsky; Ken Allen; Stanley Moses

    2010-07-01

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

  1. 77 FR 64051 - 2017 and Later Model Year Light-Duty Vehicle Greenhouse Gas Emissions and Corporate Average Fuel...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-10-18

    ... which was published in the Federal Register of Monday, October 15, 2012 (77 FR 62624). The final rule established fuel economy standards for light-duty vehicles under the Energy Policy and Conservation Act (EPCA... Model Year Light-Duty Vehicle Greenhouse Gas Emissions and Corporate Average Fuel Economy...

  2. 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

    ... REGULATION PERSONAL PROPERTY 34-MOTOR VEHICLE MANAGEMENT Obtaining Fuel Efficient Motor Vehicles § 102-34.60... average fuel economy for Government motor vehicles as follows: (a) Because there are so many motor vehicle... cylinder automatic transmission pick-up trucks, EPA rating: 24.3 mpg, plus (B) 150 Six cylinder...

  3. 40 CFR 600.510-08 - Calculation of average fuel economy.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... Where: Ealt = × 10 6 = energy efficiency while operating on alternative fuel rounded to the nearest 0.01... alternative fuel as determined in § 600.113-08(a) and (b); FEpet is the fuel economy while operated on... value of the alternative fuel; NHVpet is the net (lower) heating value of the petroleum fuel; Dalt...

  4. 40 CFR 600.510-08 - Calculation of average fuel economy.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... Where: Ealt = × 10 6 = energy efficiency while operating on alternative fuel rounded to the nearest 0.01... alternative fuel as determined in § 600.113-08(a) and (b); FEpet is the fuel economy while operated on... value of the alternative fuel; NHVpet is the net (lower) heating value of the petroleum fuel; Dalt...

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

    SciTech Connect

    Talamo, A.; Gohar, Y.; Bournos, V.; Fokov, Y.; Kiyavitskaya, H.; Routkovskaya, C.

    2012-07-01

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

  6. Irradiation performance of reduced-enrichment fuels tested under the US RERTR program

    SciTech Connect

    Snelgrove, J.L.; Hofman, G.L.; Copeland, G.L.

    1985-01-01

    Considerable progress in the irradiation testing of high-density, reduced-enrichment fuels has been made during the past year. Miniplates containing UA1, U/sub 3/Si/sub 2/, U/sub 3/Si/sub 1.5/, U/sub 3/Si, U/sub 3/SiCu, and, U/sub 6/Fe have been irradiated. Postirradiation examinations have revealed that breakway swelling has occurred in 6.4-Mg U/m/sup 3/ U/sub 3/Si plates at approx.2.8 x 10/sup 27/ fissions/m/sup 3/. U/sub 3/Si/sub 2/ plates are continuing to show satisfactory performance. The testing of full-sized fuel elements in the ORR and the SILOE reactor have continued with good results. Postirradiation examinations are confirming the satisfactory performance of these elements.

  7. 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.

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

    SciTech Connect

    Smith, Kevin Arthur; Primm, Trent

    2009-01-01

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

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

    SciTech Connect

    Ilas, Germina; Primm, Trent

    2011-05-01

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

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

    SciTech Connect

    Licht, J.; Bergeron, A.; Dionne, B.; Van den Branden, G; Kalcheva, S; Sikik, E; Koonen, E

    2015-01-01

    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, 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.

  11. Conversion of the University of Virginia reactor to low-enrichment fuel

    SciTech Connect

    Rydin, R.A.; Freeman, D.W.; Fehr, M.K.

    1989-01-01

    The University of Virginia is using the LEOPARD-LINX-TWODB computational package to assist in converting the 2-MW(thermal) University of Virginia Reactor (UVAR) to low-enrichment uranium fuel (LEU). Recent efforts have been focused on designing an upgraded LEU replacement core. The UVAR is unique in that anywhere from 16 to 27 fuel elements have been used over the years in a variety of geometrical arrangements. However, a compact and fixed core arrangement offers significant operational and experimental advantages over past practice, including a higher and more temporally constant thermal flux. The authors have concentrated on relative burnup comparisons between representative 4 {times} 4, 4 {times} 5, and 5 {times} 5 cores using 18 curved-plant/element HEU (HEU-18) and 18 and 22 flat-plate/element LEU (LEU-18 and LEU-22) fuel. It is concluded that the UVAR should be supplied with LEU-22 fuel. The authors have also explored the reactivity and power-peaking effects of separately varying the reflector compositions on various reactor faces from light water to heavy water to graphite. Thermal-hydraulic behavior has also been analyzed with a combination of the PARET and THERHYD codes, which indicates the LEU cores will be acceptable with only small modifications in UVAR safety system settings.

  12. 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

  13. Optimum Cycle Length and Discharge Burnup for Nuclear Fuel; Phase II: Results Achievable with Enrichments Greater than 5 w/o

    SciTech Connect

    J. Secker, et al

    2002-09-30

    The report evaluates increasing enrichment to achieve lower fuel cycle costs. Increasing enrichment 6 w/o does not reach the optimum point. Further increase is possible before the optimum will be reached.

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

    SciTech Connect

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

    2010-09-30

    Over a period of 30 years, more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited for nuclear data validation and to form the basis for criticality safety benchmarks. A number of the Argonne ZPR/ZPPR critical assemblies have been evaluated as ICSBEP and IRPhEP benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. ZPR-3 Assembly 11 (ZPR-3/11) was designed as a fast reactor physics benchmark experiment with an average core {sup 235}U enrichment of approximately 12 at.% and a depleted uranium reflector. Approximately 79.7% of the total fissions in this assembly occur above 100 keV, approximately 20.3% occur below 100 keV, and essentially none below 0.625 eV - thus the classification as a 'fast' assembly. This assembly is Fast Reactor Benchmark No. 8 in the Cross Section Evaluation Working Group

  15. 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

    ... (Sept. 2010) \\17\\ See 75 FR 62739 (Oct. 13, 2010). In response to the President's call to provide... Regulations, 46 FR 18026 (1981) (emphasis added). Alternatives calculated at the upper point and at the lower... agency's Corporate Average Fuel Economy program for passenger automobiles (referred to herein...

  16. 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

    ... 41 Public Contracts and Property Management 3 2012-01-01 2012-01-01 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...

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

    SciTech Connect

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

    2015-03-19

    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 and 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.

  18. Test Operation of Oxygen-Enriched Incinerator for Wastes From Nuclear Fuel Fabrication Facility

    SciTech Connect

    Kim, J.-G.; Yang, H.cC.; Park, G.-I.; Kim, I.-T.; Kim, J.-K.

    2002-02-26

    The oxygen-enriched combustion concept, which can minimize off-gas production, has been applied to the incineration of combustible uranium-containing wastes from a nuclear fuel fabrication facility. A simulation for oxygen combustion shows the off-gas production can be reduced by a factor of 6.7 theoretically, compared with conventional air combustion. The laboratory-scale oxygen enriched incineration (OEI) process with a thermal capacity of 350 MJ/h is composed of an oxygen feeding and control system, a combustion chamber, a quencher, a ceramic filter, an induced draft fan, a condenser, a stack, an off-gas recycle path, and a measurement and control system. Test burning with cleaning paper and office paper in this OEI process shows that the thermal capacity is about 320 MJ/h, 90 % of design value and the off-gas reduces by a factor of 3.5, compared with air combustion. The CO concentration for oxygen combustion is lower than that of air combustion, while the O2 concentration in off-gas is kept above 25 vol % for a simple incineration process without any grate. The NOx concentration in an off-gas stream does not reduce significantly due to air incoming by leakage, and the volume and weight reduction factors are not changed significantly, which suggests a need for an improvement in sealing.

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

    SciTech Connect

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

    2012-03-01

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

  20. Measurements of fuel pin/water hole worths and power peaking, void coefficients, and temperature coefficients for 4. 81 wt% enriched UO[sub 2] fuel rods

    SciTech Connect

    Harris, D.R.; Rohr, R.R.; Angelo, P.L.; Patrou, N.T.; Buckwheat, K.W.; Hayes, D.K.

    1990-01-01

    The Rensselaer Polytechnic Institute reactor critical facility is currently the only facility in North America providing critical measurement data in support of the light water reactor electric power industry. The reactor is fueled by 4.81 wt% [sup 235]U enriched UO[sub 2] high-density pellets in stainless steel clad fuel rods at the present time, although experiments with other fuels are being analyzed. The fuel pins are supported by inexpensive stainless steel lattice plates in a large open water tank. Three sets of lattice plates have been fabricated for fuel pins in square array with pitches 0.585, 0.613, and 0.640 in. (1.486, 0.613, and 1.656 cm, respectively) to provide a relevant range of water-to-fuel volume ratios. The measurements reported here are for the first of these, a relatively tight lattice of considerable interest for reactor physics methods for advanced fuels and reactors.

  1. Consideration of critically when directly disposing highly enriched spent nuclear fuel in unsaturated tuff: Bounding estimates

    SciTech Connect

    Rechard, R.P.; Tierney, M.S.; Sanchez, L.C.; Martell, M.-A.

    1996-05-01

    This report presents one of 2 approaches (bounding calculations) which were used in a 1994 study to examine the possibility of a criticality in a repository. Bounding probabilities, although rough, point to the difficulty of creating conditions under which a critical mass could be assembled (container corrosion, separation of neutron absorbers from fissile material, collapse or precipitation of fissile material) and how significant the geochemical and hydrologic phenomena are. The study could not conceive of a mechanism consistent with conditions under which an atomic explosion could occur. Should a criticality occur in or near a container in the future, boundary consequence calculations showed that fissions from one critical event (<10{sup 20} fissions, if similar to aqueous and metal accidents and experiments) are quite small compared to the amount of fissions represented by the spent fuel itself. If it is assumed that the containers necessary to hold the highly enriched spent fuel went critical once per day for 1 million years, creating an energy release of about 10{sup 20} fissions, the number of fissions equals about 10{sup 28}, which corresponds to only 1% of the fission inventory in a repository containing 70,000 metric tons of heavy metal, the expected size for the proposed repository at Yucca Mountain, Nevada.

  2. The Role of COMSOL Toward a Low-Enriched Uranium Fuel Design for the High Flux Isotope Reactor

    SciTech Connect

    Freels, James D; Arimilli, Rao V; Lowe, Kirk T; Bodey, Isaac T

    2009-01-01

    Design and safety analyses are underway to convert the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) from a high-enriched uranium (HEU) fuel to a low-enriched uranium (LEU) fuel. The primary constraint for the project is that the overall fuel plate dimensions and the current neutron flux performance must remain unchanged. This allows minimal impact on the facility and cost for the conversion, and provides transparency to the HFIR customer base and research projects that depend on the facility for isotopes and neutron flux. As a consequence, the LEU design demands more accuracy and less margin in the analysis efforts than the original design. Several technical disciplines are required to complete this conversion including nuclear reactor physics, heat transfer, fluid dynamics, structural mechanics, fuel fabrication, and engineering design. The role of COMSOL is to provide the fully-coupled 3D multi-physics analysis for heat transfer, turbulent flow, and structural mechanics of the fuel plates and flow channels. A goal is for COMSOL to simulate the entire fuel element array of fuel plates (171 inner, 369 outer). This paper describes the progress that has been made toward development of benchmark validation models of the existing HEU inner-element fuel plates.

  3. Enrichment of Microbial Electrolysis Cell Biocathodes from Sediment Microbial Fuel Cell Bioanodes

    PubMed Central

    Pisciotta, John M.; Zaybak, Zehra; Call, Douglas F.; Nam, Joo-Youn

    2012-01-01

    Electron-accepting (electrotrophic) biocathodes were produced by first enriching graphite fiber brush electrodes as the anodes in sediment-type microbial fuel cells (sMFCs) using two different marine sediments and then electrically inverting the anodes to function as cathodes in two-chamber bioelectrochemical systems (BESs). Electron consumption occurred at set potentials of −439 mV and −539 mV (versus the potential of a standard hydrogen electrode) but not at −339 mV in minimal media lacking organic sources of energy. Results at these different potentials were consistent with separate linear sweep voltammetry (LSV) scans that indicated enhanced activity (current consumption) below only ca. −400 mV. MFC bioanodes not originally acclimated at a set potential produced electron-accepting (electrotrophic) biocathodes, but bioanodes operated at a set potential (+11 mV) did not. CO2 was removed from cathode headspace, indicating that the electrotrophic biocathodes were autotrophic. Hydrogen gas generation, followed by loss of hydrogen gas and methane production in one sample, suggested hydrogenotrophic methanogenesis. There was abundant microbial growth in the biocathode chamber, as evidenced by an increase in turbidity and the presence of microorganisms on the cathode surface. Clone library analysis of 16S rRNA genes indicated prominent sequences most similar to those of Eubacterium limosum (Butyribacterium methylotrophicum), Desulfovibrio sp. A2, Rhodococcus opacus, and Gemmata obscuriglobus. Transfer of the suspension to sterile cathodes made of graphite plates, carbon rods, or carbon brushes in new BESs resulted in enhanced current after 4 days, demonstrating growth by these microbial communities on a variety of cathode substrates. This report provides a simple and effective method for enriching autotrophic electrotrophs by the use of sMFCs without the need for set potentials, followed by the use of potentials more negative than −400 mV. PMID:22610438

  4. Enrichment of Microbial Electrolysis Cell Biocathodes from Sediment Microbial Fuel Cell Bioanodes

    SciTech Connect

    Pisciotta, JM; Zaybak, Z; Call, DF; Nam, JY; Logan, BE

    2012-07-18

    Electron-accepting (electrotrophic) biocathodes were produced by first enriching graphite fiber brush electrodes as the anodes in sediment-type microbial fuel cells (sMFCs) using two different marine sediments and then electrically inverting the anodes to function as cathodes in two-chamber bioelectrochemical systems (BESs). Electron consumption occurred at set potentials of -439 mV and -539 mV (versus the potential of a standard hydrogen electrode) but not at -339 mV in minimal media lacking organic sources of energy. Results at these different potentials were consistent with separate linear sweep voltammetry (LSV) scans that indicated enhanced activity (current consumption) below only ca. -400 mV. MFC bioanodes not originally acclimated at a set potential produced electron-accepting (electrotrophic) biocathodes, but bioanodes operated at a set potential (+11 mV) did not. CO, was removed from cathode headspace, indicating that the electrotrophic biocathodes were autotrophic. Hydrogen gas generation, followed by loss of hydrogen gas and methane production in one sample, suggested hydrogenotrophic methanogenesis. There was abundant microbial growth in the biocathode chamber, as evidenced by an increase in turbidity and the presence of microorganisms on the cathode surface. Clone library analysis of 16S rRNA genes indicated prominent sequences most similar to those of Eubacterium limosum (Butyribacterium methylotrophicum), Desulfovibrio sp. A2, Rhodococcus opacus, and Gemmata obscuriglobus. Transfer of the suspension to sterile cathodes made of graphite plates, carbon rods, or carbon brushes in new BESs resulted in enhanced current after 4 days, demonstrating growth by these microbial communities on a variety of cathode substrates. This report provides a simple and effective method for enriching autotrophic electrotrophs by the use of sMFCs without the need for set potentials, followed by the use of potentials more negative than -400 mV.

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

    SciTech Connect

    Travelli, A.

    1983-01-01

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

  6. Systems report on the analysis of spent, highly enriched U-235 reactor fuel by delayed neutron interrogation

    SciTech Connect

    Piper, T.C.; Kirkham, R.J.

    1990-05-01

    Design aspects are briefly given of a neutron source shuffler used to measure fissile material content of spent, highly enriched reactor fuel. The mode of operation used, results of analyzing 176 fuel packages and recommended system improvements are discussed. Four measurements were made on each of the fuel packages with the mean of the 176 standard deviations being 2.03 percent of value. The maximum individual standard deviation was 9.27 percent. Appendixes concerning imprecisions introduced by counting statistics and crane speed irregularities are given. Use of an improved neutron source shuffler, an improved fuel package motion system and modernized computer system should permit system performance to be limited mainly by counting statistics, to about 1.5 percent of measured value. A stronger source could then be installed to further enhance system operation. 16 figs., 3 tabs.

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

    SciTech Connect

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

    1991-11-01

    The performance of a single-cylinder, direct-injection diesel engine was measured with intake oxygen levels of up to 35% and fuel water contents of up to 20%. Because a previous study indicated that the use of a less-expensive fuel would be more economical, two series of tests with No. 4 diesel fuel and No. 2 diesel fuel were conducted. To control the emissions of nitrogen oxides (NO{sub x}), water was introduced into the combustion process in the form of water-emulsified fuel, or the fuel injection timing was retarded. In the first series of tests, compressed oxygen was used; in the second series of tests, a hollow-tube membrane was used. Steady-state engine performance and emissions data were obtained. Test results indicated a large increase in engine power density, a slight improvement in thermal efficiency, and significant reductions in smoke and particulate-matter emissions. Although NO{sub x} emissions increased, they could be controlled by introducing water and retarding the injection timing. The results further indicated that thermal efficiency is slightly increased when moderately water-emulsified fuels are used, because a greater portion of the fuel energy is released earlier in the combustion process. Oxygen-enriched air reduced the ignition delay and caused the heat-release rate and cumulative heat-release rates to change measurably. Even at higher oxygen levels, NO{sub x} emissions decreased rapidly when the timing was retarded, and the amount of smoke and the level of particulate-matter emissions did not significantly increase. The single-cylinder engine tests confirmed the results of an earlier technical assessment and further indicated a need for a low-pressure-drop membrane specifically designed for oxygen enrichment. Extension data set indexed separately. 14 refs.

  8. Electrochemical characterization of anodic biofilms enriched with glucose and acetate in single-chamber microbial fuel cells.

    PubMed

    Yuan, Yong; Zhou, Shungui; Xu, Nan; Zhuang, Li

    2011-02-01

    This study used a simple and efficient electrochemical technique, cyclic voltammogram (CV), to quantitatively measure the electron transfer capability of anodic biofilms enriched with acetate and glucose in single-chamber microbial fuel cells (MFCs). Two pairs of distinct redox peaks were observed by CV measurements in both biofilms, identical to the CV features of a pure Geobacter strain. The CVs also revealed a higher density of electroactive species in the acetate-enriched biofilm than that in the glucose-enriched biofilm. Based on the scan rate analysis, the apparent electron transfer rate constants (k(app)) in the acetate-enriched biofilm and glucose-enriched biofilm were determined to be 0.82 and 0.15s(-1), respectively, which supported the higher power output of the MFC fed with acetate. Meanwhile, the pH dependence of the biofilms was studied by monitoring the changes of the biofilm redox peak currents and potentials. It is concluded that redox reaction of the electrochemical active species in biofilms is pH dependent, and both electrons and protons are involved in the redox reactions. PMID:21050727

  9. Conceptual Process for the Manufacture of Low-Enriched Uranium/Molybdenum Fuel for the High Flux Isotope Reactor

    SciTech Connect

    Sease, J.D.; Primm, R.T. III; Miller, J.H.

    2007-09-30

    The U.S. nonproliferation policy 'to minimize, and to the extent possible, eliminate the use of HEU in civil nuclear programs throughout the world' has resulted in the conversion (or scheduled conversion) of many of the U.S. research reactors from high-enriched uranium (HEU) to low-enriched uranium (LEU). A foil fuel appears to offer the best option for using a LEU fuel in the High Flux Isotope Reactor (HFIR) without degrading the performance of the reactor. The purpose of this document is to outline a proposed conceptual fabrication process flow sheet for a new, foil-type, 19.75%-enriched fuel for HFIR. The preparation of the flow sheet allows a better understanding of the costs of infrastructure modifications, operating costs, and implementation schedule issues associated with the fabrication of LEU fuel for HFIR. Preparation of a reference flow sheet is one of the first planning steps needed in the development of a new manufacturing capacity for low enriched fuels for U.S. research and test reactors. The flow sheet can be used to develop a work breakdown structure (WBS), a critical path schedule, and identify development needs. The reference flow sheet presented in this report is specifically for production of LEU foil fuel for the HFIR. The need for an overall reference flow sheet for production of fuel for all High Performance Research Reactors (HPRR) has been identified by the national program office. This report could provide a starting point for the development of such a reference flow sheet for a foil-based fuel for all HPRRs. The reference flow sheet presented is based on processes currently being developed by the national program for the LEU foil fuel when available, processes used historically in the manufacture of other nuclear fuels and materials, and processes used in other manufacturing industries producing a product configuration similar to the form required in manufacturing a foil fuel. The processes in the reference flow sheet are within the

  10. Daily average exposures to respirable particulate matter from combustion of biomass fuels in rural households of southern India.

    PubMed

    Balakrishnan, Kalpana; Sankar, Sambandam; Parikh, Jyothi; Padmavathi, Ramaswamy; Srividya, Kailasam; Venugopal, Vidhya; Prasad, Swarna; Pandey, Vijay Laxmi

    2002-11-01

    Indoor air pollution resulting from combustion of biomass fuels in rural households of developing countries is now recognized as a major contributor to the global burden of disease. Accurate estimation of health risks has been hampered by a paucity of quantitative exposure information. In this study we quantified exposures to respirable particulate matter from biomass-fuel combustion in 436 rural homes selected through stratified random sampling from four districts of Tamil Nadu, India. The study households are a subset of a larger sample of 5,028 households from the same districts in which socioeconomic and health information has been collected. Results of measurements for personal exposures to respirable particulate matter during cooking were reported earlier. This has been extended to calculation of 24-hr exposures with the aid of additional measurements during noncooking times and the collection of time-activity records. Concentrations of respirable particulate matter ranged from 500 to 2,000 micro g/m(3) during cooking in biomass-using households, and average 24-hr exposures ranged from 90 +/- 21 micro g/m(3) for those not involved in cooking to 231 +/- 109 micro g/m(3) for those who cooked. The 24-hr exposures were around 82 +/- 39 micro g/m(3) for those in households using clean fuels (with similar exposures across household subgroups). Fuel type, type and location of the kitchen, and the time spent near the kitchen while cooking were the most important determinants of exposure across these households among other parameters examined, including stove type, cooking duration, and smoke from neighborhood cooking. These estimates could be used to build a regional exposure database and facilitate health risk assessments. PMID:12417476

  11. Nonproliferation and safeguards aspects of fuel cycle programs in reduction of excess separated plutonium and high-enriched uranium

    SciTech Connect

    Persiani, P.J.

    1995-06-01

    The purpose of this preliminary investigation is to explore alternatives and strategies aimed at the gradual reduction of the excess inventories of separated plutonium and high-enriched uranium (HEU) in the civilian nuclear power industry. The study attempts to establish a technical and economic basis to assist in the formation of alternative approaches consistent with nonproliferation and safeguards concerns. Reference annual mass flows and inventories for a representative 1,400 Mwe Pressurized Water Reactor (PWR) fuel cycle have been investigated for three cases: the 100 percent uranium oxide UO{sub 2} fuel loading once through cycle, and the 33 percent mixed oxide MOX loading configuration for a first and second plutonium recycle. The analysis addresses fuel cycle developments; plutonium and uranium inventory and flow balances; nuclear fuel processing operations; UO{sub 2} once-through and MOX first and second recycles; and the economic incentives to draw-down the excess separated plutonium stores. The preliminary analysis explores several options in reducing the excess separated plutonium arisings and HEU, and the consequences of the interacting synergistic effects between fuel cycle processes and isotopic signatures of nuclear materials on nonproliferation and safeguards policy assessments.

  12. 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. PMID:26867100

  13. Status of highly loaded uranium-zirconium hydride Low Enriched Uranium (LEU) fuel programs

    SciTech Connect

    West, G.B.

    1984-07-01

    The LEU uranium-zirconium hydride fuel has successfully completed all essential development testing and is offered commercially in both the 20 weight percent and 45 weight percent uranium concentrations for existing TRIGA reactors and as a conversion fuel for plate-type reactors. Several reactor facilities have already begun conversion to LEU uranium-zirconium hydride fuel. Currently, U-ZrH LEU fuel is in use in a mixed plate and rod fuel configuration in the 1.0 MW reactor at the National Tsing Hua University in Taipei, Taiwan. A complete core of 20 wt-$ LEU fuel has been fabricated for the 3 MW forced flow TRIGA Mark II reactor system with pulsing capability scheduled to be completed in 1984 in Bangladesh. Also, a complete core of 20 vt-% 4-rod shrouded cluster fuel has been fabricated and shipped for conversion and upgrading of the PRR-1 reactor in the Philippines. The converted reactor will operate at 3 MW with forced cooling and have pulsing capability. Startup is now scheduled for early 1985. U-ZrH LEU fuel has also been delivered to existing TRIGA reactor facilities in Malaysia, Thailand and Yugoslavia and is being put to use as new fuel is needed to meet burnup requirements. For stepwise conversions of higher power reactors (>5 MW), where flow distribution and neutron spectrum effects are of greater importance, a detailed analysis and study is a prerequisite. A joint program involving GA, ANL, and the Cekmece Center is now in progress for analysis of a single U-ZrH LEU fuel test cluster operating in the 5 MW HEU plate-type core of the TR-2 reactor at the Cekmece Nuclear Research and Training Center in Istanbul, Turkey. Inserting this one cluster would be the initial step in a stepwise conversion of the HEU plate-type core to U-ZrH LEU fuel.

  14. 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 PAGESBeta

    Kane, J. J.; van Rooyen, I. J.; Craft, A. E.; Roney, T. J.; Morrell, S. R.

    2016-02-05

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

  15. The Melt-Dilute Treatment of Al-Base Highly Enriched DOE Spent Nuclear Fuels: Principles and Practices

    SciTech Connect

    Adams, T.M.

    1998-11-25

    The melt-dilute treatment technology program is focused on the development and implementation of a treatment technology for diluting highly enriched (>20 percent 235U) aluminum spent nuclear fuel to low enriched levels (<20 percent 235U) and qualifying the LEU Al-SNF form for geologic repository storage. In order to reduce the enrichment of these assemblies prior to ultimate geologic repository disposal, the melt-dilute technology proposes to melt these SNF assemblies and then dilute with additions of depleted uranium. The benefits accrued from this treatment process include the potential for significant volume reduction, reduced criticality potential, and the potential for enhanced SNF form characteristics. The emphasis within the development program to date has been on determining the process metallurgy and off-gas system design for the treatment of all types of Al SNF (UAlx, Al-U3O8, and Al-U3Si2). In determining the process metallurgy a wide range of alloys, representative of those expected in the Al-SNF form, have been fabricated and their product characteristics, namely microstructure, homogeneity, phase composition, and "ternary" constituent effects have been analyzed. As a result of the presence of species within the melt which will possess significant vapor pressures in the desired operating temperature range an off-gas system is necessary. Of the volitile species the one of greatest concern is 137Cs.

  16. Physical and welding metallurgy of Gd-enriched austenitic alloys for spent nuclear fuel applications. Part II, nickel base alloys.

    SciTech Connect

    Mizia, Ronald E.; Michael, Joseph Richard; Williams, David Brian; Dupont, John Neuman; Robino, Charles Victor

    2004-06-01

    The physical and welding a metallurgy of gadolinium- (Gd-) enriched Ni-based alloys has been examined using a combination of differential thermal analysis, hot ductility testing. Varestraint testing, and various microstructural characterization techniques. Three different matrix compositions were chosen that were similar to commercial Ni-Cr-Mo base alloys (UNS N06455, N06022, and N06059). A ternary Ni-Cr-Gd alloy was also examined. The Gd level of each alloy was {approx}2 wt-%. All the alloys initiated solidification by formation of primary austenite and terminated solidification by a Liquid {gamma} + Ni{sub 5}Gd eutectic-type reaction at {approx}1270 C. The solidification temperature ranges of the alloys varied from {approx}100 to 130 C (depending on alloy composition). This is a substantial reduction compared to the solidification temperature range to Gd-enriched stainless steels (360 to 400 C) that terminate solidification by a peritectic reaction at {approx}1060 C. The higher-temperature eutectic reaction that occurs in the Ni-based alloys is accompanied by significant improvements in hot ductility and solidification cracking resistance. The results of this research demonstrate that Gd-enriched Ni-based alloys are excellent candidate materials for nuclear criticality control in spent nuclear fuel storage applications that require production and fabrication of large amounts of material through conventional ingot metallurgy and fusion welding techniques.

  17. Improving Power Production in Acetate-Fed Microbial Fuel Cells via Enrichment of Exoelectrogenic Organisms in Flow-Through Systems

    SciTech Connect

    Borole, Abhijeet P; Hamilton, Choo Yieng; Vishnivetskaya, Tatiana A; Leak, David; Andras, Calin

    2009-01-01

    An exoelectrogenic, biofilm-forming microbial consortium was enriched in an acetate-fed microbial fuel cell (MFC) using a flow-through anode coupled to an air-cathode. Multiple parameters known to improve MFC performance were integrated in one design including electrode spacing, specific electrode surface area, flow-through design, minimization of dead volume within anode chamber, and control of external resistance. In addition, continuous feeding of carbon source was employed and the MFC was operated at intermittent high flows to enable removal of non-biofilm forming organisms over a period of six months. The consortium enriched using the modified design and operating conditions resulted in a power density of 345 W m-3 of net anode volume (3650 mW m-2), when coupled to a ferricyanide cathode. The enriched consortium included -, -, -Proteobacteria, Bacteroidetes and Firmicutes. Members of the order Rhodocyclaceae and Burkholderiaceae (Azospira spp. (49%), Acidovorax spp. (11%) and Comamonas spp. (7%)), dominated the microbial consortium. Denaturing gradient gel electrophoresis (DGGE) analysis based on primers selective for Archaea suggested a very low abundance of methanogens. Limiting the delivery of the carbon source via continuous feeding corresponding to the maximum cathodic oxidation rates permitted in the flow-through, air-cathode MFC resulted in coulombic efficiencies reaching 88 5.7%.

  18. Selective enrichment of electrogenic bacteria for fuel cell application: Enumerating microbial dynamics using MiSeq platform.

    PubMed

    Vamshi Krishna, K; Venkata Mohan, S

    2016-08-01

    This study is intended to examine the effect of pretreatment on selective enrichment of electrogenic bacteria from mixed culture. It has been observed that the iodopropane and heat-shock pretreatments suppress the growth of non-exoelectrons, while selecting only a limited number of strains belonging to genera Xanthomonas, Pseudomonas and Prevotella while untreated control inoculum showed more diverse community comprising of both exoelectrogens and non-exoelectrogens. High power output was observed in iodopropane (180mW/m(2)) pretreated microbial fuel cell (MFC) compared to heat-shock pretreated MFC (128mW/m(2)) and untreated control (92mW/m(2)). Coulombic efficiency of iodopropane and heat-shock pretreated MFC was higher compared to untreated control MFC, while drop in pH and volatile fatty acids (VFA) production was less in iodopropane pretreated MFC signifying the shifts in bacterial community structure toward electrogenesis instead of fermentation. These results signify the role of iodopropane and heat pretreatments on enrichment of electrogenic bacteria for fuel cell application. PMID:27061058

  19. 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

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

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-10-24

    ...The U.S. Nuclear Regulatory Commission (NRC) staff has conducted inspections of the Louisiana Energy Services (LES), LLC, National Enrichment Facility in Eunice, New Mexico, and has authorized the introduction of uranium hexafluoride (UF6) into cascades numbered 3.10, 3.11, 3.12, 4.1, 4.2, 4.3 and 4.5. In addition, the NRC verified that the systems, structures, and components......

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

    SciTech Connect

    G. Youinou; S. Bays

    2009-05-01

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

  3. Delayed fission product gamma-ray transmission through low enriched uranium dioxide fuel pin lattices in air

    NASA Astrophysics Data System (ADS)

    Trumbull, Timothy H.

    The transmission of delayed fission-product gamma rays through various arrangements of low-enriched UO2 fuel pin lattices in an air medium was studied. Experimental measurements, point-kernel and Monte Carlo photon transport calculations were performed to demonstrate the shielding effect of ordered lattices of fuel pins on the resulting gamma-ray dose to a detector outside the lattice. The variation of the gamma-ray dose on the outside of the lattice as a function of radial position, the so-called "channeling" effect, was analyzed. Techniques for performing experimental measurements and data reduction at Rensselaer Polytechnic Institute's Reactor Critical Facility (RCF) were derived. An experimental apparatus was constructed to hold the arrangements of fuel pins for the measurements. A gamma-ray spectroscopy system consisting of a sodium-iodide scintillation detector was used to collect data. Measurements were made with and without a collimator installed. A point-kernel transport code was developed to map the radial dependence of the gamma-ray flux. Input files for the Monte Carlo code, MCNP, were also developed to accurately model the experimental measurements. The results of the calculations were compared to the experimental measurements. In order to determine the delayed fission-product gamma-ray source for the calculations, a technique was developed using a previously written code, DELBG and the reactor state-point data obtained during the experimental measurements. Calculations were performed demonstrating the effects of material homogenization on the gamma-ray transmission through the fuel pin lattice. Homogeneous and heterogeneous calculations were performed for all RCF fuel pin lattices as well as for a typical commercial pressurized water reactor fuel bundle. The results of the study demonstrated the effectiveness of the experimental measurements to isolate the channeling effect of delayed fission-product gamma-rays through lattices of RCF fuel pins in an

  4. Initial assessment of radiation behavior of very-high-density low-enriched-uranium fuels.

    SciTech Connect

    Hofman, G. L.; Meyer, M. L.; Snelgrove, J. L.; Dietz, M. L.; Strain, R. V.; Kim, K. H.

    1999-10-01

    Results from the postirradiation examinations of microplates irradiated in the RERTR-1 and -2 experiments in the ATR have shown several binary and ternary U-Mo alloys to be promising candidates for use in aluminum-based dispersion fuels with uranium densities up to 8 to 9 g/cm{sup 3}. Ternary alloys of uranium, niobium, and zirconium performed poorly, however, both in terms of fuel/matrix reaction and fission-gas-bubble behavior, and have been dropped from further study. Since irradiation temperatures achieved in the present experiments (approximately 70 C)are considerably lower than might be experienced in a high-performance reactor, a new experiment is being planned with beginning-of-cycle temperatures greater than 200 C in 8-g U/cm{sup 3} fuel.

  5. Reactivity and isotopic composition of spent PWR (pressurized-water-reactor) fuel as a function of initial enrichment, burnup, and cooling time

    SciTech Connect

    Cerne, S.P.; Hermann, O.W.; Westfall, R.M.

    1987-10-01

    This study presents the reactivity loss of spent PWR fuel due to burnup in terms of the infinite lattice multiplications factor, k/sub infinity/. Calculations were performed using the SAS2 and CSAS1 control modules of the SCALE system. The k/sub infinity/ values calculated for all combinations of six enrichments, seven burnups, and five cooling times. The results are presented as a primary function of enrichment in both tabular and graphic form. An equation has been developed to estimate the tabulated values of k/sub infinity/'s by specifying enrichment, cooling time, and burnup. Atom densities for fresh fuel, and spent fuel at cooling times of 2, 10, and 20 years are included. 13 refs., 8 figs., 8 tabs.

  6. Direct Measurement of Initial Enrichment, Burn-up and Cooling Time of Spent Fuel Assembly with a Differential Die-Away Technique Based Instrument

    SciTech Connect

    Henzl, Vladimir; Swinhoe, Martyn T.; Tobin, Stephen J.

    2012-07-13

    An outline of this presentation of what a Differential Die-Away (DDA) instrument can do are: (1) Principle of operation of DDA instrument; (2) Determination of initial enrichment (IE) ({sigma} < 5%); (3) Determination of burn up (BU) ({sigma} {approx} 6%); (4) Determination of cooling time (CT) ({sigma} {approx} 20-50%); and (5) DDA instrument as a standalone device. DDA response (fresh fuel vs. spent fuel) is: (1) Fresh fuel => DDA response increases (die-away time is longer) with increasing fissile content; and (2) Spent fuel => DDA response decreases (die-away time is shorter) with higher burn-up (i.e. more neutron absorbers present).

  7. 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.

  8. Development of Enriched Borated Aluminum Alloy for Basket Material of Cask for Spent Nuclear Fuel

    SciTech Connect

    Mikio Sakai; Tadatsugu Sakaya; Hiroaki Fujiwara; Akira Sakai

    2002-07-01

    Concrete cask system is focused as the candidate one for spent fuel dry storage facilities from economic potential in Japan. Concrete cask consists of a concrete storage cask and a steel canister. A canister containing nuclear spent fuel is shipped by a transportation cask from a nuclear power plant to an interim storage facility. The canister is transferred from the transportation cask to a storage cask by a transfer cask in the storage facility. IHI has developed a concrete cask horizontal transfer system. This transfer system indicates that a canister is transferred to a storage cask horizontally. This transfer system has a merit against canister drop accident in transfer operation, i.e. spent fuel assemblies can be kept safe during the transfer operation. There are guide rails inside of the concrete cask, and the canister is installed into the storage cask with sliding on the rails. To develop the horizontal transfer system, IHI carried out a heat load test and numerical analyses by CFD. Heat load experiment was carried out by using a full-scale prototype canister, storage cask and transfer vessel. The decay heat was simulated by an electric heater installed in the canister. Assuming high burn-up spent fuel storage, heat generation was set between 20.0 kW and 25.0 kW. This experiment was focused on the concrete temperature distribution. We confirmed that the maximum concrete temperature in transfer operation period was lower than 40 deg. C (Heat generation 22.5 kW). Moreover we confirmed the maximum concrete temperature passed 24 hours with horizontal orientation was below 90 deg. C (Heat generation 22.5 kW). We analyzed the thermal performance of the concrete cask with horizontal transfer condition and normal storage condition. Thermal analyses for horizontal transfer operation were carried out based on the experimental conditions. The tendency of the analytical results was in good agreement with experimental results. The purpose of vertical thermal analysis

  9. A new fast neutron collar for safeguards inspection measurements of fresh low enriched uranium fuel assemblies containing burnable poison rods

    NASA Astrophysics Data System (ADS)

    Evans, Louise G.; Swinhoe, Martyn T.; Menlove, Howard O.; Schwalbach, Peter; Baere, Paul De; Browne, Michael C.

    2013-11-01

    Safeguards inspection measurements must be performed in a timely manner in order to detect the diversion of significant quantities of nuclear material. A shorter measurement time can increase the number of items that a nuclear safeguards inspector can reliably measure during a period of access to a nuclear facility. In turn, this improves the reliability of the acquired statistical sample, which is used to inform decisions regarding compliance. Safeguards inspection measurements should also maintain independence from facility operator declarations. Existing neutron collars employ thermal neutron interrogation for safeguards inspection measurements of fresh fuel assemblies. A new fast neutron collar has been developed for safeguards inspection measurements of fresh low-enriched uranium (LEU) fuel assemblies containing gadolinia (Gd2O3) burnable poison rods. The Euratom Fast Collar (EFC) was designed with high neutron detection efficiency to make a fast (Cd) mode measurement viable whilst meeting the high counting precision and short assay time requirements of the Euratom safeguards inspectorate. A fast mode measurement reduces the instrument sensitivity to burnable poison rod content and therefore reduces the applied poison correction, consequently reducing the dependence on the operator declaration of the poison content within an assembly. The EFC non-destructive assay (NDA) of typical modern European pressurized water reactor (PWR) fresh fuel assembly designs have been simulated using Monte Carlo N-particle extended transport code (MCNPX) simulations. Simulations predict that the EFC can achieve 2% relative statistical uncertainty on the doubles neutron counting rate for a fast mode measurement in an assay time of 600 s (10 min) with the available 241AmLi (α,n) interrogation source strength of 5.7×104 s-1. Furthermore, the calibration range of the new collar has been extended to verify 235U content in variable PWR fuel designs in the presence of up to 32

  10. Application of ENDF data to the AVR reactor with highly enriched uranium fuel and thorium feed

    SciTech Connect

    Vondy, D.R.

    1986-09-01

    Calculations were done applying ENDF data to the German AVR pebble bed reactor at KFA. Several core models were used, and the results obtained with ORNL methods for the multiplication, reaction rates, temperature coefficient of reactivity, and fuel temperature distributions are reported and compared. Only a small difference in multiplication is found for this core in going from ENDF/B-IV to ENDF/B-V cross-section data. The temperature coefficients calculated with the ENDF/B-V are somewhat smaller in magnitude. The worth of control rods was obtained and only a small difference was found with the data, but the calculated results are high compared with experiment. Neutron reaction rates with the key actinides are reported for three-dimensional core calculations.

  11. Safety evaluation report related to the evaluation of low-enriched uranium silicide-aluminum dispersion fuel for use in non-power reactors

    SciTech Connect

    Not Available

    1988-07-01

    Low-enriched uranium silicide-aluminum dispersion plate-type fuels have been extensively researched and developed under the international program, Reduced Enrichment in Research and Test Reactors. The international effort was led by Argonne National Laboratory (ANL) in the United States. This evaluation is based primarily on reports issued by ANL that discuss and summarize the developmental tests and experiments, including postirradiation examinations, of both miniature and full-sized plates of prototypical fuel compositions. This evaluation concludes that plate-type fuels suitable and acceptable for use in research and test reactors can be fabricated with U/sub 3/Si/sub 2/-Al dispersion compacts with uranium densities up to 4.8 g/cm/sup 3/. 4 refs., 1 fig.

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

    SciTech Connect

    Michael Tyacke; Dr. Igor Bolshinsky; Jeff Chamberlin

    2008-07-01

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

  13. 77 FR 62623 - 2017 and Later Model Year Light-Duty Vehicle Greenhouse Gas Emissions and Corporate Average Fuel...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-10-15

    ... 2. Off-Cycle CO 2 Credits 3. Advanced Technology Incentives for Full-Size Pickup Trucks G. Safety... Ethanol Flexible Fuel Vehicles for GHG Emissions Compliance 5. Off-cycle Technology Credits D. Technical... that the CAFE of the industry-wide combined fleet of new passenger cars and light trucks reaches...

  14. 75 FR 80430 - Passenger Car and Light Truck Average Fuel Economy Standards Request for Product Plan Information...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-12-22

    ..., plug-in hybrid, electric, and fuel cell vehicles, among others). NHTSA will also consider information...), HEVT = hybrid/electric vehicle transmission (for a BISG or CISG type hybrid, please define the actual... AB 1493 Regulation. E. Hybridization/Electrification 1. Type of Hybrid/Electric...

  15. 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

    ... Property Management Federal Property Management Regulations System (Continued) FEDERAL MANAGEMENT REGULATION PERSONAL PROPERTY 34-MOTOR VEHICLE MANAGEMENT Obtaining Fuel Efficient Motor Vehicles § 102-34.60... 41 Public Contracts and Property Management 3 2013-07-01 2013-07-01 false How do we calculate...

  16. 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

    ... Property Management Federal Property Management Regulations System (Continued) FEDERAL MANAGEMENT REGULATION PERSONAL PROPERTY 34-MOTOR VEHICLE MANAGEMENT Obtaining Fuel Efficient Motor Vehicles § 102-34.60... 41 Public Contracts and Property Management 3 2010-07-01 2010-07-01 false How do we calculate...

  17. 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

    ... Property Management Federal Property Management Regulations System (Continued) FEDERAL MANAGEMENT REGULATION PERSONAL PROPERTY 34-MOTOR VEHICLE MANAGEMENT Obtaining Fuel Efficient Motor Vehicles § 102-34.60... 41 Public Contracts and Property Management 3 2011-01-01 2011-01-01 false How do we calculate...

  18. Cross section generation and physics modeling in a feasibility study of the conversion of the high flux isotope reactor core to use low-enriched uranium fuel

    SciTech Connect

    Ellis, R. J.; Gehin, J. C.; Primm Iii, R. T.

    2006-07-01

    A computational study has been initiated at ORNL to examine the feasibility of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. The current study is limited to steady-state, nominal operation and are focused on the determination of the fuel requirements, primarily density, that are required to maintain the performance of the reactor. Reactor physics analyses are reported for a uranium-molybdenum alloy that would be substituted for the current fuel - U{sub 3}O{sub 8} mixed with aluminum. An LEU core design has been obtained and requires an increase in {sup 235}U loading of a factor of 1.9 over the current HEU fuel. These initial results indicate that the conversion from HEU to LEU results in a reduction of the thermal fluxes in the central flux trap region of approximately 9 % and in the outer beryllium reflector region of approximately 15%. Ongoing work is being performed to improve upon this initial design to further minimize the impact of conversion to LEU fuel. (authors)

  19. Partitioning of metal species during an enriched fuel combustion experiment. speciation in the gaseous and particulate phases.

    PubMed

    Pavageau, Marie-Pierre; Morin, Anne; Seby, Fabienne; Guimon, Claude; Krupp, Eva; Pécheyran, Christophe; Poulleau, Jean; Donard, Olivier F X

    2004-04-01

    Combustion processes are the most important source of metal in the atmosphere and need to be better understood to improve flue gas treatment and health impact studies. This combustion experiment was designed to study metal partitioning and metal speciation in the gaseous and particulate phases. A light fuel oil was enriched with 15 organometallic compounds of the following elements: Pb, Hg, As, Cu, Zn, Cd, Se, Sn, Mn, V, Tl, Ni, Co, Cr, and Sb. The resulting mixture was burnt in a pilot-scale fuel combustion boiler under controlled conditions. After filtration of the particles, the gaseous species were sampled in the stack through a heated sampling tube simultaneously by standardized washing bottles-based sampling techniques and cryogenically. The cryogenic samples were collected at -80 degrees C for further speciation analysis by LT/GC-ICPMS. Three species of selenium and two of mercury were evidenced as volatile species in the flue gas. Thermodynamic predictions and experiments suggest the following volatile metal species to be present in the flue gas: H2Se, CSSe, CSe2, SeCl2, Hg(0), and HgCl2. Quantification of volatile metal species in comparison between cryogenic techniques and the washing bottles-based sampling method is also discussed. Concerning metal partitioning, the results indicated that under these conditions, at least 60% (by weight) of the elements Pb, Sn, Cu, Co, Tl, Mn, V, Cr, Ni, Zn, Cd, and Sb mixed to the fuel were found in the particulate matter. For As and Se, 37 and 17%, respectively, were detected in the particles, and no particulate mercury was found. Direct metal speciation in particles was performed by XPS allowing the determination of the oxidation state of the following elements: Sb(V), Tl(III), Mn(IV), Cd(II), Zn(II), Cr(III), Ni(II), Co(II), V(V), and Cu(II). Water soluble species of inorganic Cr, As, and Se in particulate matter were determined by HPLC/ICP-MS and identified in the oxidation state Cr(III), As(V), and Se(IV). PMID

  20. Experiment Safety Assurance Package for Mixed Oxide Fuel Irradiation in an Average Power Position (I-24) in the Advanced Test Reactor

    SciTech Connect

    J. M . Ryskamp; R. C. Howard; R. C. Pedersen; S. T. Khericha

    1998-10-01

    The Fissile Material Disposition Program Light Water Reactor Mixed Oxide Fuel Irradiation Test Project Plan details a series of test irradiations designed to investigate the use of weapons-grade plutonium in MOX fuel for light water reactors (LWR) (Cowell 1996a, Cowell 1997a, Thoms 1997a). Commercial MOX fuel has been successfully used in overseas reactors for many years; however, weapons-derived test fuel contains small amounts of gallium (about 2 parts per million). A concern exists that the gallium may migrate out of the fuel and into the clad, inducing embrittlement. For preliminary out-of-pile experiments, Wilson (1997) states that intermetallic compound formation is the principal interaction mechanism between zircaloy cladding and gallium. This interaction is very limited by the low mass of gallium, so problems are not expected with the zircaloy cladding, but an in-pile experiment is needed to confirm the out-of-pile experiments. Ryskamp (1998) provides an overview of this experiment and its documentation. The purpose of this Experiment Safety Assurance Package (ESAP) is to demonstrate the safe irradiation and handling of the mixed uranium and plutonium oxide (MOX) Fuel Average Power Test (APT) experiment as required by Advanced Test Reactor (ATR) Technical Safety Requirement (TSR) 3.9.1 (LMITCO 1998). This ESAP addresses the specific operation of the MOX Fuel APT experiment with respect to the operating envelope for irradiation established by the Upgraded Final Safety Analysis Report (UFSAR) Lockheed Martin Idaho Technologies Company (LMITCO 1997a). Experiment handling activities are discussed herein.

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

    SciTech Connect

    Primm, R.T., III; Ellis, R.J.; Gehin, J.C.; Moses, D.L.; Binder, J.L.; Xoubi, N.

    2006-02-01

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

  2. Experimental evaluation of oxygen-enriched air and emulsified fuels in a single-cylinder diesel engine. Volume 1, Concept evaluation

    SciTech Connect

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

    1991-11-01

    The performance of a single-cylinder, direct-injection diesel engine was measured with intake oxygen levels of up to 35% and fuel water contents of up to 20%. Because a previous study indicated that the use of a less-expensive fuel would be more economical, two series of tests with No. 4 diesel fuel and No. 2 diesel fuel were conducted. To control the emissions of nitrogen oxides (NO{sub x}), water was introduced into the combustion process in the form of water-emulsified fuel, or the fuel injection timing was retarded. In the first series of tests, compressed oxygen was used; in the second series of tests, a hollow-tube membrane was used. Steady-state engine performance and emissions data were obtained. Test results indicated a large increase in engine power density, a slight improvement in thermal efficiency, and significant reductions in smoke and particulate-matter emissions. Although NO{sub x} emissions increased, they could be controlled by introducing water and retarding the injection timing. The results further indicated that thermal efficiency is slightly increased when moderately water-emulsified fuels are used, because a greater portion of the fuel energy is released earlier in the combustion process. Oxygen-enriched air reduced the ignition delay and caused the heat-release rate and cumulative heat-release rates to change measurably. Even at higher oxygen levels, NO{sub x} emissions decreased rapidly when the timing was retarded, and the amount of smoke and the level of particulate-matter emissions did not significantly increase. The single-cylinder engine tests confirmed the results of an earlier technical assessment and further indicated a need for a low-pressure-drop membrane specifically designed for oxygen enrichment. Extension data set indexed separately. 14 refs.

  3. Determination of the Average Aromatic Cluster Size of Fossil Fuels by Solid-State NMR at High Magnetic Field

    SciTech Connect

    Mao, Kanmi; Kennedy, Gordon J.; Althaus, Stacey M.; Pruski, Marek

    2013-01-07

    We show that the average aromatic cluster size in complex carbonaceous materials can be accurately determined using fast magic-angle spinning (MAS) NMR at a high magnetic field. To accurately quantify the nonprotonated aromatic carbon, we edited the 13C spectra using the recently reported MAS-synchronized spin–echo, which alleviated the problem of rotational recoupling of 1H-13C dipolar interactions associated with traditional dipolar dephasing experiments. The dependability of this approach was demonstrated on selected Argonne Premium coal standards, for which full sets of basic structural parameters were determined with high accuracy.

  4. 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.

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

    SciTech Connect

    Parish, T.A.

    1995-03-02

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

  6. Potential Annealing Treatments for Tailoring the Starting Microstructure of Low-Enriched U-Mo Dispersion Fuels to Optimize Performance During Irradiation

    SciTech Connect

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

    2011-12-01

    Low-enriched uranium-molybdenum 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 hours. 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.

  7. Analysis of the Reactor Physics of Low-Enrichment Fuel for the INL Advanced Test Reactor in support of RERTR

    SciTech Connect

    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 the 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.

  8. 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.

  9. 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

    NASA Astrophysics Data System (ADS)

    Favalli, A.; Vo, D.; Grogan, B.; Jansson, P.; Liljenfeldt, H.; Mozin, V.; Schwalbach, P.; Sjöland, A.; Tobin, S. J.; Trellue, H.; Vaccaro, S.

    2016-06-01

    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 [which is also a function of the variables in (1)]; (4) estimate the decay heat; and (5) determine the reactivity of spent fuel 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. The results obtained in comparison with the operator declared values, as well as the methodology developed, are discussed in detail in the paper.

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

    SciTech Connect

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

    2007-10-01

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

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

    SciTech Connect

    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 the 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)

  12. 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 PAGESBeta

    Favalli, Andrea; Vo, D.; Grogan, Brandon R.; Jansson, Peter; Liljenfeldt, Henrik; Mozin, Vladimir; Schwalbach, P.; Sjoland, A.; Tobin, Stephen J.; Trellue, Holly; et al

    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

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

    SciTech Connect

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

    1986-06-01

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

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

    SciTech Connect

    Mark DeHart; Gray S. Chang

    2012-04-01

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

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

    SciTech Connect

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

    1995-05-01

    An approach of short-notice random inspections (SNRIs) for inventory-change verification can enhance the effectiveness and efficiency of international safeguards at natural or low-enriched uranium (LEU) fuel fabrication plants. According to this approach, the plant operator declares the contents of nuclear material items before knowing if an inspection will occur to verify them. Additionally, items about which declarations are newly made should remain available for verification for an agreed time. Then a statistical inference can be made from verification results for items verified during SNRIs to the entire populations, i.e. the entire strata, even if inspectors were not present when many items were received or produced. A six-month field test of the feasibility of such SNRIs took place at the Westinghouse Electric Corporation Commercial Nuclear Fuel Division during 1993. Westinghouse personnel made daily declarations about both feed and product items, uranium hexafluoride cylinders and finished fuel assemblies, using a custom-designed computer ``mailbox``. Safeguards inspectors from the IAEA conducted eight SNRIs to verify these declarations. They arrived unannounced at the plant, in most cases immediately after travel from Canada, where the IAEA maintains a regional office. Items from both strata were verified during the SNRIs by meant of nondestructive assay equipment.

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

    SciTech Connect

    Bruce G. Schnitzler; Stanley K. Borowski

    2010-07-01

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

  17. Study of neutron physics: conversion of the University of Missouri-Rolla reactor to low-enriched fuel

    SciTech Connect

    Straka, M.; Covington, L.

    1987-01-01

    A detailed study of a fuel conversion (using LEU) has been undertaken for the University of Missouri-Rolla reactor. Results achieved with the available code package have been compared with the measured data whenever possible. The neutronic codes LEOPARD and 2DB-UM provided adequate results in most cases examined.

  18. Reactor Physics Methods and Preconceptual Core Design Analyses for Conversion of the Advanced Test Reactor to Low-Enriched Uranium Fuel Annual Report for Fiscal Year 2012

    SciTech Connect

    David W. Nigg; Sean R. Morrell

    2012-09-01

    Under the current long-term DOE policy and planning scenario, both the ATR and the ATRC will be reconfigured at an appropriate time within the next several years to operate with low-enriched uranium (LEU) fuel. This will be accomplished under the auspices of the Reduced Enrichment Research and Test Reactor (RERTR) Program, administered by the DOE National Nuclear Security Administration (NNSA). At a minimum, the internal design and composition of the fuel element plates and support structure will change, to accommodate the need for low enrichment in a manner that maintains total core excess reactivity at a suitable level for anticipated operational needs throughout each cycle while respecting all control and shutdown margin requirements and power distribution limits. The complete engineering design and optimization of LEU cores for the ATR and the ATRC will require significant multi-year efforts in the areas of fuel design, development and testing, as well as a complete re-analysis of the relevant reactor physics parameters for a core composed of LEU fuel, with possible control system modifications. Ultimately, revalidation of the computational physics parameters per applicable national and international standards against data from experimental measurements for prototypes of the new ATR and ATRC core designs will also be required for Safety Analysis Report (SAR) changes to support routine operations with LEU. This report is focused on reactor physics analyses conducted during Fiscal Year (FY) 2012 to support the initial development of several potential preconceptual fuel element designs that are suitable candidates for further study and refinement during FY-2013 and beyond. In a separate, but related, effort in the general area of computational support for ATR operations, the Idaho National Laboratory (INL) is conducting a focused multiyear effort to introduce modern high-fidelity computational reactor physics software and associated validation protocols to replace

  19. Enrichment of anodic biofilm inoculated with anaerobic or aerobic sludge in single chambered air-cathode microbial fuel cells.

    PubMed

    Gao, Chongyang; Wang, Aijie; Wu, Wei-Min; Yin, Yalin; Zhao, Yang-Guo

    2014-09-01

    Aerobic sludge after anaerobic pretreatment and anaerobic sludge were separately used as inoculum to start up air-cathode single-chamber MFCs. Aerobic sludge-inoculated MFCs arrived at 0.27 V with a maximum power density of 5.79 W m(-3), while anaerobic sludge-inoculated MFCs reached 0.21 V with 3.66 W m(-3). Microbial analysis with DGGE profiling and high-throughput sequencing indicated that aerobic sludge contained more diverse bacterial populations than anaerobic sludge. Nitrospira species dominated in aerobic sludge, while anaerobic sludge was dominated by Desulfurella and Acidithiobacillus species. Microbial community structure and composition in anodic biofilms enriched, respectively from aerobic and anaerobic sludges tended gradually to be similar. Potentially exoelectrogenic Geobacter and Anaeromusa species, biofilm-forming Zoogloea and Acinetobacter species were abundant in both anodic biofilms. This study indicated that aerobic sludge performed better for MFCs startup, and the enrichment of anodic microbial consortium with different inocula but same substrate resulted in uniformity of functional microbial communities. PMID:24973773

  20. In-plane and through-plane local and average Nusselt numbers in fibrous porous materials with different fiber layer temperatures: Gas diffusion layers for fuel cells

    NASA Astrophysics Data System (ADS)

    Sadeghifar, Hamidreza

    2016-09-01

    Convective heat transfer inside fibrous gas diffusion layers (GDLs) noticeably impacts the heat and water management of air-cooled polymer electrolyte membrane fuel cells (PEMFCs). Cutting-edge experiments have recently proved that convective heat transfer inside fibrous GDLs increases their thermal resistances considerably. However, heat transfer coefficients are difficult to measure experimentally or compute numerically for the millions of the tiny pores inside microstructural GDLs. The present study provides robust analytic models for predicting the heat transfer coefficient for both through-plane and in-plane flows inside fibrous media such as GDLs. The model is based on the unit cell approach and the integral method. Closed-form formulas are developed for local and average heat transfer coefficients. The model considers the temperature variations of the fiber layers along the medium thickness while assuming the same temperature for all the fibers in each layer. The model is well verified by COMSOL numerical data for a few pores inside a GDL. The simple, closed-form easy-to-use formulas developed in this study can be readily employed for predicting Nusselt number inside multilayer fibrous porous materials.

  1. Direct Measurement of Initial Enrichment and Burn-up of Spent Fuel Assembly with a Differential Die-Away Technique Based Instrument

    SciTech Connect

    Henzl, Vladimir; Swinhoe, Martyn T.; Tobin, Stephen J.

    2012-07-16

    A key objective of the Next Generation Safeguards Initiative (NGSI) is to utilize non-destructive assay (NDA) techniques to determine the elemental plutonium (Pu) content in a commercial-grade nuclear spent fuel assembly (SFA). In the third year of the NGSI Spent Fuel NDA project, the research focus is on the integration of a few NDA techniques. One of the reoccurring challenges to the accurate determination of Pu content has been the explicit dependence of the measured signal on the presence of neutron absorbers which build up in the assembly in accordance with its operating and irradiation history. The history of any SFA is often summarized by the parameters of burn-up (BU), initial enrichment (IE) and cooling time (CT). While such parameters can typically be provided by the operator, the ability to directly measure and verify them would significantly enhance the autonomy of the IAEA inspectorate. Within this paper, we demonstrate that an instrument based on a Differential Die-Away technique is in principle capable of direct measurement of IE and, should the CT be known, also the BU.

  2. 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.

  3. The Average of Rates and the Average Rate.

    ERIC Educational Resources Information Center

    Lindstrom, Peter

    1988-01-01

    Defines arithmetic, harmonic, and weighted harmonic means, and discusses their properties. Describes the application of these properties in problems involving fuel economy estimates and average rates of motion. Gives example problems and solutions. (CW)

  4. Preliminary Microstructural Characterization of Gadolinium-Enriched Stainless Steels for Spent Nuclear Fuel Baskets (title change from A)

    SciTech Connect

    DUPONT,J.N.; ROBINO,CHARLES V.; STEPHENS JR.,JOHN J.; MCCONNELL,PAUL E.; MIZIA,R.; BRANAGAN,D.

    2000-07-24

    Gadolinium (Gd) is a very potent neutron absorber that can potentially provide the nuclear criticality safety required for interim storage, transport, and final disposal of spent nuclear fuel. Gd could be incorporated into an alloy that can be fabricated into baskets to provide structural support, corrosion resistance, and nuclear criticality control. In particular, Gd alloyed with stainless steel has been identified as a material that may fulfill these functional requirements. However, no information is available in the open literature that describes the influence of Gd on the microstructure and resultant mechanical properties of stainless steels alloyed with Gd. Such information is vital for determination of the suitability of these types of alloys for the intended application. Characterization of Gd-stainless steel (Gd-SS) alloys is also necessary for an American Society for Testing and Materials (ASTM) material specification, subsequent code approval by the American Society of Mechanical Engineers (ASME), and regulatory approval by the Nuclear Regulatory Commission for subsequent use by the nuclear industry. The Department of Energy National Spent Nuclear Fuel Program at Idaho National Engineering and Environmental Laboratory has commissioned Lehigh University and Sandia National Laboratories to characterize the properties of a series of Gd-SS alloys to assess their suitability for the spent fuel basket application. Preliminary microstructural characterization results are presented on Gd stainless steels. Small gas tungsten arc buttons were prepared by melting 316L stainless steel with 0.1 to 10 wt.% Gd. These samples were characterized by light optical and electron optical microscopy to determine the distribution of alloying elements and volume fraction of Gd-rich phase. The results acquired to date indicate that no Gd is dissolved in the austenite matrix. Instead, the Gd was present as an interdendritic constituent, and the amount of the Gd-rich constituent

  5. Using the second law of thermodynamics for enrichment and isolation of microorganisms to produce fuel alcohols or hydrocarbons.

    PubMed

    Kohn, Richard A; Kim, Seon-Woo

    2015-10-01

    Fermentation of crops, waste biomass, or gases has been proposed as a means to produce desired chemicals and renewable fuels. The second law of thermodynamics has been shown to determine the net direction of metabolite flow in fermentation processes. In this article, we describe a process to isolate and direct the evolution of microorganisms that convert cellulosic biomass or gaseous CO2 and H2 to biofuels such as ethanol, 1-butanol, butane, or hexane (among others). Mathematical models of fermentation elucidated sets of conditions that thermodynamically favor synthesis of desired products. When these conditions were applied to mixed cultures from the rumen of a cow, bacteria that produced alcohols or alkanes were isolated. The examples demonstrate the first use of thermodynamic analysis to isolate bacteria and control fermentation processes for biofuel production among other uses. PMID:26231417

  6. Comments on the Joint Proposed Rulemaking to Establish Light-Duty Vehicle Greenhouse Gas Emission Standards and Corporate Average Fuel Economy Standards

    SciTech Connect

    Wenzel, Tom

    2009-10-27

    Tom Wenzel of Lawrence Berkeley National Laboratory comments on the joint rulemaking to establish greenhouse gas emission and fuel economy standards for light-duty vehicle, specifically on the relationship between vehicle weight and vehicle safety.

  7. Consideration of Nuclear Criticality When Directly Disposing Highly Enriched Spent Nuclear Fuel in Unsaturated Tuff - II: Geochemical Constraints

    SciTech Connect

    Rechard, Rob P.; Sanchez, Lawrence C.; Trellue, Holly R.

    2003-11-15

    This article presents several reasonable cases in which four mechanisms - dissolution, physical mixing, adsorption, and precipitation (either chemical change or evaporation) - might concentrate fissile material in and around a disposal container for radioactive waste at the proposed repository at Yucca Mountain, Nevada. The possible masses, concentrations, and volume are then compared to criticality limits. The cases examined evaluate the geologic barrier role in preventing criticality since engineered options for preventing criticality (e.g., boron or gadolinium neutron absorber in the disposal container) are not considered. The solid concentrations able to form in the natural environment are insufficient for criticality to occur because (a) solutions of {sup 235}U and {sup 239}Pu are clearly not critical; (b) physical mixing of fissile material with the entire potential iron oxide (as goethite - FeOOH) in a waste package is not critical; (c) the adsorption of {sup 239}Pu on consolidated iron oxide in a waste package is not critical; (d) the adsorption of {sup 235}U on consolidated iron oxide in a waste package is not critical when accounting for reduced adsorption because of carbonates at high pH; (e) the filtration of iron oxide colloids, containing fissile material, by the thin invert material is not critical; (f) insufficient retention through precipitation of {sup 235}U or {sup 239}Pu occurs in the invert; (g) adsorption of {sup 235}U and {sup 239}Pu on devitrified or clinoptolite-rich tuff below the repository is not critical; (h) the average precipitation/adsorption of {sup 235}U as uranyl silicates in the tuff is not critical by analogy with calcite deposition in lithophysae at Yucca Mountain; and (i) precipitation/adsorption (caused by cyclic drying) as uranyl silicates on fracture surfaces of the tuff is not critical by analogy with the oxidation of UO{sub 2}, migration of U{sup VI}, and precipitation in fractures at the Nopal I ore deposit in Mexico.

  8. Job Enrichment

    ERIC Educational Resources Information Center

    Sanders, Rick

    1970-01-01

    Job enrichment means giving people more decision-making power, more responsibility, more grasp of the totality of the job, and a sense of their own importance in the company. This article presents evidence of the successful working of this approach (Donnelly Mirrors), and the lack of success with an opposing approach (General Motors). (NL)

  9. Comments on the Joint Proposed Rulemaking to Establish Light-Duty Vehicle Greenhouse Gas Emission Standards and Corporate Average Fuel Economy Standards

    SciTech Connect

    Wenzel, Thomas P

    2009-10-27

    I appreciate the opportunity to provide comments on the joint rulemaking to establish greenhouse gas emission and fuel economy standards for light-duty vehicles. My comments are directed at the choice of vehicle footprint as the attribute by which to vary fuel economy and greenhouse gas emission standards, in the interest of protecting vehicle occupants from death or serious injury. I have made several of these points before when commenting on previous NHTSA rulemakings regarding CAFE standards and safety. The comments today are mine alone, and do not necessarily represent the views of the US Department of Energy, Lawrence Berkeley National Laboratory, or the University of California. My comments can be summarized as follows: (1) My updated analysis of casualty risk finds that, after accounting for drivers and crash location, there is a wide range in casualty risk for vehicles with the same weight or footprint. This suggests that reducing vehicle weight or footprint will not necessarily result in increased fatalities or serious injuries. (2) Indeed, the recent safety record of crossover SUVs indicates that weight reduction in this class of vehicles resulted in a reduction in fatality risks. (3) Computer crash simulations can pinpoint the effect of specific design changes on vehicle safety; these analyses are preferable to regression analyses, which rely on historical vehicle designs, and cannot fully isolate the effect of specific design changes, such as weight reduction, on crash outcomes. (4) There is evidence that automakers planned to build more large light trucks in response to the footprint-based light truck CAFE standards. Such an increase in the number of large light trucks on the road may decrease, rather than increase, overall safety.

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

    SciTech Connect

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

    2009-12-01

    operating conditions. The ultimate use of technologies tested on the engineering-scale test bed is to work with safeguards agencies to install them in operating plants (e.g., enrichment and fuel processing plants), thereby promoting new safeguards measures with minimal impact to operating plants. In addition, this system is useful in identifying features for new plants that can be incorporated as part of 'safeguards by design,' in which load cells and other monitoring technologies are specified to provide outputs for automated monitoring and inspector evaluation.

  11. Expensive Enrichment

    ERIC Educational Resources Information Center

    Resnikoff, Marvin

    1975-01-01

    This article presents an economic analysis of the nuclear fuel reprocessing industry. It indicates that while environmental safety devices have improved the working conditions, they have also added ever-increasing costs to this necessary process. (MA)

  12. 40 CFR 80.67 - Compliance on average.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 40 Protection of Environment 17 2012-07-01 2012-07-01 false Compliance on average. 80.67 Section...) REGULATION OF FUELS AND FUEL ADDITIVES Reformulated Gasoline § 80.67 Compliance on average. The requirements... with one or more of the requirements of § 80.41 is determined on average (“averaged gasoline”)....

  13. Neutron resonance averaging

    SciTech Connect

    Chrien, R.E.

    1986-10-01

    The principles of resonance averaging as applied to neutron capture reactions are described. Several illustrations of resonance averaging to problems of nuclear structure and the distribution of radiative strength in nuclei are provided. 30 refs., 12 figs.

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

    SciTech Connect

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

    2010-04-01

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

  15. 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. PMID:25902270

  16. Evaluation of Annealing Treatments for Producing Si-Rich Fuel/Matrix Interaction Layers in Low-Enriched U-Mo Dispersion Fuel Plates Rolled at a Low Temperature

    SciTech Connect

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

    2010-06-01

    During fabrication of U-7Mo dispersion fuels, exposure to relatively high temperatures affects the final microstructure of a fuel plate before it is inserted into a reactor. One impact of this high temperature exposure is a chemical interaction that can occur between dissimilar materials. For U-7Mo dispersion fuels, the U-7Mo particles will interact to some extent with the Al or Al alloy matrix to produce interaction products. It has been observed that the final irradiation behavior of a fuel plate can depend on the amount of interaction that occurs at the U-7Mo/matrix interface during fabrication, along with the type of phases that develop at this interface. For the case where a U-7Mo dispersion fuel has a Si-containing Al alloy matrix and is rolled at around 500°C, a Si-rich interaction product has been observed to form that can potentially have a positive impact on fuel performance during irradiation. This interaction product can exhibit stable irradiation behavior and it can act as a diffusion barrier to additional U-Mo/matrix interaction during irradiation. However, for U-7Mo dispersion fuels with softer claddings that are rolled at lower temperatures (e.g., near 425°C), a significant interaction layer has not been observed to form. As a result, the bulk of any interaction layer that develops in these fuels happens during irradiation, and the layer that forms may not exhibit as stable a behavior as one that is formed during fabrication. Therefore, it may be beneficial to add a heat treatment step during the fabrication of dispersion fuel plates with softer cladding alloys that will result in the formation of a uniform, Si-rich interaction layer that is a few microns thick around the U-Mo fuel particles. This type of layer would have characteristics like the one that has been observed in dispersion fuel plates with AA6061 cladding that are fabricated at 500°C, which may exhibit increased stability during irradiation. This report discusses the result of

  17. Study of a blast-furnace smelting technology which involves the injection of pulverized-coal fuel, natural gas, and an oxygen-enriched blast into the hearth

    SciTech Connect

    Ryzhenkov, A.N.; Yaroshevskii, S.L.; Zamuruev, V.P.; Popov, V.E.; Afanas'eva, Z.K.

    2006-05-15

    Studies were made of features of a blast-furnace smelting technology that involves the injection of natural gas (NG), oxygen (O{sub 2}) and pulverized-coal fuel (PCF) into the hearth. The technology has been implemented in the compensation and overcompensation regimes, which has made it possible to maintain or improve the gas dynamics of the furnace, the conditions for the reduction of iron oxides, the heating of the charge, and PCF combustion in the tuyere zone as PCF consumption is increased and coke use is decreased. Under the given conditions, with the blast having an oxygen content of 25.64-25.7%, the hearth injection of 131-138 kg PCF and 65-69 m{sup 3} NG for each ton of pig iron has made it possible to reduce coke consumption by 171-185 kg/ton pig (30.2-32.7%), reduce the consumption of comparison fuel by 36-37 kg/ton (5.2-5.3%), and lower the production cost of the pig iron by 43-49 hryvnas/ton (3.7-6.4%). Here, furnace productivity has increased 3.8-6.5%, while the quality of the conversion pig iron remains the same as before. Measures are being implemented to further increase the level and efficiency of PCF use.

  18. TEM CHARACTERIZATION OF IRRADIATED U3SI2/AL DISPERSION FUEL

    SciTech Connect

    J. Gan; B. Miller; D. Keiser; A. Robinson; P. Medvedev; D. Wachs

    2010-10-01

    The silicide dispersion fuel of U3Si2/Al has been recognized as a reasonably good performance fuel for nuclear research and test reactors except that it requires the use of high enrichment uranium. An irradiated U3Si2/Al dispersion fuel (~75% enrichment) from the high flux side of a RERTR-8 (U0R040) plate was characterized using transmission electron microscopy (TEM). The fuel plate was irradiated in the advanced test reactor (ATR) for 105 days. The average irradiation temperature and fission density of the fuel particles for the TEM sample are estimated to be approximately ~110 degrees C and 5.4 x 10-21 f/cm3. The characterization was performed using a 200KV TEM with a LaB6 filament. Detailed microstructural information along with composition analysis is obtained. The results and their implication on the performance of this silicide fuel are discussed.

  19. Geobacter, Anaeromyxobacter and Anaerolineae populations are enriched on anodes of root exudate-driven microbial fuel cells in rice field soil.

    PubMed

    Cabezas, Angela; Pommerenke, Bianca; Boon, Nico; Friedrich, Michael W

    2015-06-01

    Plant-based sediment microbial fuel cells (PMFCs) couple the oxidation of root exudates in living rice plants to current production. We analysed the composition of the microbial community on anodes from PMFC with natural rice field soil as substratum for rice by analysing 16S rRNA as an indicator of microbial activity and diversity. Terminal restriction fragment length polymorphism (TRFLP) analysis indicated that the active bacterial community on anodes from PMFCs differed strongly compared with controls. Moreover, clones related to Deltaproteobacteria and Chloroflexi were highly abundant (49% and 21%, respectively) on PMFCs anodes. Geobacter (19%), Anaeromyxobacter (15%) and Anaerolineae (17%) populations were predominant on anodes with natural rice field soil and differed strongly from those previously detected with potting soil. In open circuit (OC) control PMFCs, not allowing electron transfer, Deltaproteobacteria (33%), Betaproteobacteria (20%), Chloroflexi (12%), Alphaproteobacteria (10%) and Firmicutes (10%) were detected. The presence of an electron accepting anode also had a strong influence on methanogenic archaea. Hydrogenotrophic methanogens were more active on PMFC (21%) than on OC controls (10%), whereas acetoclastic Methanosaetaceae were more active on OC controls (31%) compared with PMFCs (9%). In conclusion, electron accepting anodes and rice root exudates selected for distinct potential anode-reducing microbial populations in rice soil inoculated PMFC. PMID:25683328

  20. Reprocessing RERTR silicide fuels

    SciTech Connect

    Rodrigues, G.C.; Gouge, A.P.

    1983-05-01

    The Reduced Enrichment Research and Test Reactor Program is one element of the United States Government's nonproliferation effort. High-density, low-enrichment, aluminum-clad uranium silicide fuels may be substituted for the highly enriched aluminum-clad alloy fuels now in use. Savannah River Laboratory has performed studies which demonstrate reprocessability of spent RERTR silicide fuels at Savannah River Plant. Results of dissolution and feed preparation tests and solvent extraction processing demonstrations with both unirradiated and irradiated uranium silicide fuels are presented.

  1. Reprocessing RERTR fuels

    SciTech Connect

    Rodrigues, G.C.

    1983-01-01

    The Reduced Enrichment Research and Test Reactor Program is one element of the United States Government's nonproliferation effort. High density, low enrichment aluminum-clad dispersed uranium compound fuels may be substituted for the highly enriched aluminum-clad aluminum-uranium alloy fuels now in use. Savannah River Laboratory has performed studies which demonstrate reprocessability of spent RERTR fuels at Savannah River Plant. Results of dissolution and feed preparation tests with both unirradiated and irradiated (up to approximately 90% burnup) fuels are presented. 13 references, 2 figures, 4 tables.

  2. Beyond Job Enrichment to Employment Enrichment

    ERIC Educational Resources Information Center

    Werther, William B., Jr.

    1975-01-01

    Employment enrichment views the total work environment confronting employees as a system consisting of two overlapping areas: worker-job and worker-organization subsystems. Job enrichment has improved the worker-job subsystem. The focus of this article is on methods of improving the worker-organization relationship. (Author/JB)

  3. The RERTR (Reduced Enrichment Research and Test Reactor) Program: Progress and plans

    SciTech Connect

    Travelli, A.

    1987-01-01

    The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. After a brief summary of the results which the RERTR Program, in collaboration with its many international partners, had achieved by the end of 1986, the activities, results, and new developments which occurred in 1987 are reviewed. Irradiation of the second miniplate series, concentrating on U/sub 3/Si/sub 2/-Al and U/sub 3/Si-Al fuels, was completed and postirradiation examinations were performed on many of its miniplates. The whole-core ORR demonstration with U/sub 3/Si/sub 2/-Al fuel at 4.8 g U/cm/sup 3/ was completed at the end of March with excellent results and with 29 elements estimated to have reached at least 40% average burnup. Good progress was made in the area of LEU usage for the production of fission /sup 99/Mo, and in the coordination of safety evaluations related to LEU conversions of US university reactors. Planned activities include testing and demonstrating advanced fuels intended to allow use of reduced enrichment uranium in very-high-performance reactors. Two candidate fuels are U/sub 3/Si-Al with 19.75% enrichment and U/sub 3/Si/sub 2/-Al with 45% enrichment. Demonstration of these fuels will include irradiation of full-size elements and, possibly, a full-core demonstration. Achievement of the final program goals is still projected for 1990. This progress could not have been possible without the close international cooperation which has existed from the beginning, and which is essential to the ultimate success of the RERTR Program.

  4. Averaging the inhomogeneous universe

    NASA Astrophysics Data System (ADS)

    Paranjape, Aseem

    2012-03-01

    A basic assumption of modern cosmology is that the universe is homogeneous and isotropic on the largest observable scales. This greatly simplifies Einstein's general relativistic field equations applied at these large scales, and allows a straightforward comparison between theoretical models and observed data. However, Einstein's equations should ideally be imposed at length scales comparable to, say, the solar system, since this is where these equations have been tested. We know that at these scales the universe is highly inhomogeneous. It is therefore essential to perform an explicit averaging of the field equations in order to apply them at large scales. It has long been known that due to the nonlinear nature of Einstein's equations, any explicit averaging scheme will necessarily lead to corrections in the equations applied at large scales. Estimating the magnitude and behavior of these corrections is a challenging task, due to difficulties associated with defining averages in the context of general relativity (GR). It has recently become possible to estimate these effects in a rigorous manner, and we will review some of the averaging schemes that have been proposed in the literature. A tantalizing possibility explored by several authors is that the corrections due to averaging may in fact account for the apparent acceleration of the expansion of the universe. We will explore this idea, reviewing some of the work done in the literature to date. We will argue however, that this rather attractive idea is in fact not viable as a solution of the dark energy problem, when confronted with observational constraints.

  5. On the Berdichevsky average

    NASA Astrophysics Data System (ADS)

    Rung-Arunwan, Tawat; Siripunvaraporn, Weerachai; Utada, Hisashi

    2016-04-01

    Through a large number of magnetotelluric (MT) observations conducted in a study area, one can obtain regional one-dimensional (1-D) features of the subsurface electrical conductivity structure simply by taking the geometric average of determinant invariants of observed impedances. This method was proposed by Berdichevsky and coworkers, which is based on the expectation that distortion effects due to near-surface electrical heterogeneities will be statistically smoothed out. A good estimation of a regional mean 1-D model is useful, especially in recent years, to be used as a priori (or a starting) model in 3-D inversion. However, the original theory was derived before the establishment of the present knowledge on galvanic distortion. This paper, therefore, reexamines the meaning of the Berdichevsky average by using the conventional formulation of galvanic distortion. A simple derivation shows that the determinant invariant of distorted impedance and its Berdichevsky average is always downward biased by the distortion parameters of shear and splitting. This means that the regional mean 1-D model obtained from the Berdichevsky average tends to be more conductive. As an alternative rotational invariant, the sum of the squared elements (ssq) invariant is found to be less affected by bias from distortion parameters; thus, we conclude that its geometric average would be more suitable for estimating the regional structure. We find that the combination of determinant and ssq invariants provides parameters useful in dealing with a set of distorted MT impedances.

  6. Nuclear mass inventory, photon dose rate and thermal decay heat of spent research reactor fuel assemblies

    SciTech Connect

    Pond, R.B.; Matos, J.E.

    1996-12-31

    This document has been prepared to assist research reactor operators possessing spent fuel containing enriched uranium of United States origin to prepare part of the documentation necessary to ship this fuel to the United States. Data are included on the nuclear mass inventory, photon dose rate, and thermal decay heat of spent research reactor fuel assemblies. Isotopic masses of U, Np, Pu and Am that are present in spent research reactor fuel are estimated for MTR, TRIGA and DIDO-type fuel assembly types. The isotopic masses of each fuel assembly type are given as functions of U-235 burnup in the spent fuel, and of initial U-235 enrichment and U-235 mass in the fuel assembly. Photon dose rates of spent MTR, TRIGA and DIDO-type fuel assemblies are estimated for fuel assemblies with up to 80% U-235 burnup and specific power densities between 0.089 and 2.857 MW/kg[sup 235]U, and for fission product decay times of up to 20 years. Thermal decay heat loads are estimated for spent fuel based upon the fuel assembly irradiation history (average assembly power vs. elapsed time) and the spent fuel cooling time.

  7. Analysis of thorium-salted fuels to improve uranium utilization in the once-through fuel cycle

    SciTech Connect

    Eschbach, E.A.; Merrill, E.T.; Prichard, A.W.

    1981-09-01

    Calculations and analyses indicate that no improvement can be achieved in uranium utilization for the once-through LWR fuel cycle over use of slightly enriched uranium by employing thorium distributed with uranium. The study included thorium additions: (1) slight amounts, (2) larger amounts, in either intimately mixed or in duplex pellets, (3) in spectrally shifted or not spectrally shifted reactors, and (4) in three- or five-year reactivity limited exposures. While thorium-uranium combinations improves the initial conversion ratio, the reactivity lifetime was not extended enough to override the additional uranium required. The effective fission cross-section of the bred /sup 233/U relative to /sup 239/Pu's in typical LWR neutron spectra is not large enough for /sup 233/U to make as great a contribution to end-of-life reactivity as /sup 239/Pu in a slightly enriched uranium fuel element. /sup 233/U's reactivity contribution relative to /sup 239/Pu's is lower in fuel configurations such as slightly enriched uranium LWR fuel loads. On the other hand, /sup 233/U's reactivity contribution appears more positive for reactors that involve lower average concentrations of thermal neutron absorbers. If /sup 238/U-thorium fuels reprocessed, the recovered /sup 233/U would increase uranium utilization, but may not reduce fuel cycle costs. The thorium-salted fuels exhibit substantially flatter reactivity characteristics with exposure time. Spectral shift helped the utilization of uranium and thorium.

  8. Reduced enrichment for research and test reactors: Proceedings

    SciTech Connect

    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.

  9. Derived enriched uranium market

    SciTech Connect

    Rutkowski, E.

    1996-12-01

    The potential impact on the uranium market of highly enriched uranium from nuclear weapons dismantling in the Russian Federation and the USA is analyzed. Uranium supply, conversion, and enrichment factors are outlined for each country; inventories are also listed. The enrichment component and conversion components are expected to cause little disruption to uranium markets. The uranium component of Russian derived enriched uranium hexafluoride is unresolved; US legislation places constraints on its introduction into the US market.

  10. Covariant approximation averaging

    NASA Astrophysics Data System (ADS)

    Shintani, Eigo; Arthur, Rudy; Blum, Thomas; Izubuchi, Taku; Jung, Chulwoo; Lehner, Christoph

    2015-06-01

    We present a new class of statistical error reduction techniques for Monte Carlo simulations. Using covariant symmetries, we show that correlation functions can be constructed from inexpensive approximations without introducing any systematic bias in the final result. We introduce a new class of covariant approximation averaging techniques, known as all-mode averaging (AMA), in which the approximation takes account of contributions of all eigenmodes through the inverse of the Dirac operator computed from the conjugate gradient method with a relaxed stopping condition. In this paper we compare the performance and computational cost of our new method with traditional methods using correlation functions and masses of the pion, nucleon, and vector meson in Nf=2 +1 lattice QCD using domain-wall fermions. This comparison indicates that AMA significantly reduces statistical errors in Monte Carlo calculations over conventional methods for the same cost.

  11. Supply of enriched uranium for research reactors

    SciTech Connect

    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 on December 31, 1988. Now, since it is apparent that these back-end problem have been solved by AEA`s ability to reprocess and the preparedness of the USA to again accept physically spent research reactor fuel the author is focusing with this paper again on the front-end of the fuel cycle on the question whether there is at all a safe supply of low and high enriched uranium for research reactors in the future.

  12. Average density in cosmology

    SciTech Connect

    Bonnor, W.B.

    1987-05-01

    The Einstein-Straus (1945) vacuole is here used to represent a bound cluster of galaxies embedded in a standard pressure-free cosmological model, and the average density of the cluster is compared with the density of the surrounding cosmic fluid. The two are nearly but not quite equal, and the more condensed the cluster, the greater the difference. A theoretical consequence of the discrepancy between the two densities is discussed. 25 references.

  13. 40 CFR 80.205 - How is the annual refinery or importer average and corporate pool average sulfur level determined?

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 40 Protection of Environment 16 2010-07-01 2010-07-01 false How is the annual refinery or importer average and corporate pool average sulfur level determined? 80.205 Section 80.205 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR PROGRAMS (CONTINUED) REGULATION OF FUELS AND FUEL ADDITIVES Gasoline Sulfur Gasoline...

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

    SciTech Connect

    Myers, Astasia

    2011-06-28

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

  15. 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

  16. 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.

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

    SciTech Connect

    Not Available

    1980-05-01

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

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

    SciTech Connect

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

    2012-06-07

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

  19. U. S. forms uranium enrichment corporation

    SciTech Connect

    Seltzer, R.

    1993-07-12

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

  20. Americans' Average Radiation Exposure

    SciTech Connect

    NA

    2000-08-11

    We live with radiation every day. We receive radiation exposures from cosmic rays, from outer space, from radon gas, and from other naturally radioactive elements in the earth. This is called natural background radiation. It includes the radiation we get from plants, animals, and from our own bodies. We also are exposed to man-made sources of radiation, including medical and dental treatments, television sets and emission from coal-fired power plants. Generally, radiation exposures from man-made sources are only a fraction of those received from natural sources. One exception is high exposures used by doctors to treat cancer patients. Each year in the United States, the average dose to people from natural and man-made radiation sources is about 360 millirem. A millirem is an extremely tiny amount of energy absorbed by tissues in the body.

  1. Profile of World Uranium Enrichment Programs - 2007

    SciTech Connect

    Laughter, Mark D

    2007-11-01

    It is generally agreed that the most difficult step in building a nuclear weapon is acquiring weapons grade fissile material, either plutonium or highly enriched uranium (HEU). Plutonium is produced in a nuclear reactor, while HEU is produced using a uranium enrichment process. Enrichment is also an important step in the civil nuclear fuel cycle, in producing low enriched uranium (LEU) for use in fuel for nuclear reactors. However, the same equipment used to produce LEU for nuclear fuel can also be used to produce HEU for weapons. Safeguards at an enrichment plant are the array of assurances and verification techniques that ensure uranium is only enriched to LEU, no undeclared LEU is produced, and no uranium is enriched to HEU or secretly diverted. There are several techniques for enriching uranium. The two most prevalent are gaseous diffusion, which uses older technology and requires a lot of energy, and gas centrifuge separation, which uses more advanced technology and is more energy efficient. Gaseous diffusion plants (GDPs) provide about 40% of current world enrichment capacity, but are being phased out as newer gas centrifuge enrichment plants (GCEPs) are constructed. Estimates of current and future enrichment capacity are always approximate, due to the constant upgrades, expansions, and shutdowns occurring at enrichment plants, largely determined by economic interests. Currently, the world enrichment capacity is approximately 53 million kg-separative work units (SWU) per year, with 22 million in gaseous diffusion and 31 million in gas centrifuge plants. Another 23 million SWU/year of capacity are under construction or planned for the near future, almost entirely using gas centrifuge separation. Other less-efficient techniques have also been used in the past, including electromagnetic and aerodynamic separations, but these are considered obsolete, at least from a commercial perspective. Laser isotope separation shows promise as a possible enrichment technique

  2. Fuel behavior comparison for a research reactor

    NASA Astrophysics Data System (ADS)

    Negut, Gh.; Mladin, M.; Prisecaru, I.; Danila, N.

    2006-06-01

    The paper presents the behavior and properties analysis of the low enriched uranium fuel, which will be loaded in the Romanian TRIGA 14 MW steady state research reactor compared with the original high enriched uranium fuel. The high and low enriched uranium fuels have similar thermal properties, but different nuclear properties. The research reactor core was modeled with both fuel materials and the reactor behavior was studied during a reactivity insertion accident. The thermal hydraulic analysis results are compared with that obtained from the safety analysis report for high enriched uranium fuel core. The low enriched uranium fuel shows a good behavior during reactivity insertion accident and a revised safety analysis report will be made for the low enriched uranium fuel core.

  3. Enrichment through Creative Arts.

    ERIC Educational Resources Information Center

    Krause, Claire S.

    The CREST (Creative Resources Enriching Student Talents) Project, an enrichment approach for elementary gifted, talented, and creative students, is described. The project is explained to incorporate an interdisciplinary approach to instruction in art and science using resources within the community. Chapter 1 outlines the project philosophy,…

  4. Experimental measurements and analytical analysis related to gas turbine heat transfer. Part 1: Time-averaged heat-flux and surface-pressure measurements on the vanes and blades of the SSME fuel-side turbine and comparison with prediction. Part 2: Phase-resolved surface-pressure and heat-flux measurements on the first blade of the SSME fuel-side turbine

    NASA Technical Reports Server (NTRS)

    1994-01-01

    Time averaged Stanton number and surface-pressure distributions are reported for the first-stage vane row, the first stage blade row, and the second stage vane row of the Rocketdyne Space Shuttle Main Engine two-stage fuel-side turbine. Unsteady pressure envelope measurements for the first blade are also reported. These measurements were made at 10 percent, 50 percent, and 90 percent span on both the pressure and suction surfaces of the first stage components. Additional Stanton number measurements were made on the first stage blade platform blade tip, and shroud, and at 50 percent span on the second vane. A shock tube was used as a short duration source of heated and pressurized air to which the turbine was subjected. Platinum thin-film heat flux gages were used to obtain the heat flux measurements, while miniature silicon-diaphragm flush-mounted pressure transducers were used to obtain the pressure measurements. The first stage vane Stanton number distributions are compared with predictions obtained using a version of STAN5 and a quasi-3D Navier-Stokes solution. This same quasi-3D N-S code was also used to obtain predictions for the first blade and the second vane.

  5. 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.

  6. Nuclear fuel, refueling, fuel handling, and licensing and regulation. Volume eleven

    SciTech Connect

    Not Available

    1986-01-01

    Volume eleven covers nuclear fuel (what is nuclear fuel, the nuclear fuel cycle, uranium mining, milling, and refining, uranium enrichment, nuclear fuel fabrication, fuel reprocessing), refueling and fuel handling (fuel assembly identification, fuel handling equipment, the fueling and refueling process, PWR refueling, BWR refueling), and licensing and regulation requirements (development of nuclear energy, federal licensing and regulatory organization, schedule for nuclear power plants, contents of reports to the Federal regulatory agency, nuclear power plant operator qualification).

  7. 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…

  8. Irradiation Experiment Conceptual Design Parameters for NBSR Fuel Conversion

    SciTech Connect

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

    2014-04-30

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

  9. Disposition of excess highly enriched uranium status and update

    SciTech Connect

    Williams, C.K. III; Arbital, J.G.

    1997-09-01

    This paper presents the status of the US DOE program charged with the disposition of excess highly enriched uranium (HEU). Approximately 174 metric tonnes of HEU, with varying assays above 20 percent, has been declared excess from US nuclear weapons. A progress report on the identification and characterization of specific batches of excess HEU is provided, and plans for processing it into commercial nuclear fuel or low-level radioactive waste are described. The resultant quantities of low enriched fuel material expected from processing are given, as well as the estimated schedule for introducing the material into the commercial reactor fuel market. 2 figs., 3 tabs.

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

    SciTech Connect

    Takaki, Naoyuki; Sagawa, Yu; Umino, Akitake; Sekimoto, Hiroshi

    2013-07-01

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

  11. 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

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

    SciTech Connect

    Brim, Cornelia P.

    2013-04-01

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

  13. Irradiation Experiment Conceptual Design Parameters for NBSR Fuel Conversion

    SciTech Connect

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

    2013-03-31

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

  14. 40 CFR 63.7522 - Can I use emissions averaging to comply with this subpart?

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... subcategories and you may not average across subcategories: (i) Units designed to burn coal/solid fossil fuel...: 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...

  15. 40 CFR 63.7522 - Can I use emissions averaging to comply with this subpart?

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... subcategories and you may not average across subcategories: (i) Units designed to burn coal/solid fossil fuel...: 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...

  16. Laser and gas centrifuge enrichment

    SciTech Connect

    Heinonen, Olli

    2014-05-09

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

  17. 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.

  18. TRIGA Mark II Criticality Benchmark Experiment with Burned Fuel

    SciTech Connect

    Persic, Andreja; Ravnik, Matjaz; Zagar, Tomaz

    2000-12-15

    The experimental results of criticality benchmark experiments performed at the Jozef Stefan Institute TRIGA Mark II reactor are presented. The experiments were performed with partly burned fuel in two compact and uniform core configurations in the same arrangements as were used in the fresh fuel criticality benchmark experiment performed in 1991. In the experiments, both core configurations contained only 12 wt% U-ZrH fuel with 20% enriched uranium. The first experimental core contained 43 fuel elements with average burnup of 1.22 MWd or 2.8% {sup 235}U burned. The last experimental core configuration was composed of 48 fuel elements with average burnup of 1.15 MWd or 2.6% {sup 235}U burned. The experimental determination of k{sub eff} for both core configurations, one subcritical and one critical, are presented. Burnup for all fuel elements was calculated in two-dimensional four-group diffusion approximation using the TRIGLAV code. The burnup of several fuel elements was measured also by the reactivity method.

  19. Nuclear fuel cycle costs

    SciTech Connect

    Burch, W.D.; Haire, M.J.; Rainey, R.H.

    1982-02-01

    The costs for the back-end of the nuclear fuel cycle, which were developed as part of the Nonproliferation Alternative Systems Assessment Program (NASAP), are presented. Total fuel cycle costs are given for the pressurized water reactor once-through and fuel recycle systems, and for the liquid-metal fast breeder reactor system. These calculations show that fuel cycle costs are a small part of the total power costs. For breeder reactors, fuel cycle costs are about half that of the present once-through system. The total power cost of the breeder reactor system is greater than that of light-water reactor at today's prices for uranium and enrichment.

  20. Profile of World Uranium Enrichment Programs-2009

    SciTech Connect

    Laughter, Mark D

    2009-04-01

    It is generally agreed that the most difficult step in building a nuclear weapon is acquiring fissile material, either plutonium or highly enriched uranium (HEU). Plutonium is produced in a nuclear reactor, whereas HEU is produced using a uranium enrichment process. Enrichment is also an important step in the civil nuclear fuel cycle, in producing low enriched uranium (LEU) for use as fuel for nuclear reactors to generate electricity. However, the same equipment used to produce LEU for nuclear reactor fuel can also be used to produce HEU for weapons. Safeguards at an enrichment plant are the array of assurances and verification techniques that ensure uranium is not diverted or enriched to HEU. There are several techniques for enriching uranium. The two most prevalent are gaseous diffusion, which uses older technology and requires a lot of energy, and gas centrifuge separation, which uses more advanced technology and is more energy efficient. Gaseous diffusion plants (GDPs) provide about 40% of current world enrichment capacity but are being phased out as newer gas centrifuge enrichment plants (GCEPs) are constructed. Estimates of current and future enrichment capacity are always approximate, due to the constant upgrades, expansions, and shutdowns occurring at enrichment plants, largely determined by economic interests. Currently, the world enrichment capacity is approximately 56 million kilogram separative work units (SWU) per year, with 22.5 million in gaseous diffusion and more than 33 million in gas centrifuge plants. Another 34 million SWU/year of capacity is under construction or planned for the near future, almost entirely using gas centrifuge separation. Other less-efficient techniques have also been used in the past, including electromagnetic and aerodynamic separations, but these are considered obsolete, at least from a commercial perspective. Laser isotope separation shows promise as a possible enrichment technique of the future but has yet to be

  1. 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

  2. The uranium cylinder assay system for enrichment plant safeguards

    SciTech Connect

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

    2010-01-01

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

  3. Averaging Models: Parameters Estimation with the R-Average Procedure

    ERIC Educational Resources Information Center

    Vidotto, G.; Massidda, D.; Noventa, S.

    2010-01-01

    The Functional Measurement approach, proposed within the theoretical framework of Information Integration Theory (Anderson, 1981, 1982), can be a useful multi-attribute analysis tool. Compared to the majority of statistical models, the averaging model can account for interaction effects without adding complexity. The R-Average method (Vidotto &…

  4. Job Enrichment in Extension.

    ERIC Educational Resources Information Center

    Fourman, Louis S.; Jones, Jo

    1997-01-01

    Interviews with 10 participants in Ohio State University's job enrichment program for midcareer extension agents found that 5 returned to their same jobs after the experience but only 2 felt challenged/renewed. Part-time participation while working made it difficult to balance responsibilities. More information and a structured orientation were…

  5. Designing job enrichment projects.

    PubMed

    Clakeley, G L

    1988-01-01

    This paper describes a management strategy for a job satisfaction program utilized in a large occupational therapy department. The goal of the program is to retain satisfied, productive employees and reduce attrition of therapists and assistants. The use of job enrichment projects for occupational therapy assistants will be presented with brief descriptions of two projects. PMID:23944880

  6. Enrichments in Biology

    ERIC Educational Resources Information Center

    Richard, Paul W.

    1969-01-01

    Emphasizes the need for enrichment materials in addition to laboratory and textbook work, particularly for interested and able students. Discusses the use of filmstrips, loop films, Innovations to Inquiry, BSCS pamphlets, newspapers, magazines and scientific periodicals, television, field trips, library resources, and programed units. (EB)

  7. Enriching the Catalog

    ERIC Educational Resources Information Center

    Tennant, Roy

    2004-01-01

    After decades of costly and time-consuming effort, nearly all libraries have completed the retrospective conversion of their card catalogs to electronic form. However, bibliographic systems still are really not much more than card catalogs on wheels. Enriched content that Amazon.com takes for granted--such as digitized tables of contents, cover…

  8. Science Student Enrichment Opportunities.

    ERIC Educational Resources Information Center

    California State Dept. of Education, Sacramento.

    This document was developed with the intention of increasing California public school students' awareness of and participation in science-related enrichment activities. Some of the activities are intended for participation by individuals, while others are meant for teams of students. These annual events are listed in chronological order for a…

  9. Enriching Number Knowledge

    ERIC Educational Resources Information Center

    Mack, Nancy K.

    2011-01-01

    Exploring number systems of other cultures can be an enjoyable learning experience that enriches students' knowledge of numbers and number systems in important ways. It helps students deepen mental computation fluency, knowledge of place value, and equivalent representations for numbers. This article describes how the author designed her…

  10. Reduced enrichment for research and test reactors: Proceedings

    SciTech Connect

    Not Available

    1988-05-01

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

  11. Averaging Internal Consistency Reliability Coefficients

    ERIC Educational Resources Information Center

    Feldt, Leonard S.; Charter, Richard A.

    2006-01-01

    Seven approaches to averaging reliability coefficients are presented. Each approach starts with a unique definition of the concept of "average," and no approach is more correct than the others. Six of the approaches are applicable to internal consistency coefficients. The seventh approach is specific to alternate-forms coefficients. Although the…

  12. Air/fuel ratio controller

    SciTech Connect

    Schechter, M.M.; Simko, A.O.

    1980-12-23

    An internal combustion engine has a fuel injection pump and an air/fuel ratio controller. The controller has a lever that is connected to the pump lever. An aneroid moves the controller lever as a function of changes in intake manifold vacuum to maintain a constant air/fuel ratio to the mixture charge. A fuel enrichment linkage is provided that modifies the movement of the fuel flow control lever by the aneroid in response to changes in manifold gas temperature levels and exhaust gas recirculation to maintain the constant air/fuel ratio. A manual override is provided to obtain a richer air/fuel ratio for maximum acceleration.

  13. Enrichment Strategies in Phosphoproteomics.

    PubMed

    Leitner, Alexander

    2016-01-01

    The comprehensive study of the phosphoproteome is heavily dependent on appropriate enrichment strategies that are most often, but not exclusively, carried out on the peptide level. In this chapter, I give an overview of the most widely used techniques. In addition to dedicated antibodies, phosphopeptides are enriched by their selective interaction with metals in the form of chelated metal ions or metal oxides. The negative charge of the phosphate group is also exploited in a variety of chromatographic fractionation methods that include different types of ion exchange chromatography, hydrophilic interaction chromatography (HILIC), and electrostatic repulsion HILIC (ERLIC) chromatography. Selected examples from the literature will demonstrate how a combination of these techniques with current high-performance mass spectrometry enables the identification of thousands of phosphorylation sites from various sample types. PMID:26584921

  14. Gas Centrifuge Enrichment Plant Safeguards System Modeling

    SciTech Connect

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

    2006-06-05

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

  15. The Averaging Problem in Cosmology

    NASA Astrophysics Data System (ADS)

    Paranjape, Aseem

    2009-06-01

    This thesis deals with the averaging problem in cosmology, which has gained considerable interest in recent years, and is concerned with correction terms (after averaging inhomogeneities) that appear in the Einstein equations when working on the large scales appropriate for cosmology. It has been claimed in the literature that these terms may account for the phenomenon of dark energy which causes the late time universe to accelerate. We investigate the nature of these terms by using averaging schemes available in the literature and further developed to be applicable to the problem at hand. We show that the effect of these terms when calculated carefully, remains negligible and cannot explain the late time acceleration.

  16. Investigation of the low enrichment conversion of the Texas A and M Nuclear Science Center Reactor

    SciTech Connect

    Reuscher, J.A.

    1988-01-01

    The use of highly enriched uranium as a fuel research reactors is of concern due to the possibility of diversion for nuclear weapons applications. The Texas A M TRIGA reactor currently uses 70% enriched uranium in a FLIP (Fuel Life Improvement Program) fuel element manufactured by General Atomics. Thus fuel also contains 1.5 weight percent of erbium as a burnable poison to prolong useful core life. US university reactors that use highly enriched uranium will be required to covert to 20% or less enrichment to satisfy Nuclear Regulatory Commission requirements for the next core loading if the fuel is available. This investigation examined the feasibility of a material alternate to uranium-zirconium hydride for LEU conversion of a TRIGA reactor. This material is a beryllium oxide uranium dioxide based fuel. The theoretical aspects of core physics analyses were examined to assess the potential advantages of the alternative fuel. A basic model was developed for the existing core configuration since it is desired to use the present fuel element grid for the replacement core. The computing approach was calibrated to the present core and then applied to a core of BeO-UO{sub 2} fuel elements. Further calculations were performed for the General Atomics TRIGA low-enriched uranium zirconium hydride fuel.

  17. Driving Speed vs Fuel Efficiency.

    ERIC Educational Resources Information Center

    Vest, Floyd

    1980-01-01

    A mathematical treatment of the relationship between driving speed and fuel efficiency is presented. The material involves applications of exponentials, logarithms, and elementary calculus, and is intended to be enrichment material for secondary and lower college mathematics classes. (MP)

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

    SciTech Connect

    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 fuel 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 the

  19. High average power pockels cell

    DOEpatents

    Daly, Thomas P.

    1991-01-01

    A high average power pockels cell is disclosed which reduces the effect of thermally induced strains in high average power laser technology. The pockels cell includes an elongated, substantially rectangular crystalline structure formed from a KDP-type material to eliminate shear strains. The X- and Y-axes are oriented substantially perpendicular to the edges of the crystal cross-section and to the C-axis direction of propagation to eliminate shear strains.

  20. Technical basis in support of the conversion of the University of Missouri Research Reactor (MURR) core from highly-enriched to low-enriched uranium - core neutron physics

    SciTech Connect

    Stillman, J.; Feldman, E.; Foyto, L; Kutikkad, K; McKibben, J C; Peters, N.; Stevens, J.

    2012-09-01

    This report contains the results of reactor design and performance for conversion of the University of Missouri Research Reactor (MURR) from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL) and the MURR Facility. The core conversion to LEU is being performed with financial support of the U. S. government.

  1. 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.

  2. 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)

  3. Clogging the enrichment bypass

    SciTech Connect

    1995-05-01

    The issue that has come to be known as the bypass is the latest dispute in the continuing saga of uranium trade between the United States and the Commonwealth of Indepenent States. Bypass deals now being conducted by US utilities have drawn heavy criticism from the US Department of Commerce but are perfectly legal under the terms of the Suspension Agreements. Urged on by the United States Enrichment Corporation, the US DOC intends to stop, or at least severely restrict, bypass deals. Using a recent agreement with Kazakhstan as a case study, this article reviews the implications of such bypass deals.

  4. Determining GPS average performance metrics

    NASA Technical Reports Server (NTRS)

    Moore, G. V.

    1995-01-01

    Analytic and semi-analytic methods are used to show that users of the GPS constellation can expect performance variations based on their location. Specifically, performance is shown to be a function of both altitude and latitude. These results stem from the fact that the GPS constellation is itself non-uniform. For example, GPS satellites are over four times as likely to be directly over Tierra del Fuego than over Hawaii or Singapore. Inevitable performance variations due to user location occur for ground, sea, air and space GPS users. These performance variations can be studied in an average relative sense. A semi-analytic tool which symmetrically allocates GPS satellite latitude belt dwell times among longitude points is used to compute average performance metrics. These metrics include average number of GPS vehicles visible, relative average accuracies in the radial, intrack and crosstrack (or radial, north/south, east/west) directions, and relative average PDOP or GDOP. The tool can be quickly changed to incorporate various user antenna obscuration models and various GPS constellation designs. Among other applications, tool results can be used in studies to: predict locations and geometries of best/worst case performance, design GPS constellations, determine optimal user antenna location and understand performance trends among various users.

  5. Vocal attractiveness increases by averaging.

    PubMed

    Bruckert, Laetitia; Bestelmeyer, Patricia; Latinus, Marianne; Rouger, Julien; Charest, Ian; Rousselet, Guillaume A; Kawahara, Hideki; Belin, Pascal

    2010-01-26

    Vocal attractiveness has a profound influence on listeners-a bias known as the "what sounds beautiful is good" vocal attractiveness stereotype [1]-with tangible impact on a voice owner's success at mating, job applications, and/or elections. The prevailing view holds that attractive voices are those that signal desirable attributes in a potential mate [2-4]-e.g., lower pitch in male voices. However, this account does not explain our preferences in more general social contexts in which voices of both genders are evaluated. Here we show that averaging voices via auditory morphing [5] results in more attractive voices, irrespective of the speaker's or listener's gender. Moreover, we show that this phenomenon is largely explained by two independent by-products of averaging: a smoother voice texture (reduced aperiodicities) and a greater similarity in pitch and timbre with the average of all voices (reduced "distance to mean"). These results provide the first evidence for a phenomenon of vocal attractiveness increases by averaging, analogous to a well-established effect of facial averaging [6, 7]. They highlight prototype-based coding [8] as a central feature of voice perception, emphasizing the similarity in the mechanisms of face and voice perception. PMID:20129047

  6. Surplus Highly Enriched Uranium Disposition Program plan

    SciTech Connect

    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.

  7. Fuel flexible fuel injector

    SciTech Connect

    Tuthill, Richard S; Davis, Dustin W; Dai, Zhongtao

    2015-02-03

    A disclosed fuel injector provides mixing of fuel with airflow by surrounding a swirled fuel flow with first and second swirled airflows that ensures mixing prior to or upon entering the combustion chamber. Fuel tubes produce a central fuel flow along with a central airflow through a plurality of openings to generate the high velocity fuel/air mixture along the axis of the fuel injector in addition to the swirled fuel/air mixture.

  8. Postirradiation examination of high-U-loaded low-enriched U/sub 3/O/sub 8/, UAl/sub 2/, and U/sub 3/Si test fuel plates

    SciTech Connect

    Gomez, J.; Morando, R.; Perez, E.E.; Giorsetti, D.R.; Copeland, G.L.; Hofmann, G.; Snelgrove, J.L.

    1984-01-01

    The scope of this work is to present an evaluation of the postirradiation examination of the second set of high-U-loaded low-enriched U/sub 3/O/sub 8/, UAl/sub 2/ and U/sub 3/Si miniature plates manufactured by the Comision Nacional de Energia Atomica (CNEA) of Argentina, and irradiated and examinated, within the framework of the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Oak Ridge National Laboratory and Argonne National Laboratory. This paper includes fabrication details of the plates, their irradiation history and the results of postirradiation examination which are compared to those of the previous test and to present results from other laboratories participating in the RERTR Program. Postirradiation examination of these plates showed satisfactory poerformance for the oxides, aluminides and silicides (except for the highest-loaded U/sub 3/Si plate) with the only indication of detrimental behavior during the slight bowing of some plates at about 80% burnup.

  9. Reduced enrichment neutronics evaluation for Texas A and M's TRIGA reactor

    SciTech Connect

    Rajalakshmi, M.J.; Reuscher, J.A. )

    1990-06-01

    The Texas A and M Nuclear Science Center reactor (NSCR) designed by General Atomics (GA) uses a fuel-life improvement program TRIGA fuel element. It is composed of 8.5 wt% uranium in U-ZrH{sub 1.6}-Er fuel with a {sup 235}U enrichment of 70 at.% and 1.5 wt% of erbium. The US Nuclear Regulatory Commission requires that the enrichment not exceed 20 at.% for the next loading, provided the fuel is available. To meet this requirement GA has developed shrouded four-rod clusters using low-enriched U-ZrH{sub 1.6}-Er fuel for TRIGA cores operating at powers up to 2 MW. This fuel contains 20 wt% of uranium with an enrichment of 20 at.% and 0.5 wt% of erbium and a homogeneous mixture of hydrogen moderator. Thermal-hydraulic calculations show the feasibility of operating the NSCR at 2 MW. The objective of this study is to assess the ability of the NSCR to operate at 2 MW using the reduced-enrichment fuel. This study also covers a three-dimensional neutronics analysis of the NSCR core using the new fuel. Results obtained are compared with results obtained with another candidate fuel, BeO-UO{sub 2}-Er.

  10. Evaluations of average level spacings

    SciTech Connect

    Liou, H.I.

    1980-01-01

    The average level spacing for highly excited nuclei is a key parameter in cross section formulas based on statistical nuclear models, and also plays an important role in determining many physics quantities. Various methods to evaluate average level spacings are reviewed. Because of the finite experimental resolution, to detect a complete sequence of levels without mixing other parities is extremely difficult, if not totally impossible. Most methods derive the average level spacings by applying a fit, with different degrees of generality, to the truncated Porter-Thomas distribution for reduced neutron widths. A method that tests both distributions of level widths and positions is discussed extensivey with an example of /sup 168/Er data. 19 figures, 2 tables.