WorldWideScience

Sample records for average fuel enrichment

  1. Thermal breeder fuel enrichment zoning

    Science.gov (United States)

    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

    Energy Technology Data Exchange (ETDEWEB)

    Roser, R. [NRG Technologies, Inc., Reno, NV (United States)

    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. Guide for the estimation of the {alpha} and {beta} coefficients in the Average enrichment equation as burnt function by fuel type; Guia para la estimacion de los coeficientes {alpha} y {beta} en la ecuacion de enriquecimiento promedio como funcion del quemado por tipo de combustible

    Energy Technology Data Exchange (ETDEWEB)

    Montes T, J.L.; Cortes C, C.C. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    1992-08-15

    The objective of the report is to determine manually or by means of a calculation sheet, the coefficients {alpha} and {beta} of the average enrichment equation as function of the fuel burnt (B) using the Lineal Reactivity Pattern, with information generated by the RECORD code of the FMS package. (Author00.

  4. Optimization of fuel rod enrichment distribution for BWR fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Hirano, Yasushi; Hida, Kazuki; Sakurada, Koichi; Yamamoto, Munenari [Toshiba Corp., Kawasaki, Kanagawa (Japan). Nuclear Engineering Lab.

    1996-09-01

    A practical method was developed for determining the optimum fuel enrichment distribution within a boiling water reactor fuel assembly. The method deals with two different optimization problems, i.e. the combinatorial optimization problem of grouping fuel rods into a given number of rod groups with the same enrichment, and the problem of determining an optimal enrichment for each fuel rod under the resultant rod-grouping pattern. In solving these problems, the primary goal is to minimize a predefined objective function over a given exposure period. The objective function used here is defined by the linear combination C{sub 1}X + C{sub 2}X{sub G}, where X and X{sub G} stand, respectively, for control variables giving constraint to the local power peaking factor and the gadolinium rod power. C{sub 1} and C{sub 2} are user-definable weighting factors to accommodate design preferences. The algorithm for solving this combinatorial optimization problem starts by finding the optimal enrichment vector without any rod-grouping, and promising candidates of rod-grouping patterns are found by exhaustive enumeration based on the resulting fuel enrichment ordering. This latter problem is solved using the method of approximation programming. A practical application is shown for a contemporary 8 x 8 Pu mixed-oxide fuel assembly with 10 gadolinium-poisoned rods. (author)

  5. Low-enriched fuel particle performance review. [UO2

    Energy Technology Data Exchange (ETDEWEB)

    Homan, F.; Nabielek, H.; Yang, L.

    1978-08-01

    The available data on low-enriched uranium (LEU) fuel particles were reviewed under the United States-Federal Republic of Germany Agreement. The most influential factors controlling the irradiation performance of LEU fuel particles were found to be plutonium transport, fission product transport, fuel particle mechanical performance, and fuel particle chemical performance.

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

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 6 2010-10-01 2010-10-01 false Determination of fuel economy values and average fuel economy. 537.9 Section 537.9 Transportation Other Regulations Relating to Transportation... ECONOMY REPORTS § 537.9 Determination of fuel economy values and average fuel economy. (a)...

  7. Cycle Average Peak Fuel Temperature Prediction Using CAPP/GAMMA+

    Energy Technology Data Exchange (ETDEWEB)

    Tak, Nam-il; Lee, Hyun Chul; Lim, Hong Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In order to obtain a cycle average maximum fuel temperature without rigorous efforts, a neutronics/thermo-fluid coupled calculation is needed with depletion capability. Recently, a CAPP/GAMMA+ coupled code system has been developed and the initial core of PMR200 was analyzed using the CAPP/GAMMA+ code system. The GAMMA+ code is a system thermo-fluid analysis code and the CAPP code is a neutronics code. The General Atomics proposed that the design limit of the fuel temperature under normal operating conditions should be a cycle-averaged maximum value. Nonetheless, the existing works of Korea Atomic Energy Research Institute (KAERI) only calculated the maximum fuel temperature at a fixed time point, e.g., the beginning of cycle (BOC) just because the calculation capability was not ready for a cycle average value. In this work, a cycle average maximum fuel temperature has been calculated using CAPP/GAMMA+ code system for the equilibrium core of PMR200. The CAPP/GAMMA+ coupled calculation was carried out for the equilibrium core of PMR 200 from BOC to EOC to obtain a cycle average peak fuel temperature. The peak fuel temperature was predicted to be 1372 .deg. C near MOC. However, the cycle average peak fuel temperature was calculated as 1181 .deg. C, which is below the design target of 1250 .deg. C.

  8. Yalina booster subcritical assembly performance with low enriched uranium fuel

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, Alberto; Gohar, Yousry, E-mail: alby@anl.gov [Argonne National Laboratory, Lemont, IL (United States)

    2011-07-01

    The YALINA Booster facility is a subcritical assembly located in Minsk, Belarus. The facility has special features that result in fast and thermal neutron spectra in different zones. The fast zone of the assembly uses a lead matrix and uranium fuels with different enrichments: 90% and 36%, 36%, or 21%. The thermal zone of the assembly contains 10% enriched uranium fuel in a polyethylene matrix. This study discusses the performance of the three YALINA Booster configurations with the different fuel enrichments. In order to maintain the same subcriticality level in the three configurations, the number of fuel rods in the thermal zone is increased as the uranium fuel enrichment in the fast zone is decreased. The maximum number of fuel rods that can be loaded in the thermal zone is about 1185. Consequently, the neutron multiplication of the configuration with 21% enriched uranium fuel in the fast zone is enhanced by changing the position of the boron carbide and the natural uranium absorber rods, located between the fast and the thermal zones, to form an annular rather than a square arrangement. (author)

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

    Science.gov (United States)

    2010-07-01

    ... fuel as determined in § 600.113-08(a) and (b); FEpet is the fuel economy while operated on petroleum... 40 Protection of Environment 29 2010-07-01 2010-07-01 false Calculation of average fuel economy... ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) ENERGY POLICY FUEL ECONOMY AND CARBON-RELATED EXHAUST EMISSIONS...

  10. 40 CFR 600.510-08 - Calculation of average fuel economy.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 29 2010-07-01 2010-07-01 false Calculation of average fuel economy...) ENERGY POLICY FUEL ECONOMY AND CARBON-RELATED EXHAUST EMISSIONS OF MOTOR VEHICLES Fuel Economy... Trucks and Passenger Automobiles)-Procedures for Determining Manufacturer's Average Fuel Economy §...

  11. 40 CFR 600.510-86 - Calculation of average fuel economy.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 29 2010-07-01 2010-07-01 false Calculation of average fuel economy...) ENERGY POLICY FUEL ECONOMY AND CARBON-RELATED EXHAUST EMISSIONS OF MOTOR VEHICLES Fuel Economy... Trucks and Passenger Automobiles)-Procedures for Determining Manufacturer's Average Fuel Economy §...

  12. 40 CFR 600.510-93 - Calculation of average fuel economy.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 29 2010-07-01 2010-07-01 false Calculation of average fuel economy...) ENERGY POLICY FUEL ECONOMY AND CARBON-RELATED EXHAUST EMISSIONS OF MOTOR VEHICLES Fuel Economy... Trucks and Passenger Automobiles)-Procedures for Determining Manufacturer's Average Fuel Economy §...

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  14. Highest average burnups achieved by MTR fuel elements of the IEA-R1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Damy, Margaret A.; Terremoto, Luis A.A.; Silva, Jose E.R.; Silva, Antonio Teixeira e; Castanheira, Myrthes; Teodoro, Celso A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear (CEN)]. E-mail: madamy@ipen.br

    2007-07-01

    Different nuclear fuels were employed in the manufacture of plate type at IPEN , usually designated as Material Testing Reactor (MTR) fuel elements. These fuel elements were used at the IEA-R1 research reactor. This work describes the main characteristics of these nuclear fuels, emphasizing the highest average burn up achieved by these fuel elements. (author)

  15. 49 CFR 525.11 - Termination of exemption; amendment of alternative average fuel economy standard.

    Science.gov (United States)

    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...

  16. Analysis of the performance of fuel cells PWR with a single enrichment and radial distribution of enrichments; Analisis del desempeno de celdas combustibles PWR con un solo enriquecimiento y con distribucion radial de enriquecimientos

    Energy Technology Data Exchange (ETDEWEB)

    Vargas, S.; Gonzalez, J. A.; Alonso, G.; Del Valle, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. Lindavista, Mexico D.F. 07738 (Mexico); Xolocostli M, J. V. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: nolosesamuel@prodigy.net.mx

    2008-07-01

    One of the main challenges in the design of fuel assemblies is the efficient use of uranium achieving burnt homogeneous of the fuel rods as well as the burnt maximum possible of the same ones to the unload. In the case of the assemblies type PWR has been decided actually for fuel assemblies with a single radial enrichment. The present work has like effect to show the because of this decision, reason why a comparison of the neutronic performance of two fuel cells takes place with the same enrichment average but one of them with radial distribution of enrichment and the other with a single enrichment equal to the average. The results shown in the present study of the behavior of the neutron flow as well as the power distribution through of assembly sustain the because of a single radial enrichment. (Author)

  17. 77 FR 18272 - Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services LLC...

    Science.gov (United States)

    2012-03-27

    ... COMMISSION Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services LLC...) staff has conducted inspections of the Louisiana Energy Services (LES), LLC, National enrichment... . SUPPLEMENTARY INFORMATION: I. Discussion The NRC staff has conducted inspections of the Louisiana...

  18. 41 CFR 102-34.55 - Are there fleet average fuel economy standards we must meet?

    Science.gov (United States)

    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...

  19. Analysis of the performance of fuel cells BWR with a single enrichment and radial distribution of enrichments; Analisis del desempeno de celdas combustibles BWR con un solo enriquecimiento y con distribucion radial de enriquecimientos

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez, J. A.; Vargas, S.; Alonso, G.; Del Valle, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. Lindavista, Mexico D. F. 07738 (Mexico); Xolocostli M, J. V. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: govaj666@hotmail.com

    2008-07-01

    The efficient use of the fuel is one of the objectives in the assemblies design of type BWR. The present tendency in the assemblies design of type BWR is through a radial distribution of enrichments. The present work has like object showing the because of this decision, for what a comparison of the neutronic performance of two fuel cells with the same enrichment average but one of them with radial distribution of enrichment and the other with a single enrichment equal to the average. The cells were analyzed with the CASMO-4 code and the obtained results of the behavior of the neutron flow and the power sustain the because of the radial distribution of enrichments. (Author)

  20. Reactivity insertion transient analysis for KUR low-enriched uranium silicide fuel core

    OpenAIRE

    Shen, Xiuzhong; Nakajima, Ken; Unesaki, Hironobu; Mishima, Kaichiro

    2013-01-01

    The purpose of this study is to realize the full core conversion from the use of High Enriched Uranium (HEU) fuels to the use of Low Enriched Uranium (LEU) fuels in Kyoto University Research Reactor (KUR). Although the conversion of nuclear energy sources is required to keep the safety margins and reactor reliability based on KUR HEU core, the uranium density (3.2 gU/cm3) and enrichment (20%) of LEU fuel (U3Si2–AL) are quite different from the uranium density (0.58 gU/cm3) and enrichment (93%...

  1. Criticality issues with highly enriched fuels in a repository environment

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, L.L. [Lockheed Martin Idaho Technologies Co., Idaho Falls, ID (United States); Sanchez, L.C.; Rath, J.S. [Sandia National Labs., Albuquerque, NM (United States)

    1998-03-01

    This paper presents preliminary analysis of a volcanic tuff repository containing a combination of low enrichment commercial spent nuclear fuels (SNF) and DOE-owned SNF packages. These SNFs were analyzed with respect to their criticality risks. Disposal of SNF packages containing significant fissile mass within a geologic repository must comply with current regulations relative to criticality safety during transportation and handling within operational facilities. However, once the repository is closed, the double contingency credits for criticality safety are subject to unremediable degradation, (e.g., water intrusion, continued presence of neutron absorbers in proximity to fissile material, and fissile material reconfiguration). The work presented in this paper focused on two attributes of criticality in a volcanic tuff repository for near-field and far-field scenarios: (1) scenario conditions necessary to have a criticality, and (2) consequences of a nuclear excursion that are components of risk. All criticality consequences are dependent upon eventual water intrusion into the repository and subsequent breach of the disposal package. Key criticality parameters necessary for a critical assembly are: (1) adequate thermal fissile mass, (2) adequate concentration of fissile material, (3) separation of neutron poison from fissile materials, and (4) sufficient neutron moderation (expressed in units of moderator to fissile atom ratios). Key results from this study indicated that the total energies released during a single excursion are minimal (comparable to those released in previous solution accidents), and the maximum frequency of occurrence is bounded by the saturation and temperature recycle times, thus resulting in small criticality risks.

  2. 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 %.

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  4. Optimization of fuel rod enrichment distribution to minimize rod power peaking throughout life within BWR fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Hirano, Yasushi; Hida, Kazuki; Sakurada, Koichi; Yamamoto, Munenari [Toshiba Corp., Kawasaki, Kanagawa (Japan). Nuclear Engineering Lab.

    1997-01-01

    A practical method was developed for determining the optimum fuel enrichment distribution within a boiling water reactor fuel assembly. The method deals with two different optimization problems, i.e. a combinatorial optimization problem grouping fuel rods into a given number of rod groups with the same enrichment, and a problem determining an optimal enrichment for each fuel rod under the resultant rod-grouping pattern. In solving these problems, the primary goal is to minimize a predefined objective function over a given exposure period. The objective function used here is defined by a linear combination: C{sub 1}X+C{sub 2}X{sub G}, where X and X{sub G} stand for a control variable to give the constraint respectively for a local power peaking factor and a gadolinium rod power, and C{sub 1} and C{sub 2} are user-definable weighting factor to accommodate the design preference. The algorithm of solving the combinatorial optimization problem starts with finding the optimal enrichment vector without any rod-grouping, and promising candidates of rod-grouping patterns are found by exhaustive enumeration based on the resulting fuel enrichment ordering, and then the latter problem is solved by using the method of approximation programming. The practical application of the present method is shown for a contemporary 8x8 Pu mixed-oxide fuel assembly with 10 gadolinium-poisoned rods. (author)

  5. 77 FR 65729 - Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services LLC...

    Science.gov (United States)

    2012-10-30

    ... COMMISSION Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services LLC... inspections of the Louisiana Energy Services (LES), LLC, National Enrichment Facility in Eunice, New Mexico... Title 10 of the Code of Federal Regulations (10 CFR) 70.32 (k) and section 193(c) of the Atomic...

  6. 78 FR 23312 - Uranium Enrichment Fuel Cycle Inspection Reports Regarding Louisiana Energy Services, National...

    Science.gov (United States)

    2013-04-18

    ... COMMISSION Uranium Enrichment Fuel Cycle Inspection Reports Regarding Louisiana Energy Services, National... Energy Services (LES), LLC, National Enrichment Facility in Eunice, New Mexico, and has authorized the... Energy Act of 1954, as amended. The introduction of UF 6 into any module of the National...

  7. 78 FR 63518 - Uranium Enrichment Fuel Cycle Inspection Reports Regarding Louisiana Energy Services, National...

    Science.gov (United States)

    2013-10-24

    ... COMMISSION Uranium Enrichment Fuel Cycle Inspection Reports Regarding Louisiana Energy Services, National... conducted inspections of the Louisiana Energy Services (LES), LLC, National Enrichment Facility in Eunice... the Atomic Energy Act of 1954, as amended. The introduction of UF 6 into any module of the...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-10-01

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

  9. Simulation studies of diesel engine performance with oxygen enriched air and water emulsified fuels

    Energy Technology Data Exchange (ETDEWEB)

    Assanis, D.N.; Baker, D. (Illinois Univ., Urbana, IL (USA)); Sekar, R.R.; Siambekos, C.T.; Cole, R.L.; Marciniak, T.J. (Argonne National Lab., IL (USA))

    1990-01-01

    A computer simulation code of a turbocharged, turbocompound diesel engine was modified to study the effects of using oxygen-enriched combustion air and water-emulsified diesel fuels. Oxygen levels of 21 percent to 40 percent by volume in the combustion air were studied. Water content in the fuel was varied from 0 percent to 50 percent mass. Simulation studies and a review and analysis of previous work in this area led to the following conclusions about expected engine performance and emissions: the power density of the engine is significantly increased by oxygen enrichment. Ignition delay and particulate emissions are reduced. Combustion temperatures and No{sub x} emissions are increased with oxygen enrichment but could be brought back to the base levels by introducing water in the fuel. The peak cylinder pressure which increases with the power output level might result in mechanical problems with engine components. Oxygen enrichment also provides an opportunity to use cheaper fuel such as No. 6 diesel fuel. Overall, the adverse effects of oxygen enrichment could be countered by the addition of water and it appears that an optimum combination of water content, oxygen level, and base diesel fuel quality may exist. This could yield improved performance and emissions characteristics compared to a state-of-the-art diesel engine. 9 refs., 8 figs.

  10. Moderator configuration options for a low-enriched uranium fueled Kilowatt-class Space Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    King, Jeffrey C., E-mail: kingjc@mines.edu [Nuclear Science and Engineering Program, Colorado School of Mines (CSM), Golden, CO (United States); Mencarini, Leonardo de Holanda; Guimaraes, Lamartine N. F., E-mail: guimaraes@ieav.cta.br, E-mail: mencarini@ieav.cta.br [Instituto de Estudos Avancados (IEAV), Sao Jose dos Campos, SP (Brazil). Divisao de Energia Nuclear

    2015-07-01

    The Brazilian Air Force, through its Institute for Advanced Studies (Instituto de Estudos Avancados, IEAv/DCTA), and the Colorado School of Mines (CSM) are studying the feasibility of a space nuclear reactor with a power of 1-5 kW{sub e} and fueled with Low-Enriched Uranium (LEU). This type of nuclear reactor would be attractive to signatory countries of the Non-Proliferation Treaty (NPT) or commercial interests. A LEU-fueled space reactor would avoid the security concerns inherent with Highly Enriched Uranium (HEU) fuel. As an initial step, the HEU-fueled Kilowatt Reactor Using Stirling Technology (KRUSTY) designed by the Los Alamos National Laboratory serves as a basis for a similar reactor fueled with LEU fuel. Using the computational code MCNP6 to predict the reactor neutronics performance, the size of the resulting reactor fueled with 19.75 wt% enriched uranium-10 wt% molybdenum alloy fuel is adjusted to match the excess reactivity of KRUSTY. Then, zirconium hydride moderator is added to the core to reduce the size of the reactor. This work presents the preliminary results of the computational modeling, with special emphasis on the comparison between homogeneous and heterogeneous moderator systems, in terms of the core diameter required to meet a specific multiplication factor (k{sub eff} = 1.035). This comparison illustrates the impact of moderator configuration on the size and performance of a LEU-fueled kilowatt-class space nuclear reactor. (author)

  11. 75 FR 25323 - Light-Duty Vehicle Greenhouse Gas Emission Standards and Corporate Average Fuel Economy Standards...

    Science.gov (United States)

    2010-05-07

    ... and Corporate Average Fuel Economy Standards; Final Rule #0;#0;Federal Register / Vol. 75, No. 88... Standards and Corporate Average Fuel Economy Standards; Final Rule AGENCY: Environmental Protection Agency... light-duty vehicles that will reduce greenhouse gas emissions and improve fuel economy. This joint...

  12. Enrichment measurement in TRIGA type fuels; Medicion de enriquecimiento en combustibles tipo Triga

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar H, F.; Mazon R, R. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2001-05-15

    The Department of Energy of the United States of North America, through the program 'Idaho Operations Nuclear Spent Fuel Program' of the Idaho National Engineering and Environmental Laboratory (INEEL), in Idaho Falls; Idaho USA, hires to Global Technologies Inc. (GTI) to develop a prototype device of detection enrichment uranium (DEU Detection of Enrichment of Uranium) to determine quantitatively the enrichment in remainder U-235 in a TRIGA fuel element at the end of it useful life. The characteristics of the prototype developed by GTI are the following ones: It allows to carry out no-destructive measurements of TRIGA type fuel. Easily transportable due to that reduced of it size. The determination of the enrichment (in grams of U-235) it is obtained with a precision of 5%. The National Institute of Nuclear Research (ININ), in its facilities of the Nuclear Center of Mexico, it has TRIGA type fuel of high and low enrichment (standard and FLIP) fresh and with burnt, it also has the infrastructure (hot cells, armor-plating of transport, etc) and qualified personnel to carry out the necessary maneuvers to prove the operation of the DEU prototype. For this its would be used standard type fuel elements and FLIP, so much fresh as with certain burnt one. In the case of the fresh fuels the measurement doesn't represent any risk, the fuels before and after the measurement its don't contain a quantity of fission products that its represent a radiological risk in its manipulation; but in the case of the fuels with burnt the handling of the same ones represents an important radiological risk reason why for its manipulation it was used the transport armor-plating and the hot cells. (Author)

  13. In-Core Fuel Managements for PWRs: Investigation on solution for optimal utilization of PWR fuel through the use of fuel assemblies with differently enriched {sup 235}U fuel pins

    Energy Technology Data Exchange (ETDEWEB)

    Caprioli, Sara

    2004-04-01

    A possibility for more efficient use of the nuclear fuel in a pressurized water reactor is investigated. The alternative proposed here consists of the implementation of PWR fuel assemblies with differently enriched {sup 235}U fuel pins. This possibility is examined in comparison with the standard assembly design. The comparison is performed both in terms of single assembly performance and in the terms of nuclear reactor core performance and fuel utility. For the evaluation of the actual performance of the new assembly types, 5 operated fuel core sequences of R3 (Ringhals' third unit), for the period 1999 - 2004 (cycles 17 - 21) were examined. For every cycle, the standard fresh fuel assemblies have been identified and taken as reference cases for the study of the new type of assemblies with differently enriched uranium rods. In every cycle, assemblies with and without burnable absorber are freshly loaded into the core. The axial enrichment distribution is kept uniform, allowing for a radial (planar) enrichment level distribution only. At an assembly level, it has been observed that the implementation of the alternative enrichment configuration can lead to lower and flatter internal peaking factor distribution with respect to the uniformly enriched reference assemblies. This can be achieved by limiting the enrichment levels distribution to a rather narrow range. The highest enrichment level chosen has the greatest impact on the power distribution of the assemblies. As it increases, the enrichment level drives the internal peaking factor to greater values than in the reference assemblies. Generally, the highest enrichment level that would allow an improvement in the power performance of the assembly lies between 3.95 w/o and 4.17 w/o. The highest possible enrichment level depends on the average enrichment of the overall assembly, which is kept constant to the average enrichment of the reference assemblies. The improvements that can be obtained at this level are

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  15. Transport of high enriched uranium fresh fuel from Yugoslavia to the Russian federation

    Directory of Open Access Journals (Sweden)

    Pešić Milan P.

    2002-01-01

    Full Text Available This paper presents the relevant data related to the recent shipment (August 2002 of fresh highly enriched uranium fuel elements from Yugoslavia back to the Russian Federation for uranium down blending. In this way, Yugoslavia gave its contribution to the Reduced Enrichment for Research and Test Reactors (RERTR Program and to the world's joint efforts to prevent possible terrorist actions against nuclear material potentially usable for the production of nuclear weapons.

  16. Verification of 235U enrichment of fresh VVER-440 fuel assemblies.

    Science.gov (United States)

    Almási, I; Nguyen, C T; Zsigrai, J; Lakosi, L; Hlavathy, Z; Nagy, P; Buglyó, N

    2012-10-01

    Enrichment of uniformly and non-uniformly enriched ("profiled") fuel assemblies in a range of 1.6-4.4% was verified by gamma-ray spectrometry at a nuclear power plant (NPP). HPGe detectors and a CdZnTe (CZT) detector, the latter fitting into the central tube of the assemblies, were used for obtaining information from outer and inner fuel rods. A procedure which has minimal impact on the NPP work was developed for verifying freshly arrived assemblies under normal operational conditions, and is now in routine use.

  17. Study of Reduced-Enrichment Uranium Fuel Possibility for Research Reactors

    Directory of Open Access Journals (Sweden)

    Ruppel V.A.

    2015-01-01

    Full Text Available Having analyzed the results obtained in the work, it is possible to conclude that the flux density of fast and thermal neutrons in the shell of fuel elements in EFA in REU-zone decreased on average by 5% for UO2 fuel and by 7% for U9%Mo fuel. Change of neutrons flux density during the cycle does not exceed 4% for both fuel types. On average the fuel burnup in reactor core during the cycle for UO2 and U9%Mo increased by 2.8%. It is 1% less that in HEU-zone, which is conditioned by higher initial loading of 235U in fuel assembly with REU fuel.

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  19. In-Core Fuel Managements for PWRs: Investigation on solution for optimal utilization of PWR fuel through the use of fuel assemblies with differently enriched {sup 235}U fuel pins

    Energy Technology Data Exchange (ETDEWEB)

    Caprioli, Sara

    2004-04-01

    A possibility for more efficient use of the nuclear fuel in a pressurized water reactor is investigated. The alternative proposed here consists of the implementation of PWR fuel assemblies with differently enriched {sup 235}U fuel pins. This possibility is examined in comparison with the standard assembly design. The comparison is performed both in terms of single assembly performance and in the terms of nuclear reactor core performance and fuel utility. For the evaluation of the actual performance of the new assembly types, 5 operated fuel core sequences of R3 (Ringhals' third unit), for the period 1999 - 2004 (cycles 17 - 21) were examined. For every cycle, the standard fresh fuel assemblies have been identified and taken as reference cases for the study of the new type of assemblies with differently enriched uranium rods. In every cycle, assemblies with and without burnable absorber are freshly loaded into the core. The axial enrichment distribution is kept uniform, allowing for a radial (planar) enrichment level distribution only. At an assembly level, it has been observed that the implementation of the alternative enrichment configuration can lead to lower and flatter internal peaking factor distribution with respect to the uniformly enriched reference assemblies. This can be achieved by limiting the enrichment levels distribution to a rather narrow range. The highest enrichment level chosen has the greatest impact on the power distribution of the assemblies. As it increases, the enrichment level drives the internal peaking factor to greater values than in the reference assemblies. Generally, the highest enrichment level that would allow an improvement in the power performance of the assembly lies between 3.95 w/o and 4.17 w/o. The highest possible enrichment level depends on the average enrichment of the overall assembly, which is kept constant to the average enrichment of the reference assemblies. The improvements that can be obtained at this level are

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

    Energy Technology Data Exchange (ETDEWEB)

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

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    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

  2. Fuel handling accident analysis for the University of Missouri Research Reactor's High Enriched Uranium to Low Enriched Uranium fuel conversion initiative

    Science.gov (United States)

    Rickman, Benjamin

    In accordance with the 1986 amendment concerning licenses for research and test reactors, the MU Research Reactor (MURR) is planning to convert from using High-Enriched Uranium (HEU) fuel to the use of Low-Enriched Uranium (LEU) fuel. Since the approval of a new LEU fuel that could meet the MURR's performance demands, the next phase of action for the fuel conversion process is to create a new Safety Analysis Report (SAR) with respect to the LEU fuel. A component of the SAR includes the Maximum Hypothetical Accident (MHA) and accidents that qualify under the class of Fuel Handling Accidents (FHA). In this work, the dose to occupational staff at the MURR is calculated for the FHAs. The radionuclide inventory for the proposed LEU fuel was calculated using the ORIGEN2 point-depletion code linked to the MURR neutron spectrum. The MURR spectrum was generated from a Monte Carlo Neutron transPort (MCNP) simulation. The coupling of these codes create MONTEBURNS, a time-dependent burnup code. The release fraction from each FHA within this analysis was established by the methodology of the 2006 HEU SAR, which was accepted by the NRC. The actual dose methodology was not recorded in the HEU SAR, so a conservative path was chosen. In compliance to NUREG 1537, when new methodology is used in a HEU to LEU analysis, it is necessary to re-evaluate the HEU accident. The Total Effective Dose Equivalent (TEDE) values were calculated in addition to the whole body dose and thyroid dose to operation personnel. The LEU FHA occupational TEDE dose was 349 mrem which is under the NRC regulatory occupational dose limit of 5 rem TEDE, and under the LEU MHA limit of 403 mrem. The re-evaluated HEU FHA occupational TEDE dose was 235 mrem, which is above the HEU MHA TEDE dose of 132 mrem. Since the new methodology produces a dose that is larger than the HEU MHA, we can safely assume that it is more conservative than the previous, unspecified dose.

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

    Energy Technology Data Exchange (ETDEWEB)

    Montierth, Leland M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-07-19

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Pope, M. A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); DeHart, M. D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Morrell, S. R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Jamison, R. K. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nef, E. C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nigg, D. W. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Montierth, Leland M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-07-19

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

  6. Progress in qualifying low-enriched U-Mo dispersion fuels

    Energy Technology Data Exchange (ETDEWEB)

    Snelgrove, J.L.; Hofman, G.L.; Hayes, S.L.; Meyer, M.K. [Argonne National Laboratory, Argonne, IL (United States)

    2001-07-01

    The U.S. Reduced Enrichment for Research and Test Reactors program is working to qualify dispersions of U-Mo alloys in aluminum with fuel-meat densities of 8 to 9 gU cm{sup -3}. Post irradiation examinations of the small fuel plates irradiated in the Advanced Test Reactor during the high-temperature RERTR-3 tests are virtually complete, and analysis of the large quantity of data obtained is underway. We have observed that the swelling of the fuel plates is stable and modest and that the swelling is dominated by the temperature-dependent interaction of the U-Mo fuel and the aluminum matrix. In order to extract detailed information about the behavior of these fuels from the data, a complex fuel-plate thermal model is being developed to account for the effects of the changing fission rate and thermal conductivity of the fuel meat during irradiation. This paper summarizes the empirical results of the post irradiation examinations and the preliminary results of the model development. In addition, the schedule for irradiation of full-sized elements in the HFR-Petten is briefly discussed. (author)

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

    Directory of Open Access Journals (Sweden)

    Shun'ichi Ishii

    Full Text Available 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

  8. 77 FR 64051 - 2017 and Later Model Year Light-Duty Vehicle Greenhouse Gas Emissions and Corporate Average Fuel...

    Science.gov (United States)

    2012-10-18

    ... Model Year Light-Duty Vehicle Greenhouse Gas Emissions and Corporate Average Fuel Economy Standards... standards to improve fuel economy and reduce greenhouse gas emissions for vehicles manufactured for sale in... and address global climate change. Need for Correction As published, the final...

  9. Criticality Calculations for a Typical Nuclear Fuel Fabrication Plant with Low Enriched Uranium

    Energy Technology Data Exchange (ETDEWEB)

    Elsayed, Hade; Nagy, Mohamed; Agamy, Said; Shaat, Mohmaed [Egyptian Atomic Energy Authority, Cairo (Egypt)

    2013-07-01

    The operations with the fissile materials such as U{sup 235} introduce the risk of a criticality accident that may be lethal to nearby personnel and can lead the facility to shutdown. Therefore, the prevention of a nuclear criticality accident should play a major role in the design of a nuclear facility. The objectives of criticality safety are to prevent a self-sustained nuclear chain reaction and to minimize the consequences. Sixty criticality accidents were occurred in the world. These are accidents divided into two categories, 22 accidents occurred in process facilities and 38 accidents occurred during critical experiments or operations with research reactor. About 21 criticality accidents including Japan Nuclear Fuel Conversion Co. (JCO) accident took place with fuel solution or slurry and only one accident occurred with metal fuel. In this study the nuclear criticality calculations have been performed for a typical nuclear fuel fabrication plant producing nuclear fuel elements for nuclear research reactors with low enriched uranium up to 20%. The calculations were performed for both normal and abnormal operation conditions. The effective multiplication factor (k{sub eff}) during the nuclear fuel fabrication process (Uranium hexafluoride - Ammonium Diuranate conversion process) was determined. Several accident scenarios were postulated and the criticalities of these accidents were evaluated. The computer code MCNP-4B which based on Monte Carlo method was used to calculate neutron multiplication factor. The criticality calculations Monte Carlo method was used to calculate neutron multiplication factor. The criticality calculations were performed for the cases of, change of moderator to fuel ratio, solution density and concentration of the solute in order to prevent or mitigate criticality accidents during the nuclear fuel fabrication process. The calculation results are analyzed and discussed.

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

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Germina [ORNL; Primm, Trent [ORNL

    2011-05-01

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

  11. Calculation of parameters for inspection planning and evaluation: low enriched uranium conversion and fuel fabrication facilities

    Energy Technology Data Exchange (ETDEWEB)

    Reardon, P.T.; Mullen, M.F.; Harms, N.L.

    1981-02-01

    As part of Task C.35 (Calculation of Parameters for Inspection Planning and Evaluation) of the US Program of Technical Assistance to IAEA Safeguards, Pacific Northwest Laboratory has performed some quantitative analyses of IAEA inspection activities at low-enriched uranium (LEU) conversion and fuel fabrication facilities. This report presents the results and conclusions of those analyses. Implementation of IAEA safeguards at LEU conversion and fuel fabrication facilities must take into account a variety of practical problems and constraints. One of the key concerns is the problem of flow verification, especially product verification. The objective of this report is to help put the problem of flow verification in perspective by presenting the results of some specific calculations of inspection effort and probability of detection for various product measurement strategies. In order to provide quantitative information about the advantages and disadvantages of the various strategies, eight specific cases were examined.

  12. Fusion solution to dispose of spent nuclear fuel, transuranic elements, and highly enriched uranium

    Energy Technology Data Exchange (ETDEWEB)

    Gohar, Yousry E-mail: gohar@anl.gov

    2001-11-01

    The disposal of the nuclear spent fuel, the transuranic elements, and the highly enriched uranium represents a major problem under investigation by the international scientific community to identify the most promising solutions. The investigation of this paper focused on achieving the top rated solution for the problem, the elimination goal, which requires complete elimination for the transuranic elements or the highly enriched uranium, and the long-lived fission products. To achieve this goal, fusion blankets with liquid carrier, molten salts or liquid metal eutectics, for the transuranic elements and the uranium isotopes are utilized. The generated energy from the fusion blankets is used to provide revenue for the system. The long-lived fission products are fabricated into fission product targets for transmutation utilizing the neutron leakage from the fusion blankets. This paper investigated the fusion blanket designs for small fusion devices and the system requirements for such application. The results show that 334 MW of fusion power from D-T plasma for 30 years with an availability factor of 0.75 can dispose of the 70,000 tons of the U.S. inventory of spent nuclear fuel generated up to the year 2015. In addition, this fusion solution eliminates the need for a geological repository site, which is a major advantage. Meanwhile, such utilization of the fusion power will provide an excellent opportunity to develop fusion energy for the future.

  13. Optimization of enrichment distributions in nuclear fuel assemblies loaded with uranium and plutonium via a modified linear programming technique

    Science.gov (United States)

    Cuevas Vivas, Gabriel Francisco

    A methodology to optimize enrichment distributions in Light Water Reactor (LWR) fuel assemblies is developed and tested. The optimization technique employed is the linear programming revised simplex method, and the fuel assembly's performance is evaluated with a neutron transport code that is also utilized in the calculation of sensitivity coefficients. The enrichment distribution optimization procedure begins from a single-value (flat) enrichment distribution until a target, maximum local power peaking factor, is achieved. The optimum rod enrichment distribution, with 1.00 for the maximum local power peaking factor and with each rod having its own enrichment, is calculated at an intermediate stage of the analysis. Later, the best locations and values for a reduced number of rod enrichments is obtained as a function of a target maximum local power peaking factor by applying sensitivity to change techniques. Finally, a shuffling process that assigns individual rod enrichments among the enrichment groups is performed. The relative rod power distribution is then slightly modified and the rod grouping redefined until the optimum configuration is attained. To verify the accuracy of the relative rod power distribution, a full computation with the neutron transport code using the optimum enrichment distribution is carried out. The results are compared and tested for assembly designs loaded with fresh Low Enriched Uranium (LEU) and plutonium Mixed OXide (MOX) fuels. MOX isotopics for both reactor-grade and weapons-grade plutonium were utilized to demonstrate the wide-range of applicability of the optimization technique. The features of the assembly designs used for evaluation purposes included burnable absorbers and internal water regions, and were prepared to resemble the configurations of modern assemblies utilized in commercial Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). In some cases, a net improvement in the relative rod power distribution or

  14. 76 FR 26996 - Notice of Intent To Prepare an Environmental Impact Statement for New Corporate Average Fuel...

    Science.gov (United States)

    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... Environmental Impact Statement for New Corporate Average Fuel Economy Standards AGENCY: National Highway...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Trellue, Holly Renee [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Fugate, Michael Lynn [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Tobin, Stephen Joesph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    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.

  17. Conceptual design and economic analysis of a light water reactor fuel enricher/regenerator. FY 1978 year-end report

    Energy Technology Data Exchange (ETDEWEB)

    Grand, P; Kouts, H J; Powell, J R; Steinberg, M; Takahashi, H

    1979-05-01

    A study has been performed to evaluate the use of high-energy particle accelerators as nuclear fuel enrichers and nuclear fuel regenerators. This builds on ideas that have been current for many years. The new study has, however, explored some novel approaches that have not been examined before. A specific conceptual system chosen for more detailed study would stretch the energy available from natural uranium by a factor of about 3, reduce the separative work requirements by a factor of about 4, and reduce the volume of spent fuel to be stored by a factor of 2, compared to the current once-through light water reactor (LWR) fuel cycle. The concept avoids the need for chemical reprocessing of spent fuel, and would permit continued use of LWR's beyond the time when limitations on fuel resources might otherwise lead to their being phased out. This concept, which is called the Linear Accelerator Fuel Enricher/Regenerator, is therefore viewed as offering a practical means of stretching the use of the nuclear fuel resource in the framework of the existing light water reactor fuel cycle. This report describes and analyzes the concept referred to. An explanation of the principles underlying the concept is given. Particular attention is devoted to engineering feasibility, proliferation resistance, and economics. It is seen that the concept draws on only proven technology as regards bothaccelerator design and the fuel irradiation process, and is adapted to existing LWR designs with no change except in fuel-handling practices. A preliminary evaluation of radiation damage, coolant options, and power conversion systems is provided. Neutronic, thermal-hydraulic, and burnup calculations are presented. An analysis is made of fuel economy. Approximate costs of electric power produced using this concept are evaluated and discussed. Estimated development costs of commercialization are provided.

  18. Surface strontium enrichment on highly active perovskites for oxygen electrocatalysis in solid oxide fuel cells

    KAUST Repository

    Crumlin, Ethan J.

    2012-01-01

    Perovskite oxides have high catalytic activities for oxygen electrocatalysis competitive to platinum at elevated temperatures. However, little is known about the oxide surface chemistry that influences the activity near ambient oxygen partial pressures, which hampers the design of highly active catalysts for many clean-energy technologies such as solid oxide fuel cells. Using in situ synchrotron-based, ambient pressure X-ray photoelectron spectroscopy to study the surface chemistry changes, we show that the coverage of surface secondary phases on a (001)-oriented La 0.8Sr 0.2CoO 3-δ (LSC) film becomes smaller than that on an LSC powder pellet at elevated temperatures. In addition, strontium (Sr) in the perovskite structure enriches towards the film surface in contrast to the pellet having no detectable changes with increasing temperature. We propose that the ability to reduce surface secondary phases and develop Sr-enriched perovskite surfaces of the LSC film contributes to its enhanced activity for O 2 electrocatalysis relative to LSC powder-based electrodes. © 2012 The Royal Society of Chemistry.

  19. Integrating engineering design improvements with exoelectrogen enrichment process to increase power output from microbial fuel cells

    Science.gov (United States)

    Borole, Abhijeet P.; Hamilton, Choo Y.; Vishnivetskaya, Tatiana A.; Leak, David; Andras, Calin; Morrell-Falvey, Jennifer; Keller, Martin; Davison, Brian

    Microbial fuel cells (MFC) hold promise as a green technology for bioenergy production. The challenge is to improve the engineering design while exploiting the ability of microbes to generate and transfer electrons directly to electrodes. A strategy using a combination of improved anode design and an enrichment process was formulated to improve power densities. The design was based on a flow-through anode with minimal dead volume and a high electrode surface area per unit volume. The strategy focused on promoting biofilm formation via a combination of forced flow through the anode, carbon limitation, and step-wise reduction of external resistance. The enrichment process resulted in development of exoelectrogenic biofilm communities dominated by Anaeromusa spp. This is the first report identifying organisms from the Veillonellaceae family in MFCs. The power density of the resulting MFC using a ferricyanide cathode reached 300 W m -3 net anode volume (3220 mW m -2), which is about a third of what is estimated to be necessary for commercial consideration. The operational stability of the MFC using high specific surface area electrodes was demonstrated by operating the MFC for a period of over four months.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  1. Enrichment of microbial electrolysis cell biocathodes from sediment microbial fuel cell bioanodes.

    Science.gov (United States)

    Pisciotta, John M; Zaybak, Zehra; Call, Douglas F; Nam, Joo-Youn; Logan, Bruce E

    2012-08-01

    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. CO(2) 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  4. Effects of inoculation sources on the enrichment and performance of anode bacterial consortia in sensor typed microbial fuel cells

    Directory of Open Access Journals (Sweden)

    Phuong Tran

    2016-01-01

    Full Text Available Microbial fuel cells are a recently emerging technology that promises a number of applications in energy recovery, environmental treatment and monitoring. In this study, we investigated the effect of inoculating sources on the enrichment of electrochemically active bacterial consortia in sensor-typed microbial fuel cells (MFCs. Several MFCs were constructed, operated with modified artificial wastewater and inoculated with different microbial sources from natural soil, natural mud, activated sludge, wastewater and a mixture of those sources. After enrichment, the MFCs inoculated with the natural soil source generated higher and more stable currents (0.53±0.03 mA, in comparisons with the MFCs inoculated with the other sources. The results from denaturing gradient gel electrophoresis (DGGE showed that there were significant changes in bacterial composition from the original inocula to the enriched consortia. Even more interestingly, Pseudomonas sp. was found dominant in the natural soil source and also in the corresponding enriched consortium. The interactions between Pseudomonas sp. and other species in such a community are probably the key for the effective and stable performance of the MFCs.

  5. 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.

    Science.gov (United States)

    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.

  6. Fuel enrichment and temperature distribution in nuclear fuel rod in (D-T) driven hybrid reactor system

    Energy Technology Data Exchange (ETDEWEB)

    Osman, Ypek [Suleyman Demirel Universitesi Muhendislik-Mimarlyk Fakultesi, Isparta (Turkey)

    2001-07-01

    In this study, melting point of the fuel rod and temperature distribution in nuclear fuel rod are investigated for different coolants under various first wall loads (P{sub w}, =5, 6, 7, 8, 9, and 10 MWm{sup -2}) in Fusion-Fission reactor fueled with 50%LWR +50%CANDU. The fusion source of neutrons of 14.1 MeV is simulated by a movable target along the main axis of cylindrical geometry as a line source. In addition, the fusion chamber was thought as a cylindrical cavity with a diameter of 300 cm that is comparatively small value. The fissile fuel zone is considered to be cooled with four different coolants, gas, flibe (Li{sub 2}BeF{sub 4}), natural lithium (Li), and eutectic lithium (Li{sub 17}Pb{sub 83}). Investigations are observed during 4 years for discrete time intervals of{delta}t= 0.5 month and by a plant factor (PF) of 75%. Volumetric ratio of coolant-to fuel is 1:1, 45.515% coolant, 45.515% fuel, 8.971% clad, in fuel zone. (author)

  7. Optimization of enrichment distributions in nuclear fuel assemblies loaded with Uranium and Plutonium via a modified linear programming technique

    Energy Technology Data Exchange (ETDEWEB)

    Cuevas Vivas, Gabriel Francisco

    1999-12-01

    A methodology to optimize enrichment distributions in Light Water Reactor (LWR) fuel assemblies is developed and tested. The optimization technique employed is the linear programming revised simplex method, and the fuel assembly's performance is evaluated with a neutron transport code that is also utilized in the calculation of sensitivity coefficients. The enrichment distribution optimization procedure begins from a single-value (flat) enrichment distribution until a target, maximum local power peaking factor, is achieved. The optimum rod enrichment distribution, with 1.00 for the maximum local power peaking factor and with each rod having its own enrichment, is calculated at an intermediate stage of the analysis. Later, the best locations and values for a reduced number of rod enrichments is obtained as a function of a target maximum local power peaking factor by applying sensitivity to change techniques. Finally, a shuffling process that assigns individual rod enrichments among the enrichment groups is performed. The relative rod power distribution is then slightly modified and the rod grouping redefined until the optimum configuration is attained. To verify the accuracy of the relative rod power distribution, a full computation with the neutron transport code using the optimum enrichment distribution is carried out. The results are compared and tested for assembly designs loaded with fresh Low Enriched Uranium (LEU) and plutonium Mixed Oxide (MOX) isotopics for both reactor-grade and weapons-grade plutonium were utilized to demonstrate the wide range of applicability of the optimization technique. The feature of the assembly designs used for evaluation purposes included burnable absorbers and internal water regions, and were prepared to resemble the configurations of modern assemblies utilized in commercial Boiling Water Reactor (BWRs) and Pressurized Water Reactors (PWRs). In some cases, a net improvement in the relative rod power distribution or in the

  8. Partially Enriched U235 for Use as Fuel in Off-Site Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Larson, C. E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    1953-08-10

    This is in reply to the memorandum of July 21, 1953, from R.W. Cook to S.R. Sapirie, requesting opinions on several questions involved in substituting uranium of low enrichment for the highly enriched material now used in the active lattice of the reactor in the Bulk Shielding Facility. The numbered paragraphs below correspond to the numbered paragraphs in Mr. Cook's memorandum.

  9. Results of Cesar II critical facility with low enriched fuel balls

    Energy Technology Data Exchange (ETDEWEB)

    Langlet, G.; Guerange, J.; Laponche, B.; Morier, F.; Neef, R.D.; Bock, H.J.; Kring, F.J.; Scherer, W.

    1972-06-15

    The Cesar facility has been transformed to load in its center a pebble bed fuel. This new Cesar assembly is called Cesar II. The program for the measurements with HTR type fuel balls is managed under a cooperation between physicists of CEA/CADARACHE and KFA/JUELICH. A description of the measuring zones of Cesar II and of the experimental results is given.

  10. Development of the on-demand fuel injection system for a light truck using the hydrogen enriched compressed natural gas (HCNG fuel

    Directory of Open Access Journals (Sweden)

    Sarawoot Watechagit

    2014-06-01

    Full Text Available The use of hydrogen as a fuel by mixing with a commercial fuel has recently been investigated continuously in order to solve the energy crisis and global warming. This article presents the results of the design and experimentation for the use of the hydrogen enriched compressed natural gas or HCNG as a fuel. The prototype vehicle a light truck equipped with a 1809 cc. gasoline engine. The proposed system is a mixing system where the compressed natural gas and the hydrogen are stored on-board and controlled separately. They are mixed as they are injected into the intake manifold right before the intake-port (Port Injection. The hydrogen supply system used in this investigation is adopted from the common system used for compressed natural gas system. The mixing ratio (%H in the total volume of HCNG ranges from 5% to 20% by volume. The performance testing is done through the Chassis Dynamometer. The results show that the power and the torque of the engine drops when using either CNG or HCNG as compared to the gasoline fuel. However, when compared to the case when using the CNG, the 10% addition of hydrogen can increase the performance by 1%. The performance, on the other hands, is reduced for other amount of hydrogen additions.

  11. Nitrogen removal in a single-chamber microbial fuel cell with nitrifying biofilm enriched at the air cathode

    KAUST Repository

    Yan, Hengjing

    2012-05-01

    Nitrogen removal is needed in microbial fuel cells (MFCs) for the treatment of most waste streams. Current designs couple biological denitrification with side-stream or combined nitrification sustained by upstream or direct aeration, which negates some of the energy-saving benefits of MFC technology. To achieve simultaneous nitrification and denitrification, without extra energy input for aeration, the air cathode of a single-chamber MFC was pre-enriched with a nitrifying biofilm. Diethylamine-functionalized polymer (DEA) was used as the Pt catalyst binder on the cathode to improve the differential nitrifying biofilm establishment. With pre-enriched nitrifying biofilm, MFCs with the DEA binder had an ammonia removal efficiency of up to 96.8% and a maximum power density of 900 ± 25 mW/m 2, compared to 90.7% and 945 ± 42 mW/m 2 with a Nafion binder. A control with Nafion that lacked nitrifier pre-enrichment removed less ammonia and had lower power production (54.5% initially, 750 mW/m 2). The nitrifying biofilm MFCs had lower Coulombic efficiencies (up to 27%) than the control reactor (up to 36%). The maximum total nitrogen removal efficiency reached 93.9% for MFCs with the DEA binder. The DEA binder accelerated nitrifier biofilm enrichment on the cathode, and enhanced system stability. These results demonstrated that with proper cathode pre-enrichment it is possible to simultaneously remove organics and ammonia in a single-chamber MFC without supplemental aeration. © 2012 Elsevier Ltd.

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

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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).

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

    Directory of Open Access Journals (Sweden)

    Varese Giovanna C

    2010-02-01

    Full Text Available Abstract 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

  15. Delayed Fission Product Gamma-Ray Transmission Through Low Enriched UO2 Fuel Pin Lattices in Air

    Energy Technology Data Exchange (ETDEWEB)

    Trumbull, TH [Rensselaer Polytechnic Inst., Troy, NY (United States)

    2004-10-18

    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

  16. Driving cycle simulation of a vehicle motored by a SI engine fueled with H sub 2 -enriched gasoline

    Energy Technology Data Exchange (ETDEWEB)

    Hacohen, J. (Exeter Univ. (GB). School of Engineering); Pinhasi, G.; Puterman, Y.; Sher, E. (Ben-Gurion Univ. of the Negev, Beersheba (IL). Dept. of Mechanical Engineering)

    1991-01-01

    In the present study, the effect of the amount of hydrogen supplement to a gasoline engine is theoretically investigated in terms of the vehicle performance running over a standard driving cycle. A rigorous mathematical model which considers the vehicle dynamics, transmission losses, the gas exchange and combustion processes inside the cylinder, as well as the chemical kinetics of the formation of the relevant products has been developed. The model was employed to predict the fuel consumption, engine power and the emission level of CO, HC, CO{sub 2} and NO{sub x} over the European ECE-15 and Japanese 10-mode driving cycles. Based on a previous work of the present authors, the hydrogen was considered to be produced by an on-board on-line auxiliary generator, which consumes a mixture of gasoline and water. It has been shown that a significant reduction in the total fuel consumption, in the order of 15 to 20% and an associated reduction in HC, CO and NO{sub x} emission levels, is achieved with only 6% of hydrogen enrichment (hydrogen to fuel mass ratio). (Author).

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

    Science.gov (United States)

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

    2010-12-01

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

  18. Average Droplet Diameter Measurement and Results for Fuel Aerosol Injected by Certain Types of the Turbojet Burners

    Institute of Scientific and Technical Information of China (English)

    TadeuszOpara

    1997-01-01

    Measurement of the diameter of the fuel aerosol droplet is very important in the design of new type burners and in diagnostic process,Diffraction method is one of the most useful measuring procedures in this case.An investigation setup is presented enabling the determination of the substituting drop diameter in fuel aerosol stream created by aeroengine injectors the results obtained for K 108-767,K 108-012,37.03.9595,16.83.0310 types are presented.

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

    Science.gov (United States)

    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

  20. HTGR Generic Technology Program: materials technology reactor; operating experience; medium-enriched-uranium fuel development. Quarterly progress report for the period ending July 31, 1978

    Energy Technology Data Exchange (ETDEWEB)

    1978-08-01

    The work reported includes the development of the materials properties data base for noncore components, plant surveillance and testing performed at Fort St. Vrain, and work to demonstrate the feasibility of using medium-enriched fuel in Fort St. Vrain. Studies and analyses plus experimental procedures and results are discussed and data are presented.

  1. HTGR Generic Technology Program. Materials technology reactor operating experience medium-enriched-uranium fuel development. Quarterly progress report for the period ending April 30, 1978

    Energy Technology Data Exchange (ETDEWEB)

    Kaae, J. L.; Lai, G. Y.; Thompson, L. D.; Sheehan, J. E.; Rosenwasser, S. N.; Johnson, W. R.; Li, C. C.; Pieren, W. R.; Smith, A. B.; Holko, K. H.; Baenteli, G. J.; Cheung, K. C.; Orr, J. D.; Potter, R. C.; Baxter, A.; Bell, W.; Lane, R.; Wunderlich, R. G.; Neylan, A. J.

    1978-05-01

    The work reported includes the development of the materials properties data base for noncore components, plant surveillance and testing performed at Fort St. Vrain, and work to demonstrate the feasibility of using medium-enriched fuel in Fort St. Vrain. Studies and analyses plus experimental procedures and results are discussed and data are presented.

  2. Study on the use of slightly enriched uranium fuel cycle in an existing CANDU 6 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yeom, Choong Sub; Kim, Hyun Dae [Institute for Advanced Engineering, Seoul (Korea, Republic of)

    1997-12-31

    To test the viability of CANFLEX-SEU bundles in an existing CANDU 6 reactor, core follow-up simulation has been carried out using the reactor fueling simulation program of the CANDU 6, RFSP computer code, and a lattice physics code, WIMS-AECL. During the core follow-up, bundle and channel powers and zone levels have been checked against their operating limits at each simulation. It is observed from the simulation results that an equilibrium core loaded with 0.9 w/o CANFLEX-SEU bundles could be refueled and maintained for 550 FPD without any significant violations in the channel and bundle power limits and the permissible operating range of the liquid zone controllers. 8 refs., 2 figs., 1 tab. (Author)

  3. Calculation of average molecular parameters, functional groups, and a surrogate molecule for heavy fuel oils using 1H and 13C NMR spectroscopy

    KAUST Repository

    Abdul Jameel, Abdul Gani

    2016-04-22

    Heavy fuel oil (HFO) is primarily used as fuel in marine engines and in boilers to generate electricity. Nuclear Magnetic Resonance (NMR) is a powerful analytical tool for structure elucidation and in this study, 1H NMR and 13C NMR spectroscopy were used for the structural characterization of 2 HFO samples. The NMR data was combined with elemental analysis and average molecular weight to quantify average molecular parameters (AMPs), such as the number of paraffinic carbons, naphthenic carbons, aromatic hydrogens, olefinic hydrogens, etc. in the HFO samples. Recent formulae published in the literature were used for calculating various derived AMPs like aromaticity factor 〖(f〗_a), C/H ratio, average paraffinic chain length (¯n), naphthenic ring number 〖(R〗_N), aromatic ring number〖 (R〗_A), total ring number〖 (R〗_T), aromatic condensation index (φ) and aromatic condensation degree (Ω). These derived AMPs help in understanding the overall structure of the fuel. A total of 19 functional groups were defined to represent the HFO samples, and their respective concentrations were calculated by formulating balance equations that equate the concentration of the functional groups with the concentration of the AMPs. Heteroatoms like sulfur, nitrogen, and oxygen were also included in the functional groups. Surrogate molecules were finally constructed to represent the average structure of the molecules present in the HFO samples. This surrogate molecule can be used for property estimation of the HFO samples and also serve as a surrogate to represent the molecular structure for use in kinetic studies.

  4. Development for analysis system of rods enrichment of nuclear fuels; Desarrollo de un sistema de analisis de enriquecimiento de barras de combustible nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Rojas C, E.L

    1998-11-01

    Nuclear industry is strongly regulated all over the world and quality assurance is important in every nuclear installation or process related with it. Nuclear fuel manufacture is not the exception. ININ was committed to manufacture four nuclear fuel bundles for the CFE nucleo electric station at Laguna Verde, Veracruz, under General Electric specifications and fulfilling all the requirements of this industry. One of the quality control requisites in nuclear fuel manufacture deals with the enrichment of the pellets inside the fuel bundle rods. To achieve the quality demanded in this aspect, the system described in this work was developed. With this system, developed at ININ it is possible to detect enrichment spikes since 0.4 % in a column of pellets with a 95 % confidence interval and to identify enrichment differences greater than 0.2 % e between homogeneous segments, also with a 95 % confidence interval. ININ delivered the four nuclear fuel bundles to CFE and these were introduced in the core of the nuclear reactor of Unit 1 in the fifth cycle. Nowadays they are producing energy and have shown a correct mechanical performance and neutronic behavior. (Author)

  5. 41 CFR 102-34.60 - How do we calculate the average fuel economy for Government motor vehicles?

    Science.gov (United States)

    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...

  6. 76 FR 74853 - 2017 and Later Model Year Light-Duty Vehicle Greenhouse Gas Emissions and Corporate Average Fuel...

    Science.gov (United States)

    2011-12-01

    .... \\4\\ Optical character recognition (OCR) is the process of converting an image of text, such as a... that the documents submitted be scanned using the Optical Character Recognition (OCR) process, thus... first phase of the National Program to regulate fuel economy and GHG emissions from U.S....

  7. Investigative studies on effect of reflector thickness on the performance of low enriched uranium-fueled miniature neutron source reactors

    Energy Technology Data Exchange (ETDEWEB)

    Odoi, H.C., E-mail: hencilod@gmail.com [School of Nuclear and Allied Sciences, University of Ghana, Legon (Ghana); Akaho, E.H.K. [Ghana Atomic Energy Commission, LG 80, Accra (Ghana); Anim-Sampong, S. [School of Nuclear and Allied Sciences, University of Ghana, Legon (Ghana); Centre for Energy Research and Training, Zaria (Nigeria); Jonah, S.A. [Centre for Energy Research and Training, Zaria (Nigeria); Nyarko, B.J.B.; Abrefah, R.G.; Ampomah-Amoako, E.; Sogbadji, R.B.M.; Lawson, I.; Birinkorang, S.A. [School of Nuclear and Allied Sciences, University of Ghana, Legon (Ghana); Ibrahim, Y.V. [Centre for Energy Research and Training, Zaria (Nigeria); Boffie, J. [School of Nuclear and Allied Sciences, University of Ghana, Legon (Ghana)

    2011-08-15

    Highlights: > Evaluation of reflector thickness required to compensate for the decrease in neutron flux due to conversion of reactor core from HEU to LEU. > Determination of neutron flux distribution along MNSR with increased reflector thickness. > Maintain the licensed reactivity of the Ghana Research Reactor-1, MNSR core. - Abstract: Neutronics analyses were performed on the 30 kW(th) GHARR-1 facility to investigate the effects on increased beryllium annular reflector thickness on nuclear criticality safety and on the neutron flux levels in the experimental channels. The investigative study was carried out using the Monte Carlo code MCNP on a hypothetical LEU UO{sub 2} core theoretically enriched to 12.6% and having the same core configuration as the present 90.2% enriched HEU U-Al core. The analyses were performed on four models consisting of a reference model with 10.2 cm annular reflector thickness and three new design modification models with increased reflector thickness of 10.3, 10.4 and 10.5 cm respectively. The simulations indicated average thermal neutron fluxes of (9.80 {+-} 0.0017)E+11 n/cm{sup 2} s in the inner irradiation channels for the reference model, indicating a 2% decrease with respect to the nominal flux of 1.00E+12 n/cm{sup 2} s. Relatively lower neutron fluxes were obtained for the modification models with an average of (9.79 {+-} 0.0017)E+11 n/cm{sup 2} s, representing losses of 2.01% and 0.01% with respect to the HEU core and reference LEU model.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  10. 76 FR 17019 - List of Approved Spent Fuel Storage Casks: HI-STORM Flood/Wind Addition

    Science.gov (United States)

    2011-03-28

    ... Reactor (BWR) fuel with high initial enrichment (up to 4.8 weight percent uranium-235 planer average...) The ability to store and transport BWR fuel with high initial enrichment (up to 4.8 weight percent... part 72, entitled ``General License for Storage of Spent Fuel at Power Reactor Sites'' (55 FR...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-02-01

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

  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

    Energy Technology Data Exchange (ETDEWEB)

    Favalli, A., E-mail: afavalli@lanl.gov [Los Alamos National Laboratory, Los Alamos, NM (United States); Vo, D. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Schwalbach, P. [European Commission, DG Energy, Euratom Safeguards Luxemburg, Luxemburg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company, Stockholm (Sweden); Tobin, S.J.; Trellue, H. [Los Alamos National Laboratory, Los Alamos, NM (United States); Vaccaro, S. [European Commission, DG Energy, Euratom Safeguards Luxemburg, Luxemburg (Luxembourg)

    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 {sup 137}Cs count rate and the {sup 154}Eu/{sup 137}Cs, {sup 134}Cs/{sup 137}Cs, {sup 106}Ru/{sup 137}Cs, and {sup 144}Ce/{sup 137}Cs 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.

  13. 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

    Science.gov (United States)

    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.

  14. Treatment and electricity harvesting from sulfate/sulfide-containing wastewaters using microbial fuel cell with enriched sulfate-reducing mixed culture

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Duu-Jong, E-mail: cedean@mail.ntust.edu.tw [Department of Chemical Engineering, National Taiwan University, Taipei, Taiwan (China); Department of Chemical Engineering, National Taiwan University of Science and Technology, Taipei, Taiwan (China); Lee, Chin-Yu [Department of Chemical Engineering, National Taiwan University, Taipei, Taiwan (China); Chang, Jo-Shu [Department of Chemical Engineering, National Cheng Kung University, Tainan, Taiwan (China); Center for Bioscience and Biotechnology, National Cheng Kung University, Tainan, Taiwan (China); Research Center for Energy Technology and Strategy, National Cheng Kung University, Tainan, Taiwan (China)

    2012-12-15

    Highlights: Black-Right-Pointing-Pointer We started up microbial fuel cell (MFC) using enriched sulfate-reducing mixed culture. Black-Right-Pointing-Pointer Sulfate-reducing bacteria and anode-respiring bacteria were enriched in anodic biofilms. Black-Right-Pointing-Pointer The MFC effectively remove sulfate to elementary sulfur in the presence of lactate. Black-Right-Pointing-Pointer The present device can treat sulfate laden wastewaters with electricity harvesting. - Abstract: Anaerobic treatment of sulfate-laden wastewaters can produce excess sulfide, which is corrosive to pipelines and is toxic to incorporated microorganisms. This work started up microbial fuel cell (MFC) using enriched sulfate-reducing mixed culture as anodic biofilms and applied the so yielded MFC for treating sulfate or sulfide-laden wastewaters. The sulfate-reducing bacteria in anodic biofilm effectively reduced sulfate to sulfide, which was then used by neighboring anode respiring bacteria (ARB) as electron donor for electricity production. The presence of organic carbons enhanced MFC performance since the biofilm ARB were mixotrophs that need organic carbon to grow. The present device introduces a route for treating sulfate laden wastewaters with electricity harvesting.

  15. A sensitivity study on DUPIC fuel composition

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok; Roh, Gyu Hong

    1997-01-01

    In DUPIC fuel cycle, the spent pressurized water reactor (PWR) fuel is refabricated as a DUPIC fuel by a dry process. Because the spent PWR fuel composition depends on the initial enrichment and burnup condition of PWR fuel, the composition of a DUPIC fuel is not uniquely defined. Therefore, for the purpose of reducing the effects of such a composition heterogeneity on core performance, a composition adjustment of DUPIC fuel was studies. The composition adjustment was made in two steps: mixing two spent PWR fuel assemblies of higher and lower {sup 239}Pu contents and blending in fresh uranium with the mixed spent PWR fuels. Because the fuel and core performances depend on both the absolute amount of fissile isotopes and the ratio of major fissile isotope contents, a parametric study was performed to determine the reference compositions of {sup 235}U and {sup 239}Pu. The reference enrichments of {sup 235}U and {sup 239}Pu were determined such that the DUPIC core performance is comparable to that of a natural uranium core with high spent PWR fuel utilization and low fuel cycle cost. Under this condition, it is possible to utilize 90% of spent PWR fuels as the DUPIC fuel formula. On average, the amounts of slightly enriched and depleted uranium used for blending correspond to 8.6% and 10.6%, respectively, of the mass of candidate spent PWR fuels. (author). 16 refs., 30 tabs., 9 figs.

  16. Modelling of HTR (High Temperature Reactor Pebble-Bed 10 MW to Determine Criticality as A Variations of Enrichment and Radius of the Fuel (Kernel With the Monte Carlo Code MCNP4C

    Directory of Open Access Journals (Sweden)

    Hammam Oktajianto

    2014-12-01

    Full Text Available Gas-cooled nuclear reactor is a Generation IV reactor which has been receiving significant attention due to many desired characteristics such as inherent safety, modularity, relatively low cost, short construction period, and easy financing. High temperature reactor (HTR pebble-bed as one of type of gas-cooled reactor concept is getting attention. In HTR pebble-bed design, radius and enrichment of the fuel kernel are the key parameter that can be chosen freely to determine the desired value of criticality. This paper models HTR pebble-bed 10 MW and determines an effective of enrichment and radius of the fuel (Kernel to get criticality value of reactor. The TRISO particle coated fuel particle which was modelled explicitly and distributed in the fuelled region of the fuel pebbles using a Simple-Cubic (SC lattice. The pebble-bed balls and moderator balls distributed in the core zone using a Body-Centred Cubic lattice with assumption of a fresh fuel by the fuel enrichment was 7-17% at 1% range and the size of the fuel radius was 175-300 µm at 25 µm ranges. The geometrical model of the full reactor is obtained by using lattice and universe facilities provided by MCNP4C. The details of model are discussed with necessary simplifications. Criticality calculations were conducted by Monte Carlo transport code MCNP4C and continuous energy nuclear data library ENDF/B-VI. From calculation results can be concluded that an effective of enrichment and radius of fuel (Kernel to achieve a critical condition was the enrichment of 15-17% at a radius of 200 µm, the enrichment of 13-17% at a radius of 225 µm, the enrichments of 12-15% at radius of 250 µm, the enrichments of 11-14% at a radius of 275 µm and the enrichment of 10-13% at a radius of 300 µm, so that the effective of enrichments and radii of fuel (Kernel can be considered in the HTR 10 MW. Keywords—MCNP4C, HTR, enrichment, radius, criticality 

  17. Volatile behaviour of enrichment uranium in the total nuclear fuel price; Volatilidad de los mercados de Uranio enriquecido. Impactos sobre el coste de combustible

    Energy Technology Data Exchange (ETDEWEB)

    Arnaiz, J.; Inchausti, J. M.; Tarin, F.

    2004-07-01

    In this article the historical high volatile behaviour of the total nuclear fuel price is evaluated quantitatively and it is concluded that it has been due mainly to the fluctuations of the price of the principal components of enriched uranium (concentrates and enrichment). In order to avoid the negative effects of this volatiles behaviour as far as possible, a basic strategy in the uranium procurement activities is recommended (union of buyers, diversification of supplier, stock management, optimisation of contract portfolio and suitable currency management that guarantees a reliable uranium supply at reasonable prices. These guidelines are those that ENUSA has been following on behalf of the Spanish Utilities in the Commission of Uranium Procurement (CAU in Spanish). (Author) 11 refs.

  18. The effect of alternative fuel combustion in the cement kiln main burner on production capacity and improvement with oxygen enrichment.

    OpenAIRE

    Ariyaratne, W.K.Hiromi; Melaaen, Morten Christian; Tokheim, Lars-André

    2013-01-01

    A mathematical model based on a mass and energy balance for the combustion in a cement rotary kiln was developed. The model was used to investigate the impact of replacing about 45 % of the primary coal energy by different alternative fuels. Refuse derived fuel, waste wood, solid hazardous waste and liquid hazardous waste were used in the modeling. The results showed that in order to keep the kiln temperature unchanged, and thereby maintain the required clinker quality, the production capa...

  19. State Averages

    Data.gov (United States)

    U.S. Department of Health & Human Services — A list of a variety of averages for each state or territory as well as the national average, including each quality measure, staffing, fine amount and number of...

  20. 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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  1. 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

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Bodey, Isaac T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Curtis, Franklin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Arimilli, Rao V. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ekici, Kivanc [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Freels, James D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Bodey, Isaac T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Curtis, Franklin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Arimilli, Rao V. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ekici, Kivanc [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Freels, James D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-11-01

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

  4. LMFBR type reactor core and its fuel exchange method

    Energy Technology Data Exchange (ETDEWEB)

    Ishibashi, Yoko; Koyama, Jun-ichi; Aoyama, Motoo; Haikawa, Katsumasa; Yamanaka, Akihiro

    1996-08-20

    Upon initial loading, two kinds of fuel assemblies including first fuel assemblies having a highest enrichment degree and second fuel assemblies having a lowest enrichment degree are loaded. The average fuel enrichment degree of an upper region of the first fuel assembly is made greater than that of the lower region. The reactivity of the lower region of the first fuel assembly is made lower than that of the upper portion to reduce power peak. Upon transfer from a first cycle to a second cycle, at least one of the second fuel assemblies is exchanged by the same number of the third fuel assemblies. In this case, an average fuel enrichment degree of the upper region of the third fuel assembly is made greater than that of the lower region to suppress the reactivity in the lower region of the third fuel assembly lower than the reactivity in the upper region thereby reducing the power peak. Thus, the upper power peak over the entire reactor core is moderated thereby capable of ensuring the reactor shut down margin without deteriorating the same. (N.H.)

  5. Investigative studies on the effects of cadmium rabbits on high enriched uranium-fueled and low enriched uranium-fueled cores of Ghana Research Reactor-1 using MCNP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Boffie, J., E-mail: jboffie@yahoo.com [Department of Nuclear Engineering and Material Science, School of Nuclear and Allied Sciences (SNAS), University of Ghana, P.O. Box AE 1, Atomic Energy, Accra (Ghana); National Nuclear Research Institute, Ghana Atomic Energy Commission, P.O. Box LG 80, Legon, Accra (Ghana); Akaho, E.H.K. [Department of Nuclear Engineering and Material Science, School of Nuclear and Allied Sciences (SNAS), University of Ghana, P.O. Box AE 1, Atomic Energy, Accra (Ghana); Nyarko, B.J.B.; Odoi, H.C.; Tuffour-Achampong, K.; Abrefah, R.G. [Department of Nuclear Engineering and Material Science, School of Nuclear and Allied Sciences (SNAS), University of Ghana, P.O. Box AE 1, Atomic Energy, Accra (Ghana); National Nuclear Research Institute, Ghana Atomic Energy Commission, P.O. Box LG 80, Legon, Accra (Ghana)

    2013-12-15

    Highlights: • The operating parameters for both the HEU core and proposed LEU core were similar. • The length of the Cd in the capsules must be increased for its use in the LEU core. • Cd rabbits can emergently be used to shut down MNSRs. - Abstract: Miniature Neutron Source Reactors (MNSRs) are noted to be among highly safe research reactors. However, because of its use of one control rod for reactivity control and shutdown purposes, alternative methods of shutting it down are important. The Ghana MNSR uses four cadmium rabbits of approximate dimensions 6.5 cm × 5.0 cm × 0.1 cm and mass of 9.48 g each to emergently shut down the reactor. The Monte Carlo N-Particle code; version 5, (MCNP5) was used to design the high enriched uranium (HEU) and low enriched uranium (LEU) cores of the MNSR with four cadmium rabbits inserted in four inner irradiation sites of each core. The operating parameters and shutdown parameters for both cores with the central control rod (CCR) either fully withdrawn or fully inserted had similar results with the HEU core having slightly better results in terms of safety. However, the results show that the four inserted cadmium rabbits make the HEU core subcritical whiles in the LEU core, it still remains critical (k{sub eff} = 1.00005 ± 0.00007). The length of the cadmium material in each cadmium rabbit must therefore be increased by at least 0.5 cm in order to attain subcriticality (k{sub eff} = 0.99989 ± 0.00006) and shutdown margin of 0.11 mk when inserted in the LEU core.

  6. Review of 15 years: high-density low-enriched UMo dispersion fuel development for research reactors in Europe

    Energy Technology Data Exchange (ETDEWEB)

    Berghe, S. van den [SCK.CEN, Nuclear Materials Science Institute (NMS), Boeretan (Belgium); Lemoine, P [Commissariat a l' Eergie Atomique, CEA Saclay, Yvette Cedex (France)

    2014-04-15

    This review aims to provide a synthesis of the knowledge generated and the lessons learned in roughly 15 years of UMo dispersion fuel R and D in Europe through a series of irradiation experiments. A lot of irradiations were also performed outside of Europe, particularly in the USA, Russia, Canada, Korea and Argentina. In addition, a large number of out-of-pile investigations were done throughout the world, providing support to the understanding of the phenomena governing the UMo behaviour in pile. However, the focus of this article will be on the irradiations and Post-Irradiation Examination (PIE) results obtained in European experiments. The introduction of the article provides a historic overview of the evolution and progress in the high density UMo dispersion fuel development. The ensuing sections then provide further details on the various phases of the development, from the UMo dispersion in a pure Al matrix through the addition of Si to the matrix to address the interaction layer formation and finally to the more advanced solutions to the excessive swelling encountered in various experiments. This review was based only on published results or results that are currently in the process of being published.

  7. Extension of the lean limit through hydrogen enrichment of a LFG-fueled spark-ignition engine and emissions reduction

    Energy Technology Data Exchange (ETDEWEB)

    Kornbluth, Kurt; Greenwood, Jason; McCaffrey, Zach; Vernon, David; Erickson, Paul [Mechanical and Aerospace Engineering Department, UC Davis, CA 95616 (United States)

    2010-02-15

    In this experimental investigation the affect of hydrogen addition to a landfill gas-fueled naturally-aspirated spark-ignition engine was explored. Hydrogen concentrations of 0%, 30%, 40%, and 50% by volume were added to simulated landfill gas (60% CH{sub 4} and 40% CO{sub 2}). Efficiency, coefficient of variance of indicated mean effective pressure, and CO emissions were measured from near stoichiometric mixtures up to the lean operating limit. Engine-out NOx emissions were compared to predicted future best available control technology targets for NOx emissions in landfill gas-to-energy projects. From this study, it was determined that with 40% hydrogen by volume untreated exhaust NOx emissions can meet the 0.22 g/kWh NOx target while retaining 95% of baseline power and low CO emissions. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Wenzel, Tom [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)

    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.

  9. Emerging Technologies for the Production of Renewable Liquid Transport Fuels from Biomass Sources Enriched in Plant Cell Walls

    Directory of Open Access Journals (Sweden)

    Hwei-Ting Tan

    2016-12-01

    Full Text Available Plant cell walls are composed predominantly of cellulose, a range of non-cellulosic polysaccharides and lignin. The walls account for a large proportion not only of crop residues such as wheat straw and sugarcane bagasse, but also of residues of the timber industry and specialist grasses and other plants being grown specifically for biofuel production. The polysaccharide components of plant cell walls have long been recognized as an extraordinarily large source of fermentable sugars that might be used for the production of bioethanol and other renewable liquid transport fuels. Estimates place annual plant cellulose production from captured light energy in the order of hundreds of billions of tonnes. Lignin is synthesised in the same order of magnitude and, as a very large polymer of phenylpropanoid residues, lignin is also an abundant, high energy macromolecule. However, one of the major functions of these cell wall constituents in plants is to provide the extreme tensile and compressive strengths that enable plants to resist the forces of gravity and a broad range of other mechanical forces. Over millions of years these wall constituents have evolved under natural selection to generate extremely tough and resilient biomaterials. The rapid degradation of these tough cell wall composites to fermentable sugars is therefore a difficult task and has significantly slowed the development of a viable lignocellulose-based biofuels industry. However, good progress has been made in overcoming this so-called recalcitrance of lignocellulosic feedstocks for the biofuels industry, through modifications to the lignocellulose itself, innovative pre-treatments of the biomass, improved enzymes and the development of superior yeasts and other microorganisms for the fermentation process. Nevertheless, it has been argued that bioethanol might not be the best or only biofuel that can be generated from lignocellulosic biomass sources and that hydrocarbons with

  10. Reynolds-averaged Navier-Stokes analysis of the flow through a model rocket-based combined-cycle engine with an independently-fueled ramjet stream

    Science.gov (United States)

    Bond, Ryan Bomar

    A new concept for the low speed propulsion mode in rocket based combined cycle (RBCC) engines has been developed as part of the NASA GTX program. This concept, called the independent ramjet stream (IRS) cycle, is a variation of the traditional ejector ramjet (ER) design and involves the injection of hydrogen fuel directly into the air stream, where it is ignited by the rocket plume. Experiments and computational fluid dynamics (CFD) are currently being used to evaluate the feasibility of the new design. In this work, a Navier-Stokes code valid for general reactive flows is applied to the model engine under cold flow, ejector ramjet, and IRS cycle operation. Pressure distributions corresponding to cold-flow and ejector ramjet operation are compared with experimental data. The engine response under independent ramjet stream cycle operation is examined for different reaction models and grid sizes. The engine response to variations in fuel injection is also examined. Mode transition simulations are also analyzed both with and without a nitrogen purge of the rocket. The solutions exhibit a high sensitivity to both grid resolution and reaction mechanism, but they do indicate that thermal throat ramjet operation is possible through the injection and burning of additional fuel into the air stream. The solutions also indicate that variations in fuel injection location can affect the position of the thermal throat. The numerical simulations predicted successful mode transition both with and without a nitrogen purge of the rocket; however, the reliability of the mode transition results cannot be established without experimental data to validate the reaction mechanism.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  12. Evaluating the effectiveness of dilution of the recovered uranium with depleted uranium and low-enriched uranium to obtain fuel for VVER reactors

    Science.gov (United States)

    Smirnov, A. Yu; Sulaberidze, G. A.; Dudnikov, A. A.; Nevinitsa, V. A.

    2016-09-01

    The possibility of the recovered uranium enrichment in a cascade of gas centrifuges with three feed flows (depleted uranium, low-enriched uranium, recovered uranium) with simultaneous dilution of U-232,234,236 isotopes was shown. A series of numerical experiments were performed for different content of U-235 in low-enriched uranium. It has been demonstrated that the selected combination of diluents can simultaneously reduce the cost of separative work and the consumption of natural uranium, not only with respect to the previously used multi-flow cascade schemes, but also in comparison to the standard cascade for uranium enrichment.

  13. Uranium Conversion & Enrichment

    Energy Technology Data Exchange (ETDEWEB)

    Karpius, Peter Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-02-06

    The isotopes of uranium that are found in nature, and hence in ‘fresh’ Yellowcake’, are not in relative proportions that are suitable for power or weapons applications. The goal of conversion then is to transform the U3O8 yellowcake into UF6. Conversion and enrichment of uranium is usually required to obtain material with enough 235U to be usable as fuel in a reactor or weapon. The cost, size, and complexity of practical conversion and enrichment facilities aid in nonproliferation by design.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    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.

  15. Integrated Aircraft Fuel Tank Inerting and Compartment Fire Suppression System. Volume 2. Evaluation of Nitrogen-Enriched Air as a Fire Suppressant

    Science.gov (United States)

    1983-04-01

    speculation that such an ideal condition might not occur in an actual test situation, further tests were conducted to determine whether the...TYPE: POOL FUEL TYPE: .JP4 FUEL FLOW RATE; N/A EXTINGUISHANT: C02 RUN1Cn AIR TEMPERATURE: VARIOUS 4 .. i00 PUNtOS RUN117 .110 2 30 4065 VOLUME I

  16. The Software Design for252Cf Neutron Activation Fuel Rod 235U Enrichment Inspecting Equipment%252Cf中子活化核燃料棒235U富集度检测设备的软件设计

    Institute of Scientific and Technical Information of China (English)

    张雷; 刘明; 马金波

    2013-01-01

    It introduces the software design for 252Cf neutron activation fuel red235U enrichment inspecting equipment.It used multithread technique to control Advantech PCI-1780 counter/timer card,and collect γ-ray signal from the six-path detectors.Process and analyze the collected data can exactly check the actual 235U enrichment and abnormal pellets in the nuclear fuel rods.The software can measure the actual 235U enrichment and judge whether there are abnormal pellets in the nuclear fuel rods accurately,and send customizing messages to PLC which complete automatic sorting,at 6 m/min detection speed.Now the software is used on nondestructive test equipment in Nuclear Fuel Element Factory.%介绍了252Cf中子活化核燃料棒235U富集度检测设备的软件设计,该软件采用多线程技术控制研华PCI-1780采集卡定时采集六路探测器输出的经252Cf中子活化后235U裂变产物的γ射线信号,针对采集数据的特性,进行相应的处理和分析,可以检测出核燃料棒的实际235U富集度以及有无异常芯块.该软件经过实验验证在检测速度为6时,能够准确测量核燃料棒的实际235U富集度值并判断棒中是否混有异常芯块,同时向PLC发送相应信号实现自动分选.目前已应用在核燃料元件厂的核燃料棒235U富集度无损检测设备上.

  17. Cycle-by-cycle Variations in a Direct Injection Hydrogen Enriched Compressed Natural Gas Engine Employing EGR at Relative Air-Fuel Ratios.

    Directory of Open Access Journals (Sweden)

    Olalekan Wasiu Saheed

    2014-07-01

    Full Text Available Since the pressure development in a combustion chamber is uniquely related to the combustion process, substantial variations in the combustion process on a cycle-by-cycle basis are occurring. To this end, an experimental study of cycle-by-cycle variation in a direct injection spark ignition engine fueled with natural gas-hydrogen blends combined with exhaust gas recirculation at relative air-fuel ratios was conducted. The impacts of relative air-fuel ratios (i.e. λ = 1.0, 1.2, 1.3 and 1.4 which represent stoichiometric, moderately lean, lean and very lean mixtures respectively, hydrogen fractions and EGR rates were studied. The results showed that increasing the relative air-fuel ratio increases the COVIMEP. The behavior is more pronounced at the larger relative air-fuel ratios. More so, for a specified EGR rate; increasing the hydrogen fractions decreases the maximum COVIMEP value just as increasing in EGR rates increases the maximum COVIMEP value. (i.e. When percentage EGR rates is increased from 0% to 17% and 20% respectively. The maximum COVIMEP value increases from 6.25% to 6.56% and 8.30% respectively. Since the introduction of hydrogen gas reduces the cycle-by-cycle combustion variation in engine cylinder; thus it can be concluded that addition of hydrogen into direct injection compressed natural gas engine employing EGR at various relative air-fuel ratios is a viable approach to obtain an improved combustion quality which correspond to lower coefficient of variation in imep, (COVIMEP in a direct injection compressed natural gas engine employing EGR at relative air-fuel ratios.

  18. Behaviour of the reactivity for BWR fuel cells; Comportamiento de la reactividad para celdas de combustible BWR

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez, J. A.; Alonso, G.; Delfin, A.; Vargas, S. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Del Valle G, E., E-mail: galonso@inin.gob.mx [IPN, Escuela Superior de Fisica y Matematicas, U. P. Adolfo Lopez Mateos, Col. Lindavista, 07738 Mexico D. F. (Mexico)

    2011-11-15

    In this work the behaviour of the reactivity of a fuel assembly type BWR was studied, the objective is to obtain some expressions that consider the average enrichment of U-235 and the gadolinium concentration like a function of the fuel cells burnt. Also, the applicability of the lineal reactivity model was analyzed for fuel cells type BWR. The analysis was carried out with the CASMO-4 code. (Author)

  19. 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

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.; Ross, Kyle W. (Los Alamos National Laboratory, Los Alamos, NM); 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.

  1. Radiological and nuclear safety aspects in the fabrication of 1.8% enriched U O{sub 2} fuel rods for the RA-8 critical facility; Aspectos de seguridad radiologica y nuclear en la fabricacion de barras combustibles, con U O{sub 2} enriquecido al 1.8%, para la facilidad critica RA-8

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas, Hugo; Becarra, Fabian; Herrero, Jorge; Luna, Manuel; Perez, Aldo [Comision Nacional de Energia Atomica, (Argentina). Centro Atomico Constituyentes

    1997-10-01

    The neutronic behavioral study of the fuel for the future nuclear reactor CAREM required to mount critical facility with 1.8% enriched U O{sub 2} fuel rods. The present work describes the various operation and production processes, the safety and radioprotection systems, the administrative procedures and the associated radiological controls. Also, the results obtained in the area and personal monitoring and waste generation are detailed. (author). 10 refs., 4 figs., 1 tab.

  2. Field Measurement of Am241 and Total Uranium at a Mixed Oxide Fuel Facility with Variable Uranium Enrichments Ranging from 0.3% to 97% U235

    Energy Technology Data Exchange (ETDEWEB)

    Conway, K. C.

    2002-02-28

    The uranium and transuranic content of site soils and building rubble can be accurately measured using a NaI(Tl) well counter, without significant soil preparation. Accurate measurements of total uranium in uranium-transuranic mixtures can be made, despite a wide range (0.3% to 97%) of uranium enrichment, sample mass, and activity concentrations. The appropriate uranium scaling factors needed to include the undetected uranium isotopes, particularly U 234 can be readily determined on a sample by sample basis as a part of the field analysis, by comparing the relative response of the U 235 186 keV peak versus the K shell X rays of U 238 , U 235, and their immediate ingrowth daughters. The ratio of the two results is a sensitive and accurate predictor of the uranium enrichment and scaling factors. The case study will illustrate how NaI(Tl) gamma spectrometry was used to provide rapid turnaround uranium and transuranic activity levels for soil and building rubble with sample by sample determination of the appropriate scaling factor to include the U234 and Uranium238 content.

  3. Composition heterogeneity analysis for DUPIC fuel(I) - Statistical analysis

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-08-01

    The fuel composition heterogeneity effect on reactor performance parameters was assessed by refueling simulations for three DUPIC fuel options of fuel composition heterogeneity control: the fissile content adjustment, the reactivity control by slightly enriched and depleted uranium, and the reactivity control by natural uranium. For each DUPIC fuel option, the simulations were performed using 30 heterogeneous fuel types which were determined by the agglomerative hierarchical clustering method. The heterogeneity effect was considered during the refueling simulation by randomly selecting fuel types for the refueling operation. The refueling simulations of the heterogeneous core have shown that the key performance parameters such as the maximum channel power (MCP), maximum bundle power (MBP), and channel power peaking factor (CPPF) are close to those of the core that has single fuel type. For the three DUPIC fuel options, the uncertainties of MCP, MBP, and CPPF due to the fuel composition heterogeneity are less than 0.6, 1.5 and 0.8%, respectively, including the uncertainty of the group-average fuel property. This study has shown that the three DUPIC fuel options reduces the composition heterogeneity effectively and the zone power control system has a sufficient margin to adjust the perturbations cased by the fuel composition heterogeneity. 15 refs., 28 figs.,10 tabs. (Author)

  4. Structure and thermal properties of as-fabricated U-7Mo/Mg and U-10Mo/Mg low-enriched uranium research reactor fuels

    Science.gov (United States)

    Kulakov, Mykola; Saoudi, Mouna; Piro, Markus H. A.; Donaberger, Ronald L.

    2017-02-01

    Aluminum-clad U-7Mo/Mg and U-10Mo/Mg pin-type mini-elements (with a core uranium loading of 4.5 gU/cm3) have been fabricated at the Canadian Nuclear Laboratories for experimental tests and ultimately for use in research and test reactors. In this study, the microstructure and phase composition of unirradiated U-7Mo/Mg and U-10Mo/Mg fuel cores were analyzed using optical and scanning electron microscopy, and neutron powder diffraction. Thermal properties were characterized using a combination of experimental measurements and thermodynamic calculations. The thermal diffusivity was measured using the laser flash method. The temperature-dependent specific heat capacities were calculated based on the linear rule of mixture using the weight fraction of different crystalline phases and their specific heat capacity values taken from the literature. The thermal conductivity was then calculated using the measured thermal diffusivity, the measured density and the calculated specific heat capacity. The resulting thermal conductivity is practically identical for both types of fuel. The in-reactor temperatures were predicted using conjugate heat transfer simulations.

  5. Investigation of deuterium cross section data by integral testing: ZED-2 measurements of high-enriched uranium fuel substituted into a natural uranium core

    Energy Technology Data Exchange (ETDEWEB)

    Atfield, J.E.; Kozier, K.S.; Roubtsov, D.; Zeller, M.B. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2012-07-01

    Historical ZED-2 measurements of an HEU fuel rod substituted into a lattice of NU rods were analysed to determine their reactivity sensitivity to differences between the neutron elastic scattering cross-sections of deuterium from different evaluated nuclear data libraries. The differences in the deuterium nuclear data concern the angular probability distribution at neutron energies below 3.2 MeV. These ZED-2 experiments were selected due to the presence of HEU fuel in D{sub 2}O, since analyses of other critical experiments involving solutions of HEU fluoride in D{sub 2}O show substantial sensitivity (~10 mk) to these differences in the deuterium nuclear data. This analysis shows that the existing ZED-2 HEU experiments are insufficiently sensitive to resolve the discrepancy between the different deuterium data libraries. Further analysis of hypothetical configurations with high sensitivity shows that the sensitivity to the angular probability distribution of deuterium is strongly correlated with the leakage of fast neutrons, and it is recommended that further experiments to address this deuterium nuclear data issue be designed/evaluated to maximize this quantity. (author)

  6. Corporate average fuel consumption (CAFC) and its limit standard of passenger vehicle in China%中国乘用车企业平均燃料消耗量(CAFC)及其限值标准

    Institute of Scientific and Technical Information of China (English)

    马冬; 安锋; 康利平; Robert Earley

    2012-01-01

    提高中国汽车行业燃料经济性水平有利于节能减排。该文基于权威部门数据和标准,研究了中国汽车市场上乘用车的企业平均燃料消耗量(CAFC)。结果发现:2011年度中国汽车乘用车CAFC平均为7.5L/(100km),生产车型总体满足中国《乘用车燃料消耗量限值》第二阶段(GB19578—2004,2004-09—02)限值标准,尚不满足第三阶段(GB27999—2011,2011—12—30)目标值。与合资品牌相比,自主品牌汽车平均CAFC较低,但是由于产品技术相对落后,CAFC实际值与目标值的比值反而更大,面临较大达标压力。与国产汽车相比,进口汽车平均CAFC较高,而且CAFC实际值与目标值的比值也很大,因此第三阶段达标形势更加严峻。%To promote the fuel economy level of passenger vehicle industry in China is beneficial to energy saving and emission reduction. The Corporate Average Fuel Consumption (CAFC) of passenger vehicles in China market was investigated based on the authority's data and standards. The results show that the passenger-vehicle CAFC in China market in the year of 2011 is 7.5 L/(100 km), which overall meets the target value of Phase 2 in the "Limits of Fuel Consumption for Passenger Cars" (GB 19578-2004, 2004-09-02) of China, but does not meet the target of Phase 3 (GB 27999-2011, 2011-12-30). The vehicle companies with independent-brands have lower real CAFC values than those with the joint-venture-brands, but have a higher ratio of the real value to the CAFC target standard, so they have more work to do to meet the Phase 3 standard Compared with domestic vehicles, imported passenger vehicles have higher CAFC real values and a higher ratio of the real value to the target standard, which shows a severe challenge to the domestic vehicles.

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

    Science.gov (United States)

    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.

  8. FUEL CELL ELECTRODE MATERIALS

    Science.gov (United States)

    FUEL CELL ELECTRODE MATERIALS. RAW MATERIAL SELECTION INFLUENCES POLARIZATION BUT IS NOT A SINGLE CONTROLLING FACTOR. AVAILABLE...DATA INDICATES THAT AN INTERRELATIONSHIP OF POROSITY, AVERAGE PORE VOLUME, AND PERMEABILITY CONTRIBUTES TO ELECTRODE FUEL CELL BEHAVIOR.

  9. Irradiation Experiment Conceptual Design Parameters for NBSR Fuel Conversion

    Energy Technology Data Exchange (ETDEWEB)

    Brown, N. R. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Brown, N. R. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Baek, J. S [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Hanson, A. L. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Cuadra, A. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Cheng, L. Y. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Diamond, D. J. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.

    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.

  10. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Science.gov (United States)

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

    2011-12-01

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

  12. 垃圾衍生燃料富氧燃烧污染物排放特性%Pollutants emission characteristics of refuse derived fuel in oxygen-enriched combustion

    Institute of Scientific and Technical Information of China (English)

    李延吉; 姜璐; 赵宁; 李玉龙; 李润东; 池涌

    2013-01-01

    Combustion and emission characteristics of refuse derived fuel ( RDF ) were experimental studied in a tubular high temperature furnace. The results showed that; 1 The increase of plastic proportion in RDF and oxygen concentration in combustion air lead the increase of NOX emission; but under the pure oxygen environment of RDF combustion, NOx emission is greatly reduced; 2 SO2 concentration increases with increasing plastic proportion in RDF and oxygen concentration in combustion air,; 3 Oxygen concentration in combustion air does not have significance effect on CO emission when it is less than 80% in oxygen-enriched combustion,. In pure oxygen combustion, increasing the plastic ratio can reduce the CO emissions; CO emissions tended to decrease with increasing oxygen concentration, CO emissions is minimum in pure oxygen condition; 4 The emissions of NOx, SO2 and CO are lower than national standard, which indicated that the RDF combustion with oxygen enrichment is beneficial for the c pollutant emissions reduction.%为了解决城市生活垃圾直接焚烧产生的二次污染问题,将城市生活垃圾制成垃圾衍生燃料(RDF),在高温管式炉内进行富氧燃烧污染物排放特性研究.结果表明:塑料比例增加,燃烧过程中NOx浓度增大;氧浓度增加,NOx浓度增大;但纯氧条件下RDF燃烧,NOx浓度大大降低.塑料比例增加,燃烧过程中SO2浓度增大;氧浓度增加,SO2浓度降低.当氧浓度为80%时,CO浓度相差不大,纯氧时,塑料比例增大,CO浓度减小;氧浓度增大,CO浓度呈减小趋势,纯氧时CO浓度最小.NOx、SO2、CO的浓度均低于国家标准,说明RDF富氧燃烧有利于降低污染物排放浓度.

  13. HTGR fuel and fuel cycle technology

    Energy Technology Data Exchange (ETDEWEB)

    Lotts, A.L.; Coobs, J.H.

    1976-08-01

    The status of fuel and fuel cycle technology for high-temperature gas-cooled reactors (HTGRs) is reviewed. The all-ceramic core of the HTGRs permits high temperatures compared with other reactors. Core outlet temperatures of 740/sup 0/C are now available for the steam cycle. For advanced HTGRs such as are required for direct-cycle power generation and for high-temperature process heat, coolant temperatures as high as 1000/sup 0/C may be expected. The paper discusses the variations of HTGR fuel designs that meet the performance requirements and the requirements of the isotopes to be used in the fuel cycle. Also discussed are the fuel cycle possibilities, which include the low-enrichment cycle, the Th-/sup 233/U cycle, and plutonium utilization in either cycle. The status of fuel and fuel cycle development is summarized.

  14. Aggregation and Averaging.

    Science.gov (United States)

    Siegel, Irving H.

    The arithmetic processes of aggregation and averaging are basic to quantitative investigations of employment, unemployment, and related concepts. In explaining these concepts, this report stresses need for accuracy and consistency in measurements, and describes tools for analyzing alternative measures. (BH)

  15. On Averaging Rotations

    DEFF Research Database (Denmark)

    Gramkow, Claus

    1999-01-01

    In this article two common approaches to averaging rotations are compared to a more advanced approach based on a Riemannian metric. Very offten the barycenter of the quaternions or matrices that represent the rotations are used as an estimate of the mean. These methods neglect that rotations belong...... approximations to the Riemannian metric, and that the subsequent corrections are inherient in the least squares estimation. Keywords: averaging rotations, Riemannian metric, matrix, quaternion...

  16. Supply of enriched uranium for research reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-08-01

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

  17. Irradiation Experiment Conceptual Design Parameters for NBSR Fuel Conversion

    Energy Technology Data Exchange (ETDEWEB)

    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.

  18. Your Average Nigga

    Science.gov (United States)

    Young, Vershawn Ashanti

    2004-01-01

    "Your Average Nigga" contends that just as exaggerating the differences between black and white language leaves some black speakers, especially those from the ghetto, at an impasse, so exaggerating and reifying the differences between the races leaves blacks in the impossible position of either having to try to be white or forever struggling to…

  19. On Averaging Rotations

    DEFF Research Database (Denmark)

    Gramkow, Claus

    2001-01-01

    In this paper two common approaches to averaging rotations are compared to a more advanced approach based on a Riemannian metric. Very often the barycenter of the quaternions or matrices that represent the rotations are used as an estimate of the mean. These methods neglect that rotations belong...

  20. Profile of World Uranium Enrichment Programs-2009

    Energy Technology Data Exchange (ETDEWEB)

    Laughter, Mark D [ORNL

    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. Negative Average Preference Utilitarianism

    Directory of Open Access Journals (Sweden)

    Roger Chao

    2012-03-01

    Full Text Available For many philosophers working in the area of Population Ethics, it seems that either they have to confront the Repugnant Conclusion (where they are forced to the conclusion of creating massive amounts of lives barely worth living, or they have to confront the Non-Identity Problem (where no one is seemingly harmed as their existence is dependent on the “harmful” event that took place. To them it seems there is no escape, they either have to face one problem or the other. However, there is a way around this, allowing us to escape the Repugnant Conclusion, by using what I will call Negative Average Preference Utilitarianism (NAPU – which though similar to anti-frustrationism, has some important differences in practice. Current “positive” forms of utilitarianism have struggled to deal with the Repugnant Conclusion, as their theory actually entails this conclusion; however, it seems that a form of Negative Average Preference Utilitarianism (NAPU easily escapes this dilemma (it never even arises within it.

  2. Profile of World Uranium Enrichment Programs - 2007

    Energy Technology Data Exchange (ETDEWEB)

    Laughter, Mark D [ORNL

    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

  3. Advanced PWR in-core fuel management with optimized gadolinia fuel designs

    Energy Technology Data Exchange (ETDEWEB)

    Berger, H.D.; Neufert, A. [Siemens AG / Power Generation KWU, Nuclear Fuel Cycle, Erlangen (Germany)

    1999-07-01

    Utilities operating LWRs require fuel assemblies and in-core fuel management service, which ensure safe, flexible and cost-effective production of electricity. With the reliability of the fuel having been always the most important requirement, advanced measures to minimize fuel cycle costs are receiving increasing attention in the light of the pressure on costs within the de-regulated power generation markets. The role of in-core fuel management in supporting the goal to minimize fuel cycle costs consists in the development of more demanding core loading strategies, i.e. in the first place more advanced low leakage loading patterns. A prerequisite for this type of loading pattern is the use of an optimized burnable absorber design. Gadolinia as integrated burnable absorber is a very effective means for limiting the critical boron concentration and power peaking factors. Siemens has accumulated extensive experience with Gd-fuel for almost 20 years with e.g. more than 5500 Gd-FA's delivered for PWRs and irradiated up to 65 MWd/kg{sub HM}. Current development efforts for optimizing Gd-fuel are focused on the reduction of the inherent penalties of today's Gd-Fa designs, i.e. reduced average FA enrichment and heavy metal content as well as residual reactivity binding. The most effective way to overcome these drawbacks is the reduction of the Gd{sub 2}O{sub 3} concentration to values of approximately 2 w/o, for which according to recent measurements of the heat conductivity of modern Gd-fuels the reduction of the fissile content in the Gd-rods is no longer necessary. Various feasibility studies have been performed to evaluate the consequences of low-Gd designs for both Siemens PWRs and Non-Siemens PWRs, for which more restrictive boundary conditions with respect to critical boron concentration and peaking factors have to be fulfilled. These studies as well as the first realization of an extended reactor cycle using a low Gd-Fa reload design confirm that the in

  4. Fuel rod behavior under normal operating conditions in Super Fast Reactor with high power density

    Energy Technology Data Exchange (ETDEWEB)

    Ju, Haitao, E-mail: haitaoju@gmail.com [Science and Technology on Reactor System Design Technology Laboratory, Chengdu, Sichuan 610041 (China); Ishiwatari, Yuki [Department of Nuclear Engineering and Management, The University of Tokyo, Hongo, Bunkyo, Tokyo 113-8656 (Japan); Oka, Yoshiaki [Joint Department of Nuclear Energy, Waseda University, Totsukamachi, Shinjuku, Tokyo 169-8050 (Japan)

    2015-08-15

    Highlights: • The improved core of Super Fast Reactor with high power density is analyzed. • We analyzed four types of the limiting fuel rods. • The influence of Pu enrichment and compressive stress to yield strength ratio are analyzed. • The improved fuel rod design of the new core is suggested. - Abstract: A Super Fast Reactor is a pressure-vessel type, fast spectrum SuperCritical Water Reactor (SCWR) which is presently researched in a Japanese project. A preliminary core has an average power density of 158.8 W/cc. However one of the most important advantages of the Super Fast Reactor is the higher power density compared to the thermal spectrum SCWR, which reduces the capital cost. After the sensitivity analyses on the fuel rod configurations, the fuel assembly configurations and the core configurations, an improved core with an average power density of 294.8 W/cc is designed by 3-D neutronic/thermal-hydraulic coupled calculations. In order to ensure the fuel rod integrity of new core design with high power density, the fuel rod behaviors under normal operating condition are analyzed using fuel performance code FEMAXI-6. The power histories of each fuel rod are taken from the neutronics calculation results in the core design. The cladding surface temperature histories are generated from the thermal-hydraulic calculation results in the core design. Four types of the limiting fuel rods, individually with the Maximum Cladding Surface Temperature (MCST), Maximum Power Peak (MPP), Maximum Discharge Burnup (MDB) and Different Coolant Flow Pattern (DCFP), are chosen to cover all the fuel rods in the core. The available design range of the fuel rod design parameters, such as initial gas plenum pressure, gas plenum position, gas plenum length, grain size and gap size, are found out in order to satisfy the following design criteria: (1) Maximum fuel centerline temperature should be less than 1900 °C. (2) Maximum cladding stress in circumferential direction should

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  6. The physics scheme design for In-Hospital Neutron Irradiator-Mark 1 reactor loaded low-enrichment fuel%低浓化医院中子照射器(IHNI-1)堆芯的物理方案设计

    Institute of Scientific and Technical Information of China (English)

    江新标; 张文首; 高集金; 李义国; 周永茂

    2009-01-01

    采用蒙特卡罗程序MCNP/4B模拟计算了功率为30 kw的低浓化医院中子照射器的堆芯物理参数,设计了合理的堆芯布置方案、~(235)U富集度、控制棒价值、后备反应性和停堆深度,得到固有安全性较高、寿期达10年且无需换料、采用低浓化UO_2燃料的医院中子照射器的堆芯物理设计方案,为后续反应堆工程设计以及硼中子俘获治疗肿瘤用中子束的设计提供理论依据.%The core physics properties of 30 Kw In-Hospital Neutron Irradiator-Mark l(IHNI-l) reactor loaded low-enrichment fuel are simulated by using MCNP/4B Monte Carlo code. The arrangement scheme of core, ~(235)U enrichment, control rods' value, excess reactivity and shutdown margin are also reasonably introduced. The results show that the physics scheme of IHNI-1 reactor which possesses some particular characteristics, such as high inherent safety, 10 years lifetime without any refueling and low-enrichment UO_2 fuel, has been discovered in the paper and it will provide theoretic basis for reactor engineering design and neutron beam design for Boron Neutron Capture Therapy in the future.

  7. FY2016 Ceramic Fuels Development Annual Highlights

    Energy Technology Data Exchange (ETDEWEB)

    Mcclellan, Kenneth James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-01-24

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

  8. FY2015 ceramic fuels development annual highlights

    Energy Technology Data Exchange (ETDEWEB)

    Mcclellan, Kenneth James [Los Alamos National Laboratory (LANL), Los Alamos, NM (United States)

    2015-09-22

    Key challenges for the Advanced Fuels Campaign are the development of fuel technologies to enable major increases in fuel performance (safety, reliability, power and burnup) beyond current technologies, and development of characterization methods and predictive fuel performance models to enable more efficient development and licensing of advanced fuels. Ceramic fuel development activities for fiscal year 2015 fell within the areas of 1) National and International Technical Integration, 2) Advanced Accident Tolerant Ceramic Fuel Development, 3) Advanced Techniques and Reference Materials Development, and 4) Fabrication of Enriched Ceramic Fuels. High uranium density fuels were the focus of the ceramic fuels efforts. Accomplishments for FY15 primarily reflect the prioritization of identification and assessment of new ceramic fuels for light water reactors which have enhanced accident tolerance while also maintaining or improving normal operation performance, and exploration of advanced post irradiation examination techniques which will support more efficient testing and qualification of new fuel systems.

  9. Production of enriched U O{sub 2} fuel pellets for the RA-8 critical facility; Produccion de pastillas combustibles de U O{sub 2} enriquecido para la facilidad critica RA-8

    Energy Technology Data Exchange (ETDEWEB)

    Marajofsky, A.; Thern, G.; Beuter, O.; Altamirano, J.; Perez, Lidia; Taboada, H.; Arnaldo, A.; Benitez, A. [Comision Nacional de Energia Atomica, Buenos Aires (Argentina)

    1997-10-01

    The fabrication of U O{sub 2} fuel pellets of l.8 and 3.4% of U-235 for the RA-8 critical facility involved the scaling-up and the adaptation of the existing fabrication plant to operate with more stringent safety requirements. The installation of new equipment and the construction of a U O{sub 1}2 powder mixing area is described. A new powder press of NCNC control and a full-size sintering furnace were put into operation satisfactorily and up to the present, fuel pellets for 2000 fuel rods were produced. Description of the process development, licensing procedures, quality control and final results are described. (author). 3 refs., 7 figs.

  10. Exploratory Design of a Reactor/Fuel Cycle Using Spent Nuclear Fuel Without Conventional Reprocessing - 13579

    Energy Technology Data Exchange (ETDEWEB)

    Bertch, Timothy C.; Schleicher, Robert W.; Rawls, John D. [General Atomics 3550 General Atomics Court San Diego, CA 92130 (United States)

    2013-07-01

    General Atomics has started design of a waste to energy nuclear reactor (EM2) that can use light water reactor (LWR) spent nuclear fuel (SNF). This effort addresses two problems: using an advanced small reactor with long core life to reduce nuclear energy overnight cost and providing a disposal path for LWR SNF. LWR SNF is re-fabricated into new EM2 fuel using a dry voloxidation process modeled on AIROX/ OREOX processes which remove some of the fission products but no heavy metals. By not removing all of the fission products the fuel remains self-protecting. By not separating heavy metals, the process remains proliferation resistant. Implementation of Energy Multiplier Module (EM2) fuel cycle will provide low cost nuclear energy while providing a long term LWR SNF disposition path which is important for LWR waste confidence. With LWR waste confidence recent impacts on reactor licensing, an alternate disposition path is highly relevant. Centered on a reactor operating at 250 MWe, the compact electricity generating system design maximizes site flexibility with truck transport of all system components and available dry cooling features that removes the need to be located near a body of water. A high temperature system using helium coolant, electricity is efficiently produced using an asynchronous high-speed gas turbine while the LWR SNF is converted to fission products. Reactor design features such as vented fuel and silicon carbide cladding support reactor operation for decades between refueling, with improved fuel utilization. Beyond the reactor, the fuel cycle is designed so that subsequent generations of EM2 reactor fuel will use the previous EM2 discharge, providing its own waste confidence plus eliminating the need for enrichment after the first generation. Additional LWR SNF is added at each re-fabrication to replace the removed fission products. The fuel cycle uses a dry voloxidation process for both the initial LWR SNF re-fabrication and later for EM2

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

    Science.gov (United States)

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

    2016-10-01

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

  12. Verification of 235U mass content in nuclear fuel plates by an absolute method

    Science.gov (United States)

    El-Gammal, W.

    2007-01-01

    Nuclear Safeguards is referred to a verification System by which a State can control all nuclear materials (NM) and nuclear activities under its authority. An effective and efficient Safeguards System must include a system of measurements with capabilities sufficient to verify such NM. Measurements of NM using absolute methods could eliminate the dependency on NM Standards, which are necessary for other relative or semi-absolute methods. In this work, an absolute method has been investigated to verify the 235U mass content in nuclear fuel plates of Material Testing Reactor (MTR) type. The most intense gamma-ray signature at 185.7 keV emitted after α-decay of the 235U nuclei was employed in the method. The measuring system (an HPGe-spectrometer) was mathematically calibrated for efficiency using the general Monte Carlo transport code MCNP-4B. The calibration results and the measured net count rate were used to estimate the 235U mass content in fuel plates at different detector-to-fuel plate distances. Two sets of fuel plates, containing natural and low enriched uranium, were measured at the Fuel Fabrication Facility. Average accuracies for the estimated 235U masses of about 2.62% and 0.3% are obtained for the fuel plates containing natural and low enriched uranium; respectively, with a precision of about 3%.

  13. Physical Theories with Average Symmetry

    OpenAIRE

    Alamino, Roberto C.

    2013-01-01

    This Letter probes the existence of physical laws invariant only in average when subjected to some transformation. The concept of a symmetry transformation is broadened to include corruption by random noise and average symmetry is introduced by considering functions which are invariant only in average under these transformations. It is then shown that actions with average symmetry obey a modified version of Noether's Theorem with dissipative currents. The relation of this with possible violat...

  14. Verification Calculation Results to Validate the Procedures and Codes for Pin-by-Pin Power Computation in VVER Type Reactors with MOX Fuel Loading

    Energy Technology Data Exchange (ETDEWEB)

    Chizhikova, Z.N.; Kalashnikov, A.G.; Kapranova, E.N.; Korobitsyn, V.E.; Manturov, G.N.; Tsiboulia, A.A.

    1998-12-01

    One of the important problems for ensuring the VVER type reactor safety when the reactor is partially loaded with MOX fuel is the choice of appropriate physical zoning to achieve the maximum flattening of pin-by-pin power distribution. When uranium fuel is replaced by MOX one provided that the reactivity due to fuel assemblies is kept constant, the fuel enrichment slightly decreases. However, the average neutron spectrum fission microscopic cross-section for {sup 239}Pu is approximately twice that for {sup 235}U. Therefore power peaks occur in the peripheral fuel assemblies containing MOX fuel which are aggravated by the interassembly water. Physical zoning has to be applied to flatten the power peaks in fuel assemblies containing MOX fuel. Moreover, physical zoning cannot be confined to one row of fuel elements as is the case with a uniform lattice of uranium fuel assemblies. Both the water gap and the jump in neutron absorption macroscopic cross-sections which occurs at the interface of fuel assemblies with different fuels make the problem of calculating space-energy neutron flux distribution more complicated since it increases nondiffusibility effects. To solve this problem it is necessary to update the current codes, to develop new codes and to verify all the codes including nuclear-physical constants libraries employed. In so doing it is important to develop and validate codes of different levels--from design codes to benchmark ones. This paper presents the results of the burnup calculation for a multiassembly structure, consisting of MOX fuel assemblies surrounded by uranium dioxide fuel assemblies. The structure concerned can be assumed to model a fuel assembly lattice symmetry element of the VVER-1000 type reactor in which 1/4 of all fuel assemblies contains MOX fuel.

  15. TMI Fuel Characteristics for Disposal Criticality Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Larry L. Taylor

    2003-09-01

    This report documents the reported contents of the Three Mile Island Unit 2 (TMI-2) canisters. proposed packaging, and degradation scenarios expected in the repository. Most fuels within the U.S. Department of Energy spent nuclear fuel inventory deal with highly enriched uranium, that in most cases require some form of neutronic poisoning inside the fuel canister. The TMI-2 fuel represents a departure from these fuel forms due to its lower enrichment (2.96% max.) values and the disrupted nature of the fuel itself. Criticality analysis of these fuel canisters has been performed over the years to reflect conditions expected during transit from the reactor to the Idaho National Engineering and Environmental Laboratory, water pool storage,1 and transport/dry-pack storage at Idaho Nuclear Technology and Engineering Center.2,3 None of these prior analyses reflect the potential disposal conditions for this fuel inside a postclosure repository.

  16. Corrosion Minimization for Research Reactor Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Eric Shaber; Gerard Hofman

    2005-06-01

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

  17. Average Convexity in Communication Situations

    NARCIS (Netherlands)

    Slikker, M.

    1998-01-01

    In this paper we study inheritance properties of average convexity in communication situations. We show that the underlying graph ensures that the graphrestricted game originating from an average convex game is average convex if and only if every subgraph associated with a component of the underlyin

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  19. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    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.

  20. Sampling Based Average Classifier Fusion

    Directory of Open Access Journals (Sweden)

    Jian Hou

    2014-01-01

    fusion algorithms have been proposed in literature, average fusion is almost always selected as the baseline for comparison. Little is done on exploring the potential of average fusion and proposing a better baseline. In this paper we empirically investigate the behavior of soft labels and classifiers in average fusion. As a result, we find that; by proper sampling of soft labels and classifiers, the average fusion performance can be evidently improved. This result presents sampling based average fusion as a better baseline; that is, a newly proposed classifier fusion algorithm should at least perform better than this baseline in order to demonstrate its effectiveness.

  1. Fuel lattice design in a boiling water reactor using an ant-colony-based system

    Energy Technology Data Exchange (ETDEWEB)

    Montes, Jose Luis, E-mail: joseluis.montes@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carr. Mexico-Toluca S/N, La Marquesa, Ocoyoacac, Estado de Mexico, CP 52750 (Mexico); Facultad de Ciencias, Universidad Autonoma del Estado de Mexico (Mexico); Francois, Juan-Luis, E-mail: juan.luis.francois@gmail.com [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532, Jiutepec, Mor., CP 62550 (Mexico); Ortiz, Juan Jose, E-mail: juanjose.ortiz@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carr. Mexico-Toluca S/N, La Marquesa, Ocoyoacac, Estado de Mexico, CP 52750 (Mexico); Martin-del-Campo, Cecilia, E-mail: cecilia.martin.del.campo@gmail.com [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532, Jiutepec, Mor., CP 62550 (Mexico); Perusquia, Raul, E-mail: raul.perusquia@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carr. Mexico-Toluca S/N, La Marquesa, Ocoyoacac, Estado de Mexico, CP 52750 (Mexico)

    2011-06-15

    Research highlights: > We present an ant-colony-based system for BWR fuel lattice design and optimization. > Assessment of candidate solutions at 0.0 MWd/kg {sup 235}U seems to have a limited scope. > Suitable heuristic rules enable more realistic fuel lattice designs. > The election of the objective has a large impact in CPU time. > ACS enables an important decrease of the initial average U-235 enrichment. - Abstract: This paper presents a new approach to deal with the boiling water reactor radial fuel lattice design. The goal is to optimize the distribution of both, the fissionable material, and the reactivity control poison material inside the fuel lattice at the beginning of its life. An ant-colony-based system was used to search for either: the optimum location of the poisoned pin inside the lattice, or the U{sup 235} enrichment and Gd{sub 2}O{sub 3} concentrations. In the optimization process, in order to know the parameters of the candidate solutions, the neutronic simulator CASMO-4 transport code was used. A typical 10 x 10 BWR fuel lattice with an initial average U{sup 235} enrichment of 4.1%, used in the current operation of Laguna Verde Nuclear Power Plant was taken as a reference. With respect to that reference lattice, it was possible to decrease the average U{sup 235} enrichment up to 3.949%, this obtained value represents a decrease of 3.84% with respect to the reference U{sup 235} enrichment; whereas, the k-infinity was inside the {+-}100 pcm's range, and there was a difference of 0.94% between the local power peaking factor and the lattice reference value. Particular emphasis was made on defining the objective function which is used for making the assessment of candidate solutions. In a typical desktop personal computer, about four hours of CPU time were necessary for the algorithm to fulfill the goals of the optimization process. The results obtained with the application of the implemented system showed that the proposed approach represents a

  2. A Model to Predict Thermal Conductivity of Irradiated U-Mo Dispersion Fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-01

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

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

    Science.gov (United States)

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

    2016-05-01

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

  4. Profile of World Uranium Enrichment Programs-2009

    Energy Technology Data Exchange (ETDEWEB)

    Laughter, Mark D [ORNL

    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

  5. Physical Theories with Average Symmetry

    CERN Document Server

    Alamino, Roberto C

    2013-01-01

    This Letter probes the existence of physical laws invariant only in average when subjected to some transformation. The concept of a symmetry transformation is broadened to include corruption by random noise and average symmetry is introduced by considering functions which are invariant only in average under these transformations. It is then shown that actions with average symmetry obey a modified version of Noether's Theorem with dissipative currents. The relation of this with possible violations of physical symmetries, as for instance Lorentz invariance in some quantum gravity theories, is briefly commented.

  6. The Study of Numerical Simulation of Oxygen-‎enriched Burner System

    Directory of Open Access Journals (Sweden)

    Yuesheng Fan

    2010-12-01

    Full Text Available In order to reduce overall fuel consumption, or partially substitute a “valuable” fuel with a ‎poor one, in electric power plant boilers, oxygen enrichment of combustion air can be very ‎effective. The paper proposes an oxygen-enriched ignition system which based on the ‎existing pulverized coal fired boiler ignition devices. Small coal particle is suitable for this ‎system. The new burner includes inside, outside and middle casings. And it transfer heat in ‎two ways of downstream and upstream. The burner has authorized a patent in China. A ‎numerical simulation theory were used to analysis it. The results indicate that: it can ‎increase the maximum burning velocity ‎ ‎ and the average burning ‎velocity ‎, and decrease ignition temperature Ti and burnout temperature Tb of ‎pulverized coal. In addition, the pulverized coal fired boilers are easier to be ignited and the ‎comprehensive combustibility index S is improved. At the same time, it demonstrates that it ‎is an effective way to warm-up the pulverized coal in ignition of the boiler in the power ‎plant.‎

  7. 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

    Energy Technology Data Exchange (ETDEWEB)

    Stillman, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Feldman, E. [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Foyto, L [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Kutikkad, K [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; McKibben, J C [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Peters, N. [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Stevens, J. [Argonne National Lab. (ANL), Argonne, IL (United States)

    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.

  8. Uranium Enrichment Reduction in the PGSFR Core Design

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chihyung; Hartanto, Donny; Kim, Yonghee [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    Korea is currently developing the so-called Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) to investigate and demonstrate the capability of TRU transmutation. However, since fuel recycling technology is still at early development in Korea and also due to lack of experience in burning TRU in a fast reactor, the initial core of PGSFR is loaded with low-enriched uranium (LEU) fuel. Several test assemblies containing TRU fuels are supposed to be irradiated and tested for future TRU fuel developments. The uranium enrichment in the LEU PGSFR core is high, about 19.20%, due to large neutron leakage and low conversion ratio. In this paper, the required uranium enrichment is reduced by replacing the reflector material and modifying the reflector geometry in order to decrease the fuel cost of the LEU PGSFR core. PbO is chosen as the reflector material to replace the current HT9 and an inverted reflector assembly is also investigated in this study. It is shown that longer cycle length, higher fuel burnup and flattening power distribution can be achieved with PbO reflector and enhanced neutron leakage can be handled by the optimization of shielding material or core geometry. PbO reflector with inverted geometry is suggest in this research and by using inverted PbO reflector, core performance can be improved while leakage is negligibly enhanced than conventional pin type reflector assembly. Research about reducing the uranium enrichment more by increasing the uranium content in the uranium fuel which is U-10Zr now or increasing the smeared density which is currently 75% can be considered as a future work. Detailed analysis about multi-batch fuel management should be carried out since currently it is done approximately by using linear reactivity theory. Also, analysis for PGSFR with various reflector materials like LME, liquid lead will be carried out and the chemical reaction of those materials including PbO with sodium should be carefully investigated.

  9. Quantized average consensus with delay

    NARCIS (Netherlands)

    Jafarian, Matin; De Persis, Claudio

    2012-01-01

    Average consensus problem is a special case of cooperative control in which the agents of the network asymptotically converge to the average state (i.e., position) of the network by transferring information via a communication topology. One of the issues of the large scale networks is the cost of co

  10. Development of a Monolithic Research Reactor Fuel Type at Argonne National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Clark, C.R.; Briggs, R.J.

    2004-10-06

    The Reduced Enrichment for Research and Test Reactors (RERTR) program has been tasked with the conversion of research reactors from highly enriched to low-enriched uranium (LEU). To convert several high power reactors, monolithic fuel, a new fuel type, is being developed. This fuel type replaces the standard fuel dispersion with a fuel alloy foil, which allows for fuel densities far in excess of that found in dispersion fuel. The single-piece fuel foil also contains a significantly lower interface area between the fuel and the aluminum in the plate than the standard fuel type, limiting the amount of detrimental fuel-aluminum interaction that can occur. Implementation of monolithic fuel is dependant on the development of a suitable fabrication method as traditional roll-bonding techniques are inadequate.

  11. 77 FR 33253 - Regulatory Guide 8.24, Revision 2, Health Physics Surveys During Enriched Uranium-235 Processing...

    Science.gov (United States)

    2012-06-05

    ... COMMISSION Regulatory Guide 8.24, Revision 2, Health Physics Surveys During Enriched Uranium-235 Processing... CONTACT: Gregory Chapman, Uranium Enrichment Branch, Division of Fuel Cycle Safety and Safeguards, Office.... Introduction Revision 2 of Regulatory Guide 8.24, ``Health Physics Surveys During Enriched...

  12. Enriching the Catalog

    Science.gov (United States)

    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…

  13. Gaussian moving averages and semimartingales

    DEFF Research Database (Denmark)

    Basse-O'Connor, Andreas

    2008-01-01

    In the present paper we study moving averages (also known as stochastic convolutions) driven by a Wiener process and with a deterministic kernel. Necessary and sufficient conditions on the kernel are provided for the moving average to be a semimartingale in its natural filtration. Our results...... are constructive - meaning that they provide a simple method to obtain kernels for which the moving average is a semimartingale or a Wiener process. Several examples are considered. In the last part of the paper we study general Gaussian processes with stationary increments. We provide necessary and sufficient...

  14. Spent fuel characteristics & disposal considerations

    Energy Technology Data Exchange (ETDEWEB)

    Oversby, V.M.

    1996-06-01

    The fuel used in commercial nuclear power reactors is uranium, generally in the form of an oxide. The gas-cooled reactors developed in England use metallic uranium enclosed in a thin layer of Magnox. Since this fuel must be processed into a more stable form before disposal, we will not consider the characteristics of the Magnox spent fuel. The vast majority of the remaining power reactors in the world use uranium dioxide pellets in Zircaloy cladding as the fuel material. Reactors that are fueled with uranium dioxide generally use water as the moderator. If ordinary water is used, the reactors are called Light Water Reactors (LWR), while if water enriched in the deuterium isotope of hydrogen is used, the reactors are called Heavy Water reactors. The LWRs can be either pressurized reactors (PWR) or boiling water reactors (BWR). Both of these reactor types use uranium that has been enriched in the 235 isotope to about 3.5 to 4% total abundance. There may be minor differences in the details of the spent fuel characteristics for PWRs and BWRs, but for simplicity we will not consider these second-order effects. The Canadian designed reactor (CANDU) that is moderated by heavy water uses natural uranium without enrichment of the 235 isotope as the fuel. These reactors run at higher linear power density than LWRs and produce spent fuel with lower total burn-up than LWRs. Where these difference are important with respect to spent fuel management, we will discuss them. Otherwise, we will concentrate on spent fuel from LWRs.

  15. Characterization of used nuclear fuel with multivariate analysis for process monitoring

    Science.gov (United States)

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

    2014-01-01

    This paper presents initial development of a reactor-type classifier that is used to select a reactor-specific partial least squares model to predict used nuclear fuel burnup. Nuclide activities for prototypic used fuel samples were generated in ORIGEN-ARP and used to investigate techniques to characterize used nuclear fuel in terms of reactor type (pressurized or boiling water reactor) and burnup. A variety of reactor type classification algorithms, including k-nearest neighbors, linear and quadratic discriminant analyses, and support vector machines, were evaluated to differentiate used fuel from pressurized and boiling water reactors. Then, reactor type-specific partial least squares models were developed to predict the burnup of the fuel. Using these reactor type-specific models instead of a model trained for all light water reactors improved the accuracy of burnup predictions. The developed classification and prediction models were combined and applied to a large dataset that included eight fuel assembly designs, two of which were not used in training the models, and spanned the range of the initial 235U enrichment, cooling time, and burnup values expected of future commercial used fuel for reprocessing. Error rates were consistent across the range of considered enrichment, cooling time, and burnup values. Average absolute relative errors in burnup predictions for validation data both within and outside the training space were 0.0574% and 0.0597%, respectively. The errors seen in this work are artificially low, because the models were trained, optimized, and tested on simulated, noise-free data. However, these results indicate that the developed models may generalize well to new data and that the proposed approach constitutes a viable first step in developing a fuel characterization algorithm based on gamma spectra.

  16. Comparative Analysis of Structural Changes In U-Mo Dispersed Fuel of Full-Size Fuel Elements And Mini-Rods Irradiated In The MIR Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Izhutov, Aleksey L.; Iakovlev, Valeriy V.; Novoselov, Andrey E. and others

    2013-12-15

    The paper summarizes the irradiation test and post-irradiation examination (PIE) data for the U-Mo low-enriched fuel that was irradiated in the MIR reactor under the RERTR Program. The PIE data were analyzed for both full-size fuel rods and mini-rods with atomized powder dispersed in Al matrix as well as with additions of 2%, 5% and 13% of silicon in the matrix and ZrN protective coating on the fuel particles. The full-size fuel rods were irradiated up to an average burnup of ∼ 60%{sup 235}U; the mini-rods were irradiated to an average burnup of ∼ 85%{sup 235}U. The presented data show a significant increase of the void fraction in the U-Mo alloy as the U-235 burnup rises from ∼ 40% up to ∼ 85%. The effect of irradiation test conditions and U-235 burnup were analyzed with regard to the formation of an interaction layer between the matrix and fuel particles as well as generation of porosity in the U-Mo alloy. Shown here are changes in distribution of U fission products as the U-235 burnup increases from ∼ 40% up to ∼ 85%.

  17. Radial optimization of a BWR fuel cell using genetic algorithms; Optimizacion radial de una celda de combustible BWR usando algoritmos geneticos

    Energy Technology Data Exchange (ETDEWEB)

    Martin del Campo M, C.; Carmona H, R.; Oropeza C, I.P. [UNAM, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)]. e-mail: cmcm@fi-b.unam.mx

    2006-07-01

    The development of the application of the Genetic Algorithms (GA) to the optimization of the radial distribution of enrichment in a cell of fuel of a BWR (Boiling Water Reactor) is presented. The optimization process it was ties to the HELIOS simulator, which is a transport code of neutron simulation of fuel cells that has been validated for the calculation of nuclear banks for BWRs. With heterogeneous radial designs can improve the radial distribution of the power, for what the radial design of fuel has a strong influence in the global design of fuel recharges. The optimum radial distribution of fuel bars is looked for with different enrichments of U{sup 235} and contents of consumable poison. For it is necessary to define the representation of the solution, the objective function and the implementation of the specific optimization process to the solution of the problem. The optimization process it was coded in 'C' language, it was automated the creation of the entrances to the simulator, the execution of the simulator and the extraction, in the exit of the simulator, of the parameters that intervene in the objective function. The objective function includes four parameters: average enrichment of the cell, average gadolinia concentration of the cell, peak factor of radial power and k-infinite multiplication factor. To be able to calculate the parameters that intervene in the objective function, the one evaluation process of GA was ties to the HELIOS code executed in a Compaq Alpha workstation. It was applied to the design of a fuel cell of 10 x 10 that it can be employee in the fuel assemble designs that are used at the moment in the Laguna Verde Nucleo electric Central. Its were considered 10 different fuel compositions which four contain gadolinia. Three heuristic rules that consist in prohibiting the placement of bars with gadolinia in the ends of the cell, to place the compositions with the smallest enrichment in the corners of the cell and to fix

  18. Vocal attractiveness increases by averaging.

    Science.gov (United States)

    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.

  19. Nuclear Fuel Reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Harold F. McFarlane; Terry Todd

    2013-11-01

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

  20. 76 FR 72984 - Revised Application for a License To Export High-Enriched Uranium

    Science.gov (United States)

    2011-11-28

    ... COMMISSION Revised Application for a License To Export High-Enriched Uranium The application for a license to export high-enriched Uranium has been revised as noted below. Notice of this application was previously... kilograms To fabricate fuel France. Security Complex; October 18, Uranium (93.35%). uranium (174.0...

  1. Fuel Fabrication and Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Karpius, Peter Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-02-02

    The uranium from the enrichment plant is still in the form of UF6. UF6 is not suitable for use in a reactor due to its highly corrosive chemistry as well as its phase diagram. UF6 is converted into UO2 fuel pellets, which are in turn placed in fuel rods and assemblies. Reactor designs are variable in moderators, coolants, fuel, performance etc.The dream of energy ‘too-cheap to meter’ is no more, and now the nuclear power industry is pushing ahead with advanced reactor designs.

  2. Averaged Electroencephalic Audiometry in Infants

    Science.gov (United States)

    Lentz, William E.; McCandless, Geary A.

    1971-01-01

    Normal, preterm, and high-risk infants were tested at 1, 3, 6, and 12 months of age using averaged electroencephalic audiometry (AEA) to determine the usefulness of AEA as a measurement technique for assessing auditory acuity in infants, and to delineate some of the procedural and technical problems often encountered. (KW)

  3. Ergodic averages via dominating processes

    DEFF Research Database (Denmark)

    Møller, Jesper; Mengersen, Kerrie

    2006-01-01

    We show how the mean of a monotone function (defined on a state space equipped with a partial ordering) can be estimated, using ergodic averages calculated from upper and lower dominating processes of a stationary irreducible Markov chain. In particular, we do not need to simulate the stationary ...

  4. Heavy ozone enrichments from MIPAS limb emission spectra

    Directory of Open Access Journals (Sweden)

    A. Dudhia

    2009-11-01

    Full Text Available Vertical enrichment profiles of stratospheric heavy ozone (asymmetric and symmetric 50O3 isotopomers have been derived from the Michelson Interferometer for Passive Atmospheric Sounding (MIPAS. As a continuously operating satellite instrument, MIPAS has an advantage over previous measurements in providing global coverage over an extended time period, allowing the variations in enrichment to be determined. However, since the spectral features of the isotopomers are comparable with the magnitude of the instrument noise, a sequential estimation retrieval scheme has been used to average the measurements zonally.

    The magnitude of observed enrichments is ~10% in stratosphere, a value consistent with previous observations and lab measurements. The data indicate that asymmetric heavy ozone is significantly more enriched than the symmetric isotopomer and show latitudinal and vertical variations in both the asymmetric and symmetric 50O3 enrichments. Enrichments decrease with decreasing temperature, as previously noted, at midlatidudes and polar regions, while they increase with decreasing temperature in equatorial regions.

    MIPAS measurements contain also information on the asymmetric and symmetric 49O3 isotopomers. The enrichments of 49O3, obtained with the same sequential estimator approach, are much larger than the enrichments of 50O3, in contraddictions with previous results which predict larger enrichments for heavier isotopomers.

  5. Measurement of the average lifetime of b hadrons

    Science.gov (United States)

    Adriani, O.; Aguilar-Benitez, M.; Ahlen, S.; Alcaraz, J.; Aloisio, A.; Alverson, G.; Alviggi, M. G.; Ambrosi, G.; An, Q.; Anderhub, H.; Anderson, A. L.; Andreev, V. P.; Angelescu, T.; Antonov, L.; Antreasyan, D.; Arce, P.; Arefiev, A.; Atamanchuk, A.; Azemoon, T.; Aziz, T.; Baba, P. V. K. S.; Bagnaia, P.; Bakken, J. A.; Ball, R. C.; Banerjee, S.; Bao, J.; Barillère, R.; Barone, L.; Baschirotto, A.; Battiston, R.; Bay, A.; Becattini, F.; Bechtluft, J.; Becker, R.; Becker, U.; Behner, F.; Behrens, J.; Bencze, Gy. L.; Berdugo, J.; Berges, P.; Bertucci, B.; Betev, B. L.; Biasini, M.; Biland, A.; Bilei, G. M.; Bizzarri, R.; Blaising, J. J.; Bobbink, G. J.; Bock, R.; Böhm, A.; Borgia, B.; Bosetti, M.; Bourilkov, D.; Bourquin, M.; Boutigny, D.; Bouwens, B.; Brambilla, E.; Branson, J. G.; Brock, I. C.; Brooks, M.; Bujak, A.; Burger, J. D.; Burger, W. J.; Busenitz, J.; Buytenhuijs, A.; Cai, X. D.; Capell, M.; Caria, M.; Carlino, G.; Cartacci, A. M.; Castello, R.; Cerrada, M.; Cesaroni, F.; Chang, Y. H.; Chaturvedi, U. K.; Chemarin, M.; Chen, A.; Chen, C.; Chen, G.; Chen, G. M.; Chen, H. F.; Chen, H. S.; Chen, M.; Chen, W. Y.; Chiefari, G.; Chien, C. Y.; Choi, M. T.; Chung, S.; Civinini, C.; Clare, I.; Clare, R.; Coan, T. E.; Cohn, H. O.; Coignet, G.; Colino, N.; Contin, A.; Costantini, S.; Cotorobai, F.; Cui, X. T.; Cui, X. Y.; Dai, T. S.; D'Alessandro, R.; de Asmundis, R.; Degré, A.; Deiters, K.; Dénes, E.; Denes, P.; DeNotaristefani, F.; Dhina, M.; DiBitonto, D.; Diemoz, M.; Dimitrov, H. R.; Dionisi, C.; Ditmarr, M.; Djambazov, L.; Dova, M. T.; Drago, E.; Duchesneau, D.; Duinker, P.; Duran, I.; Easo, S.; El Mamouni, H.; Engler, A.; Eppling, F. J.; Erné, F. C.; Extermann, P.; Fabbretti, R.; Fabre, M.; Falciano, S.; Fan, S. J.; Fackler, O.; Fay, J.; Felcini, M.; Ferguson, T.; Fernandez, D.; Fernandez, G.; Ferroni, F.; Fesefeldt, H.; Fiandrini, E.; Field, J. H.; Filthaut, F.; Fisher, P. H.; Forconi, G.; Fredj, L.; Freudenreich, K.; Friebel, W.; Fukushima, M.; Gailloud, M.; Galaktionov, Yu.; Gallo, E.; Ganguli, S. N.; Garcia-Abia, P.; Gele, D.; Gentile, S.; Gheordanescu, N.; Giagu, S.; Goldfarb, S.; Gong, Z. F.; Gonzalez, E.; Gougas, A.; Goujon, D.; Gratta, G.; Gruenewald, M.; Gu, C.; Guanziroli, M.; Guo, J. K.; Gupta, V. K.; Gurtu, A.; Gustafson, H. R.; Gutay, L. J.; Hangarter, K.; Hartmann, B.; Hasan, A.; Hauschildt, D.; He, C. F.; He, J. T.; Hebbeker, T.; Hebert, M.; Hervé, A.; Hilgers, K.; Hofer, H.; Hoorani, H.; Hu, G.; Hu, G. Q.; Ille, B.; Ilyas, M. M.; Innocente, V.; Janssen, H.; Jezequel, S.; Jin, B. N.; Jones, L. W.; Josa-Mutuberria, I.; Kasser, A.; Khan, R. A.; Kamyshkov, Yu.; Kapinos, P.; Kapustinsky, J. S.; Karyotakis, Y.; Kaur, M.; Khokhar, S.; Kienzle-Focacci, M. N.; Kim, J. K.; Kim, S. C.; Kim, Y. G.; Kinnison, W. W.; Kirkby, A.; Kirkby, D.; Kirsch, S.; Kittel, W.; Klimentov, A.; Klöckner, R.; König, A. C.; Koffeman, E.; Kornadt, O.; Koutsenko, V.; Koulbardis, A.; Kraemer, R. W.; Kramer, T.; Krastev, V. R.; Krenz, W.; Krivshich, A.; Kuijten, H.; Kumar, K. S.; Kunin, A.; Landi, G.; Lanske, D.; Lanzano, S.; Lebedev, A.; Lebrun, P.; Lecomte, P.; Lecoq, P.; Le Coultre, P.; Lee, D. M.; Lee, J. S.; Lee, K. Y.; Leedom, I.; Leggett, C.; Le Goff, J. M.; Leiste, R.; Lenti, M.; Leonardi, E.; Li, C.; Li, H. T.; Li, P. J.; Liao, J. Y.; Lin, W. T.; Lin, Z. Y.; Linde, F. L.; Lindemann, B.; Lista, L.; Liu, Y.; Lohmann, W.; Longo, E.; Lu, Y. S.; Lubbers, J. M.; Lübelsmeyer, K.; Luci, C.; Luckey, D.; Ludovici, L.; Luminari, L.; Lustermann, W.; Ma, J. M.; Ma, W. G.; MacDermott, M.; Malik, R.; Malinin, A.; Maña, C.; Maolinbay, M.; Marchesini, P.; Marion, F.; Marin, A.; Martin, J. P.; Martinez-Laso, L.; Marzano, F.; Massaro, G. G. G.; Mazumdar, K.; McBride, P.; McMahon, T.; McNally, D.; Merk, M.; Merola, L.; Meschini, M.; Metzger, W. J.; Mi, Y.; Mihul, A.; Mills, G. B.; Mir, Y.; Mirabelli, G.; Mnich, J.; Möller, M.; Monteleoni, B.; Morand, R.; Morganti, S.; Moulai, N. E.; Mount, R.; Müller, S.; Nadtochy, A.; Nagy, E.; Napolitano, M.; Nessi-Tedaldi, F.; Newman, H.; Neyer, C.; Niaz, M. A.; Nippe, A.; Nowak, H.; Organtini, G.; Pandoulas, D.; Paoletti, S.; Paolucci, P.; Pascale, G.; Passaleva, G.; Patricelli, S.; Paul, T.; Pauluzzi, M.; Paus, C.; Pauss, F.; Pei, Y. J.; Pensotti, S.; Perret-Gallix, D.; Perrier, J.; Pevsner, A.; Piccolo, D.; Pieri, M.; Piroué, P. A.; Plasil, F.; Plyaskin, V.; Pohl, M.; Pojidaev, V.; Postema, H.; Qi, Z. D.; Qian, J. M.; Qureshi, K. N.; Raghavan, R.; Rahal-Callot, G.; Rancoita, P. G.; Rattaggi, M.; Raven, G.; Razis, P.; Read, K.; Ren, D.; Ren, Z.; Rescigno, M.; Reucroft, S.; Ricker, A.; Riemann, S.; Riemers, B. C.; Riles, K.; Rind, O.; Rizvi, H. A.; Ro, S.; Rodriguez, F. J.; Roe, B. P.; Röhner, M.; Romero, L.; Rosier-Lees, S.; Rosmalen, R.; Rosselet, Ph.; van Rossum, W.; Roth, S.; Rubbia, A.; Rubio, J. A.; Rykaczewski, H.; Sachwitz, M.; Salicio, J.; Salicio, J. M.; Sanders, G. S.; Santocchia, A.; Sarakinos, M. S.; Sartorelli, G.; Sassowsky, M.; Sauvage, G.; Schegelsky, V.; Schmitz, D.; Schmitz, P.; Schneegans, M.; Schopper, H.; Schotanus, D. J.; Shotkin, S.; Schreiber, H. J.; Shukla, J.; Schulte, R.; Schulte, S.; Schultze, K.; Schwenke, J.; Schwering, G.; Sciacca, C.; Scott, I.; Sehgal, R.; Seiler, P. G.; Sens, J. C.; Servoli, L.; Sheer, I.; Shen, D. Z.; Shevchenko, S.; Shi, X. R.; Shumilov, E.; Shoutko, V.; Son, D.; Sopczak, A.; Soulimov, V.; Spartiotis, C.; Spickermann, T.; Spillantini, P.; Starosta, R.; Steuer, M.; Stickland, D. P.; Sticozzi, F.; Stone, H.; Strauch, K.; Stringfellow, B. C.; Sudhakar, K.; Sultanov, G.; Sun, L. Z.; Susinno, G. F.; Suter, H.; Swain, J. D.; Syed, A. A.; Tang, X. W.; Taylor, L.; Terzi, G.; Ting, Samuel C. C.; Ting, S. M.; Tonutti, M.; Tonwar, S. C.; Tóth, J.; Tsaregorodtsev, A.; Tsipolitis, G.; Tully, C.; Tung, K. L.; Ulbricht, J.; Urbán, L.; Uwer, U.; Valente, E.; Van de Walle, R. T.; Vetlitsky, I.; Viertel, G.; Vikas, P.; Vikas, U.; Vivargent, M.; Vogel, H.; Vogt, H.; Vorobiev, I.; Vorobyov, A. A.; Vuilleumier, L.; Wadhwa, M.; Wallraff, W.; Wang, C.; Wang, C. R.; Wang, X. L.; Wang, Y. F.; Wang, Z. M.; Warner, C.; Weber, A.; Weber, J.; Weill, R.; Wenaus, T. J.; Wenninger, J.; White, M.; Willmott, C.; Wittgenstein, F.; Wright, D.; Wu, S. X.; Wynhoff, S.; Wysłouch, B.; Xie, Y. Y.; Xu, J. G.; Xu, Z. Z.; Xue, Z. L.; Yan, D. S.; Yang, B. Z.; Yang, C. G.; Yang, G.; Ye, C. H.; Ye, J. B.; Ye, Q.; Yeh, S. C.; Yin, Z. W.; You, J. M.; Yunus, N.; Yzerman, M.; Zaccardelli, C.; Zaitsev, N.; Zemp, P.; Zeng, M.; Zeng, Y.; Zhang, D. H.; Zhang, Z. P.; Zhou, B.; Zhou, G. J.; Zhou, J. F.; Zhu, R. Y.; Zichichi, A.; van der Zwaan, B. C. C.; L3 Collaboration

    1993-11-01

    The average lifetime of b hadrons has been measured using the L3 detector at LEP, running at √ s ≈ MZ. A b-enriched sample was obtained from 432538 hadronic Z events collected in 1990 and 1991 by tagging electrons and muons from semileptonic b hadron decays. From maximum likelihood fits to the electron and muon impact parameter distributions, the average b hadron lifetime was measured to be τb = (1535 ± 35 ± 28) fs, where the first error is statistical and the second includes both the experimental and the theoretical systematic uncertainties.

  6. REVIEW OF FAST FLUX TEST FACILITY (FFTF) FUEL EXPERIMENTS FOR STORAGE IN INTERIM STORAGE CASKS (ISC)

    Energy Technology Data Exchange (ETDEWEB)

    CHASTAIN, S.A.

    2005-10-24

    Appendix H, Section H.3.3.10.11 of the Final Safety Analysis Report (FSAR), provides the limits to be observed for fueled components authorized for storage in the Fast Flux Test Facility (FFTF) spent fuel storage system. Currently, the authorization basis allows standard driver fuel assemblies (DFA), as described in the FSAR Chapter 17, Section 17.5.3.1, to be stored provided decay power per assembly is {le} 250 watts, post-irradiation time is four years minimum, average assembly burn-up is 150,000 MWD/MTHM maximum and the pre-irradiation enrichment is 29.3% maximum (per H.3.3.10.11). In addition, driver evaluation (DE), core characterizer assemblies (CCA), and run-to-cladding-breach (RTCB) assemblies are included based on their similarities to a standard DFA. Ident-69 pin containers with fuel pins from these DFAs can also be stored. Section H.3.3.10.11 states that fuel types outside the specification criteria above will be addressed on a case-by-case basis. There are many different types of fuel and blanket experiments that were irradiated in the FFTF which now require offload to the spent fuel storage system. Two reviews were completed for a portion of these special type fuel components to determine if placement into the Core Component Container (CCC)/Interim Storage Cask (ISC) would require any special considerations or changes to the authorization basis. Project mission priorities coupled with availability of resources and analysts prevented these evaluations from being completed as a single effort. Areas of review have included radiological accident release consequences, radiological shielding adequacy, criticality safety, thermal limits, confinement, and stress. The results of these reviews are available in WHC-SD-FF-RPT-005, Rev. 0 and 1, ''Review of FFTF Fuel Experiments for Storage at ISA'', (Reference I), which subsequently allowed a large portion of these components to be included in the authorization basis (Table H.3.3-21). The

  7. Dependability in Aggregation by Averaging

    CERN Document Server

    Jesus, Paulo; Almeida, Paulo Sérgio

    2010-01-01

    Aggregation is an important building block of modern distributed applications, allowing the determination of meaningful properties (e.g. network size, total storage capacity, average load, majorities, etc.) that are used to direct the execution of the system. However, the majority of the existing aggregation algorithms exhibit relevant dependability issues, when prospecting their use in real application environments. In this paper, we reveal some dependability issues of aggregation algorithms based on iterative averaging techniques, giving some directions to solve them. This class of algorithms is considered robust (when compared to common tree-based approaches), being independent from the used routing topology and providing an aggregation result at all nodes. However, their robustness is strongly challenged and their correctness often compromised, when changing the assumptions of their working environment to more realistic ones. The correctness of this class of algorithms relies on the maintenance of a funda...

  8. High average power supercontinuum sources

    Indian Academy of Sciences (India)

    J C Travers

    2010-11-01

    The physical mechanisms and basic experimental techniques for the creation of high average spectral power supercontinuum sources is briefly reviewed. We focus on the use of high-power ytterbium-doped fibre lasers as pump sources, and the use of highly nonlinear photonic crystal fibres as the nonlinear medium. The most common experimental arrangements are described, including both continuous wave fibre laser systems with over 100 W pump power, and picosecond mode-locked, master oscillator power fibre amplifier systems, with over 10 kW peak pump power. These systems can produce broadband supercontinua with over 50 and 1 mW/nm average spectral power, respectively. Techniques for numerical modelling of the supercontinuum sources are presented and used to illustrate some supercontinuum dynamics. Some recent experimental results are presented.

  9. Measuring Complexity through Average Symmetry

    OpenAIRE

    Alamino, Roberto C.

    2015-01-01

    This work introduces a complexity measure which addresses some conflicting issues between existing ones by using a new principle - measuring the average amount of symmetry broken by an object. It attributes low (although different) complexity to either deterministic or random homogeneous densities and higher complexity to the intermediate cases. This new measure is easily computable, breaks the coarse graining paradigm and can be straightforwardly generalised, including to continuous cases an...

  10. Proliferation resistance fuel cycle technology

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. S.; Ko, W. I

    1999-02-01

    The issues of dual use in nuclear technology are analysed for nuclear fuel cycle with special focus on uranium enrichment and spent fuel reprocessing which are considered as the most sensitive components in terms of vulnerability to diversion. Technical alternatives to mitigrate the vulnerability, as has been analysed in depth during the NASAP and INFCE era in the late seventies, are reviewed to characterize the DUPIC fuel cycle alternative. On the other hand, the new realities in nuclear energy including the disposition of weapon materials as a legacy of cold war are recast in an angle of nuclear proliferation resistance and safeguards with a discussion on the concept of spent fuel standard concept and its compliance with the DUPIC fuel cycle technology. (author)

  11. Application of genetic algorithms and CASMO to fuel optimization of BWRs; Aplicacion de algoritmos geneticos y CASMO a la optimizacion de combustible de BWRs

    Energy Technology Data Exchange (ETDEWEB)

    Carmona H, R.; Martin del Campo M, C.; Oropeza C, I.P. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, Jiutepec, Morelos 62550 (Mexico)]. e-mail: rockbert@ieee.org

    2008-07-01

    It was developed a system for the optimization of the radial distribution of enrichment in a fuel cell of a boiling water reactor based on genetic algorithms (GA's). The objective function includes four parameters: Average of the cell enrichment, average of gadolinium concentration of the cell, radial peak power factor and multiplication k-infinite factor. In order to be able to calculate the parameters that take part in the objective function, the process of evaluation of GA's was tied to the code CASMO-4, which is a code of transport in neutronic simulation groups of fuel assemblies that have been validated and it is used thoroughly for the calculation of nuclear data banks for boiling water reactors. A good radial distribution of fuel rods looks for, with different enrichment of U{sup 2}35 and contents of consumable poison in gadolinium form. For it is necessary to define the representation of the solution, the objective function and the implementation of the specific optimization process to the solution problem. The optimization process was codified in language C in the operating system LINUX. It was automated the creation of the entrances to the simulator, the execution of simulator CASMO-4 and the obtaining of the parameters that take part in the objective function from the exit of the simulator. It was applied to the fuel cell design of lOxlO that can be used in the fuel designs which are used at the moment in the nuclear power plant of Laguna Verde. They were considered 10 different fuel compositions from which four contain gadolinium. Three heuristic rules were applied: the peripheral positions of the fuel cell cannot contain burning poison, are placed the compositions with the smallest enrichment in the cell corners and, it is fixed the placement of the water rods. Nevertheless, the placement of the rods with gadolinium cell inside left free. Designs were obtained that complete with the wanted reactivity and the radial peak power factor. The

  12. Economic study of fuel scenarios for a reload; Estudio economico de escenarios de combustible para una recarga

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz S, J. J.; Castillo M, J. A.; Montes T, J. L.; Perusquia del C, R., E-mail: juanjose.ortiz@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    In this work the results to plan different scenarios for designing a nuclear fuel reload are shown. Given a reload with specific energy requirements, the objective is to verify the feasibility of using either a greater number of fresh fuel with less uranium enrichment, or otherwise reduce the number of fresh fuel assemblies and therefore they have a higher average uranium enrichment. For the study a cycle balance 18-month basis with 112 fresh assemblies divided into two lots, with energy produced of 10,075 Mwd/Tu was used. For the designs under the mentioned scenarios, the heuristic techniques known as taboo search and neural networks were used. To verify the feasibility of obtained reloads an economic study of the reload costs was performed. The results showed that is possible to design reloads under the two scenarios, but was more complicated decrease the amount of fresh fuel assemblies. In both scenarios was possible to reduce manufacturing costs of fuel and according to purely static calculation, it would be possible to increase the energy produced. (Author)

  13. 78 FR 77650 - Low Enriched Uranium From France: Continuation of Antidumping Duty Order

    Science.gov (United States)

    2013-12-24

    ... defined as enriched uranium dioxide (UO 2 ), whether or not contained in nuclear fuel rods or assemblies... form, such as UO 2 , or fabricated into nuclear fuel assemblies, regardless of the means by which the... of the LEU for consumption by the end- user in a nuclear reactor outside the United States....

  14. 40 CFR 1051.720 - How do I calculate my average emission level or emission credits?

    Science.gov (United States)

    2010-07-01

    ... AGENCY (CONTINUED) AIR POLLUTION CONTROLS CONTROL OF EMISSIONS FROM RECREATIONAL ENGINES AND VEHICLES Averaging, Banking, and Trading for Certification § 1051.720 How do I calculate my average emission level or... vehicles in the engine family times the average internal surface area of the vehicles' fuel tanks....

  15. Fuel Economy Testing and Data

    Science.gov (United States)

    EPA’s Fuel Economy pages provide information on current standards and how federal agencies work to enforce those laws, testing for national Corporate Average Fuel Economy or CAFE standards, and what you can do to reduce your own vehicle emissions.

  16. Effectiveness of bioremediation for the Prestige fuel spill : a summary of case studies

    Energy Technology Data Exchange (ETDEWEB)

    Gallego, J.R. [Oviedo Univ., Asturias (Spain); Gonzalez-Rojas, E.; Pelaez, A.I.; Sanchez, J [Oviedo Univ., Asturias (Spain). Inst. de Biotecnologia de Asturias; Garcia-Martinez, M.J.; Llamas, J.F. [Univ. Polictenica de Madrid, Madrid (Spain). Laboratorio de Estratigrafia Biomolecular

    2006-07-01

    This paper described novel bioremediation strategies used to remediate coastal areas in Spain impacted by the Prestige fuel oil spill in 2002. The bioremediation techniques were applied after hot pressurized water washing was used to remove hydrocarbons adhering to shorelines and rocks. Bioremediation strategies included monitored natural attenuation as well as accelerating biodegradation by stimulating indigenous populations through the addition of exogenous microbial populations. The sites selected for bioremediation were rocky shorelines of heterogenous granitic sediments with grain sizes ranging from sands to huge boulders; limestone-sandstone pebbles and cobbles; and fuel-coated limestone cliffs. Total surface area covered by the fuel was determined through the use of image analysis calculations. A statistical measurement of the fuel layer thickness was calculated by averaging the weights of multiple-fuel sampling increments. Bioremediation products included the use of oleophilic fertilizers; a biodegradable surfactant; and a microbial seeding agent. Determinations of saturate, aromatic, resins, and asphaltene (SARA) were performed using maltenes extraction and liquid chromatography. Microbial plating and selective enrichment with fuel as the sole carbon source were used to monitor the evolution of microbial populations in a variety of experiments. It was concluded that the biostimulation technique enhanced the efficiency of the in situ oleophilic fertilizers. 17 refs., 2 tabs., 6 figs.

  17. Oxygen potential measurements in high burnup LWR U0 2 fuel

    Science.gov (United States)

    Matzke, Hj.

    1995-05-01

    A miniature solid state galvanic cell was used to measure the oxygen potential Δ overlineG( O2) of reactor irradiated U0 2 fuel at different burnups in the range of 28 to ⩾ 150 GWd d/t M. This very high burnup was achieved in the rim region of a fuel with a cross section average burnup of 75 GWd d/t M. The fuels had different enrichments and therefore different contributions of fission of 235U and 239Pu. The temperature range covered was 900 to 1350 K. None of the fuels showed a significant oxidation. Rather, if allowance is made for the dissolved rare earth fission products and the Pu formed during irradiation, some of the fuels were very slightly substoichiometric and the highest possible degree of oxidation corresponded to U0 2.001. In general, the Δ overlineG( O2) at 750°C was about -400 kJ/mol, corresponding to the Δ overlineG( O2) of the reaction Mo + O 2 → MoO 2. The implication of these results which are in contrast to commonly assumed ideas that U0 2 fuel oxidizes due to burnup, are discussed and the importance of the fission product Mo and of the zircaloy clad as oxygen buffers is outlined.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Jue, Jan-Fong, E-mail: dennis.keiser@inl.gov; Keiser, Dennis D.; Breckenridge, Cynthia R.; Moore, Glenn A.; Meyer, Mitchell K.

    2014-05-01

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

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

    Science.gov (United States)

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

    2014-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1980-01-01

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

  2. Sensitivity and System Response of Pin Power Peaking in VVER-1000 Fuel Assembly Using TSUNAMI-2D

    Science.gov (United States)

    Frybort, J.

    2014-04-01

    Pin power peaking in a VVER-1000 fuel assembly and its sensitivity and uncertainty was analyzed by TSUNAMI-2D code. Several types of fuel assemblies were considered. They differ in number and position of gadolinium fuel pins. The calculations were repeated for several fuel compositions obtained by fuel depletion calculation. The results are quantified sensitivity data, which can be used for enrichment profiling.

  3. Applicability of the SCALE code system to MOX fuel transport systems for criticality safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Yamamoto, Toshihiro; Naito, Yoshitaka [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Hayashi, Toshiaki; Takasugi, Masahiro; Natsume, Toshihiro; Tsuda, Kazuaki

    1996-11-01

    In order to ascertain feasibilities of the SCALE code system for MOX fuel transport systems, criticality analyses were performed for MOX fuel (Pu enrichment; 3.0 wt.%) criticality experiments at JAERI`s TCA and for infinite fuel rod arrays as parameters of Pu enrichment and lattice pitch. The comparison with a combination of the continuous energy Monte Carlo code MCNP and JENDL-3.2 indicated that the SCALE code system with GAM-THERMOS 123-group library can produce feasible results. Though HANSEN-ROACH 16-group library gives poorer results for MOS fuel transport systems, the errors are conservative except for high enriched fuels. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

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

  5. Automated refueling simulations of a CANDU for the exploitation of thorium fuels

    Science.gov (United States)

    Holmes, Bradford

    CANDU nuclear reactors are in a unique circumstance where they are able to utilize and exploit a number of different fuel options to provide power as a utility. Thorium, a fertile isotope found naturally, is one option that should be explored. Thorium is more abundant than uranium, which is the typical fuel in the reactor and the availability of thorium makes nuclear energy desirable to more countries. This document contains the culmination of a project that explores, tests, and analyzes the feasibility of using thorium in a CANDU reactor. The project first develops a set of twodimensional lattice and three dimensional control rod simulations using the DRAGON Version 4 nuclear physics codes. This step is repeated for many concentrations of thorium. The data generated in these steps is then used to determine a functional enrichment of thorium. This is done via a procedural elimination and optimization of certain key parameters including but not limited to average exit burnup and reactivity evolution. For the purposes of this project, an enrichment of 1 % thorium was found viable. Full core calculations were done using the DONJON 4 code. CANFUEL, a program which simulates the refueling operations of a CANDU reactor for this fuel type was developed and ran for a simulation period of one hundred days. The program and the fuel selection met all selected requirements for the entirety of the simulation period. CANFUEL requires optimization for fuel selection before it can be used extensively. The fuel selection was further scrutinized when a reactivity insertion event was simulated. The adjuster rod 11 withdrawal from the core was analyzed and compared to classical CANDU results in order to ensure no significant deviations or unwanted evolutions were encountered. For this case, the simulation results were deemed acceptable with no significant deviations from the classical CANDU case.

  6. Accident Analyses for Conversion of the University of Missouri Research Reactor (MURR) from Highly-Enriched to Low-Enriched Uranium

    Energy Technology Data Exchange (ETDEWEB)

    Stillman, J. A. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Div., Research and Test Reactor Dept.; Feldman, E. E. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Div., Research and Test Reactor Dept.; Wilson, E. H. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Div., Research and Test Reactor Dept.; Foyto, L. P. [Univ. of Missouri, Columbia, MO (United States). Research Reactor; Kutikkad, K. [Univ. of Missouri, Columbia, MO (United States). Research Reactor; McKibben, J. C. [Univ. of Missouri, Columbia, MO (United States). Research Reactor; Peters, N. J. [Univ. of Missouri, Columbia, MO (United States). Research Reactor; Cowherd, W. M. [Univ. of Missouri, Columbia, MO (United States). College of Engineering, Nuclear Engineering Program; Rickman, B. [Univ. of Missouri, Columbia, MO (United States). College of Engineering, Nuclear Engineering Program

    2014-12-01

    This report contains the results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion 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), the MURR Facility, and the Nuclear Engineering Program – College of Engineering, University of Missouri-Columbia. The core conversion to LEU is being performed with financial support from the U. S. government. This report contains the results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion 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), the MURR Facility, and the Nuclear Engineering Program – College of Engineering, University of Missouri-Columbia. The core conversion to LEU is being performed with financial support from the U. S. government. In the framework of non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context most research and test reactors, both domestic and international, have started a program of conversion to the use of LEU fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (U-Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MURR. This report presents the results of a study of core behavior under a set of accident conditions for MURR cores fueled with HEU U-Alx dispersion fuel or LEU monolithic U-Mo alloy fuel with 10 wt% Mo

  7. Fossil Fuels.

    Science.gov (United States)

    Crank, Ron

    This instructional unit is one of 10 developed by students on various energy-related areas that deals specifically with fossil fuels. Some topics covered are historic facts, development of fuels, history of oil production, current and future trends of the oil industry, refining fossil fuels, and environmental problems. Material in each unit may…

  8. Initial report on characterization of excess highly enriched uranium

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-07-01

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

  9. MONOLITHIC FUEL FABRICATION PROCESS DEVELOPMENT AT THE IDAHO NATIONAL LABORATORY_

    Energy Technology Data Exchange (ETDEWEB)

    G. A. Moore; F. J. Rice; N. E. Woolstenhulme; J-F. Jue; B. H. Park; S. E. Steffler; N. P. Hallinan; M. D. Chapple; M. C. Marshall; B. L. Mackowiak; C. R. Clark; B. H. Rabin

    2009-11-01

    Full-size/prototypic U10Mo monolithic fuel-foils and aluminum clad fuel plates are being developed at the Idaho National Laboratory’s (INL) Materials and Fuels Complex (MFC). These efforts are focused on realizing Low Enriched Uranium (LEU) high density monolithic fuel plates for use in High Performance Research and Test Reactors. The U10Mo fuel foils under development afford a fuel meat density of ~16 gU/cc and thus have the potential to facilitate LEU conversions without any significant reactor-performance penalty. An overview is provided of the ongoing monolithic UMo fuel development effort, including application of a zirconium barrier layer on fuel foils, fabrication scale-up efforts, and development of complex/graded fuel foils. Fuel plate clad bonding processes to be discussed include: Hot Isostatic Pressing (HIP) and Friction Bonding (FB).

  10. Fuel distribution

    Energy Technology Data Exchange (ETDEWEB)

    Tison, R.R.; Baker, N.R.; Blazek, C.F.

    1979-07-01

    Distribution of fuel is considered from a supply point to the secondary conversion sites and ultimate end users. All distribution is intracity with the maximum distance between the supply point and end-use site generally considered to be 15 mi. The fuels discussed are: coal or coal-like solids, methanol, No. 2 fuel oil, No. 6 fuel oil, high-Btu gas, medium-Btu gas, and low-Btu gas. Although the fuel state, i.e., gas, liquid, etc., can have a major impact on the distribution system, the source of these fuels (e.g., naturally-occurring or coal-derived) does not. Single-source, single-termination point and single-source, multi-termination point systems for liquid, gaseous, and solid fuel distribution are considered. Transport modes and the fuels associated with each mode are: by truck - coal, methanol, No. 2 fuel oil, and No. 6 fuel oil; and by pipeline - coal, methane, No. 2 fuel oil, No. 6 oil, high-Btu gas, medium-Btu gas, and low-Btu gas. Data provided for each distribution system include component makeup and initial costs.

  11. DESIGN STUDY FOR A LOW-ENRICHED URANIUM CORE FOR THE HIGH FLUX ISOTOPE REACTOR, ANNUAL REPORT FOR FY 2010

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-02-01

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

  12. CLEAN: CLustering Enrichment ANalysis

    Directory of Open Access Journals (Sweden)

    Medvedovic Mario

    2009-07-01

    Full Text Available Abstract Background Integration of biological knowledge encoded in various lists of functionally related genes has become one of the most important aspects of analyzing genome-wide functional genomics data. In the context of cluster analysis, functional coherence of clusters established through such analyses have been used to identify biologically meaningful clusters, compare clustering algorithms and identify biological pathways associated with the biological process under investigation. Results We developed a computational framework for analytically and visually integrating knowledge-based functional categories with the cluster analysis of genomics data. The framework is based on the simple, conceptually appealing, and biologically interpretable gene-specific functional coherence score (CLEAN score. The score is derived by correlating the clustering structure as a whole with functional categories of interest. We directly demonstrate that integrating biological knowledge in this way improves the reproducibility of conclusions derived from cluster analysis. The CLEAN score differentiates between the levels of functional coherence for genes within the same cluster based on their membership in enriched functional categories. We show that this aspect results in higher reproducibility across independent datasets and produces more informative genes for distinguishing different sample types than the scores based on the traditional cluster-wide analysis. We also demonstrate the utility of the CLEAN framework in comparing clusterings produced by different algorithms. CLEAN was implemented as an add-on R package and can be downloaded at http://Clusteranalysis.org. The package integrates routines for calculating gene specific functional coherence scores and the open source interactive Java-based viewer Functional TreeView (FTreeView. Conclusion Our results indicate that using the gene-specific functional coherence score improves the reproducibility of the

  13. Post irradiation examination of thermal reactor fuels

    Science.gov (United States)

    Sah, D. N.; Viswanathan, U. K.; Ramadasan, E.; Unnikrishnan, K.; Anantharaman, S.

    2008-12-01

    The post irradiation examination (PIE) facility at the Bhabha Atomic Research Centre (BARC) has been in operation for more than three decades. Over these years this facility has been utilized for examination of experimental fuel pins and fuels from commercial power reactors operating in India. In a program to assess the performance of (U,Pu)O 2 MOX fuel prior to its introduction in commercial reactors, three experimental MOX fuel clusters irradiated in the pressurized water loop (PWL) of CIRUS up to burnup of 16 000 MWd/tU were examined. Fission gas release from these pins was measured by puncture test. Some of these fuel pins in the cluster contained controlled porosity pellets, low temperature sintered (LTS) pellets, large grain size pellets and annular pellets. PIE has also been carried out on natural UO 2 fuel bundles from Indian PHWRs, which included two high burnup (˜15 000 MWd/tU) bundles. Salient investigations carried out consisted of visual examination, leak testing, axial gamma scanning, fission gas analysis, microstructural examination of fuel and cladding, β, γ autoradiography of the fuel cross-section and fuel central temperature estimation from restructuring. A ThO 2 fuel bundle irradiated in Kakrapar Atomic Power Station (KAPS) up to a nominal fuel burnup of ˜11 000 MWd/tTh was also examined to evaluate its in-pile performance. The performance of the BWR fuel pins of Tarapur Atomic Power Stations (TAPS) was earlier assessed by carrying out PIE on 18 fuel elements selected from eight fuel assemblies irradiated in the two reactors. The burnup of these fuel elements varied from 5000 to 29 000 MWd/tU. This paper provides a brief review of some of the fuels examined and the results obtained on the performance of natural UO 2, enriched UO 2, MOX, and ThO 2 fuels.

  14. Feasibility study of boiling water reactor core based on thorium-uranium fuel concept

    Energy Technology Data Exchange (ETDEWEB)

    Nunez-Carrera, Alejandro [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779, Col Narvarte, 03020 Mexico D.F. (Mexico); Francois Lacouture, Juan Luis; Martin del Campo, Cecilia [Universidad Nacional Autonoma de Mexico, Facultad de Ingenieria, Paseo Cuauhnahuac 8532, Jiutepec, Mor. (Mexico); Espinosa-Paredes, Gilberto [Area de Ingenieria en Recursos Energeticos, Universidad Autonoma Metropolitana Iztapalapa, Apartado Postal 55-534, Mexico D.F. 09340 (Mexico)], E-mail: gepe@xanum.uam.mx

    2008-01-15

    The design of a boiling water reactor (BWR) equilibrium core using the thorium-uranium (blanket-seed) concept in the same integrated fuel assembly is presented in this paper. The lattice design uses the thorium conversion capability to {sup 233}U in a BWR spectrum. A core design was developed to achieve an equilibrium cycle of one effective full power year in a standard BWR with a reload of 104 fuel assemblies designed with an average {sup 235}U enrichment of 7.5 w/o in the seed sub-lattice. The main core operating parameters were obtained. It was observed that the analyzed parameters behave like those obtained in a standard BWR. The economic analysis shows that the fuel cycle cost of the proposed core design can be competitive with a standard uranium core design. Finally, a comparison of the toxicity of the spent fuel showed that the toxicity is lower in the thorium cycle than in other fuel cycles (UO{sub 2} and MOX uranium and plutonium) in the case of the once through cycle for light water reactors (LWR)

  15. Data mining in the study of nuclear fuel cells; Mineria de datos en el estudio de celdas de combustible nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Medina P, J. A. [Universidad Autonoma de Campeche, Av. Agustin Melgar s/n, Col. Buenavista, 24039 San Francisco de Campeche, Campeche (Mexico); Ortiz S, J. J.; Castillo, A.; Montes T, J. L.; Perusquia, R., E-mail: j.angel.mp@hotmail.com [ININ, Departamento de Sistemas Nucleares, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2015-09-15

    In this paper is presented a study of data mining application in the analysis of fuel cells and their performance within a nuclear boiling water reactor. A decision tree was used to fulfill questions of the type If (condition) and Then (conclusion) to classify if the fuel cells will have good performance. The performance is measured by compliance or not of the cold shutdown margin, the rate of linear heat generation and the average heat generation in a plane of the reactor. It is assumed that the fuel cells are simulated in the reactor under a fuel reload and rod control patterns pre designed. 18125 fuel cells were simulated according to a steady-state calculation. The decision tree works on a target variable which is one of the three mentioned before. To analyze this objective, the decision tree works with a set of attribute variables. In this case, the attributes are characteristics of the cell as number of gadolinium rods, rods number with certain uranium enrichment mixed with a concentration of gadolinium, etc. The found model was able to predict the execution or not of the shutdown margin with a precision of around 95%. However, the other two variables showed lower percentages due to few learning cases of the model in which these variables were or were not achieved. Even with this inconvenience, the model is quite reliable and can be used in way coupled in optimization systems of fuel cells. (Author)

  16. Advanced Fuel Cycle Cost Basis

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-12-01

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

  17. Advanced Fuel Cycle Cost Basis

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-04-01

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

  18. Advanced Fuel Cycle Cost Basis

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-03-01

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

  19. Nuclear fuel supply: challenges and opportunities

    Energy Technology Data Exchange (ETDEWEB)

    Lowen, S. [Cameco Corp., Saskatoon, Saskatchewan (Canada)

    2006-07-01

    Prices of uranium, conversion services and enrichment services have all significantly increased in the last few years. These price increases have generally been driven by a tightening in the supply of these products and services, mostly due to long lead times required to bring these products and services to the market. This paper will describe the various steps in the nuclear fuel cycle for natural and enriched uranium fuel, will discuss the development of the front-end fuel cycle for low void reactivity fuel, and will address the challenges faced in the long-term supply of each component, particularly in the light of potential demand increases as a result of a nuclear renaissance. The opportunities for new capacity and uranium production will be outlined and the process required to achieve sufficient new supply will be discussed. (author)

  20. Corrosion Evaluation of RERTR Uranium Molybdenum Fuel

    Energy Technology Data Exchange (ETDEWEB)

    A K Wertsching

    2012-09-01

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

  1. Interpreting Sky-Averaged 21-cm Measurements

    Science.gov (United States)

    Mirocha, Jordan

    2015-01-01

    Within the first ~billion years after the Big Bang, the intergalactic medium (IGM) underwent a remarkable transformation, from a uniform sea of cold neutral hydrogen gas to a fully ionized, metal-enriched plasma. Three milestones during this epoch of reionization -- the emergence of the first stars, black holes (BHs), and full-fledged galaxies -- are expected to manifest themselves as extrema in sky-averaged ("global") measurements of the redshifted 21-cm background. However, interpreting these measurements will be complicated by the presence of strong foregrounds and non-trivialities in the radiative transfer (RT) modeling required to make robust predictions.I have developed numerical models that efficiently solve the frequency-dependent radiative transfer equation, which has led to two advances in studies of the global 21-cm signal. First, frequency-dependent solutions facilitate studies of how the global 21-cm signal may be used to constrain the detailed spectral properties of the first stars, BHs, and galaxies, rather than just the timing of their formation. And second, the speed of these calculations allows one to search vast expanses of a currently unconstrained parameter space, while simultaneously characterizing the degeneracies between parameters of interest. I find principally that (1) physical properties of the IGM, such as its temperature and ionization state, can be constrained robustly from observations of the global 21-cm signal without invoking models for the astrophysical sources themselves, (2) translating IGM properties to galaxy properties is challenging, in large part due to frequency-dependent effects. For instance, evolution in the characteristic spectrum of accreting BHs can modify the 21-cm absorption signal at levels accessible to first generation instruments, but could easily be confused with evolution in the X-ray luminosity star-formation rate relation. Finally, (3) the independent constraints most likely to aide in the interpretation

  2. Fuel Cells

    DEFF Research Database (Denmark)

    Smith, Anders; Pedersen, Allan Schrøder

    2014-01-01

    Fuel cells have been the subject of intense research and development efforts for the past decades. Even so, the technology has not had its commercial breakthrough yet. This entry gives an overview of the technological challenges and status of fuel cells and discusses the most promising applications...... of the different types of fuel cells. Finally, their role in a future energy supply with a large share of fluctuating sustainable power sources, e.g., solar or wind, is surveyed....

  3. 7 CFR 1209.12 - On average.

    Science.gov (United States)

    2010-01-01

    ... 7 Agriculture 10 2010-01-01 2010-01-01 false On average. 1209.12 Section 1209.12 Agriculture Regulations of the Department of Agriculture (Continued) AGRICULTURAL MARKETING SERVICE (MARKETING AGREEMENTS....12 On average. On average means a rolling average of production or imports during the last two...

  4. Novel Membranes and Processes for Oxygen Enrichment

    Energy Technology Data Exchange (ETDEWEB)

    Lin, Haiqing

    2011-11-15

    The overall goal of this project is to develop a membrane process that produces air containing 25-35% oxygen, at a cost of $25-40/ton of equivalent pure oxygen (EPO2). Oxygen-enriched air at such a low cost will allow existing air-fueled furnaces to be converted economically to oxygen-enriched furnaces, which in turn will improve the economic and energy efficiency of combustion processes significantly, and reduce the cost of CO{sub 2} capture and sequestration from flue gases throughout the U.S. manufacturing industries. During the 12-month Concept Definition project: We identified a series of perfluoropolymers (PFPs) with promising oxygen/nitrogen separation properties, which were successfully made into thin film composite membranes. The membranes showed oxygen permeance as high as 1,200 gpu and oxygen/nitrogen selectivity of 3.0, and the permeance and selectivity were stable over the time period tested (60 days). We successfully scaled up the production of high-flux PFP-based membranes, using MTR's commercial coaters. Two bench-scale spiral-wound modules with countercurrent designs were made and parametric tests were performed to understand the effect of feed flow rate and pressure, permeate pressure and sweep flow rate on the membrane module separation properties. At various operating conditions that modeled potential industrial operating conditions, the module separation properties were similar to the pure-gas separation properties in the membrane stamps. We also identified and synthesized new polymers [including polymers of intrinsic microporosity (PIMs) and polyimides] with higher oxygen/nitrogen selectivity (3.5-5.0) than the PFPs, and made these polymers into thin film composite membranes. However, these membranes were susceptible to severe aging; pure-gas permeance decreased nearly six-fold within two weeks, making them impractical for industrial applications of oxygen enrichment. We tested the effect of oxygen-enriched air on NO{sub x} emissions

  5. High burnup effects in WWER fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Smirnov, V.; Smirnov, A. [RRC Research Institute of Atomic Reactors, Dimitrovqrad (Russian Federation)

    1996-03-01

    Since 1987 at the Research Institute of Atomic Reactors, the examinations of the WWER spent fuel assemblies has been carried out. These investigations are aimed to gain information on WWER spent fuel conditions in order to validate the fuel assemblies use during the 3 and 4 year fuel cycle in the WWER-440 and WWER-1000 units. At present time, the aim is to reach an average fuel burnup of 55 MWd/kgU. According to this aim, a new investigation program on the WWER spent fuel elements is started. The main objectives of this program are to study the high burnup effects and their influence on the WWER fuel properties. This paper presented the main statistical values of the WWER-440 and WWER-1000 reactors` fuel assemblies and their fragment parameters. Average burnup of fuel in the investigated fuel assemblies was in the range of 13 to 49.7 MWd/kgU. In this case, the numer of fuel cycles was from 1 to 4 during operation of the fuel assemblies.

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

    Energy Technology Data Exchange (ETDEWEB)

    None

    2010-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    None

    2011-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    None

    2016-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    None

    2012-12-01

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

  10. A sensitivity study on neutronic properties of DUPIC fuel

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok; Roh, Gyu Hong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    A sensitivity study has been done to determine the composition of DUPIC fuel from the viewpoint of neutronics fuel design. The spent PWR fuel compositions were generated and fissile contents adjusted by blending fresh uranium after mixing two spent PWR fuel assemblies. The {sup 239}Pu and {sup 235}U enrichments of DUPIC fuel were adjusted by controlling the amount of fresh uranium feed and the ratio of slightly enriched and depleted uranium in the feed uranium. Based on the material balance calculation, it is recommended that DUPIC fuel composition be such that spent PWR fuel utilization is more than 90%. A sensitivity study on the temperature reactivity coefficient of DUPIC fuel and shown that it is desirable to increase the {sup 239}Pu and {sup 235}U contents to reduce both the fuel and coolant temperature coefficients. On the other hand, refueling simulations of the DUPIC core have shown that the channel power peaking factor, which is a measure of the reactor trip margin, increases with the total fissile content. Considering these neutronic characteristics of the DUPIC fuel, it is recommended to have enrichments of 0.45 and 1.00 wt% for {sup 239}Pu and {sup 235}U, respectively. 3 refs., 2 tabs. (Author)

  11. Fast facility spent-fuel and waste assay instrument. [Fluorinel Dissolution and Fuel Storage (FAST) Facility

    Energy Technology Data Exchange (ETDEWEB)

    Eccleston, G.W.; Johnson, S.S.; Menlove, H.O.; Van Lyssel, T.; Black, D.; Carlson, B.; Decker, L.; Echo, M.W.

    1983-01-01

    A delayed-neutron assay instrument was installed in the Fluorinel Dissolution and Fuel Storage Facility at Idaho National Engineering Laboratory. The dual-assay instrument is designed to measure both spent fuel and waste solids that are produced from fuel processing. A set of waste standards, fabricated by Los Alamos using uranium supplied by Exxon Nuclear Idaho Company, was used to calibrate the small-sample assay region of the instrument. Performance testing was completed before installation of the instrument to determine the effects of uranium enrichment, hydrogenous materials, and neutron poisons on assays. The unit was designed to measure high-enriched uranium samples in the presence of large neutron backgrounds. Measurements indicate that the system can assay low-enriched uranium samples with moderate backgrounds if calibrated with proper standards.

  12. Research reactor de-fueling and fuel shipment

    Energy Technology Data Exchange (ETDEWEB)

    Ice, R.D.; Jawdeh, E.; Strydom, J.

    1998-08-01

    Planning for the Georgia Institute of Technology Research Reactor operations during the 1996 Summer Olympic Games began in early 1995. Before any details could be outlined, several preliminary administrative decisions had to be agreed upon by state, city, and university officials. The two major administrative decisions involving the reactor were (1) the security level and requirements and (2) the fuel status of the reactor. The Georgia Tech Research Reactor (GTRR) was a heavy-water moderated and cooled reactor, fueled with high-enriched uranium. The reactor was first licensed in 1964 with an engineered lifetime of thirty years. The reactor was intended for use in research applications and as a teaching facility for nuclear engineering students and reactor operators. Approximately one year prior to the olympics, the Georgia Tech administration decided that the GTRR fuel would be removed. In addition, a heightened, beyond regulatory requirements, security system was to be implemented. This report describes the scheduling, operations, and procedures.

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

    Energy Technology Data Exchange (ETDEWEB)

    Paul A. Demkowicz; John D. Hunn; Robert N. Morris; Charles A. Baldwin; Philip L. Winston; Jason M. Harp; Scott A. Ploger; Tyler Gerczak; Isabella J. van Rooyen; Fred C. Montgomery; Chinthaka M. Silva

    2014-10-01

    The AGR-1 experiment contained 72 low-enriched uranium oxide/uranium carbide TRISO-coated particle fuel compacts in six capsules irradiated to burnups of 11.2 to 19.5% FIMA, with zero TRISO coating failures detected during the irradiation. The irradiation performance of the fuel–including the extent of fission product release and the evolution of kernel and coating microstructures–was evaluated based on detailed examination of the irradiation capsules, the fuel compacts, and individual particles. Fractional release of 110mAg from the fuel compacts was often significant, with capsule-average values ranging from 0.01 to 0.38. Analysis of silver release from individual compacts indicated that it was primarily dependent on fuel temperature history. Europium and strontium were released in small amounts through intact coatings, but were found to be significantly retained in the outer pyrocrabon and compact matrix. The capsule-average fractional release from the compacts was 1×10 4 to 5×10 4 for 154Eu and 8×10 7 to 3×10 5 for 90Sr. The average 134Cs release from compacts was <3×10 6 when all particles maintained intact SiC. An estimated four particles out of 2.98×105 experienced partial cesium release due to SiC failure during the irradiation, driving 134Cs release in two capsules to approximately 10 5. Identification and characterization of these particles has provided unprecedented insight into the nature and causes of SiC coating failure in high-quality TRISO fuel. In general, changes in coating morphology were found to be dominated by the behavior of the buffer and inner pyrolytic carbon (IPyC), and infrequently observed SiC layer damage was usually related to cracks in the IPyC. Palladium attack of the SiC layer was relatively minor, except for the particles that released cesium during irradiation, where SiC corrosion was found adjacent to IPyC cracks. Palladium, silver, and uranium were found in the SiC layer of irradiated particles, and characterization

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-02-01

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

  15. Measurements and analysis of critical assemblies for research reactors with mixed enrichments

    Energy Technology Data Exchange (ETDEWEB)

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

    1982-01-01

    As part of the RERTR Program whole-core demonstration in the Ford Nuclear Reactor (FNR) at the University of Michigan, data have been obtained which will allow more extensive validation of neutronics methods for whole-core calculations of an equilibrium high-enriched-uranium (HEU) core and a fresh low-enriched-uranium (LEU) core. A series of experiments designed to provide the data needed for these validations has been performed in the Pool Critical Assembly (PCA) at the Oak Ridge National laboratory (ORNL). This paper reports the results of the measurements and of the subsequent validation calculations performed at ANL. Measurements were made on approximately 20 different critical configurations in the PCA during the period June 15 to 26, 1981. The normal PCA fuel elements contained high-enriched uranium (HEU, 93 wt% /sup 235/U) while the reduced-enrichment fuel elements, obtained for irradiation testing in the Oak Ridge Research Reactor (ORR) under the fuel demonstration activity of the RERTR Program, contained either medium-enriched uranium (MEU, 45 wt% /sup 235/U) or low-enriched uranium (LEU, 19.8 wt% /sup 235/U).

  16. Release of segregated nuclides from spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, L.H.; Tait, J.C. [Atomic Energy Canada Ltd., Pinawa, MB (Canada). Whiteshell Laboratories

    1997-10-01

    The potential release of fission and activation products from spent nuclear fuel into groundwater after container failure in the Swedish deep repository is discussed. Data from studies of fission gas release from representative Swedish BWR fuel are used to estimate the average fission gas release for the spent fuel population. Information from a variety of leaching studies on LWR and CANDU fuel are then reviewed as a basis for estimating the fraction of the inventory of key radionuclides that could be released preferentially (the Instant Release Fraction of IRF) upon failure of the fuel cladding. The uncertainties associated with these estimates are discussed. 33 refs, 6 figs, 3 tabs.

  17. Quadratic reactivity fuel cycle model

    Energy Technology Data Exchange (ETDEWEB)

    Lewins, J.D.

    1985-11-01

    For educational purposes it is highly desirable to provide simple yet realistic models for fuel cycle and fuel economy. In particular, a lumped model without recourse to detailed spatial calculations would be very helpful in providing the student with a proper understanding of the purposes of fuel cycle calculations. A teaching model for fuel cycle studies based on a lumped model assuming the summability of partial reactivities with a linear dependence of reactivity usefully illustrates fuel utilization concepts. The linear burnup model does not satisfactorily represent natural enrichment reactors. A better model, showing the trend of initial plutonium production before subsequent fuel burnup and fission product generation, is a quadratic fit. The study of M-batch cycles, reloading 1/Mth of the core at end of cycle, is now complicated by nonlinear equations. A complete account of the asymptotic cycle for any order of M-batch refueling can be given and compared with the linear model. A complete account of the transient cycle can be obtained readily in the two-batch model and this exact solution would be useful in verifying numerical marching models. It is convenient to treat the parabolic fit rho = 1 - tau/sup 2/ as a special case of the general quadratic fit rho = 1 - C/sub tau/ - (1 - C)tau/sup 2/ in suitably normalized reactivity and cycle time units. The parabolic results are given in this paper.

  18. RERTR Fuel Developmemt and Qualification Plan

    Energy Technology Data Exchange (ETDEWEB)

    Dan Wachs

    2007-01-01

    In late 2003 it became evident that U-Mo aluminum fuels under development exhibited significant fuel performance problems under the irradiation conditions required for conversion of most high-powered research reactors. Solutions to the fuel performance issue have been proposed and show promise in early testing. Based on these results, a Reduced Enrichment Research and Test Reactor (RERTR) program strategy has been mapped to allow generic fuel qualification to occur prior to the end of FY10 and reactor conversion to occur prior to the end of FY14. This strategy utilizes a diversity of technologies, test conditions, and test types. Scoping studies using miniature fuel plates will be completed in the time frame of 2006-2008. Irradiation of larger specimens will occur in the Advanced Test Reactor (ATR) in the United States, the Belgian Reactor-2 (BR2) reactor in Belgium, and in the OSIRIS reactor in France in 2006-2009. These scoping irradiation tests provide a large amount of data on the performance of advanced fuel types under irradiation and allow the down selection of technology for larger scale testing during the final stages of fuel qualification. In conjunction with irradiation testing, fabrication processes must be developed and made available to commercial fabricators. The commercial fabrication infrastructure must also be upgraded to ensure a reliable low enriched uranium (LEU) fuel supply. Final qualification of fuels will occur in two phases. Phase I will obtain generic approval for use of dispersion fuels with density less than 8.5 g-U/cm3. In order to obtain this approval, a larger scale demonstration of fuel performance and fabrication technology will be necessary. Several Materials Test Reactor (MTR) plate-type fuel assemblies will be irradiated in both the High Flux Reactor (HFR) and the ATR (other options include the BR2 and Russian Research Reactor, Dmitrovgrad, Russia [MIR] reactors) in 2008-2009. Following postirradiation examination, a report

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

    Energy Technology Data Exchange (ETDEWEB)

    Hanson A. L.; Diamond D.

    2014-06-30

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

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

    Energy Technology Data Exchange (ETDEWEB)

    William Anderson; James Tulenko; Bradley Rearden; Gary Harms

    2008-09-11

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

  1. Accurate Switched-Voltage voltage averaging circuit

    OpenAIRE

    金光, 一幸; 松本, 寛樹

    2006-01-01

    Abstract ###This paper proposes an accurate Switched-Voltage (SV) voltage averaging circuit. It is presented ###to compensated for NMOS missmatch error at MOS differential type voltage averaging circuit. ###The proposed circuit consists of a voltage averaging and a SV sample/hold (S/H) circuit. It can ###operate using nonoverlapping three phase clocks. Performance of this circuit is verified by PSpice ###simulations.

  2. Spectral averaging techniques for Jacobi matrices

    CERN Document Server

    del Rio, Rafael; Schulz-Baldes, Hermann

    2008-01-01

    Spectral averaging techniques for one-dimensional discrete Schroedinger operators are revisited and extended. In particular, simultaneous averaging over several parameters is discussed. Special focus is put on proving lower bounds on the density of the averaged spectral measures. These Wegner type estimates are used to analyze stability properties for the spectral types of Jacobi matrices under local perturbations.

  3. Enrichment of light hydrocarbon mixture

    Science.gov (United States)

    Yang; Dali; Devlin, David; Barbero, Robert S.; Carrera, Martin E.; Colling, Craig W.

    2010-08-10

    Light hydrocarbon enrichment is accomplished using a vertically oriented distillation column having a plurality of vertically oriented, nonselective micro/mesoporous hollow fibers. Vapor having, for example, both propylene and propane is sent upward through the distillation column in between the hollow fibers. Vapor exits neat the top of the column and is condensed to form a liquid phase that is directed back downward through the lumen of the hollow fibers. As vapor continues to ascend and liquid continues to countercurrently descend, the liquid at the bottom of the column becomes enriched in a higher boiling point, light hydrocarbon (propane, for example) and the vapor at the top becomes enriched in a lower boiling point light hydrocarbon (propylene, for example). The hollow fiber becomes wetted with liquid during the process.

  4. Accident Analyses for Conversion of the University of Missouri Research Reactor (MURR) from Highly-Enriched to Low-Enriched Uranium

    Energy Technology Data Exchange (ETDEWEB)

    Stillman, J. A. [Argonne National Lab. (ANL), Argonne, IL (United States); Feldman, E. E. [Argonne National Lab. (ANL), Argonne, IL (United States); Jaluvka, D. [Argonne National Lab. (ANL), Argonne, IL (United States); Wilson, E. H. [Argonne National Lab. (ANL), Argonne, IL (United States); Foyto, L. P. [Univ. of Missouri, Columbia, MO (United States); Kutikkad, K. [Univ. of Missouri, Columbia, MO (United States); McKibben, J. C. [Univ. of Missouri, Columbia, MO (United States); Peters, N. J. [Univ. of Missouri, Columbia, MO (United States)

    2017-02-01

    This report contains the results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion 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 in the Research and Test Reactor Department at the Argonne National Laboratory (ANL) and the MURR Facility. MURR LEU conversion is part of an overall effort to develop and qualify high-density fuel within the U.S. High Performance Research Reactor Conversion (USHPRR) program conducted by the U.S. Department of Energy National Nuclear Security Administration’s Office of Material Management and Minimization (M3).

  5. 31 CFR 540.316 - Uranium enrichment.

    Science.gov (United States)

    2010-07-01

    ... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false Uranium enrichment. 540.316 Section... FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.316 Uranium enrichment. The term uranium enrichment means the process...

  6. Average-Time Games on Timed Automata

    OpenAIRE

    Jurdzinski, Marcin; Trivedi, Ashutosh

    2009-01-01

    An average-time game is played on the infinite graph of configurations of a finite timed automaton. The two players, Min and Max, construct an infinite run of the automaton by taking turns to perform a timed transition. Player Min wants to minimise the average time per transition and player Max wants to maximise it. A solution of average-time games is presented using a reduction to average-price game on a finite graph. A direct consequence is an elementary proof of determinacy for average-tim...

  7. Grassmann Averages for Scalable Robust PCA

    DEFF Research Database (Denmark)

    Hauberg, Søren; Feragen, Aasa; Black, Michael J.

    2014-01-01

    arbitrarily corrupt the results. Unfortunately, state-of-the-art approaches for robust PCA do not scale beyond small-to-medium sized datasets. To address this, we introduce the Grassmann Average (GA), which expresses dimensionality reduction as an average of the subspaces spanned by the data. Because averages...... to vectors (subspaces) or elements of vectors; we focus on the latter and use a trimmed average. The resulting Trimmed Grassmann Average (TGA) is particularly appropriate for computer vision because it is robust to pixel outliers. The algorithm has low computational complexity and minimal memory requirements...

  8. Reformer Fuel Injector

    Science.gov (United States)

    Suder, Jennifer L.

    2004-01-01

    Today's form of jet engine power comes from what is called a gas turbine engine. This engine is on average 14% efficient and emits great quantities of green house gas carbon dioxide and air pollutants, Le. nitrogen oxides and sulfur oxides. The alternate method being researched involves a reformer and a solid oxide fuel cell (SOFC). Reformers are becoming a popular area of research within the industry scale. NASA Glenn Research Center's approach is based on modifying the large aspects of industry reforming processes into a smaller jet fuel reformer. This process must not only be scaled down in size, but also decrease in weight and increase in efficiency. In comparison to today's method, the Jet A fuel reformer will be more efficient as well as reduce the amount of air pollutants discharged. The intent is to develop a 10kW process that can be used to satisfy the needs of commercial jet engines. Presently, commercial jets use Jet-A fuel, which is a kerosene based hydrocarbon fuel. Hydrocarbon fuels cannot be directly fed into a SOFC for the reason that the high temperature causes it to decompose into solid carbon and Hz. A reforming process converts fuel into hydrogen and supplies it to a fuel cell for power, as well as eliminating sulfur compounds. The SOFC produces electricity by converting H2 and CO2. The reformer contains a catalyst which is used to speed up the reaction rate and overall conversion. An outside company will perform a catalyst screening with our baseline Jet-A fuel to determine the most durable catalyst for this application. Our project team is focusing on the overall research of the reforming process. Eventually we will do a component evaluation on the different reformer designs and catalysts. The current status of the project is the completion of buildup in the test rig and check outs on all equipment and electronic signals to our data system. The objective is to test various reformer designs and catalysts in our test rig to determine the most

  9. Assessment of Nuclear Fuels using Radiographic Thickness Measurement Method

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-11-01

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

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

    Science.gov (United States)

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

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Mohammed, Abdul Aziz, E-mail: azizM@uniten.edu.my; Rahman, Shaik Mohmmed Haikhal Abdul [Universiti Tenaga Nasional. Jalan Ikram-UNITEN, 43000 Kajang, Selangor (Malaysia); Pauzi, Anas Muhamad, E-mail: anas@uniten.edu.my; Zin, Muhamad Rawi Muhammad; Jamro, Rafhayudi; Idris, Faridah Mohamad [Malaysian Nuclear Agency, Bangi, 43000 Kajang, Selangor (Malaysia)

    2016-01-22

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

  12. Thermogenic Effect from Nutritionally Enriched Coffee Consumption

    Directory of Open Access Journals (Sweden)

    Jennings Peter F

    2006-06-01

    Full Text Available Abstract Background The purpose of this study was to examine the effect of nutritionally enriched JavaFit™ (JF coffee (450 mg of caffeine, 1200 mg of garcinia cambogia, 360 mg of citrus aurantium extract, and 225 mcg of chromium polynicotinate on resting oxygen uptake (VO2, respiratory exchange ratio (RER, heart rate (HR, and blood pressure (BP in healthy and physically active individuals. Method Ten subjects (8 male, 2 female; 20.9 ± 1.7 y; 178.1 ± 10.4 cm; 71.8 ± 12.1 kg underwent two testing sessions administered in a randomized and double-blind fashion. During each session, subjects reported to the Human Performance Laboratory after at least 3-h post-absorptive state and were provided either 354 ml (1.5 cups of freshly brewed JF or commercially available caffeinated coffee (P. Subjects then rested in a semi-recumbent position for three hours. VO2 and HR were determined every 5 min during the first 30 min and every 10 min during the next 150 min. BP was determined every 15 min during the first 30 min and every 30 min thereafter. Area under the curve (AUC analysis was computed for VO2, whereas a session-average was calculated for RER, HR and BP. Results Initial analysis revealed no significant differences. However, seven of the ten subjects were considered responders to JF (had a higher AUC for VO2during JF than P. Statistical analysis showed the difference between JF and P (12% to be significantly different in these responders. In addition, the average systolic BP was higher (p Conclusion It appears that consuming a nutritionally-enriched coffee beverage may increase resting energy expenditure in individuals that are sensitive to the caffeine and herbal coffee supplement. In addition, this supplement also appears to affect cardiovascular dynamics by augmenting systolic arterial blood pressure.

  13. Fuel cells

    Directory of Open Access Journals (Sweden)

    D. N. Srivastava

    1962-05-01

    Full Text Available The current state of development of fuel cells as potential power sources is reviewed. Applications in special fields with particular reference to military requirements are pointed out.

  14. Future Fuels

    Science.gov (United States)

    2006-04-01

    Storage Devices, Fuel Management, Gasification, Fischer-Tropsch, Syngas , Hubberts’s Peak UNCLAS UNCLAS UNCLAS UU 80 Dr. Sujata Millick (703) 696...prices ever higher, and perhaps lead to intermittent fuel shortages as production fluctuates. Clearly, this competition for resources also provides oil...producers multiple options for selling their products, and raises the possibility that the US could face shortages resulting from shifts in

  15. Uranium enrichment: a competitive market in the future?

    Energy Technology Data Exchange (ETDEWEB)

    Marques, Andre Ferreira; Honaiser, Eduardo Henrique Rangel [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil)]. E-mail: 20-1@ctemsp.mar.mil.br

    2005-07-01

    Uranium enrichment is the costly step in the nuclear fuel cycle. It has born as a an activity for the military in the 40s, financed by governments, such as the United States (US) and the former Soviet Union. Later, other major nations have joined them in the nuclear weapons development. The activity of enrichment was done in each country that developed nuclear weapons, and the nuclear weapons countries, especially the US and Soviet Union, dictated the mined uranium market. In the 70s, with the growth of the commercial use of nuclear energy, uranium enrichment started to be treated as a market, which gradually have structured itself, strongly influenced by the historical background. Today, the market is an oligopoly of four major government-owned (or government-influenced) companies. In this paper, the trends in the enrichment market are identified, focusing on competitiveness. Through the conduction of a market analysis (past and future), and the study of the market structure evolution, a more competitive market is shown, but still influenced by the governmental participation. Competitiveness is dictated by government support, verticalization capacity, and, mainly by technological advantages. (author)

  16. Fabrication of uranium dioxide fuel pellets in support of a SLOWPOKE-2 research reactor HEU to LEU core conversion

    Energy Technology Data Exchange (ETDEWEB)

    Bergeron, A. [Aomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2014-07-01

    The International Centre for Environmental and Nuclear Sciences (ICENS) at the University of the West Indies in Jamaica operates a SLOWPOKE-2 research reactor that is currently fuelled with highly-enriched uranium (HEU). As part of the Global Threat Reduction Initiative, Atomic Energy of Canada Ltd. has been subcontracted to fabricate low-enriched uranium (LEU) fuel for the ICENS SLOWPOKE-2. The low enriched uranium core consists of a fuel cage containing uranium dioxide fuelled elements. This paper describes the fabrication of the low-enriched uranium dioxide fuel pellets for the SLOWPOKE-2 core conversion. (author)

  17. Criticality safety and fuel segregation at Zircatec - past present and future - part I

    Energy Technology Data Exchange (ETDEWEB)

    Hameed, R.; Pant, A. [Zircatec Precision Industries, Port Hope, Ontario (Canada)

    2005-07-01

    In response to the Power Uprate program at Bruce Power, Zircatec has committed to introduce, by Spring 2006 a new manufacturing line for the production of 43 element Bruce LVRF bundles containing Slightly Enriched Uranium (SEU) with a centre pin of blended dysprosia/urania (BDU). This is a new fuel design and is the first change in fuel design since the introduction of the current 37 element fuel over 20 years ago. Zircatec has been involved with the manufacture of enriched fuel for at least 30 years. In the recent past ({approx} 20 years) enriched manufacture at Zircatec has been limited to small special orders for test purposes; this has included the major development work conducted to support the evolution of the current design. This presentation will discuss methods that have been used for small scale manufacture of special fuels including those with multiple enrichment. Specific examples will be given as pertaining to the manufacture of 26 LVRF Demonstration Irradiation bundles. (author)

  18. WIDTHS AND AVERAGE WIDTHS OF SOBOLEV CLASSES

    Institute of Scientific and Technical Information of China (English)

    刘永平; 许贵桥

    2003-01-01

    This paper concerns the problem of the Kolmogorov n-width, the linear n-width, the Gel'fand n-width and the Bernstein n-width of Sobolev classes of the periodicmultivariate functions in the space Lp(Td) and the average Bernstein σ-width, averageKolmogorov σ-widths, the average linear σ-widths of Sobolev classes of the multivariatequantities.

  19. Bacterioplankton responses to iron enrichment during the SAGE experiment

    Science.gov (United States)

    Kuparinen, J.; Hall, J.; Ellwood, M.; Safi, K.; Peloquin, J.; Katz, D.

    2011-03-01

    We studied the microbial food web in the upper 100 m of the water column in iron-limited sub-Antarctic HNLC waters south-east of New Zealand in the SAGE experiment in 2004, with focus on bacterioplankton. Samples were collected daily from inside and outside the iron enriched patch. Short term enrichment experiments were conducted on board in 4 L polycarbonate bottles with water outside the iron enriched patch to study single and combined effects of micronutrient additions on microbial food web. Low bacterial growth was recorded in the study area with community turnover times of 50 h or more during the study period. Measurements of bacterial standing stocks and production rates in the study show minor responses to the large scale iron enrichment, with increase in rates and stocks after the first enrichment and at the end of the study period after the third iron enrichment when solar radiation increased and wind mixing decreased. The average daily bacterial production rates were 31.5 and 33.7 mgCm -2 d -1 for the OUT and IN stations, respectively; thus overall there was not a significant difference between the control and the iron-enriched patch. In the bottle experiments bacterial thymidine incorporation showed responses to single iron and silicic acid enrichments and a major growth response to the combined iron and sucrose enrichments. Phytoplankton chlorophyll- a showed clear stimulation by single additions of iron and silicic acid and silicic acid enhanced the iron impact. Cobalt additions had no effect on bacteria growth and a negative effect on phytoplankton growth. Low bacterial in situ growth rates and the enrichment experiments suggest that bacteria are co-limited by iron and carbon, and that bacterial iron uptake is dependent on carbon supply by the food web. With the high iron quota (μmol Fe mol C -1) bacteria may scavenge considerable amounts of the excess iron, and thus influence the relative importance of the microbial food web as a carbon sink.

  20. NOAA Average Annual Salinity (3-Zone)

    Data.gov (United States)

    California Department of Resources — The 3-Zone Average Annual Salinity Digital Geography is a digital spatial framework developed using geographic information system (GIS) technology. These salinity...

  1. Stochastic averaging of quasi-Hamiltonian systems

    Institute of Scientific and Technical Information of China (English)

    朱位秋

    1996-01-01

    A stochastic averaging method is proposed for quasi-Hamiltonian systems (Hamiltonian systems with light dampings subject to weakly stochastic excitations). Various versions of the method, depending on whether the associated Hamiltonian systems are integrable or nonintegrable, resonant or nonresonant, are discussed. It is pointed out that the standard stochastic averaging method and the stochastic averaging method of energy envelope are special cases of the stochastic averaging method of quasi-Hamiltonian systems and that the results obtained by this method for several examples prove its effectiveness.

  2. Fuel lattice design using heuristics and new strategies

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz S, J. J.; Castillo M, J. A.; Torres V, M.; Perusquia del Cueto, R. [ININ, Carretera Mexico-Toluca s/n, Ocoyoacac 52750, Estado de Mexico (Mexico); Pelta, D. A. [ETS Ingenieria Informatica y Telecomunicaciones, Universidad de Granada, Daniel Saucedo Aranda s/n, 18071 Granada (Spain); Campos S, Y., E-mail: juanjose.ortiz@inin.gob.m [IPN, Escuela Superior de Fisica y Matematicas, Unidad Profesional Adolfo Lopez Mateos, Edif. 9, 07738 Mexico D. F. (Mexico)

    2010-10-15

    This work show some results of the fuel lattice design in BWRs when some allocation pin rod rules are not taking into account. Heuristics techniques like Path Re linking and Greedy to design fuel lattices were used. The scope of this work is to search about how do classical rules in design fuel lattices affect the heuristics techniques results and the fuel lattice quality. The fuel lattices quality is measured by Power Peaking Factor and Infinite Multiplication Factor at the beginning of the fuel lattice life. CASMO-4 code to calculate these parameters was used. The analyzed rules are the following: pin rods with lowest uranium enrichment are only allocated in the fuel lattice corner, and pin rods with gadolinium cannot allocated in the fuel lattice edge. Fuel lattices with and without gadolinium in the main diagonal were studied. Some fuel lattices were simulated in an equilibrium cycle fuel reload, using Simulate-3 to verify their performance. So, the effective multiplication factor and thermal limits can be verified. The obtained results show a good performance in some fuel lattices designed, even thought, the knowing rules were not implemented. A fuel lattice performance and fuel lattice design characteristics analysis was made. To the realized tests, a dell workstation was used, under Li nux platform. (Author)

  3. 75 FR 7525 - Application for a License To Export High-Enriched Uranium

    Science.gov (United States)

    2010-02-19

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Application for a License To Export High-Enriched Uranium Pursuant to 10 CFR 110.70(c) ``Public... fuel France; Belgium. Security Complex, February 2, Uranium (93.35%). uranium (87.3 elements in...

  4. 75 FR 15743 - Application for a License To Export High-Enriched Uranium

    Science.gov (United States)

    2010-03-30

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Application for a License To Export High-Enriched Uranium Pursuant to 10 CFR 110.70(c) ``Public... fabricate fuel France. Complex, March 3, 2010. Uranium (93.35%). uranium (149.36 elements in March 9,...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-01-31

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

  6. Proceedings of the 1990 International Meeting on Reduced Enrichment for Research and Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

    The global effort to reduce, and possibly, eliminate the international traffic in highly-enriched uranium caused by its use in research reactors requires extensive cooperation and free exchange of information among all participants. To foster this free exchange of information, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the thirteenth of a series which began in 1978. The common effort brought together, past, a large number of specialists from many countries. On hundred twenty-three participants from 26 countries, including scientists, reactor operators, and personnel from commercial fuel suppliers, research centers, and government organizations, convened in Newport, Rhode Island to discuss their results, their activities, and their plans relative to converting research reactors to low-enriched fuels. As more and more reactors convert to the use of low-enriched uranium, the emphasis of our effort has begun to shift from research and development to tasks more directly related to implementation of the new fuels and technologies that have been developed, and to refinements of those fuels and technologies. It is appropriate, for this reason, that the emphasis of this meeting was placed on safety and on conversion experiences. This individual papers in this report have been cataloged separately.

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

    Energy Technology Data Exchange (ETDEWEB)

    Aden, V.G.; Kartashov, E.F.; Lukichev, V.A. [and others

    1997-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-03-01

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

  9. Strategy for Used Fuel Acquisition

    Energy Technology Data Exchange (ETDEWEB)

    Steven C. Marschman; Chris Rusch

    2013-09-01

    The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), Office of Fuel Cycle Technology, has established the Used Fuel Disposition Campaign (UFDC) to conduct the research and development activities related to storage, transportation, and disposal of used nuclear fuel and high-level radioactive waste. The mission of the UFDC is to identify alternatives and conduct scientific research and technology development to enable storage, transportation and disposal of used nuclear fuel (UNF) and wastes generated by existing and future nuclear fuel cycles. The Storage and Transportation staffs within the UFDC are responsible for addressing issues regarding the extended or long-term storage of UNF and its subsequent transportation. The near-term objectives of the Storage and Transportation task are to use a science-based approach to develop the technical bases to support the continued safe and secure storage of UNF for extended periods, subsequent retrieval, and transportation. While both wet and dry storage have been shown to be safe options for storing UNF, the focus of the program is on dry storage at reactor or centralized locations. Because limited information is available on the properties of high burnup fuel (exceeding 45 gigawatt-days per metric tonne of uranium [GWd/MTU]), and because much of the fuel currently discharged from today’s reactors exceeds this burnup threshold, a particular emphasis of this program is on high burnup fuels. Since high burnup used fuels have only been loaded into dry storage systems in the past decade or so, these materials are available to the UFDC for testing in only very limited quantities. Much of what is available has come via NRC testing programs. Some of these fuels may have achieved "high burnup," but that does not mean they were designed for high burnup use (e.g. lower enrichments, smaller plenum spaces, extra reactor cycles). The handling and transfer of these materials from utility to laboratory has not always been

  10. Dynamic Multiscale Averaging (DMA) of Turbulent Flow

    Energy Technology Data Exchange (ETDEWEB)

    Richard W. Johnson

    2012-09-01

    A new approach called dynamic multiscale averaging (DMA) for computing the effects of turbulent flow is described. The new method encompasses multiple applications of temporal and spatial averaging, that is, multiscale operations. Initially, a direct numerical simulation (DNS) is performed for a relatively short time; it is envisioned that this short time should be long enough to capture several fluctuating time periods of the smallest scales. The flow field variables are subject to running time averaging during the DNS. After the relatively short time, the time-averaged variables are volume averaged onto a coarser grid. Both time and volume averaging of the describing equations generate correlations in the averaged equations. These correlations are computed from the flow field and added as source terms to the computation on the next coarser mesh. They represent coupling between the two adjacent scales. Since they are computed directly from first principles, there is no modeling involved. However, there is approximation involved in the coupling correlations as the flow field has been computed for only a relatively short time. After the time and spatial averaging operations are applied at a given stage, new computations are performed on the next coarser mesh using a larger time step. The process continues until the coarsest scale needed is reached. New correlations are created for each averaging procedure. The number of averaging operations needed is expected to be problem dependent. The new DMA approach is applied to a relatively low Reynolds number flow in a square duct segment. Time-averaged stream-wise velocity and vorticity contours from the DMA approach appear to be very similar to a full DNS for a similar flow reported in the literature. Expected symmetry for the final results is produced for the DMA method. The results obtained indicate that DMA holds significant potential in being able to accurately compute turbulent flow without modeling for practical

  11. Winter fuels report

    Energy Technology Data Exchange (ETDEWEB)

    1995-02-17

    The Winter Fuels Report is intended to provide consise, timely information to the industry, the press, policymakers, consumers, analysts, and State and local governments on the following topics: Distillate fuel oil net production, imports and stocks on a US level and for all Petroleum Administration for Defense Districts (PADD) and product supplied on a US level; Natural gas supply and disposition and underground storage for the US and consumption for all PADD`s as well as selected National average prices; Residential and wholesale pricing data for heating oil and propane for those States participating in the joint Energy Information Administration (EIA)/State Heating Oil and Propane Program; Crude oil and petroleum price comparisons for the US and selected cities; and A 6-10 Day and 30-Day outlook for temperature and precipitation and US total heating degree days by city.

  12. Actinides record, power calculations and activity for present isotopes in the spent fuel of a BWR; Historial de actinidos y calculos de potencia y actividad para isotopos presentes en el combustible gastado de un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Enriquez C, P.; Ramirez S, J. R.; Lucatero, M. A., E-mail: pastor.enriquez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-10-15

    The administration of spent fuel is one of the more important stages of the nuclear fuel cycle, and this has become a problem of supreme importance in countries that possess nuclear reactors. Due to this in this work, the study on the actinides record and present fission products to the discharge of the irradiated fuel in a light water reactor type BWR is shown, to quantify the power and activity that emit to the discharge and during the cooling time. The analysis was realized on a fuel assembly type 10 x 10 with an enrichment average of 3.69 wt % in U-235 and the assembly simulation assumes four cycles of operation of 18 months each one and presents an exposition of 47 G Wd/Tm to the discharge. The module OrigenArp of the Scale 6 code is the computation tool used for the assembly simulation and to obtain the results on the actinides record presents to the fuel discharge. The study covers the following points: a) Obtaining of the plutonium vector used in the fuel production of mixed oxides, and b) Power calculation and activity for present actinides to the discharge. The results presented in this work, correspond at the same time immediate of discharge (0 years) and to a cooling stage in the irradiated fuel pool (5 years). (Author)

  13. Fuel lattice design in a boiling water reactor using a knowledge-based automation system

    Energy Technology Data Exchange (ETDEWEB)

    Tung, Wu-Hsiung, E-mail: wstong@iner.gov.tw; Lee, Tien-Tso; Kuo, Weng-Sheng; Yaur, Shung-Jung

    2015-11-15

    Highlights: • An automation system was developed for the fuel lattice radial design of BWRs. • An enrichment group peaking equalizing method is applied to optimize the design. • Several heuristic rules and restrictions are incorporated to facilitate the design. • The CPU time for the system to design a 10x10 lattice was less than 1.2 h. • The beginning-of-life LPF was improved from 1.319 to 1.272 for one of the cases. - Abstract: A knowledge-based fuel lattice design automation system for BWRs is developed and applied to the design of 10 × 10 fuel lattices. The knowledge implemented in this fuel lattice design automation system includes the determination of gadolinium fuel pin location, the determination of fuel pin enrichment and enrichment distribution. The optimization process starts by determining the gadolinium distribution based on the pin power distribution of a flat enrichment lattice and some heuristic rules. Next, a pin power distribution flattening and an enrichment grouping process are introduced to determine the enrichment of each fuel pin enrichment type and the initial enrichment distribution of a fuel lattice design. Finally, enrichment group peaking equalizing processes are performed to achieve lower lattice peaking. Several fuel lattice design constraints are also incorporated in the automation system such that the system can accomplish a design which meets the requirements of practical use. Depending on the axial position of the lattice, a different method is applied in the design of the fuel lattice. Two typical fuel lattices with U{sup 235} enrichment of 4.471% and 4.386% were taken as references. Application of the method demonstrates that improved lattice designs can be achieved through the enrichment grouping and the enrichment group peaking equalizing method. It takes about 11 min and 1 h 11 min of CPU time for the automation system to accomplish two design cases on an HP-8000 workstation, including the execution of CASMO-4

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-11-01

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

  15. Average sampling theorems for shift invariant subspaces

    Institute of Scientific and Technical Information of China (English)

    2000-01-01

    The sampling theorem is one of the most powerful results in signal analysis. In this paper, we study the average sampling on shift invariant subspaces, e.g. wavelet subspaces. We show that if a subspace satisfies certain conditions, then every function in the subspace is uniquely determined and can be reconstructed by its local averages near certain sampling points. Examples are given.

  16. Testing linearity against nonlinear moving average models

    NARCIS (Netherlands)

    de Gooijer, J.G.; Brännäs, K.; Teräsvirta, T.

    1998-01-01

    Lagrange multiplier (LM) test statistics are derived for testing a linear moving average model against an additive smooth transition moving average model. The latter model is introduced in the paper. The small sample performance of the proposed tests are evaluated in a Monte Carlo study and compared

  17. Averaging Einstein's equations : The linearized case

    NARCIS (Netherlands)

    Stoeger, William R.; Helmi, Amina; Torres, Diego F.

    2007-01-01

    We introduce a simple and straightforward averaging procedure, which is a generalization of one which is commonly used in electrodynamics, and show that it possesses all the characteristics we require for linearized averaging in general relativity and cosmology for weak-field and perturbed FLRW situ

  18. Average Transmission Probability of a Random Stack

    Science.gov (United States)

    Lu, Yin; Miniatura, Christian; Englert, Berthold-Georg

    2010-01-01

    The transmission through a stack of identical slabs that are separated by gaps with random widths is usually treated by calculating the average of the logarithm of the transmission probability. We show how to calculate the average of the transmission probability itself with the aid of a recurrence relation and derive analytical upper and lower…

  19. Average excitation potentials of air and aluminium

    NARCIS (Netherlands)

    Bogaardt, M.; Koudijs, B.

    1951-01-01

    By means of a graphical method the average excitation potential I may be derived from experimental data. Average values for Iair and IAl have been obtained. It is shown that in representing range/energy relations by means of Bethe's well known formula, I has to be taken as a continuously changing fu

  20. 49 CFR 531.5 - Fuel economy standards.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 6 2010-10-01 2010-10-01 false Fuel economy standards. 531.5 Section 531.5... ADMINISTRATION, DEPARTMENT OF TRANSPORTATION PASSENGER AUTOMOBILE AVERAGE FUEL ECONOMY STANDARDS § 531.5 Fuel economy standards. (a) Except as provided in paragraph (e) of this section, each manufacturer of...

  1. New results on averaging theory and applications

    Science.gov (United States)

    Cândido, Murilo R.; Llibre, Jaume

    2016-08-01

    The usual averaging theory reduces the computation of some periodic solutions of a system of ordinary differential equations, to find the simple zeros of an associated averaged function. When one of these zeros is not simple, i.e., the Jacobian of the averaged function in it is zero, the classical averaging theory does not provide information about the periodic solution associated to a non-simple zero. Here we provide sufficient conditions in order that the averaging theory can be applied also to non-simple zeros for studying their associated periodic solutions. Additionally, we do two applications of this new result for studying the zero-Hopf bifurcation in the Lorenz system and in the Fitzhugh-Nagumo system.

  2. The effects of oxygen-enriched intake air on FFV exhaust emissions using M85

    Energy Technology Data Exchange (ETDEWEB)

    Poola, R.B.; Sekar, R.; Ng, H.K. [Argonne National Lab., IL (United States); Baudino, J.H. [Autoresearch Labs., Inc., Chicago, IL (United States); Colucci, C.P. [National Renewable Energy Lab., Golden, CO (United States)

    1996-05-01

    This paper presents results of emission tests of a flexible fuel vehicle (FFV) powered by an SI engine, fueled by M85 (methanol), and supplied with oxygen-enriched intake air containing 21, 23, and 25 vol% O2. Engine-out total hydrocarbons (THCs) and unburned methanol were considerably reduced in the entire FTP cycle when the O2 content of the intake air was either 23 or 25%. However, CO emissions did not vary much, and NOx emissions were higher. HCHO emissions were reduced by 53% in bag 1, 84% in bag 2, and 59% in bag 3 of the FTP cycle with 25% oxygen-enriched intake air. During cold-phase FTP,reductions of 42% in THCs, 40% in unburned methanol, 60% in nonmethane hydrocarbons, and 45% in nonmethane organic gases (NMOGs) were observed with 25% enriched air; NO{sub x} emissions increased by 78%. Converter-out emissions were also reduced with enriched air but to a lesser degree. FFVs operating on M85 that use 25% enriched air during only the initial 127 s of cold-phase FTP or that use 23 or 25% enriched air during only cold-phase FTP can meet the reactivity-adjusted NMOG, CO, NO{sub x}, and HCHO emission standards of the transitional low-emission vehicle.

  3. Enrichment of lanthanides in aragonite

    Institute of Scientific and Technical Information of China (English)

    瞿成利; 路波; 刘刚

    2009-01-01

    Using the constant addition technique,the coprecipitation of lanthanum,gadolinium,and lutetium with aragonite in seawater was experimentally investigated at 25 ℃.Their concentrations in aragonite overgrowths were determined by inductive coupled plasma mass spectrometer.All these lanthanides were strongly enriched in aragonite overgrowths.The amount of lanthanum,gadolinium,and lutetium incorporated into aragonite accounted for 57%-99%,50%-89%,and 40%-91% of their initial total amount,respectively.With the in...

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

    Energy Technology Data Exchange (ETDEWEB)

    Kanda, Keiji; Nishihara, Hedeaki [Kyoto Univ., Osaka (Japan); Shirai, Eiji; Oyamada, Rokuro; Sanokawa, Konomo [Japan Atomic Energy Research Institute, Tokyo (Japan)

    1997-08-01

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

  5. Enrichment and aggression in primates.

    Science.gov (United States)

    Honess, P E; Marin, C M

    2006-01-01

    There is considerable evidence that primates housed under impoverished conditions develop behavioural abnormalities, including, in the most extreme example, self-harming behaviour. This has implications for all contexts in which primates are maintained in captivity from laboratories to zoos since by compromising the animals' psychological well-being and allowing them to develop behavioural abnormalities their value as appropriate educational and research models is diminished. This review examines the extensive body of literature documenting attempts to improve living conditions with a view to correcting behavioural abnormalities and housing primates in such a way that they are encouraged to exhibit a more natural range and proportion of behaviours, including less self-directed and social aggression. The results of housing, feeding, physical, sensory and social enrichment efforts are examined with specific focus on their effect on aggressive behaviour and variation in their use and efficacy. It is concluded that while inappropriate or poorly distributed enrichment may encourage aggressive competition, enrichment that is species, sex, age and background appropriate can dramatically reduce aggression, can eliminate abnormal behaviour and substantially improve the welfare of primates maintained in captivity.

  6. Fuel cells:

    DEFF Research Database (Denmark)

    Sørensen, Bent

    2013-01-01

    A brief overview of the progress in fuel cell applications and basic technology development is presented, as a backdrop for discussing readiness for penetration into the marketplace as a solution to problems of depletion, safety, climate or environmental impact from currently used fossil and nucl......A brief overview of the progress in fuel cell applications and basic technology development is presented, as a backdrop for discussing readiness for penetration into the marketplace as a solution to problems of depletion, safety, climate or environmental impact from currently used fossil...

  7. Solar fuels

    Energy Technology Data Exchange (ETDEWEB)

    Bolton, J.R.

    1978-11-17

    The paper is concerned with (1) the thermodynamic and kinetic limits for the photochemical conversion and storage of solar energy as it is received on the earth's surface, and (2) the evaluation of a number of possible photochemical reactions with particular emphasis on the production of solar hydrogen from water. Procedures for generating hydrogen fuel are considered. Topics examined include the general requirements for a fuel-generation reaction, the photochemical reaction, limits on the conversion of light energy to chemical energy, an estimate of chemical storage efficiency, and the water decomposition reaction.

  8. Study of an ADS Loaded with Thorium and Reprocessed Fuel

    Directory of Open Access Journals (Sweden)

    Graiciany de Paula Barros

    2012-01-01

    Full Text Available Accelerator-driven systems (ADSs are investigated for long-lived fission product transmutation and fuel regeneration. The aim of this paper is to investigate the nuclear fuel evolution and the neutronic parameters of a lead-cooled accelerator-driven system used for fuel breeding. The fuel used in some fuel rods was T232hO2 for U233 production. In the other fuel rods was used a mixture based upon Pu-MA, removed from PWR-spent fuel, reprocessed by GANEX, and finally spiked with thorium or depleted uranium. The use of reprocessed fuel ensured the use of T232hO2 without the initial requirement of U233 enrichment. In this paper was used the Monte Carlo code MCNPX 2.6.0 that presents the depletion/burnup capability, combining an ADS source and kcode-mode (for criticality calculations. The multiplication factor (keff evolution, the neutron energy spectra in the core at BOL, and the nuclear fuel evolution during the burnup were evaluated. The results indicated that the combined use of T232hO2 and reprocessed fuel allowed U233 production without the initial requirement of U233 enrichment.

  9. Winters fuels report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-10-27

    The outlook for distillate fuel oil this winter is for increased demand and a return to normal inventory patterns, assuming a resumption of normal, cooler weather than last winter. With industrial production expected to grow slightly from last winter`s pace, overall consumption is projected to increase 3 percent from last winter, to 3.4 million barrels per day during the heating season (October 1, 1995-March 31, 1996). Much of the supply win come from stock drawdowns and refinery production. Estimates for the winter are from the Energy Information Administration`s (EIA) 4th Quarter 1995 Short-Tenn Energy Outlook (STEO) Mid-World Oil Price Case forecast. Inventories in place on September 30, 1995, of 132 million barrels were 9 percent below the unusually high year-earlier level. Inventories of high-sulfur distillate fuel oil, the principal type used for heating, were 13 percent lower than a year earlier. Supply problems are not anticipated because refinery production and the ready availability of imports should be adequate to meet demand. Residential heating off prices are expected to be somewhat higher than last winter`s, as the effects of lower crude oil prices are offset by lower distillate inventories. Heating oil is forecast to average $0.92 per gallon, the highest price since the winter of 1992-93. Diesel fuel (including tax) is predicted to be slightly higher than last year at $1.13 per gallon. This article focuses on the winter assessment for distillate fuel oil, how well last year`s STEO winter outlook compared to actual events, and expectations for the coming winter. Additional analyses include regional low-sulfur and high-sulfur distillate supply, demand, and prices, and recent trends in distillate fuel oil inventories.

  10. Container concept for the PBMR fuel plant

    Energy Technology Data Exchange (ETDEWEB)

    Braehler, G.; Kelly, M.; Sauer, L.; Welbers, P. [NUKEM Technologies GmbH (Germany)

    2009-07-01

    This presentation describes the types and the characteristics of the containers uniquely designed for and used in the PBMR Fuel Plant in Pelindaba, South Africa. The PBMR Fuel Plant is designed to manufacture so called Fuel Spheres for the High Temperature Reactor which shall be built in Koeberg near Cape Town. The PBMR fuel is based on a proven, high quality German fuel design consisting of 10% enriched Uranium triple-coated isotropic (LEU-TRISO) particles contained in a moulded graphite sphere. A coated particle comprises a kernel of Uranium dioxide surrounded by different layers. A fuel sphere consists of 9 g of Uranium (some 15 000 particles) and has a diameter of 60 mm; the total mass of a fuel sphere is 210 g. During normal operation the PBMR core contains a load of 456,000 fuel spheres. The PBMR Fuel plant is designed to produce 270,000 fuel spheres a year with the option to doubling the throughput. Two basic types of containers are foreseen for the plant, i.e. the Safe Geometry Containers for transportation and storage of Uranium bearing material and Containers in non restricted geometry for solid raw and auxiliary materials required by the different processes. Unlike other concepts in fuel element plants, the Safe Geometry Containers are allowed to be stored, from a criticality point of view, everywhere in the plant. Special identification measures, physical and IT based, prevent wrong material being processed. The sub-critical configuration is still maintained after the specified fire scenario for the PBMR Fuel Plant. (orig.)

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

    Science.gov (United States)

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

    2012-06-01

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

  12. Fuel Cell and Battery Powered Forklifts

    DEFF Research Database (Denmark)

    Zhang, Zhe; Mortensen, Henrik H.; Jensen, Jes Vestervang

    2013-01-01

    A hydrogen-powered materials handling vehicle with a fuel cell combines the advantages of diesel/LPG and battery powered vehicles. Hydrogen provides the same consistent power and fast refueling capability as diesel and LPG, whilst fuel cells provide energy efficient and zero emission Electric...... propulsion similar to batteries. In this paper, the performance of a forklift powered by PEM fuel cells and lead acid batteries as auxiliary energy source is introduced and investigated. In this electromechanical propulsion system with hybrid energy/power sources, fuel cells will deliver average power...

  13. Network-based functional enrichment

    Directory of Open Access Journals (Sweden)

    Poirel Christopher L

    2011-11-01

    Full Text Available Abstract Background Many methods have been developed to infer and reason about molecular interaction networks. These approaches often yield networks with hundreds or thousands of nodes and up to an order of magnitude more edges. It is often desirable to summarize the biological information in such networks. A very common approach is to use gene function enrichment analysis for this task. A major drawback of this method is that it ignores information about the edges in the network being analyzed, i.e., it treats the network simply as a set of genes. In this paper, we introduce a novel method for functional enrichment that explicitly takes network interactions into account. Results Our approach naturally generalizes Fisher’s exact test, a gene set-based technique. Given a function of interest, we compute the subgraph of the network induced by genes annotated to this function. We use the sequence of sizes of the connected components of this sub-network to estimate its connectivity. We estimate the statistical significance of the connectivity empirically by a permutation test. We present three applications of our method: i determine which functions are enriched in a given network, ii given a network and an interesting sub-network of genes within that network, determine which functions are enriched in the sub-network, and iii given two networks, determine the functions for which the connectivity improves when we merge the second network into the first. Through these applications, we show that our approach is a natural alternative to network clustering algorithms. Conclusions We presented a novel approach to functional enrichment that takes into account the pairwise relationships among genes annotated by a particular function. Each of the three applications discovers highly relevant functions. We used our methods to study biological data from three different organisms. Our results demonstrate the wide applicability of our methods. Our algorithms are

  14. Experimental Demonstration of Squeezed State Quantum Averaging

    CERN Document Server

    Lassen, Mikael; Sabuncu, Metin; Filip, Radim; Andersen, Ulrik L

    2010-01-01

    We propose and experimentally demonstrate a universal quantum averaging process implementing the harmonic mean of quadrature variances. The harmonic mean protocol can be used to efficiently stabilize a set of fragile squeezed light sources with statistically fluctuating noise levels. The averaged variances are prepared probabilistically by means of linear optical interference and measurement induced conditioning. We verify that the implemented harmonic mean outperforms the standard arithmetic mean strategy. The effect of quantum averaging is experimentally tested both for uncorrelated and partially correlated noise sources with sub-Poissonian shot noise or super-Poissonian shot noise characteristics.

  15. Averaged Lema\\^itre-Tolman-Bondi dynamics

    CERN Document Server

    Isidro, Eddy G Chirinos; Piattella, Oliver F; Zimdahl, Winfried

    2016-01-01

    We consider cosmological backreaction effects in Buchert's averaging formalism on the basis of an explicit solution of the Lema\\^itre-Tolman-Bondi (LTB) dynamics which is linear in the LTB curvature parameter and has an inhomogeneous bang time. The volume Hubble rate is found in terms of the volume scale factor which represents a derivation of the simplest phenomenological solution of Buchert's equations in which the fractional densities corresponding to average curvature and kinematic backreaction are explicitly determined by the parameters of the underlying LTB solution at the boundary of the averaging volume. This configuration represents an exactly solvable toy model but it does not adequately describe our "real" Universe.

  16. FREQUENTIST MODEL AVERAGING ESTIMATION: A REVIEW

    Institute of Scientific and Technical Information of China (English)

    Haiying WANG; Xinyu ZHANG; Guohua ZOU

    2009-01-01

    In applications, the traditional estimation procedure generally begins with model selection.Once a specific model is selected, subsequent estimation is conducted under the selected model without consideration of the uncertainty from the selection process. This often leads to the underreporting of variability and too optimistic confidence sets. Model averaging estimation is an alternative to this procedure, which incorporates model uncertainty into the estimation process. In recent years, there has been a rising interest in model averaging from the frequentist perspective, and some important progresses have been made. In this paper, the theory and methods on frequentist model averaging estimation are surveyed. Some future research topics are also discussed.

  17. Averaging of Backscatter Intensities in Compounds

    Science.gov (United States)

    Donovan, John J.; Pingitore, Nicholas E.; Westphal, Andrew J.

    2002-01-01

    Low uncertainty measurements on pure element stable isotope pairs demonstrate that mass has no influence on the backscattering of electrons at typical electron microprobe energies. The traditional prediction of average backscatter intensities in compounds using elemental mass fractions is improperly grounded in mass and thus has no physical basis. We propose an alternative model to mass fraction averaging, based of the number of electrons or protons, termed “electron fraction,” which predicts backscatter yield better than mass fraction averaging. PMID:27446752

  18. Average-passage flow model development

    Science.gov (United States)

    Adamczyk, John J.; Celestina, Mark L.; Beach, Tim A.; Kirtley, Kevin; Barnett, Mark

    1989-01-01

    A 3-D model was developed for simulating multistage turbomachinery flows using supercomputers. This average passage flow model described the time averaged flow field within a typical passage of a bladed wheel within a multistage configuration. To date, a number of inviscid simulations were executed to assess the resolution capabilities of the model. Recently, the viscous terms associated with the average passage model were incorporated into the inviscid computer code along with an algebraic turbulence model. A simulation of a stage-and-one-half, low speed turbine was executed. The results of this simulation, including a comparison with experimental data, is discussed.

  19. Changing mortality and average cohort life expectancy

    DEFF Research Database (Denmark)

    Schoen, Robert; Canudas-Romo, Vladimir

    2005-01-01

    of survivorship. An alternative aggregate measure of period mortality which has been seen as less sensitive to period changes, the cross-sectional average length of life (CAL) has been proposed as an alternative, but has received only limited empirical or analytical examination. Here, we introduce a new measure......, the average cohort life expectancy (ACLE), to provide a precise measure of the average length of life of cohorts alive at a given time. To compare the performance of ACLE with CAL and with period and cohort life expectancy, we first use population models with changing mortality. Then the four aggregate...

  20. N-Reactor (U-metal) Fuel Characteristics for Disposal Criticality Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, Larry Lorin

    2000-05-01

    DOE-owned spent nuclear fuels encompass many fuel types. In an effort to facilitate criticality analysis for these various fuel types, they were categorized into nine characteristic fuel groups with emphasis on fuel matrix composition. Out of each fuel group, a representative fuel type was chosen for analysis as a bounding case within that fuel group. Generally, burnup data, fissile enrichments, and total fuel and fissile mass govern the selection of the representative or candidate fuel within that group. Additionally, the criticality analysis will also require data to support design of the canister internals, thermal, and radiation shielding. The purpose of this report is to consolidate and provide in a concise format, material and information/data needed to perform supporting analyses to qualify N-Reactor fuels for acceptance into the designated repository. The N Reactor fuels incorporate zirconium cladding and uranium metal with unique fabrication details in terms of physical size, and method of construction. The fuel construction and post-irradiation handling have created attendant issues relative to cladding failure in the underwater storage environment. These fuels were comprised of low-enriched metal (0.947 to 1.25 wt% 235U) that were originally intended to generate weapons-grade plutonium for national defense. Modifications in subsequent fuel design and changes in the mode of reactor operation in later years were focused more toward power production.

  1. Pyroprocessing of Fast Flux Test Facility Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-01

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

  2. NEAT: an efficient network enrichment analysis test

    OpenAIRE

    Signorelli, Mirko; Vinciotti, Veronica; Wit, Ernst C

    2016-01-01

    Background Network enrichment analysis is a powerful method, which allows to integrate gene enrichment analysis with the information on relationships between genes that is provided by gene networks. Existing tests for network enrichment analysis deal only with undirected networks, they can be computationally slow and are based on normality assumptions. Results We propose NEAT, a test for network enrichment analysis. The test is based on the hypergeometric distribution, which naturally arises ...

  3. Transport fuel

    DEFF Research Database (Denmark)

    Ronsse, Frederik; Jørgensen, Henning; Schüßler, Ingmar

    2014-01-01

    Worldwide, the use of transport fuel derived from biomass increased four-fold between 2003 and 2012. Mainly based on food resources, these conventional biofuels did not achieve the expected emission savings and contributed to higher prices for food commod - ities, especially maize and oilseeds...

  4. Fuel Cells

    Science.gov (United States)

    Hawkins, M. D.

    1973-01-01

    Discusses the theories, construction, operation, types, and advantages of fuel cells developed by the American space programs. Indicates that the cell is an ideal small-scale power source characterized by its compactness, high efficiency, reliability, and freedom from polluting fumes. (CC)

  5. Characterization of surface sediments from the Beijing-Hangzhou Grand Canal (Zaozhuang section), China: assessment of beryllium enrichment, biological effect, and mobility.

    Science.gov (United States)

    Zhuang, Wen; Chen, Qing; Gao, Xuelu; Zhou, Fengxia; Wang, Mantang; Liu, Yongxia

    2016-07-01

    The South-to-North Water Diversion Project is one of the world's largest water diversion projects, benefiting seven million people in China. The Zaozhuang section of the Beijing-Hangzhou Grand Canal is an important part of this project. This paper investigated the enrichment, biological effect, and mobility of beryllium (Be) in surface sediments of the Zaozhuang section. Results showed that high values were found in Tai'erzhuang District, Zaozhuang city, and the areas near the inlet of the Nansihu Lake, which might have been influenced by local human activities including metallurgy, burning of fossil fuels, and transportation. Four geochemical fractions of Be were obtained: acid-soluble fraction, reducible fraction, oxidizable fraction, and residual fraction. The non-residual fractions (the sum of the first three) accounted for 72.5 ∼ 96.1 % of the total amount of Be. Acid-soluble fraction might be mainly influenced by human activities, with the strongest mobility and bio-availability, accounting for 4.1 ∼ 44.7 % of the total amount, with an average of 20.2 %. Enrichment factor (EF) showed minor to moderate enrichment in some regions; adverse effect index (AEI) also showed that there were high levels of Be in some regions, which might have negative impacts on organisms. Generally, mobility, EF, and AEI of elements are carried out separately. But the results of this study indicated that a comprehensive assessment on the enrichment, mobility, and biological effects of Be caused by human activities is necessary in understanding the environmental risks of Be.

  6. A method for limitation of probability of accumulation of fuel elements claddings damage in WWER

    OpenAIRE

    Sergey N. Pelykh; Mark V. Nikolsky; S. D. Ryabchikov

    2014-01-01

    The aim is to reduce the probability of accumulation of fuel elements claddings damage by developing a method to control the properties of the fuel elements on stages of design and operation of WWER. An averaged over the fuel assembly WWER-1000 fuel element is considered. The probability of depressurization of fuel elements claddings is found. The ability to predict the reliability of claddings by controlling the factors that determine the properties of the fuel elements is proved. The expedi...

  7. A practical guide to averaging functions

    CERN Document Server

    Beliakov, Gleb; Calvo Sánchez, Tomasa

    2016-01-01

    This book offers an easy-to-use and practice-oriented reference guide to mathematical averages. It presents different ways of aggregating input values given on a numerical scale, and of choosing and/or constructing aggregating functions for specific applications. Building on a previous monograph by Beliakov et al. published by Springer in 2007, it outlines new aggregation methods developed in the interim, with a special focus on the topic of averaging aggregation functions. It examines recent advances in the field, such as aggregation on lattices, penalty-based aggregation and weakly monotone averaging, and extends many of the already existing methods, such as: ordered weighted averaging (OWA), fuzzy integrals and mixture functions. A substantial mathematical background is not called for, as all the relevant mathematical notions are explained here and reported on together with a wealth of graphical illustrations of distinct families of aggregation functions. The authors mainly focus on practical applications ...

  8. Rotational averaging of multiphoton absorption cross sections

    Energy Technology Data Exchange (ETDEWEB)

    Friese, Daniel H., E-mail: daniel.h.friese@uit.no; Beerepoot, Maarten T. P.; Ruud, Kenneth [Centre for Theoretical and Computational Chemistry, University of Tromsø — The Arctic University of Norway, N-9037 Tromsø (Norway)

    2014-11-28

    Rotational averaging of tensors is a crucial step in the calculation of molecular properties in isotropic media. We present a scheme for the rotational averaging of multiphoton absorption cross sections. We extend existing literature on rotational averaging to even-rank tensors of arbitrary order and derive equations that require only the number of photons as input. In particular, we derive the first explicit expressions for the rotational average of five-, six-, and seven-photon absorption cross sections. This work is one of the required steps in making the calculation of these higher-order absorption properties possible. The results can be applied to any even-rank tensor provided linearly polarized light is used.

  9. Rotational averaging of multiphoton absorption cross sections

    Science.gov (United States)

    Friese, Daniel H.; Beerepoot, Maarten T. P.; Ruud, Kenneth

    2014-11-01

    Rotational averaging of tensors is a crucial step in the calculation of molecular properties in isotropic media. We present a scheme for the rotational averaging of multiphoton absorption cross sections. We extend existing literature on rotational averaging to even-rank tensors of arbitrary order and derive equations that require only the number of photons as input. In particular, we derive the first explicit expressions for the rotational average of five-, six-, and seven-photon absorption cross sections. This work is one of the required steps in making the calculation of these higher-order absorption properties possible. The results can be applied to any even-rank tensor provided linearly polarized light is used.

  10. Sea Surface Temperature Average_SST_Master

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Sea surface temperature collected via satellite imagery from http://www.esrl.noaa.gov/psd/data/gridded/data.noaa.ersst.html and averaged for each region using ArcGIS...

  11. MN Temperature Average (1961-1990) - Line

    Data.gov (United States)

    Minnesota Department of Natural Resources — This data set depicts 30-year averages (1961-1990) of monthly and annual temperatures for Minnesota. Isolines and regions were created using kriging and...

  12. MN Temperature Average (1961-1990) - Polygon

    Data.gov (United States)

    Minnesota Department of Natural Resources — This data set depicts 30-year averages (1961-1990) of monthly and annual temperatures for Minnesota. Isolines and regions were created using kriging and...

  13. Spacetime Average Density (SAD) Cosmological Measures

    CERN Document Server

    Page, Don N

    2014-01-01

    The measure problem of cosmology is how to obtain normalized probabilities of observations from the quantum state of the universe. This is particularly a problem when eternal inflation leads to a universe of unbounded size so that there are apparently infinitely many realizations or occurrences of observations of each of many different kinds or types, making the ratios ambiguous. There is also the danger of domination by Boltzmann Brains. Here two new Spacetime Average Density (SAD) measures are proposed, Maximal Average Density (MAD) and Biased Average Density (BAD), for getting a finite number of observation occurrences by using properties of the Spacetime Average Density (SAD) of observation occurrences to restrict to finite regions of spacetimes that have a preferred beginning or bounce hypersurface. These measures avoid Boltzmann brain domination and appear to give results consistent with other observations that are problematic for other widely used measures, such as the observation of a positive cosmolo...

  14. Monthly snow/ice averages (ISCCP)

    Data.gov (United States)

    National Aeronautics and Space Administration — September Arctic sea ice is now declining at a rate of 11.5 percent per decade, relative to the 1979 to 2000 average. Data from NASA show that the land ice sheets in...

  15. Appeals Council Requests - Average Processing Time

    Data.gov (United States)

    Social Security Administration — This dataset provides annual data from 1989 through 2015 for the average processing time (elapsed time in days) for dispositions by the Appeals Council (AC) (both...

  16. Average Annual Precipitation (PRISM model) 1961 - 1990

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — This map layer shows polygons of average annual precipitation in the contiguous United States, for the climatological period 1961-1990. Parameter-elevation...

  17. Average Vegetation Growth 1990 - Direct Download

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — This map layer is a grid map of 1990 average vegetation growth for Alaska and the conterminous United States. The nominal spatial resolution is 1 kilometer and the...

  18. Average Vegetation Growth 1997 - Direct Download

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — This map layer is a grid map of 1997 average vegetation growth for Alaska and the conterminous United States. The nominal spatial resolution is 1 kilometer and the...

  19. Average Vegetation Growth 1992 - Direct Download

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — This map layer is a grid map of 1992 average vegetation growth for Alaska and the conterminous United States. The nominal spatial resolution is 1 kilometer and the...

  20. Average Vegetation Growth 2001 - Direct Download

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — This map layer is a grid map of 2001 average vegetation growth for Alaska and the conterminous United States. The nominal spatial resolution is 1 kilometer and the...

  1. Average Vegetation Growth 1995 - Direct Download

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — This map layer is a grid map of 1995 average vegetation growth for Alaska and the conterminous United States. The nominal spatial resolution is 1 kilometer and the...

  2. Average Vegetation Growth 2000 - Direct Download

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — This map layer is a grid map of 2000 average vegetation growth for Alaska and the conterminous United States. The nominal spatial resolution is 1 kilometer and the...

  3. Average Vegetation Growth 1998 - Direct Download

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — This map layer is a grid map of 1998 average vegetation growth for Alaska and the conterminous United States. The nominal spatial resolution is 1 kilometer and the...

  4. Average Vegetation Growth 1994 - Direct Download

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — This map layer is a grid map of 1994 average vegetation growth for Alaska and the conterminous United States. The nominal spatial resolution is 1 kilometer and the...

  5. Enriching Music and Language Arts Experiences

    Science.gov (United States)

    Flohr, John W.

    2006-01-01

    The article focuses on enriching music and language arts experiences of students. Music can enrich literature and language arts, poetry, theater arts, transitions, science, and math, as well as help meet special learner needs. A well-understood example of enrichment is the alphabet song. A music or classroom teacher using the alphabet song helps…

  6. How Did the IGM Become Enriched?

    CERN Document Server

    Aguirre, A; Aguirre, Anthony; Schaye, Joop

    2006-01-01

    The enrichment of the intergalactic medium with heavy elements is a process that lies at the nexus of poorly-understood aspects of physical cosmology. We review current understanding of the processes that may remove metals from galaxies, the basic predictions of these models, the key observational constraints on enrichment, and how intergalactic enrichment may be used to test cosmological simulations.

  7. Symmetric Euler orientation representations for orientational averaging.

    Science.gov (United States)

    Mayerhöfer, Thomas G

    2005-09-01

    A new kind of orientation representation called symmetric Euler orientation representation (SEOR) is presented. It is based on a combination of the conventional Euler orientation representations (Euler angles) and Hamilton's quaternions. The properties of the SEORs concerning orientational averaging are explored and compared to those of averaging schemes that are based on conventional Euler orientation representations. To that aim, the reflectance of a hypothetical polycrystalline material with orthorhombic crystal symmetry was calculated. The calculation was carried out according to the average refractive index theory (ARIT [T.G. Mayerhöfer, Appl. Spectrosc. 56 (2002) 1194]). It is shown that the use of averaging schemes based on conventional Euler orientation representations leads to a dependence of the result from the specific Euler orientation representation that was utilized and from the initial position of the crystal. The latter problem can be overcome partly by the introduction of a weighing factor, but only for two-axes-type Euler orientation representations. In case of a numerical evaluation of the average, a residual difference remains also if a two-axes type Euler orientation representation is used despite of the utilization of a weighing factor. In contrast, this problem does not occur if a symmetric Euler orientation representation is used as a matter of principle, while the result of the averaging for both types of orientation representations converges with increasing number of orientations considered in the numerical evaluation. Additionally, the use of a weighing factor and/or non-equally spaced steps in the numerical evaluation of the average is not necessary. The symmetrical Euler orientation representations are therefore ideally suited for the use in orientational averaging procedures.

  8. Average Bandwidth Allocation Model of WFQ

    Directory of Open Access Journals (Sweden)

    Tomáš Balogh

    2012-01-01

    Full Text Available We present a new iterative method for the calculation of average bandwidth assignment to traffic flows using a WFQ scheduler in IP based NGN networks. The bandwidth assignment calculation is based on the link speed, assigned weights, arrival rate, and average packet length or input rate of the traffic flows. We prove the model outcome with examples and simulation results using NS2 simulator.

  9. Reprocessing of LEU U-Mo Dispersion and Monolithic Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Vandegrift, G.F.; Jerden, J.; Stepinski, D.C.; Figueroa, J.; Williamson, M.A.; Kleeck, M.A. Van; Blaskovitz, R.J.; Ziegler, A.J.; Maggos, L.E.; Swanson, J.; Fortner, J.; Bakel, A.J. [Chemical Sciences and Engineering Division, Argonne National Laboratory, 9700 S. Cass Ave., Argonne, IL 60439 (United States)

    2011-07-01

    For conversion of high-performance research reactors from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel, a fuel material with a higher density than uranium aluminide is required. Development studies are underway to develop U-Mo dispersion and monolithic fuels for conversion of several high- performance reactors. For dispersion fuels, development is narrowing down to a composition of U-7Mo dispersed in an aluminium matrix containing {approx}5% silicon. For monolithic fuels to be used in high performance research reactors in the United States, a zirconium-bonded U-10Mo foil appears to be the fuel of choice. For conversion to be realized a back-end disposition path is required for both fuels; one disposition pathway is reprocessing. Argonne National Laboratory is developing a pyroprocess for reprocessing spent monolithic fuel. Pyroprocessing was chosen over conventional aqueous solvent extraction due to the necessity of adding fluoride to the fuel-dissolution solution in order to dissolve the zirconium bonding layer on the U-Mo fuel. The proposed flowsheet and development activities will be described. A literature survey points to the ability to reprocess U-Mo dispersion fuels by an aqueous process, but due to several special characteristics of the fuel, the solvent-extraction flowsheets will be a departure from that normally used for the reprocessing of power reactor fuel. Special concerns that must be addressed in reprocessing these fuels are, for example, the low solubilities of uranyl molybdate, molybdic acid, and silicic acid in nitric acid solutions. This paper will address these concerns and development activities required to overcome them. (author)

  10. 40 CFR 86.1817-05 - Complete heavy-duty vehicle averaging, trading, and banking program.

    Science.gov (United States)

    2010-07-01

    ... in clean-fuel vehicles as specified in 40 CFR part 88 are not eligible for this program..., trading, and banking program. 86.1817-05 Section 86.1817-05 Protection of Environment ENVIRONMENTAL... Complete heavy-duty vehicle averaging, trading, and banking program. (a) General. (1) Complete...

  11. Water-moderated reactor fuel cladding reliability study

    OpenAIRE

    Бакутяк, Елена Викторовна; Пелых, Сергей Николаевич

    2014-01-01

    Considering the fuel element, averaged by fuel assembly (FA) of water-moderated reactor with the power of 1000 MW (VVER-1000), the number of fuel elements with the greatest cladding failure probability after 4 operation years at Khmelnitsky NPP-2 (KNPP-2) is found. This will allow to calculate the fuel cladding failure probability and determine the most likely cladding damages, which will enable to improve the performance and economic indexes of VVER.The novelty of the paper lies in calculati...

  12. Thermal analysis of IRT-T reactor fuel elements

    OpenAIRE

    Naymushin, Artem Georgievich; Chertkov, Yuri Borisovich; Lebedev, Ivan Igorevich; Anikin, Mikhail Nikolaevich

    2015-01-01

    The article describes the method and results of thermo-physical calculations of IRT-T reactor core. Heat fluxes, temperatures of cladding, fuel meat and coolant were calculated for height of core, azimuth directions of FA and each fuel elements in FA. Average calculated values of uniformity factor of energy release distribution for height of fuel assemblies were shown in this research. Onset nucleate boiling temperature and ONB-ratio were calculated. Shows that temperature regimes of fuel ele...

  13. Accelerating the Reduction of Excess Russian Highly Enriched Uranium

    Energy Technology Data Exchange (ETDEWEB)

    Benton, J; Wall, D; Parker, E; Rutkowski, E

    2004-02-18

    This paper presents the latest information on one of the Accelerated Highly Enriched Uranium (HEU) Disposition initiatives that resulted from the May 2002 Summit meeting between Presidents George W. Bush and Vladimir V. Putin. These initiatives are meant to strengthen nuclear nonproliferation objectives by accelerating the disposition of nuclear weapons-useable materials. The HEU Transparency Implementation Program (TIP), within the National Nuclear Security Administration (NNSA) is working to implement one of the selected initiatives that would purchase excess Russian HEU (93% 235U) for use as fuel in U.S. research reactors over the next ten years. This will parallel efforts to convert the reactors' fuel core from HEU to low enriched uranium (LEU) material, where feasible. The paper will examine important aspects associated with the U.S. research reactor HEU purchase. In particular: (1) the establishment of specifications for the Russian HEU, and (2) transportation safeguard considerations for moving the HEU from the Mayak Production Facility in Ozersk, Russia, to the Y-12 National Security Complex in Oak Ridge, TN.

  14. Laser Isotope Enrichment for Medical and Industrial Applications

    Energy Technology Data Exchange (ETDEWEB)

    Leonard Bond

    2006-07-01

    Laser Isotope Enrichment for Medical and Industrial Applications by Jeff Eerkens (University of Missouri), Jay Kunze (Idaho State University), and Leonard Bond (Idaho National Laboratory) The principal isotope enrichment business in the world is the enrichment of uranium for commercial power reactor fuels. However, there are a number of other needs for separated isotopes. Some examples are: 1) Pure isotopic targets for irradiation to produce medical radioisotopes. 2) Pure isotopes for semiconductors. 3) Low neutron capture isotopes for various uses in nuclear reactors. 4) Isotopes for industrial tracer/identification applications. Examples of interest to medicine are targets to produce radio-isotopes such as S-33, Mo-98, Mo-100, W-186, Sn-112; while for MRI diagnostics, the non-radioactive Xe-129 isotope is wanted. For super-semiconductor applications some desired industrial isotopes are Si-28, Ga-69, Ge-74, Se-80, Te-128, etc. An example of a low cross section isotope for use in reactors is Zn-68 as a corrosion inhibitor material in nuclear reactor primary systems. Neutron activation of Ar isotopes is of interest in industrial tracer and diagnostic applications (e.g. oil-logging). . In the past few years there has been a sufficient supply of isotopes in common demand, because of huge Russian stockpiles produced with old electromagnetic and centrifuge separators previously used for uranium enrichment. Production of specialized isotopes in the USA has been largely accomplished using old ”calutrons” (electromagnetic separators) at Oak Ridge National Laboratory. These methods of separating isotopes are rather energy inefficient. Use of lasers for isotope separation has been considered for many decades. None of the proposed methods have attained sufficient proof of principal status to be economically attractive to pursue commercially. Some of the authors have succeeded in separating sulfur isotopes using a rather new and different method, known as condensation

  15. Enrichment planting without soil treatment

    Energy Technology Data Exchange (ETDEWEB)

    Hagner, Mats

    1998-12-31

    Where enrichment planting had been carried out with either of the two species Picea abies and Pinus contorta, the survival of the planted seedlings was at least as good as after planting in a normal clear cut area treated with soil scarification. This was in spite of the fact that the seedlings were placed shallow in the humus layer without any soil treatment. However, they were sheltered from insects by treatment before planting. Where enrichment planting was carried out with Pinus sylvestris the survival in dense forest was poor, but in open forest the survival was good. The growth of planted seedlings was enhanced by traditional clearing and soil treatment. However, this was for Pinus sylvestris not enough to compensate for the loss of time, 1-2 years, caused by arrangement of soil scarification. The growth of seedlings planted under crown cover was directly related to basal area of retained trees. However, the variation in height growth among individual seedlings was very big, which meant that some seedlings grow well also under a fairly dense forest cover. The pioneer species Pinus sylvestris reacted more strongly to basal area of retained trees than did the shade tolerant species Picea abies. Enrichment planting seems to be a necessary tool for preserving volume productivity, at places where fairly intensive harvest of mature trees has been carried out in stands of ordinary forest type in central Sweden. If double seedlings, with one Picea abies and one Pinus sylvestris, are used, the probability for long term establishment is enhanced 13 refs, 20 figs, 4 tabs

  16. Boron enrichment in martian clay.

    Directory of Open Access Journals (Sweden)

    James D Stephenson

    Full Text Available We have detected a concentration of boron in martian clay far in excess of that in any previously reported extra-terrestrial object. This enrichment indicates that the chemistry necessary for the formation of ribose, a key component of RNA, could have existed on Mars since the formation of early clay deposits, contemporary to the emergence of life on Earth. Given the greater similarity of Earth and Mars early in their geological history, and the extensive disruption of Earth's earliest mineralogy by plate tectonics, we suggest that the conditions for prebiotic ribose synthesis may be better understood by further Mars exploration.

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

    Energy Technology Data Exchange (ETDEWEB)

    Ade, Brian J [ORNL; Gauld, Ian C [ORNL

    2011-10-01

    In currently operating commercial nuclear power plants (NPP), there are two main types of nuclear fuel, low enriched uranium (LEU) fuel, and mixed-oxide uranium-plutonium (MOX) fuel. The LEU fuel is made of pure uranium dioxide (UO{sub 2} or UOX) and has been the fuel of choice in commercial light water reactors (LWRs) for a number of years. Naturally occurring uranium contains a mixture of different uranium isotopes, primarily, {sup 235}U and {sup 238}U. {sup 235}U is a fissile isotope, and will readily undergo a fission reaction upon interaction with a thermal neutron. {sup 235}U has an isotopic concentration of 0.71% in naturally occurring uranium. For most reactors to maintain a fission chain reaction, the natural isotopic concentration of {sup 235}U must be increased (enriched) to a level greater than 0.71%. Modern nuclear reactor fuel assemblies contain a number of fuel pins potentially having different {sup 235}U enrichments varying from {approx}2.0% to {approx}5% enriched in {sup 235}U. Currently in the United States (US), all commercial nuclear power plants use UO{sub 2} fuel. In the rest of the world, UO{sub 2} fuel is still commonly used, but MOX fuel is also used in a number of reactors. MOX fuel contains a mixture of both UO{sub 2} and PuO{sub 2}. Because the plutonium provides the fissile content of the fuel, the uranium used in MOX is either natural or depleted uranium. PuO{sub 2} is added to effectively replace the fissile content of {sup 235}U so that the level of fissile content is sufficiently high to maintain the chain reaction in an LWR. Both reactor-grade and weapons-grade plutonium contains a number of fissile and non-fissile plutonium isotopes, with the fraction of fissile and non-fissile plutonium isotopes being dependent on the source of the plutonium. While only RG plutonium is currently used in MOX, there is the possibility that WG plutonium from dismantled weapons will be used to make MOX for use in US reactors. Reactor-grade plutonium

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-04-01

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

  19. Core analysis during transition from 37-element fuel to CANFLEX-NU fuel in CANDU 6

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Chang Joon; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    An 1200-day time-dependent fuel-management for the transition from 37-element fuel to CANFLEX-NU fuel in a CANDU 6 reactor has been simulated to show the compatibility of the CANFLEX-NU fuel with the reactor operation. The simulation calculations were carried out with the RFSP code, provided by cell averaged fuel properties obtained from the POWDERPUFS-V code. The refueling scheme for both fuels was an eight bundle shift at a time. The simulation results show that the maximum channel and bundle powers were maintained below the license limit of the CANDU 6. This indicates that the CANFLEX-NU fuel bundle is compatible with the CANDU 6 reactor operation during the transition period. 3 refs., 2 figs., 1 tab. (Author)

  20. 14 CFR Appendix M to Part 25 - Fuel Tank System Flammability Reduction Means

    Science.gov (United States)

    2010-01-01

    .... (a) The Fleet Average Flammability Exposure of each fuel tank, as determined in accordance with... failures of the FRM that occur in service that could increase any fuel tank's Fleet Average Flammability... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Fuel Tank System Flammability...

  1. Modern new nuclear fuel characteristics and radiation protection aspects.

    Science.gov (United States)

    Terry, Ian R

    2005-01-01

    The glut of fissile material from reprocessing plants and from the conclusion of the cold war has provided the opportunity to design new fuel types to beneficially dispose of such stocks by generating useful power. Thus, in addition to the normal reactor core complement of enriched uranium fuel assemblies, two other types are available on the world market. These are the ERU (enriched recycled uranium) and the MOX (mixed oxide) fuel assemblies. Framatome ANP produces ERU fuel assemblies by taking feed material from reprocessing facilities and blending this with highly enriched uranium from other sources. MOX fuel assemblies contain plutonium isotopes, thus exploiting the higher neutron yield of the plutonium fission process. This paper describes and evaluates the gamma, spontaneous and alpha reaction neutron source terms of these non-irradiated fuel assembly types by defining their nuclear characteristics. The dose rates which arise from these terms are provided along with an overview of radiation protection aspects for consideration in transporting and delivering such fuel assemblies to power generating utilities.

  2. Neutronics and Fuel Performance Evaluation of Accident Tolerant Fuel under Normal Operation Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Xu Wu; Piyush Sabharwall; Jason Hales

    2014-07-01

    This report details the analysis of neutronics and fuel performance analysis for enhanced accident tolerance fuel, with Monte Carlo reactor physics code Serpent and INL’s fuel performance code BISON, respectively. The purpose is to evaluate two of the most promising candidate materials, FeCrAl and Silicon Carbide (SiC), as the fuel cladding under normal operating conditions. Substantial neutron penalty is identified when FeCrAl is used as monolithic cladding for current oxide fuel. From the reactor physics standpoint, application of the FeCrAl alloy as coating layer on surface of zircaloy cladding is possible without increasing fuel enrichment. Meanwhile, SiC brings extra reactivity and the neutron penalty is of no concern. Application of either FeCrAl or SiC could be favorable from the fuel performance standpoint. Detailed comparison between monolithic cladding and hybrid cladding (cladding + coating) is discussed. Hybrid cladding is more practical based on the economics evaluation during the transition from current UO2/zircaloy to Accident Tolerant Fuel (ATF) system. However, a few issues remain to be resolved, such as the creep behavior of FeCrAl, coating spallation, inter diffusion with zirconium, etc. For SiC, its high thermal conductivity, excellent creep resistance, low thermal neutron absorption cross section, irradiation stability (minimal swelling) make it an excellent candidate materials for future nuclear fuel/cladding system.

  3. Average Temperatures in the Southwestern United States, 2000-2015 Versus Long-Term Average

    Data.gov (United States)

    U.S. Environmental Protection Agency — This indicator shows how the average air temperature from 2000 to 2015 has differed from the long-term average (1895–2015). To provide more detailed information,...

  4. Averaged controllability of parameter dependent conservative semigroups

    Science.gov (United States)

    Lohéac, Jérôme; Zuazua, Enrique

    2017-02-01

    We consider the problem of averaged controllability for parameter depending (either in a discrete or continuous fashion) control systems, the aim being to find a control, independent of the unknown parameters, so that the average of the states is controlled. We do it in the context of conservative models, both in an abstract setting and also analysing the specific examples of the wave and Schrödinger equations. Our first result is of perturbative nature. Assuming the averaging probability measure to be a small parameter-dependent perturbation (in a sense that we make precise) of an atomic measure given by a Dirac mass corresponding to a specific realisation of the system, we show that the averaged controllability property is achieved whenever the system corresponding to the support of the Dirac is controllable. Similar tools can be employed to obtain averaged versions of the so-called Ingham inequalities. Particular attention is devoted to the 1d wave equation in which the time-periodicity of solutions can be exploited to obtain more precise results, provided the parameters involved satisfy Diophantine conditions ensuring the lack of resonances.

  5. Novel Accident-Tolerant Fuel Meat and Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Robert D. Mariani; Pavel G Medvedev; Douglas L Porter; Steven L Hayes; James I. Cole; Xian-Ming Bai

    2013-09-01

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

  6. Environmental enrichment for primates in laboratories

    Science.gov (United States)

    Buchanan-Smith, H. M.

    2010-06-01

    Environmental enrichment is a critical component of Refinement, one of the 3Rs underlying humane experimentation on animals. In this paper I discuss why primates housed in laboratories, which often have constraints of space and study protocols, are a special case for enrichment. I outline a framework for categorising the different types of enrichment, using the marmoset as a case study, and summarise the methods used to determine what animals want/prefer. I briefly review the arguments that enrichment does not negatively affect experimental outcomes. Finally I focus on complexity and novelty, choice and control, the underlying features of enrichment that makes it successful, and how combined with a thorough understanding of natural history we can put effective enrichment into practice in laboratories. Throughout the paper I emphasise the need to evaluate enrichment to ensure it is having the desired effect.

  7. Foundations for the definition of MOX fuel quality requirements

    Science.gov (United States)

    Bairiot, H.; Deramaix, P.; Mostin, N.; Trauwaert, E.; Vanderborck, Y.

    1991-02-01

    The quality of uranium-plutonium mixed oxide (MOX) fuel, as of any nuclear fuel, depends on the design optimization and on the fabrication process stability. The design optimization is essentially based on feed-back from irradiation experience through engineering assessment of the results; the stability of the process is necessary to justify minimal uncertainty margins in the fuel design. Since MOX fuel is quite similar to UO 2 fuel, the lessons learned from UO 2 fuels can complement the MOX experimental data base. MOX is however different from UO 2 fuel in some respects, among others: - the industrial fabrication scale is a factor 10 lower than for UO 2 fuel, - the fuel enrichment process takes place in the manufacturing plant, - the radioactivity of Pu imposes handling constraints, - Pu ages quite rapidly, altering its isotopic composition during storage, - the incorporation of Pu alters the material physics and neutronic characteristics of the fuel. In this perspective, the paper outlines some quality attributes for which MOX fuel may or even must depart from UO 2 fuel.

  8. Preliminary Calculation on a Spent Fuel Pool Accident using GOTHIC

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jaehwan; Choi, Yu Jung; Hong, Tae Hyub; Kim, Hyeong-Taek [KHNP-CRI, Daejeon (Korea, Republic of)

    2015-10-15

    The probability of an accident happening at the spent fuel pool was believed to be quite low until the 2011 Fukushima accident occurred. Notably, large amount of spent fuel are normally stored in the spent fuel pool for a long time compared to the amount of fuel in the reactor core and the total heat released from the spent fuel is high enough to boil the water of the spent fuel pool when the cooling system does not operate. In addition, the enrichment and the burnup of the fuel have both increased in the past decade and heat generation from the spent fuel thereby has also increased. The failure of the cooling system at the spent fuel pool (hereafter, a loss-of-cooling accident) is one of the principal hypothetical causes of an accident that could occur at the spent fuel pool. In this paper, the preliminary calculation of a loss-of-cooling accident was performed. In this paper, the preliminary calculation of a loss-of cooling accident was performed with GOTHIC. The calculation results show boiling away of water in the spent fuel pool due to the loss-of-cooling accident and similar thermal performance of the spent fuel pool with previous research results.

  9. Stochastic Averaging and Stochastic Extremum Seeking

    CERN Document Server

    Liu, Shu-Jun

    2012-01-01

    Stochastic Averaging and Stochastic Extremum Seeking develops methods of mathematical analysis inspired by the interest in reverse engineering  and analysis of bacterial  convergence by chemotaxis and to apply similar stochastic optimization techniques in other environments. The first half of the text presents significant advances in stochastic averaging theory, necessitated by the fact that existing theorems are restricted to systems with linear growth, globally exponentially stable average models, vanishing stochastic perturbations, and prevent analysis over infinite time horizon. The second half of the text introduces stochastic extremum seeking algorithms for model-free optimization of systems in real time using stochastic perturbations for estimation of their gradients. Both gradient- and Newton-based algorithms are presented, offering the user the choice between the simplicity of implementation (gradient) and the ability to achieve a known, arbitrary convergence rate (Newton). The design of algorithms...

  10. Books average previous decade of economic misery.

    Science.gov (United States)

    Bentley, R Alexander; Acerbi, Alberto; Ormerod, Paul; Lampos, Vasileios

    2014-01-01

    For the 20(th) century since the Depression, we find a strong correlation between a 'literary misery index' derived from English language books and a moving average of the previous decade of the annual U.S. economic misery index, which is the sum of inflation and unemployment rates. We find a peak in the goodness of fit at 11 years for the moving average. The fit between the two misery indices holds when using different techniques to measure the literary misery index, and this fit is significantly better than other possible correlations with different emotion indices. To check the robustness of the results, we also analysed books written in German language and obtained very similar correlations with the German economic misery index. The results suggest that millions of books published every year average the authors' shared economic experiences over the past decade.

  11. High Average Power Yb:YAG Laser

    Energy Technology Data Exchange (ETDEWEB)

    Zapata, L E; Beach, R J; Payne, S A

    2001-05-23

    We are working on a composite thin-disk laser design that can be scaled as a source of high brightness laser power for tactical engagement and other high average power applications. The key component is a diffusion-bonded composite comprising a thin gain-medium and thicker cladding that is strikingly robust and resolves prior difficulties with high average power pumping/cooling and the rejection of amplified spontaneous emission (ASE). In contrast to high power rods or slabs, the one-dimensional nature of the cooling geometry and the edge-pump geometry scale gracefully to very high average power. The crucial design ideas have been verified experimentally. Progress this last year included: extraction with high beam quality using a telescopic resonator, a heterogeneous thin film coating prescription that meets the unusual requirements demanded by this laser architecture, thermal management with our first generation cooler. Progress was also made in design of a second-generation laser.

  12. Books average previous decade of economic misery.

    Directory of Open Access Journals (Sweden)

    R Alexander Bentley

    Full Text Available For the 20(th century since the Depression, we find a strong correlation between a 'literary misery index' derived from English language books and a moving average of the previous decade of the annual U.S. economic misery index, which is the sum of inflation and unemployment rates. We find a peak in the goodness of fit at 11 years for the moving average. The fit between the two misery indices holds when using different techniques to measure the literary misery index, and this fit is significantly better than other possible correlations with different emotion indices. To check the robustness of the results, we also analysed books written in German language and obtained very similar correlations with the German economic misery index. The results suggest that millions of books published every year average the authors' shared economic experiences over the past decade.

  13. Benchmarking statistical averaging of spectra with HULLAC

    Science.gov (United States)

    Klapisch, Marcel; Busquet, Michel

    2008-11-01

    Knowledge of radiative properties of hot plasmas is important for ICF, astrophysics, etc When mid-Z or high-Z elements are present, the spectra are so complex that one commonly uses statistically averaged description of atomic systems [1]. In a recent experiment on Fe[2], performed under controlled conditions, high resolution transmission spectra were obtained. The new version of HULLAC [3] allows the use of the same model with different levels of details/averaging. We will take advantage of this feature to check the effect of averaging with comparison with experiment. [1] A Bar-Shalom, J Oreg, and M Klapisch, J. Quant. Spectros. Rad. Transf. 65, 43 (2000). [2] J. E. Bailey, G. A. Rochau, C. A. Iglesias et al., Phys. Rev. Lett. 99, 265002-4 (2007). [3]. M. Klapisch, M. Busquet, and A. Bar-Shalom, AIP Conference Proceedings 926, 206-15 (2007).

  14. Cosmic structure, averaging and dark energy

    CERN Document Server

    Wiltshire, David L

    2013-01-01

    These lecture notes review the theoretical problems associated with coarse-graining the observed inhomogeneous structure of the universe at late epochs, of describing average cosmic evolution in the presence of growing inhomogeneity, and of relating average quantities to physical observables. In particular, a detailed discussion of the timescape scenario is presented. In this scenario, dark energy is realized as a misidentification of gravitational energy gradients which result from gradients in the kinetic energy of expansion of space, in the presence of density and spatial curvature gradients that grow large with the growth of structure. The phenomenology and observational tests of the timescape model are discussed in detail, with updated constraints from Planck satellite data. In addition, recent results on the variation of the Hubble expansion on < 100/h Mpc scales are discussed. The spherically averaged Hubble law is significantly more uniform in the rest frame of the Local Group of galaxies than in t...

  15. Purity and Enrichment of Laser-Microdissected Midbrain Dopamine Neurons

    Directory of Open Access Journals (Sweden)

    Amanda L. Brown

    2013-01-01

    Full Text Available The ability to microdissect individual cells from the nervous system has enormous potential, as it can allow for the study of gene expression in phenotypically identified cells. However, if the resultant gene expression profiles are to be accurately ascribed, it is necessary to determine the extent of contamination by nontarget cells in the microdissected sample. Here, we show that midbrain dopamine neurons can be laser-microdissected to a high degree of enrichment and purity. The average enrichment for tyrosine hydroxylase (TH gene expression in the microdissected sample relative to midbrain sections was approximately 200-fold. For the dopamine transporter (DAT and the vesicular monoamine transporter type 2 (Vmat2, average enrichments were approximately 100- and 60-fold, respectively. Glutamic acid decarboxylase (Gad65 expression, a marker for GABAergic neurons, was several hundredfold lower than dopamine neuron-specific genes. Glial cell and glutamatergic neuron gene expression were not detected in microdissected samples. Additionally, SN and VTA dopamine neurons had significantly different expression levels of dopamine neuron-specific genes, which likely reflects functional differences between the two cell groups. This study demonstrates that it is possible to laser-microdissect dopamine neurons to a high degree of cell purity. Therefore gene expression profiles can be precisely attributed to the targeted microdissected cells.

  16. Fuel control system for dual fuel engines

    Energy Technology Data Exchange (ETDEWEB)

    Helmich, M.J.; Ryan, W.P.; Marvin, D.H.

    1987-11-24

    A fuel governing system for an engine adapted for operation on a first fuel and a second fuel is described comprising: a first fuel governing system including a spontaneous motion metering means; and a second fuel governing system, the second fuel governing system further comprising: means for providing a first signal indicative of position of the first fuel metering means, which signal approximates total load on the engine, means for providing a second signal of the selected percentage of first fuel relative to total load, means for controlling flow of the second fuel to the engine, which flow causes reflective displacement of the first fuel metering means, means for determining the difference between the first signal and the second signal, which difference is indicative of distance the first fuel metering means must be moved to attain the selected percentage of first fuel relative to total load, and means for causing operation of the means for controlling flow of the second fuel to the engine to cause displacement of the first fuel metering means equal to the distance the first fuel metering means must be moved to attain the selected percentage of first fuel relative to total load.

  17. Feasibility of Burning Civilian Grade Pu in the Modular HTR with Th Fuel Cycle

    Institute of Scientific and Technical Information of China (English)

    许云林; 经荥清

    2001-01-01

    The Modular High Temperature Gas-Cooled Reactor (HTR) can be usedto burn plutonium fuel to reduce Pu stockpiles because of its inherent safety characteristics and ability to burn a variety of fuel mixtures. The equilibrium core is calculated and analyzed for Pu enriched fuel. Fuel spheres with 7 g heavy metal including the civilian grade Pu and thorium are loaded into the reactor. An enrichment of 11% is chosen to provide the desired equilibrium core reactivity. The fuel and moderator temperature coefficients are both negative. The maximum fuel element temperature during normal operation and during a loss of coolant accident is less than 1500 ℃. 92% of 239Pu will be burnt during nomal operation. Therefore, a thorium fuel cycle in the modular HTR is an effective method for burning civilian grade plutonium.

  18. Model averaging and muddled multimodel inferences.

    Science.gov (United States)

    Cade, Brian S

    2015-09-01

    Three flawed practices associated with model averaging coefficients for predictor variables in regression models commonly occur when making multimodel inferences in analyses of ecological data. Model-averaged regression coefficients based on Akaike information criterion (AIC) weights have been recommended for addressing model uncertainty but they are not valid, interpretable estimates of partial effects for individual predictors when there is multicollinearity among the predictor variables. Multicollinearity implies that the scaling of units in the denominators of the regression coefficients may change across models such that neither the parameters nor their estimates have common scales, therefore averaging them makes no sense. The associated sums of AIC model weights recommended to assess relative importance of individual predictors are really a measure of relative importance of models, with little information about contributions by individual predictors compared to other measures of relative importance based on effects size or variance reduction. Sometimes the model-averaged regression coefficients for predictor variables are incorrectly used to make model-averaged predictions of the response variable when the models are not linear in the parameters. I demonstrate the issues with the first two practices using the college grade point average example extensively analyzed by Burnham and Anderson. I show how partial standard deviations of the predictor variables can be used to detect changing scales of their estimates with multicollinearity. Standardizing estimates based on partial standard deviations for their variables can be used to make the scaling of the estimates commensurate across models, a necessary but not sufficient condition for model averaging of the estimates to be sensible. A unimodal distribution of estimates and valid interpretation of individual parameters are additional requisite conditions. The standardized estimates or equivalently the t

  19. An approximate analytical approach to resampling averages

    DEFF Research Database (Denmark)

    Malzahn, Dorthe; Opper, M.

    2004-01-01

    Using a novel reformulation, we develop a framework to compute approximate resampling data averages analytically. The method avoids multiple retraining of statistical models on the samples. Our approach uses a combination of the replica "trick" of statistical physics and the TAP approach for appr......Using a novel reformulation, we develop a framework to compute approximate resampling data averages analytically. The method avoids multiple retraining of statistical models on the samples. Our approach uses a combination of the replica "trick" of statistical physics and the TAP approach...

  20. A singularity theorem based on spatial averages

    Indian Academy of Sciences (India)

    J M M Senovilla

    2007-07-01

    Inspired by Raychaudhuri's work, and using the equation named after him as a basic ingredient, a new singularity theorem is proved. Open non-rotating Universes, expanding everywhere with a non-vanishing spatial average of the matter variables, show severe geodesic incompletness in the past. Another way of stating the result is that, under the same conditions, any singularity-free model must have a vanishing spatial average of the energy density (and other physical variables). This is very satisfactory and provides a clear decisive difference between singular and non-singular cosmologies.

  1. Model averaging and muddled multimodel inferences

    Science.gov (United States)

    Cade, Brian S.

    2015-01-01

    Three flawed practices associated with model averaging coefficients for predictor variables in regression models commonly occur when making multimodel inferences in analyses of ecological data. Model-averaged regression coefficients based on Akaike information criterion (AIC) weights have been recommended for addressing model uncertainty but they are not valid, interpretable estimates of partial effects for individual predictors when there is multicollinearity among the predictor variables. Multicollinearity implies that the scaling of units in the denominators of the regression coefficients may change across models such that neither the parameters nor their estimates have common scales, therefore averaging them makes no sense. The associated sums of AIC model weights recommended to assess relative importance of individual predictors are really a measure of relative importance of models, with little information about contributions by individual predictors compared to other measures of relative importance based on effects size or variance reduction. Sometimes the model-averaged regression coefficients for predictor variables are incorrectly used to make model-averaged predictions of the response variable when the models are not linear in the parameters. I demonstrate the issues with the first two practices using the college grade point average example extensively analyzed by Burnham and Anderson. I show how partial standard deviations of the predictor variables can be used to detect changing scales of their estimates with multicollinearity. Standardizing estimates based on partial standard deviations for their variables can be used to make the scaling of the estimates commensurate across models, a necessary but not sufficient condition for model averaging of the estimates to be sensible. A unimodal distribution of estimates and valid interpretation of individual parameters are additional requisite conditions. The standardized estimates or equivalently the

  2. Model Year 2016 Fuel Economy Guide: EPA Fuel Economy Estimates

    Energy Technology Data Exchange (ETDEWEB)

    None

    2015-11-01

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

  3. Model Year 2015 Fuel Economy Guide: EPA Fuel Economy Estimates

    Energy Technology Data Exchange (ETDEWEB)

    None

    2014-12-01

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

  4. Model Year 2010 Fuel Economy Guide: EPA Fuel Economy Estimates

    Energy Technology Data Exchange (ETDEWEB)

    None

    2009-10-14

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

  5. Model Year 2014 Fuel Economy Guide: EPA Fuel Economy Estimates

    Energy Technology Data Exchange (ETDEWEB)

    None

    2013-12-01

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

  6. Model Year 2008 Fuel Economy Guide: EPA Fuel Economy Estimates

    Energy Technology Data Exchange (ETDEWEB)

    None

    2007-10-01

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

  7. Model Year 2009 Fuel Economy Guide: EPA Fuel Economy Estimates

    Energy Technology Data Exchange (ETDEWEB)

    None

    2008-10-01

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

  8. Model Year 2007 Fuel Economy Guide: EPA Fuel Economy Estimates

    Energy Technology Data Exchange (ETDEWEB)

    None

    2007-10-01

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

  9. Model Year 2005 Fuel Economy Guide: EPA Fuel Economy Estimates

    Energy Technology Data Exchange (ETDEWEB)

    None

    2004-11-01

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

  10. Model Year 2006 Fuel Economy Guide: EPA Fuel Economy Estimates

    Energy Technology Data Exchange (ETDEWEB)

    None

    2005-11-01

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

  11. Reference Spent Fuel and Its Source Terms for a Design of Deep Geological Disposal System

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Dong Keun

    2005-12-15

    In this study, current status and future trend of domestic spent fuels were analyzed to propose reference spent nuclear fuel. And then, source terms needed for design of a deep geological disposal system were calculated using ORIGEN-ARP. The reference spent fuels selected based on assembly physical dimension, inventory projection, trend of initial enrichment of 235U, discharge burnup are as follows; The 17x17 Korean Optimized Fuel Assembly with initial enrichment of 4.0 wt.% 235U and discharge burnup of 45 GWD/MTU was adopted as a low-burnup representative fuel. For the high-burnup representative fuel, 16x16 Korean Standard Fuel Assembly with initial enrichment of 4.5 wt.% 235U and discharge burnup of 55 GWD/MTU was chosen. CANDU fuel with initial enrichment of 0.711 wt.% 235U and discharge burnup of 7.5 GWD/MTU was also considered. For these reference fuels, decay heat, radiation intensity and spectrum, nuclide concentration, and individual nuclide radioactivity were calculated using ORIGEN-ARP for a disposal system design. It is expected that the source terms estimated in this study will be applied to the disposal system development in the future.

  12. Quantum Averaging of Squeezed States of Light

    DEFF Research Database (Denmark)

    Squeezing has been recognized as the main resource for quantum information processing and an important resource for beating classical detection strategies. It is therefore of high importance to reliably generate stable squeezing over longer periods of time. The averaging procedure for a single qu...

  13. Generalized Jackknife Estimators of Weighted Average Derivatives

    DEFF Research Database (Denmark)

    Cattaneo, Matias D.; Crump, Richard K.; Jansson, Michael

    With the aim of improving the quality of asymptotic distributional approximations for nonlinear functionals of nonparametric estimators, this paper revisits the large-sample properties of an important member of that class, namely a kernel-based weighted average derivative estimator. Asymptotic...

  14. Bayesian Model Averaging for Propensity Score Analysis

    Science.gov (United States)

    Kaplan, David; Chen, Jianshen

    2013-01-01

    The purpose of this study is to explore Bayesian model averaging in the propensity score context. Previous research on Bayesian propensity score analysis does not take into account model uncertainty. In this regard, an internally consistent Bayesian framework for model building and estimation must also account for model uncertainty. The…

  15. High average-power induction linacs

    Energy Technology Data Exchange (ETDEWEB)

    Prono, D.S.; Barrett, D.; Bowles, E.; Caporaso, G.J.; Chen, Yu-Jiuan; Clark, J.C.; Coffield, F.; Newton, M.A.; Nexsen, W.; Ravenscroft, D.

    1989-03-15

    Induction linear accelerators (LIAs) are inherently capable of accelerating several thousand amperes of /approximately/ 50-ns duration pulses to > 100 MeV. In this paper we report progress and status in the areas of duty factor and stray power management. These technologies are vital if LIAs are to attain high average power operation. 13 figs.

  16. Discontinuities and hysteresis in quantized average consensus

    NARCIS (Netherlands)

    Ceragioli, Francesca; Persis, Claudio De; Frasca, Paolo

    2011-01-01

    We consider continuous-time average consensus dynamics in which the agents’ states are communicated through uniform quantizers. Solutions to the resulting system are defined in the Krasowskii sense and are proven to converge to conditions of ‘‘practical consensus’’. To cope with undesired chattering

  17. On averaging methods for partial differential equations

    NARCIS (Netherlands)

    Verhulst, F.

    2001-01-01

    The analysis of weakly nonlinear partial differential equations both qualitatively and quantitatively is emerging as an exciting eld of investigation In this report we consider specic results related to averaging but we do not aim at completeness The sections and contain important material which

  18. A Functional Measurement Study on Averaging Numerosity

    Science.gov (United States)

    Tira, Michael D.; Tagliabue, Mariaelena; Vidotto, Giulio

    2014-01-01

    In two experiments, participants judged the average numerosity between two sequentially presented dot patterns to perform an approximate arithmetic task. In Experiment 1, the response was given on a 0-20 numerical scale (categorical scaling), and in Experiment 2, the response was given by the production of a dot pattern of the desired numerosity…

  19. Bayesian Averaging is Well-Temperated

    DEFF Research Database (Denmark)

    Hansen, Lars Kai

    2000-01-01

    Bayesian predictions are stochastic just like predictions of any other inference scheme that generalize from a finite sample. While a simple variational argument shows that Bayes averaging is generalization optimal given that the prior matches the teacher parameter distribution the situation...

  20. Average utility maximization: A preference foundation

    NARCIS (Netherlands)

    A.V. Kothiyal (Amit); V. Spinu (Vitalie); P.P. Wakker (Peter)

    2014-01-01

    textabstractThis paper provides necessary and sufficient preference conditions for average utility maximization over sequences of variable length. We obtain full generality by using a new algebraic technique that exploits the richness structure naturally provided by the variable length of the sequen

  1. Full averaging of fuzzy impulsive differential inclusions

    Directory of Open Access Journals (Sweden)

    Natalia V. Skripnik

    2010-09-01

    Full Text Available In this paper the substantiation of the method of full averaging for fuzzy impulsive differential inclusions is studied. We extend the similar results for impulsive differential inclusions with Hukuhara derivative (Skripnik, 2007, for fuzzy impulsive differential equations (Plotnikov and Skripnik, 2009, and for fuzzy differential inclusions (Skripnik, 2009.

  2. Materials for high average power lasers

    Energy Technology Data Exchange (ETDEWEB)

    Marion, J.E.; Pertica, A.J.

    1989-01-01

    Unique materials properties requirements for solid state high average power (HAP) lasers dictate a materials development research program. A review of the desirable laser, optical and thermo-mechanical properties for HAP lasers precedes an assessment of the development status for crystalline and glass hosts optimized for HAP lasers. 24 refs., 7 figs., 1 tab.

  3. Independence, Odd Girth, and Average Degree

    DEFF Research Database (Denmark)

    Löwenstein, Christian; Pedersen, Anders Sune; Rautenbach, Dieter;

    2011-01-01

      We prove several tight lower bounds in terms of the order and the average degree for the independence number of graphs that are connected and/or satisfy some odd girth condition. Our main result is the extension of a lower bound for the independence number of triangle-free graphs of maximum...

  4. A dynamic analysis of moving average rules

    NARCIS (Netherlands)

    C. Chiarella; X.Z. He; C.H. Hommes

    2006-01-01

    The use of various moving average (MA) rules remains popular with financial market practitioners. These rules have recently become the focus of a number empirical studies, but there have been very few studies of financial market models where some agents employ technical trading rules of the type use

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

    Science.gov (United States)

    Perton, M.; Lévesque, D.; Monchalin, J.-P.; Lord, M.; Smith, J. A.; Rabin, B. H.

    2013-01-01

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

  6. Enrichment and characterization of sulfate reducing, naphthalene degrading microorganisms

    Science.gov (United States)

    Steffen, Kümmel; Florian-Alexander, Herbst; Márcia, Duarte; Dietmar, Pieper; Jana, Seifert; Bergen Martin, von; Hans-Hermann, Richnow; Carsten, Vogt

    2014-05-01

    Polycyclic aromatic hydrocarbons (PAH) are pollutants of great concern due to their potential toxicity, mutagenicity and carcinogenicity. PAH are widely distributed in the environment by accidental discharges during the transport, use and disposal of petroleum products, and during forest and grass fires. Caused by their hydrophobic nature, PAH basically accumulate in sediments from where they are slowly released into the groundwater. Although generally limited by the low water solubility of PAH, microbial degradation is one of the major mechanisms leading to the complete clean-up of PAH-contaminated sites. Whereas organisms and biochemical pathways responsible for the aerobic breakdown of PAH are well known, anaerobic PAH biodegradation is less understood; only a few anaerobic PAH degrading cultures have been described. We studied the anaerobic PAH degradation in a microcosm approach to enrich anaerobic PAH degraders. Anoxic groundwater and sediment samples were used as inoculum. Groundwater samples were purchased from the erstwhile gas works facility and a former wood impregnation site. In contrast, sources of sediment samples were a former coal refining area and an old fuel depot. Samples were incubated in anoxic mineral salt medium with naphthalene as sole carbon source and sulfate as terminal electron acceptor. Grown cultures were characterized by feeding with 13C-labeled naphthalene, 16S rRNA gene sequencing using an Illumina® approach, and functional proteome analyses. Finally, six enrichment cultures able to degrade naphthalene under anoxic conditions were established. First results point to a dominance of identified sequences affiliated to the freshwater sulfate-reducing strain N47, which is a known anaerobic naphthalene degrader, in four out of the six enrichments. In those enrichments, peptides related to the pathway of anoxic naphthalene degradation in N47 were abundant. Overall the data underlines the importance of Desulfobacteria for natural

  7. Macstor dry spent fuel storage system

    Energy Technology Data Exchange (ETDEWEB)

    Pare, F. E. [Atomic Energy of Canada Limited, Montreal (Canada)

    1996-04-15

    AECL, a Canadian Grown Corporation established since 1952, is unique among the world's nuclear organizations. It is both supplier of research reactors and heavy water moderated CANDU power reactors as well as operator of extensive nuclear research facilities. As part of its mandate, AECL has developed products and conceptual designs for the short, intermediate and long term storage and disposal of spent nuclear fuel. AECL has also assumed leadership in the area of dry storage of spent fuel. This Canadian Crown Corporation first started to look into dry storage for the management of its spent nuclear fuel in the early 1970's. After developing silo-like structures called concrete canisters for the storage of its research reactor enriched uranium fuel, AECL went on to perfect that technology for spent CANDU natural uranium fuel. In 1989 AECL teamed up with Trans nuclear, Inc.,(TN), a US based member of the international Trans nuclear Group, to extend its dry storage technology to LWR spent fuel. This association combines AECL's expertise and many years experience in the design of spent fuel storage facilities with TN's proven capabilities of processing, transportation, storage and handling of LWR spent fuel. From the early AECL-designed unventilated concrete canisters to the advanced MACSTOR concept - Modular Air-Cooled Canister Storage - now available also for LWR fuel - dry storage is proving to be safe, economical, practical and, most of all, well accepted by the general public. AECL's experience with different fuels and circumstances has been conclusive.

  8. Solid recovered fuels in the cement industry--semi-automated sample preparation unit as a means for facilitated practical application.

    Science.gov (United States)

    Aldrian, Alexia; Sarc, Renato; Pomberger, Roland; Lorber, Karl E; Sipple, Ernst-Michael

    2016-03-01

    One of the challenges for the cement industry is the quality assurance of alternative fuel (e.g., solid recovered fuel, SRF) in co-incineration plants--especially for inhomogeneous alternative fuels with large particle sizes (d95⩾100 mm), which will gain even more importance in the substitution of conventional fuels due to low production costs. Existing standards for sampling and sample preparation do not cover the challenges resulting from these kinds of materials. A possible approach to ensure quality monitoring is shown in the present contribution. For this, a specially manufactured, automated comminution and sample divider device was installed at a cement plant in Rohožnik. In order to prove its practical suitability with methods according to current standards, the sampling and sample preparation process were validated for alternative fuel with a grain size >30 mm (i.e., d95=approximately 100 mm), so-called 'Hotdisc SRF'. Therefore, series of samples were taken and analysed. A comparison of the analysis results with the yearly average values obtained through a reference investigation route showed good accordance. Further investigations during the validation process also showed that segregation or enrichment of material throughout the comminution plant does not occur. The results also demonstrate that compliance with legal standards regarding the minimum sample amount is not sufficient for inhomogeneous and coarse particle size alternative fuels. Instead, higher sample amounts after the first particle size reduction step are strongly recommended in order to gain a representative laboratory sample.

  9. Design and Testing of Prototypic Elements Containing Monolithic Fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-10-01

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

  10. Study on the fuel cycle cost of gas turbine high temperature reactor (GTHTR300). Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Takei, Masanobu; Katanishi, Shoji; Nakata, Tetsuo; Kunitomi, Kazuhiko [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Oda, Takefumi; Izumiya, Toru [Nuclear Fuel Industries, Ltd., Tokyo (Japan)

    2002-11-01

    In the basic design of gas turbine high temperature reactor (GTHTR300), reduction of the fuel cycle cost has a large benefit of improving overall plant economy. Then, fuel cycle cost was evaluated for GTHTR300. First, of fuel fabrication for high-temperature gas cooled reactor, since there was no actual experience with a commercial scale, a preliminary design for a fuel fabrication plant with annual processing of 7.7 ton-U sufficient four GTHTR300 was performed, and fuel fabrication cost was evaluated. Second, fuel cycle cost was evaluated based on the equilibrium cycle of GTHTR300. The factors which were considered in this cost evaluation include uranium price, conversion, enrichment, fabrication, storage of spent fuel, reprocessing, and waste disposal. The fuel cycle cost of GTHTR300 was estimated at about 1.07 yen/kWh. If the back-end cost of reprocessing and waste disposal is included and assumed to be nearly equivalent to LWR, the fuel cycle cost of GTHTR300 was estimated to be about 1.31 yen/kWh. Furthermore, the effects on fuel fabrication cost by such of fuel specification parameters as enrichment, the number of fuel types, and the layer thickness were considered. Even if the enrichment varies from 10 to 20%, the number of fuel types change from 1 to 4, the 1st layer thickness of fuel changes by 30 {mu}m, or the 2nd layer to the 4th layer thickness of fuel changes by 10 {mu}m, the impact on fuel fabrication cost was evaluated to be negligible. (author)

  11. Application of oxygen-enriched combustion for locomotive diesel engines. Phase 1

    Energy Technology Data Exchange (ETDEWEB)

    Poola, R.B.; Sekar, R.R.; Assanis, D.N.

    1996-09-01

    A thermodynamic simulation is used to study the effects of oxygen-enriched intake air on the performance and nitrogen oxide (NO) emissions of a locomotive diesel engine. The parasitic power of the air separation membrane required to supply the oxygen-enriched air is also estimated. For a given constraint on peak cylinder pressure, the gross and net power outputs of an engine operating under different levels of oxygen enrichment are compared with those obtained when a high-boost turbocharged engine is used. A 4% increase in peak cylinder pressure can result in an increase in net engine power of approximately 13% when intake air with an oxygen content of 28% by volume is used and fuel injection timing is retarded by 4 degrees. When the engine is turbocharged to a higher inlet boost, the same increase in peak cylinder pressure improves power by only 4%. If part of the significantly higher exhaust enthalpies available as a result of oxygen enrichment are recovered, the power requirements of the air separator membrane can be met, resulting in substantial net power improvements. Oxygen enrichment reduces particulate and visible smoke emissions but increases NO emissions. However, a combination of retarded fuel injection timing and post-treatment of exhaust gases may be adequate to meet the locomotive diesel engine NO{sub x} standards. Exhaust gas after-treatment and heat recovery would be required to realize the full potential of oxygen enrichment. Economic analysis shows that oxygen-enrichment technology is economically feasible and provides high returns on investment. The study also indicates the strong influence of membrane parasitic requirements and exhaust energy recovery on economic benefits. To obtain an economic advantage while using a membrane with higher parasitic power requirements, it is necessary to recover a part of the exhaust energy.

  12. The cycle of the nuclear fuel used in EDF power plants; Le cycle du combustible nucleaire utilise dans les centrales EDF

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-11-15

    This document briefly indicates the different stages of the nuclear fuel cycle, from the purchase of natural uranium to waste storage. It also indicates the main responsibilities of EDF regarding this fuel cycle (to secure supplies, to organise material transportation, to process and store used fuels and associated wastes). It presents the different associated processes: uranium extraction, purification and concentration, conversion or fluoridation, enrichment. It briefly describes the fuel assembly fabrication, and indicates the main uranium producers in the world. Other addressed steps are: the transportation of fuel assembly, fuel loading, and spent fuel management, the processing of spent fuel and radioactive wastes

  13. Aviation fuels outlook

    Science.gov (United States)

    Momenthy, A. M.

    1980-01-01

    Options for satisfying the future demand for commercial jet fuels are analyzed. It is concluded that the most effective means to this end are to attract more refiners to the jet fuel market and encourage development of processes to convert oil shale and coal to transportation fuels. Furthermore, changing the U.S. refineries fuel specification would not significantly alter jet fuel availability.

  14. ANALYSIS OF THE FACTORS AFFECTING THE AVERAGE

    Directory of Open Access Journals (Sweden)

    Carmen BOGHEAN

    2013-12-01

    Full Text Available Productivity in agriculture most relevantly and concisely expresses the economic efficiency of using the factors of production. Labour productivity is affected by a considerable number of variables (including the relationship system and interdependence between factors, which differ in each economic sector and influence it, giving rise to a series of technical, economic and organizational idiosyncrasies. The purpose of this paper is to analyse the underlying factors of the average work productivity in agriculture, forestry and fishing. The analysis will take into account the data concerning the economically active population and the gross added value in agriculture, forestry and fishing in Romania during 2008-2011. The distribution of the average work productivity per factors affecting it is conducted by means of the u-substitution method.

  15. Time-average dynamic speckle interferometry

    Science.gov (United States)

    Vladimirov, A. P.

    2014-05-01

    For the study of microscopic processes occurring at structural level in solids and thin biological objects, a method of dynamic speckle interferometry successfully applied. However, the method has disadvantages. The purpose of the report is to acquaint colleagues with the method of averaging in time in dynamic speckle - interferometry of microscopic processes, allowing eliminating shortcomings. The main idea of the method is the choice the averaging time, which exceeds the characteristic time correlation (relaxation) the most rapid process. The method theory for a thin phase and the reflecting object is given. The results of the experiment on the high-cycle fatigue of steel and experiment to estimate the biological activity of a monolayer of cells, cultivated on a transparent substrate is given. It is shown that the method allows real-time visualize the accumulation of fatigue damages and reliably estimate the activity of cells with viruses and without viruses.

  16. Averaged Extended Tree Augmented Naive Classifier

    Directory of Open Access Journals (Sweden)

    Aaron Meehan

    2015-07-01

    Full Text Available This work presents a new general purpose classifier named Averaged Extended Tree Augmented Naive Bayes (AETAN, which is based on combining the advantageous characteristics of Extended Tree Augmented Naive Bayes (ETAN and Averaged One-Dependence Estimator (AODE classifiers. We describe the main properties of the approach and algorithms for learning it, along with an analysis of its computational time complexity. Empirical results with numerous data sets indicate that the new approach is superior to ETAN and AODE in terms of both zero-one classification accuracy and log loss. It also compares favourably against weighted AODE and hidden Naive Bayes. The learning phase of the new approach is slower than that of its competitors, while the time complexity for the testing phase is similar. Such characteristics suggest that the new classifier is ideal in scenarios where online learning is not required.

  17. Trajectory averaging for stochastic approximation MCMC algorithms

    CERN Document Server

    Liang, Faming

    2010-01-01

    The subject of stochastic approximation was founded by Robbins and Monro [Ann. Math. Statist. 22 (1951) 400--407]. After five decades of continual development, it has developed into an important area in systems control and optimization, and it has also served as a prototype for the development of adaptive algorithms for on-line estimation and control of stochastic systems. Recently, it has been used in statistics with Markov chain Monte Carlo for solving maximum likelihood estimation problems and for general simulation and optimizations. In this paper, we first show that the trajectory averaging estimator is asymptotically efficient for the stochastic approximation MCMC (SAMCMC) algorithm under mild conditions, and then apply this result to the stochastic approximation Monte Carlo algorithm [Liang, Liu and Carroll J. Amer. Statist. Assoc. 102 (2007) 305--320]. The application of the trajectory averaging estimator to other stochastic approximation MCMC algorithms, for example, a stochastic approximation MLE al...

  18. Fabrication of Monolithic RERTR Fuels by Hot Isostatic Pressing

    Energy Technology Data Exchange (ETDEWEB)

    Jan-Fong Jue; Blair H. Park; Curtis R. Clark; Glenn A. Moore; Dennis D. Keiser, Jr.

    2010-11-01

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

  19. Average Annual Rainfall over the Globe

    Science.gov (United States)

    Agrawal, D. C.

    2013-01-01

    The atmospheric recycling of water is a very important phenomenon on the globe because it not only refreshes the water but it also redistributes it over land and oceans/rivers/lakes throughout the globe. This is made possible by the solar energy intercepted by the Earth. The half of the globe facing the Sun, on the average, intercepts 1.74 ×…

  20. The Ghirlanda-Guerra identities without averaging

    CERN Document Server

    Chatterjee, Sourav

    2009-01-01

    The Ghirlanda-Guerra identities are one of the most mysterious features of spin glasses. We prove the GG identities in a large class of models that includes the Edwards-Anderson model, the random field Ising model, and the Sherrington-Kirkpatrick model in the presence of a random external field. Previously, the GG identities were rigorously proved only `on average' over a range of temperatures or under small perturbations.

  1. Hydrogen, nitrogen and syngas enriched diesel combustion

    OpenAIRE

    Christodoulou, Fanos

    2014-01-01

    This thesis was submitted for the degree of Doctor of Philosophy and awarded by Brunel University On-board hydrogen and syngas production is considered as a transition solution from fossil fuel to hydrogen powered vehicles until problems associated with hydrogen infrastructure, distribution and storage are resolved. A hydrogen- or syngas-rich stream, which substitutes part of the main hydrocarbon fuel, can be produced by supplying diesel fuel in a fuel-reforming reactor, integrated within ...

  2. Spatial averaging infiltration model for layered soil

    Institute of Scientific and Technical Information of China (English)

    HU HePing; YANG ZhiYong; TIAN FuQiang

    2009-01-01

    To quantify the influences of soil heterogeneity on infiltration, a spatial averaging infiltration model for layered soil (SAI model) is developed by coupling the spatial averaging approach proposed by Chen et al. and the Generalized Green-Ampt model proposed by Jia et al. In the SAI model, the spatial heterogeneity along the horizontal direction is described by a probability distribution function, while that along the vertical direction is represented by the layered soils. The SAI model is tested on a typical soil using Monte Carlo simulations as the base model. The results show that the SAI model can directly incorporate the influence of spatial heterogeneity on infiltration on the macro scale. It is also found that the homogeneous assumption of soil hydraulic conductivity along the horizontal direction will overestimate the infiltration rate, while that along the vertical direction will underestimate the infiltration rate significantly during rainstorm periods. The SAI model is adopted in the spatial averaging hydrological model developed by the authors, and the results prove that it can be applied in the macro-scale hydrological and land surface process modeling in a promising way.

  3. Unscrambling The "Average User" Of Habbo Hotel

    Directory of Open Access Journals (Sweden)

    Mikael Johnson

    2007-01-01

    Full Text Available The “user” is an ambiguous concept in human-computer interaction and information systems. Analyses of users as social actors, participants, or configured users delineate approaches to studying design-use relationships. Here, a developer’s reference to a figure of speech, termed the “average user,” is contrasted with design guidelines. The aim is to create an understanding about categorization practices in design through a case study about the virtual community, Habbo Hotel. A qualitative analysis highlighted not only the meaning of the “average user,” but also the work that both the developer and the category contribute to this meaning. The average user a represents the unknown, b influences the boundaries of the target user groups, c legitimizes the designer to disregard marginal user feedback, and d keeps the design space open, thus allowing for creativity. The analysis shows how design and use are intertwined and highlights the developers’ role in governing different users’ interests.

  4. A simple algorithm for averaging spike trains.

    Science.gov (United States)

    Julienne, Hannah; Houghton, Conor

    2013-02-25

    Although spike trains are the principal channel of communication between neurons, a single stimulus will elicit different spike trains from trial to trial. This variability, in both spike timings and spike number can obscure the temporal structure of spike trains and often means that computations need to be run on numerous spike trains in order to extract features common across all the responses to a particular stimulus. This can increase the computational burden and obscure analytical results. As a consequence, it is useful to consider how to calculate a central spike train that summarizes a set of trials. Indeed, averaging responses over trials is routine for other signal types. Here, a simple method for finding a central spike train is described. The spike trains are first mapped to functions, these functions are averaged, and a greedy algorithm is then used to map the average function back to a spike train. The central spike trains are tested for a large data set. Their performance on a classification-based test is considerably better than the performance of the medoid spike trains.

  5. Spatial averaging infiltration model for layered soil

    Institute of Scientific and Technical Information of China (English)

    2009-01-01

    To quantify the influences of soil heterogeneity on infiltration, a spatial averaging infiltration model for layered soil (SAI model) is developed by coupling the spatial averaging approach proposed by Chen et al. and the Generalized Green-Ampt model proposed by Jia et al. In the SAI model, the spatial hetero- geneity along the horizontal direction is described by a probability distribution function, while that along the vertical direction is represented by the layered soils. The SAI model is tested on a typical soil using Monte Carlo simulations as the base model. The results show that the SAI model can directly incorporate the influence of spatial heterogeneity on infiltration on the macro scale. It is also found that the homogeneous assumption of soil hydraulic conductivity along the horizontal direction will overes- timate the infiltration rate, while that along the vertical direction will underestimate the infiltration rate significantly during rainstorm periods. The SAI model is adopted in the spatial averaging hydrological model developed by the authors, and the results prove that it can be applied in the macro-scale hy- drological and land surface process modeling in a promising way.

  6. Geomagnetic effects on the average surface temperature

    Science.gov (United States)

    Ballatore, P.

    Several results have previously shown as the solar activity can be related to the cloudiness and the surface solar radiation intensity (Svensmark and Friis-Christensen, J. Atmos. Sol. Terr. Phys., 59, 1225, 1997; Veretenenkoand Pudovkin, J. Atmos. Sol. Terr. Phys., 61, 521, 1999). Here, the possible relationships between the averaged surface temperature and the solar wind parameters or geomagnetic activity indices are investigated. The temperature data used are the monthly SST maps (generated at RAL and available from the related ESRIN/ESA database) that represent the averaged surface temperature with a spatial resolution of 0.5°x0.5° and cover the entire globe. The interplanetary data and the geomagnetic data are from the USA National Space Science Data Center. The time interval considered is 1995-2000. Specifically, possible associations and/or correlations of the average temperature with the interplanetary magnetic field Bz component and with the Kp index are considered and differentiated taking into account separate geographic and geomagnetic planetary regions.

  7. Disk-averaged synthetic spectra of Mars

    CERN Document Server

    Tinetti, G; Fong, W; Meadows, V S; Snively, H; Velusamy, T; Crisp, David; Fong, William; Meadows, Victoria S.; Snively, Heather; Tinetti, Giovanna; Velusamy, Thangasamy

    2004-01-01

    The principal goal of the NASA Terrestrial Planet Finder (TPF) and ESA Darwin mission concepts is to directly detect and characterize extrasolar terrestrial (Earth-sized) planets. This first generation of instruments is expected to provide disk-averaged spectra with modest spectral resolution and signal-to-noise. Here we use a spatially and spectrally resolved model of the planet Mars to study the detectability of a planet's surface and atmospheric properties from disk-averaged spectra as a function of spectral resolution and wavelength range, for both the proposed visible coronograph (TPF-C) and mid-infrared interferometer (TPF-I/Darwin) architectures. At the core of our model is a spectrum-resolving (line-by-line) atmospheric/surface radiative transfer model which uses observational data as input to generate a database of spatially-resolved synthetic spectra for a range of illumination conditions (phase angles) and viewing geometries. Results presented here include disk averaged synthetic spectra, light-cur...

  8. Disk-averaged synthetic spectra of Mars

    Science.gov (United States)

    Tinetti, Giovanna; Meadows, Victoria S.; Crisp, David; Fong, William; Velusamy, Thangasamy; Snively, Heather

    2005-01-01

    The principal goal of the NASA Terrestrial Planet Finder (TPF) and European Space Agency's Darwin mission concepts is to directly detect and characterize extrasolar terrestrial (Earthsized) planets. This first generation of instruments is expected to provide disk-averaged spectra with modest spectral resolution and signal-to-noise. Here we use a spatially and spectrally resolved model of a Mars-like planet to study the detectability of a planet's surface and atmospheric properties from disk-averaged spectra. We explore the detectability as a function of spectral resolution and wavelength range, for both the proposed visible coronograph (TPFC) and mid-infrared interferometer (TPF-I/Darwin) architectures. At the core of our model is a spectrum-resolving (line-by-line) atmospheric/surface radiative transfer model. This model uses observational data as input to generate a database of spatially resolved synthetic spectra for a range of illumination conditions and viewing geometries. The model was validated against spectra recorded by the Mars Global Surveyor-Thermal Emission Spectrometer and the Mariner 9-Infrared Interferometer Spectrometer. Results presented here include disk-averaged synthetic spectra, light curves, and the spectral variability at visible and mid-infrared wavelengths for Mars as a function of viewing angle, illumination, and season. We also considered the differences in the spectral appearance of an increasingly ice-covered Mars, as a function of spectral resolution, signal-to-noise and integration time for both TPF-C and TPFI/ Darwin.

  9. In Comparative Analysis for Fuel Burnup of Fuel Assembly Designs for the 300 kW Small Medical Reactor

    Science.gov (United States)

    Sambuu, Odmaa; Nanzad, Norov

    2009-03-01

    A 300 kW small medical reactor was designed to be used for boron neutron capture therapy (BNCT) at KAIST in 1996 [1]. In this paper, analysis for the core life cycle of the original design of the BNCT facility and modifications of the fuel assembly configuration and enrichment to get a proper life cycle were performed and a criticality, neutron flux distribution and fuel burnup calculations were carried out.

  10. Environmental enrichment in farm, zoo, companion and experimental animals

    Directory of Open Access Journals (Sweden)

    Vučinić Marijana

    2009-01-01

    Full Text Available The paper deals with environmental enrichment for domestic animals at farms, animals in zoos, experimental animals and pet animals. Also, the paper defines and describes different strategies of environmental enrichment. Environmental enrichment is a simple and effective mean of prevention of boredom, behavioral disorders as well as an effective mean of improving animal welfare in farm, zoo, companion and experimental animals. Different items and materials may be used for environmental enrichment. They need to be evaluated for use by taking into account the following: the species of an animal, its needs, habits and capabilities, the type of an enrichment device, the device's ability to stimulate the animal's interest and the safety of the device. Enrichment programmes should always include two forms of enrichment: behavioral enrichment and environmental enrichment. Enrichment comes in many forms such as structural or physical enrichment, sensory enrichment (auditory and olfactory stimulation, dietary enrichment, manipulatable enrichment and social enrichment.

  11. Burnup determination of a fuel element concerning different cooling times; Seguimiento del quemado de un elemento combustible, para diferentes tiempos de enfriamento

    Energy Technology Data Exchange (ETDEWEB)

    Henriquez, C.; Navarro, G.; Pereda, C.; Mutis, O. [Comision Chilena de Energia Nuclear, Santiago (Chile). Dept. de Aplicaciones Nucleares. Unidad de Reactores; Terremoto, Luis A.A.; Zeituni, Carlos A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear

    2002-07-01

    In this work we report a complete set of measurements and some relevant results regarding the burnup process of a fuel element containing low enriched nuclear fuel. This fuel element was fabricated at the Plant of Fuel Elements of the Chilean Nuclear Energy Commission (CCHEN). Measurements were carried out using gamma-ray spectroscopy and the absolute burnup of the fuel element was determined. (author)

  12. Alternate fuels; Combustibles alternos

    Energy Technology Data Exchange (ETDEWEB)

    Romero Paredes R, Hernando; Ambriz G, Juan Jose [Universidad Autonoma Metropolitana. Iztapalapa (Mexico)

    2003-07-01

    In the definition and description of alternate fuels we must center ourselves in those technological alternatives that allow to obtain compounds that differ from the traditional ones, in their forms to be obtained. In this article it is tried to give an overview of alternate fuels to the conventional derivatives of petroleum and that allow to have a clear idea on the tendencies of modern investigation and the technological developments that can be implemented in the short term. It is not pretended to include all the tendencies and developments of the present world, but those that can hit in a relatively short term, in accordance with agreed with the average life of conventional fuels. Nevertheless, most of the conversion principles are applicable to the spectrum of carbonaceous or cellulosic materials which are in nature, are cultivated or wastes of organic origin. Thus one will approach them in a successive way, the physical, chemical and biological conversions that can take place in a production process of an alternate fuel or the same direct use of the fuel such as burning the sweepings derived from the forests. [Spanish] En la definicion y descripcion de combustibles alternos nos debemos centrar en aquellas alternativas tecnologicas que permitan obtener compuestos que difieren de los tradicionales, al menos en sus formas de ser obtenidos. En este articulo se pretende dar un panorama de los combustibles alternos a los convencionales derivados del petroleo y que permita tener una idea clara sobre las tendencias de la investigacion moderna y los desarrollos tecnologicos que puedan ser implementados en el corto plazo. No se pretende abarcar todas las tendencias y desarrollos del mundo actual, sino aquellas que pueden impactar en un plazo relativamente corto, acordes con la vida media de los combustibles convencionales. Sin embargo, la mayor parte de los principios de conversion son aplicables al espectro de materiales carbonaceos o celulosicos los cuales se

  13. Compact Fuel Element Environment Test

    Science.gov (United States)

    Bradley, D. E.; Mireles, O. R.; Hickman, R. R.; Broadway, J. W.

    2012-01-01

    Deep space missions with large payloads require high specific impulse (I(sub sp)) and relatively high thrust to achieve mission goals in reasonable time frames. Conventional, storable propellants produce average I(sub sp). Nuclear thermal rockets (NTRs) capable of high I(sub sp) thrust have been proposed. NTR employs heat produced by fission reaction to heat and therefore accelerate hydrogen, which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3,000 K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited. The primary concern is the mechanical failure of fuel elements that employ high melting point metals, ceramics, or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. It is not necessary to include fissile material in test samples intended to explore high-temperature hydrogen exposure of the structural support matrices. A small-scale test bed designed to heat fuel element samples via noncontact radio frequency heating and expose samples to hydrogen for typical mission durations has been developed to assist in optimal material and manufacturing process selection without employing fissile material. This Technical Memorandum details the test bed design and results of testing conducted to date.

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

    Energy Technology Data Exchange (ETDEWEB)

    Nick Soelberg; Steve Piet

    2010-11-01

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

  15. Fuel processors for fuel cell APU applications

    Science.gov (United States)

    Aicher, T.; Lenz, B.; Gschnell, F.; Groos, U.; Federici, F.; Caprile, L.; Parodi, L.

    The conversion of liquid hydrocarbons to a hydrogen rich product gas is a central process step in fuel processors for auxiliary power units (APUs) for vehicles of all kinds. The selection of the reforming process depends on the fuel and the type of the fuel cell. For vehicle power trains, liquid hydrocarbons like gasoline, kerosene, and diesel are utilized and, therefore, they will also be the fuel for the respective APU systems. The fuel cells commonly envisioned for mobile APU applications are molten carbonate fuel cells (MCFC), solid oxide fuel cells (SOFC), and proton exchange membrane fuel cells (PEMFC). Since high-temperature fuel cells, e.g. MCFCs or SOFCs, can be supplied with a feed gas that contains carbon monoxide (CO) their fuel processor does not require reactors for CO reduction and removal. For PEMFCs on the other hand, CO concentrations in the feed gas must not exceed 50 ppm, better 20 ppm, which requires additional reactors downstream of the reforming reactor. This paper gives an overview of the current state of the fuel processor development for APU applications and APU system developments. Furthermore, it will present the latest developments at Fraunhofer ISE regarding fuel processors for high-temperature fuel cell APU systems on board of ships and aircrafts.

  16. GSPEL - Fuel Cell Laboratory

    Data.gov (United States)

    Federal Laboratory Consortium — The Fuel Cell Lab (FCL) Provides testing for technology readiness of fuel cell systems The FCL investigates, tests and verifies the performance of fuel-cell systems...

  17. GSPEL - Fuel Cell Laboratory

    Data.gov (United States)

    Federal Laboratory Consortium — The Fuel Cell Lab (FCL)Provides testing for technology readiness of fuel cell systems The FCL investigates, tests and verifies the performance of fuel-cell systems...

  18. Fuel cells: A survey

    Science.gov (United States)

    Crowe, B. J.

    1973-01-01

    A survey of fuel cell technology and applications is presented. The operating principles, performance capabilities, and limitations of fuel cells are discussed. Diagrams of fuel cell construction and operating characteristics are provided. Photographs of typical installations are included.

  19. Bayesian Model Averaging and Weighted Average Least Squares : Equivariance, Stability, and Numerical Issues

    NARCIS (Netherlands)

    De Luca, G.; Magnus, J.R.

    2011-01-01

    This article is concerned with the estimation of linear regression models with uncertainty about the choice of the explanatory variables. We introduce the Stata commands bma and wals which implement, respectively, the exact Bayesian Model Averaging (BMA) estimator and the Weighted Average Least Squa

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

    Science.gov (United States)

    Syarifah, Ratna Dewi; Suud, Zaki

    2015-09-01

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

  1. Bootstrapping Density-Weighted Average Derivatives

    DEFF Research Database (Denmark)

    Cattaneo, Matias D.; Crump, Richard K.; Jansson, Michael

    Employing the "small bandwidth" asymptotic framework of Cattaneo, Crump, and Jansson (2009), this paper studies the properties of a variety of bootstrap-based inference procedures associated with the kernel-based density-weighted averaged derivative estimator proposed by Powell, Stock, and Stoker......" variance estimator derived from the "small bandwidth" asymptotic framework. The results of a small-scale Monte Carlo experiment are found to be consistent with the theory and indicate in particular that sensitivity with respect to the bandwidth choice can be ameliorated by using the "robust...

  2. The average free volume model for liquids

    CERN Document Server

    Yu, Yang

    2014-01-01

    In this work, the molar volume thermal expansion coefficient of 59 room temperature ionic liquids is compared with their van der Waals volume Vw. Regular correlation can be discerned between the two quantities. An average free volume model, that considers the particles as hard core with attractive force, is proposed to explain the correlation in this study. A combination between free volume and Lennard-Jones potential is applied to explain the physical phenomena of liquids. Some typical simple liquids (inorganic, organic, metallic and salt) are introduced to verify this hypothesis. Good agreement from the theory prediction and experimental data can be obtained.

  3. Phase-averaged transport for quasiperiodic Hamiltonians

    CERN Document Server

    Bellissard, J; Schulz-Baldes, H

    2002-01-01

    For a class of discrete quasi-periodic Schroedinger operators defined by covariant re- presentations of the rotation algebra, a lower bound on phase-averaged transport in terms of the multifractal dimensions of the density of states is proven. This result is established under a Diophantine condition on the incommensuration parameter. The relevant class of operators is distinguished by invariance with respect to symmetry automorphisms of the rotation algebra. It includes the critical Harper (almost-Mathieu) operator. As a by-product, a new solution of the frame problem associated with Weyl-Heisenberg-Gabor lattices of coherent states is given.

  4. Fluctuations of wavefunctions about their classical average

    CERN Document Server

    Bénet, L; Hernandez-Saldana, H; Izrailev, F M; Leyvraz, F; Seligman, T H

    2003-01-01

    Quantum-classical correspondence for the average shape of eigenfunctions and the local spectral density of states are well-known facts. In this paper, the fluctuations of the quantum wavefunctions around the classical value are discussed. A simple random matrix model leads to a Gaussian distribution of the amplitudes whose width is determined by the classical shape of the eigenfunction. To compare this prediction with numerical calculations in chaotic models of coupled quartic oscillators, we develop a rescaling method for the components. The expectations are broadly confirmed, but deviations due to scars are observed. This effect is much reduced when both Hamiltonians have chaotic dynamics.

  5. Sparsity averaging for radio-interferometric imaging

    CERN Document Server

    Carrillo, Rafael E; Wiaux, Yves

    2014-01-01

    We propose a novel regularization method for compressive imaging in the context of the compressed sensing (CS) theory with coherent and redundant dictionaries. Natural images are often complicated and several types of structures can be present at once. It is well known that piecewise smooth images exhibit gradient sparsity, and that images with extended structures are better encapsulated in wavelet frames. Therefore, we here conjecture that promoting average sparsity or compressibility over multiple frames rather than single frames is an extremely powerful regularization prior.

  6. A sixth order averaged vector field method

    OpenAIRE

    Li, Haochen; Wang, Yushun; Qin, Mengzhao

    2014-01-01

    In this paper, based on the theory of rooted trees and B-series, we propose the concrete formulas of the substitution law for the trees of order =5. With the help of the new substitution law, we derive a B-series integrator extending the averaged vector field (AVF) method to high order. The new integrator turns out to be of order six and exactly preserves energy for Hamiltonian systems. Numerical experiments are presented to demonstrate the accuracy and the energy-preserving property of the s...

  7. Fluctuations of wavefunctions about their classical average

    Energy Technology Data Exchange (ETDEWEB)

    Benet, L [Centro Internacional de Ciencias, Ciudad Universitaria, Chamilpa, Cuernavaca (Mexico); Flores, J [Centro Internacional de Ciencias, Ciudad Universitaria, Chamilpa, Cuernavaca (Mexico); Hernandez-Saldana, H [Centro Internacional de Ciencias, Ciudad Universitaria, Chamilpa, Cuernavaca (Mexico); Izrailev, F M [Centro Internacional de Ciencias, Ciudad Universitaria, Chamilpa, Cuernavaca (Mexico); Leyvraz, F [Centro Internacional de Ciencias, Ciudad Universitaria, Chamilpa, Cuernavaca (Mexico); Seligman, T H [Centro Internacional de Ciencias, Ciudad Universitaria, Chamilpa, Cuernavaca (Mexico)

    2003-02-07

    Quantum-classical correspondence for the average shape of eigenfunctions and the local spectral density of states are well-known facts. In this paper, the fluctuations of the quantum wavefunctions around the classical value are discussed. A simple random matrix model leads to a Gaussian distribution of the amplitudes whose width is determined by the classical shape of the eigenfunction. To compare this prediction with numerical calculations in chaotic models of coupled quartic oscillators, we develop a rescaling method for the components. The expectations are broadly confirmed, but deviations due to scars are observed. This effect is much reduced when both Hamiltonians have chaotic dynamics.

  8. Grassmann Averages for Scalable Robust PCA

    OpenAIRE

    2014-01-01

    As the collection of large datasets becomes increasingly automated, the occurrence of outliers will increase—“big data” implies “big outliers”. While principal component analysis (PCA) is often used to reduce the size of data, and scalable solutions exist, it is well-known that outliers can arbitrarily corrupt the results. Unfortunately, state-of-the-art approaches for robust PCA do not scale beyond small-to-medium sized datasets. To address this, we introduce the Grassmann Average (GA), whic...

  9. Future aviation fuels overview

    Science.gov (United States)

    Reck, G. M.

    1980-01-01

    The outlook for aviation fuels through the turn of the century is briefly discussed and the general objectives of the NASA Lewis Alternative Aviation Fuels Research Project are outlined. The NASA program involves the evaluation of potential characteristics of future jet aircraft fuels, the determination of the effects of those fuels on engine and fuel system components, and the development of a component technology to use those fuels.

  10. Student science enrichment training program

    Energy Technology Data Exchange (ETDEWEB)

    Sandhu, S.S.

    1994-08-01

    This is a report on the Student Science Enrichment Training Program, with special emphasis on chemical and computer science fields. The residential summer session was held at the campus of Claflin College, Orangeburg, SC, for six weeks during 1993 summer, to run concomitantly with the college`s summer school. Fifty participants selected for this program, included high school sophomores, juniors and seniors. The students came from rural South Carolina and adjoining states which, presently, have limited science and computer science facilities. The program focused on high ability minority students, with high potential for science engineering and mathematical careers. The major objective was to increase the pool of well qualified college entering minority students who would elect to go into science, engineering and mathematical careers. The Division of Natural Sciences and Mathematics and engineering at Claflin College received major benefits from this program as it helped them to expand the Departments of Chemistry, Engineering, Mathematics and Computer Science as a result of additional enrollment. It also established an expanded pool of well qualified minority science and mathematics graduates, which were recruited by the federal agencies and private corporations, visiting Claflin College Campus. Department of Energy`s relationship with Claflin College increased the public awareness of energy related job opportunities in the public and private sectors.

  11. Thorium fuel performance assessment in HTRs

    Energy Technology Data Exchange (ETDEWEB)

    Allelein, H.-J. [Forschungszentrum Jülich, D-52425 Jülich (Germany); RWTH Aachen, D-52072 Aachen (Germany); Kania, M.J.; Nabielek, H. [Forschungszentrum Jülich, D-52425 Jülich (Germany); Verfondern, K., E-mail: k.verfondern@fz-juelich.de [Forschungszentrum Jülich, D-52425 Jülich (Germany)

    2014-05-01

    Thorium as a nuclear fuel is receiving renewed interest, because of its widespread availability and the good irradiation performance of Th and mixed (Th,U) oxide compounds as fuels in nuclear power systems. Early HTR development employed thorium together with high-enriched uranium. After 1980, most HTR fuel systems switched to low-enriched uranium. After completing fuel development for AVR and THTR with BISO coated particles, the German program expanded efforts on a new program utilizing thorium and high-enriched uranium TRISO coated particles for advanced HTR concepts for process heat applications (PNP) and direct-cycle electricity production (HHT). The combination of LTI inner and outer pyrocarbon layers surrounding a strong, stable SiC layer greatly improved manufacturing conditions and the subsequent contamination and defective particle fractions in production fuel elements. In addition, this combination provided improved mechanical strength and a higher degree of solid fission product retention, not known previously with HTI-BISO coatings. The improved performance of the HEU (Th,U)O{sub 2} TRISO fuel system was successfully demonstrated in three primary areas of development: manufacturing, irradiation testing under normal operating conditions, and accident simulation testing. In terms of demonstrating performance for advanced HTR applications, the experimental failure statistic from manufacture and irradiation testing are significantly below the coated particle requirements specified for PNP and HHT designs at the time. Covering a range to 1300 °C in normal operations and 1600 °C in accidents, with burnups up to 13% FIMA and fast fluences to 8 × 10{sup 25} m{sup −2} (E > 16 fJ), the results exceed the design limits on manufacturing and operational requirements for the German HTR Modul concept, which were: <6.5 × 10{sup −5} for manufacturing; <2 × 10{sup −4} for normal operating conditions; and <5 × 10{sup −4} for accident conditions. These

  12. Neutron intensity of fast reactor spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Takamatsu, Misao; Aoyama, Takafumi [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-03-01

    Neutron intensity of spent fuel of the JOYO Mk-II core with a burnup of 62,500 MWd/t and cooling time of 5.2 years was measured at the spent fuel storage pond. The measured data were compared with the calculated values based on the JOYO core management code system `MAGI`, and the average C/E approximately 1.2 was obtained. It was found that the axial neutron intensity didn`t simply follow the burnup distribution, and the neutron intensity was locally increased at the bottom end of the fuel region due to an accumulation of {sup 244}Cm. (author)

  13. Catalytic Fuel Conversion Facility

    Data.gov (United States)

    Federal Laboratory Consortium — This facility enables unique catalysis research related to power and energy applications using military jet fuels and alternative fuels. It is equipped with research...

  14. Fuel cell

    Energy Technology Data Exchange (ETDEWEB)

    Enomoto, Hirofumi.

    1989-05-22

    This invention aims to maintain a long-term operation with stable cell output characteristics by uniformly supplying an electrolyte from the reserver to the matrix layer over the entire matrix layer, and further to prevent the excessive wetting of the catalyst layer by smoothly absorbing the volume change of the electrolyte, caused by the repeated stop/start-up of the fuel cell, within the reserver system. For this purpose, in this invention, an electrolyte transport layer, which connects with an electrolyte reservor formed at the electrode end, is partly formed between the electrode material and the catalyst layer; a catalyst layer, which faces the electrolyte transport layer, has through-holes, which connect to the matrix, dispersely distributed. The electrolyte-transport layer is a thin sheet of a hydrophilic fibers which are non-wovens of such fibers as carbon, silicon carbide, silicon nitride or inorganic oxides. 11 figs.

  15. Technology Roadmap: Fuel Economy of Road Vehicles

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-07-01

    This roadmap explores the potential improvement of existing technologies to enhance the average fuel economy of motorised vehicles; the roadmap’s vision is to achieve a 30% to 50% reduction in fuel use per kilometre from new road vehicles including 2-wheelers, LDV s and HDV s) around the world in 2030, and from the stock of all vehicles on the road by 2050. This achievement would contribute to significant reductions in GHG emissions and oil use, compared to a baseline projection. Different motorised modes are treated separately, with a focus on LDV s, HDV s and powered two-wheelers. A section on in-use fuel economy also addresses technical and nontechnical parameters that could allow fuel economy to drastically improve over the next decades. Technology cost analysis and payback time show that significant progress can be made with low or negative cost for fuel-efficient vehicles over their lifetime use. Even though the latest data analysed by the IEA for fuel economy between 2005 and 2008 showed that a gap exists in achieving the roadmap’s vision, cutting the average fuel economy of road motorised vehicles by 30% to 50% by 2030 is achievable, and the policies and technologies that could help meet this challenge are already deployed in many places around the world.

  16. Fuel Management Strategies for a Possible Future LEU Core of a TRIGA Mark II Vienna

    Energy Technology Data Exchange (ETDEWEB)

    Khan, R.; Villa, M.; Steinhauser, G.; Boeck, H. [Vienna University of Technology-Atominstitut (Austria)

    2011-07-01

    The Vienna University of Technology/Atominstitut (VUT/ATI) operates a TRIGA Mark II research reactor. It is operated with a completely mixed core of three different types of fuel. Due to the US fuel return program, the ATI have to return its High Enriched Uranium (HEU) fuel latest by 2019. As an alternate, the Low Enrich Uranium (LEU) fuel is under consideration. The detailed results of the core conversion study are presented at the RRFM 2011 conference. This paper describes the burn up calculations of the new fuel to predict the future burn up behavior and core life time. It also develops an effective and optimized fuel management strategy for a possible future operation of the TRIGA Mark II with a LEU core. This work is performed by the combination of MCNP5 and diffusion based neutronics code TRIGLAV. (author)

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

    Science.gov (United States)

    Collette, R.; King, J.; Buesch, C.; Keiser, D. D.; Williams, W.; Miller, B. D.; Schulthess, J.

    2016-07-01

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

  18. MACHINE PROTECTION FOR HIGH AVERAGE CURRENT LINACS

    Energy Technology Data Exchange (ETDEWEB)

    Jordan, Kevin; Allison, Trent; Evans, Richard; Coleman, James; Grippo, Albert

    2003-05-01

    A fully integrated Machine Protection System (MPS) is critical to efficient commissioning and safe operation of all high current accelerators. The Jefferson Lab FEL [1,2] has multiple electron beam paths and many different types of diagnostic insertion devices. The MPS [3] needs to monitor both the status of these devices and the magnet settings which define the beam path. The matrix of these devices and beam paths are programmed into gate arrays, the output of the matrix is an allowable maximum average power limit. This power limit is enforced by the drive laser for the photocathode gun. The Beam Loss Monitors (BLMs), RF status, and laser safety system status are also inputs to the control matrix. There are 8 Machine Modes (electron path) and 8 Beam Modes (average power limits) that define the safe operating limits for the FEL. Combinations outside of this matrix are unsafe and the beam is inhibited. The power limits range from no beam to 2 megawatts of electron beam power.

  19. Trajectory averaging for stochastic approximation MCMC algorithms

    KAUST Repository

    Liang, Faming

    2010-10-01

    The subject of stochastic approximation was founded by Robbins and Monro [Ann. Math. Statist. 22 (1951) 400-407]. After five decades of continual development, it has developed into an important area in systems control and optimization, and it has also served as a prototype for the development of adaptive algorithms for on-line estimation and control of stochastic systems. Recently, it has been used in statistics with Markov chain Monte Carlo for solving maximum likelihood estimation problems and for general simulation and optimizations. In this paper, we first show that the trajectory averaging estimator is asymptotically efficient for the stochastic approximation MCMC (SAMCMC) algorithm under mild conditions, and then apply this result to the stochastic approximation Monte Carlo algorithm [Liang, Liu and Carroll J. Amer. Statist. Assoc. 102 (2007) 305-320]. The application of the trajectory averaging estimator to other stochastic approximationMCMC algorithms, for example, a stochastic approximation MLE algorithm for missing data problems, is also considered in the paper. © Institute of Mathematical Statistics, 2010.

  20. Local average height distribution of fluctuating interfaces

    Science.gov (United States)

    Smith, Naftali R.; Meerson, Baruch; Sasorov, Pavel V.

    2017-01-01

    Height fluctuations of growing surfaces can be characterized by the probability distribution of height in a spatial point at a finite time. Recently there has been spectacular progress in the studies of this quantity for the Kardar-Parisi-Zhang (KPZ) equation in 1 +1 dimensions. Here we notice that, at or above a critical dimension, the finite-time one-point height distribution is ill defined in a broad class of linear surface growth models unless the model is regularized at small scales. The regularization via a system-dependent small-scale cutoff leads to a partial loss of universality. As a possible alternative, we introduce a local average height. For the linear models, the probability density of this quantity is well defined in any dimension. The weak-noise theory for these models yields the "optimal path" of the interface conditioned on a nonequilibrium fluctuation of the local average height. As an illustration, we consider the conserved Edwards-Wilkinson (EW) equation, where, without regularization, the finite-time one-point height distribution is ill defined in all physical dimensions. We also determine the optimal path of the interface in a closely related problem of the finite-time height-difference distribution for the nonconserved EW equation in 1 +1 dimension. Finally, we discuss a UV catastrophe in the finite-time one-point distribution of height in the (nonregularized) KPZ equation in 2 +1 dimensions.

  1. Intensity contrast of the average supergranule

    CERN Document Server

    Langfellner, J; Gizon, L

    2016-01-01

    While the velocity fluctuations of supergranulation dominate the spectrum of solar convection at the solar surface, very little is known about the fluctuations in other physical quantities like temperature or density at supergranulation scale. Using SDO/HMI observations, we characterize the intensity contrast of solar supergranulation at the solar surface. We identify the positions of ${\\sim}10^4$ outflow and inflow regions at supergranulation scales, from which we construct average flow maps and co-aligned intensity and magnetic field maps. In the average outflow center, the maximum intensity contrast is $(7.8\\pm0.6)\\times10^{-4}$ (there is no corresponding feature in the line-of-sight magnetic field). This corresponds to a temperature perturbation of about $1.1\\pm0.1$ K, in agreement with previous studies. We discover an east-west anisotropy, with a slightly deeper intensity minimum east of the outflow center. The evolution is asymmetric in time: the intensity excess is larger 8 hours before the reference t...

  2. 21 CFR 137.350 - Enriched rice.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 2 2010-04-01 2010-04-01 false Enriched rice. 137.350 Section 137.350 Food and... Related Products § 137.350 Enriched rice. (a) The foods for which definitions and standards of identity are prescribed by this section are forms of milled rice (except rice coated with talc and glucose...

  3. A Component Analysis of Marriage Enrichment.

    Science.gov (United States)

    Buston, Beverley G.; And Others

    Although marriage enrichment programs have been shown to be effective for many couples, a multidimensional approach to assessment is needed in investigating these groups. The components of information and social support in successful marriage enrichment programs were compared in a completely crossed 2 x 2 factorial design with repeated measures.…

  4. Inoculation Stress Hypothesis of Environmental Enrichment

    Science.gov (United States)

    Crofton, Elizabeth J.; Zhang, Yafang; Green, Thomas A.

    2014-01-01

    One hallmark of psychiatric conditions is the vast continuum of individual differences in susceptibility vs. resilience resulting from the interaction of genetic and environmental factors. The environmental enrichment paradigm is an animal model that is useful for studying a range of psychiatric conditions, including protective phenotypes in addiction and depression models. The major question is how environmental enrichment, a non-drug and non-surgical manipulation, can produce such robust individual differences in such a wide range of behaviors. This paper draws from a variety of published sources to outline a coherent hypothesis of inoculation stress as a factor producing the protective enrichment phenotypes. The basic tenet suggests that chronic mild stress from living in a complex environment and interacting non-aggressively with conspecifics can inoculate enriched rats against subsequent stressors and/or drugs of abuse. This paper reviews the enrichment phenotypes, mulls the fundamental nature of environmental enrichment vs. isolation, discusses the most appropriate control for environmental enrichment, and challenges the idea that cortisol/corticosterone equals stress. The intent of the inoculation stress hypothesis of environmental enrichment is to provide a scaffold with which to build testable hypotheses for the elucidation of the molecular mechanisms underlying these protective phenotypes and thus provide new therapeutic targets to treat psychiatric/neurological conditions. PMID:25449533

  5. Proliferation Resistant Nuclear Reactor Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Gray, L W; Moody, K J; Bradley, K S; Lorenzana, H E

    2011-02-18

    Global appetite for fission power is projected to grow dramatically this century, and for good reason. Despite considerable research to identify new sources of energy, fission remains the most plentiful and practical alternative to fossil fuels. The environmental challenges of fossil fuel have made the fission power option increasingly attractive, particularly as we are forced to rely on reserves in ecologically fragile or politically unstable corners of the globe. Caught between a globally eroding fossil fuel reserve as well as the uncertainty and considerable costs in the development of fusion power, most of the world will most likely come to rely on fission power for at least the remainder of the 21st century. Despite inevitable growth, fission power faces enduring challenges in sustainability and security. One of fission power's greatest hurdles to universal acceptance is the risk of potential misuse for nefarious purposes of fissionable byproducts in spent fuel, such as plutonium. With this issue in mind, we have discussed intrinsic concepts in this report that are motivated by the premise that the utility, desirability, and applicability of nuclear materials can be reduced. In a general sense, the intrinsic solutions aim to reduce or eliminate the quantity of existing weapons usable material; avoid production of new weapons-usable material through enrichment, breeding, extraction; or employ engineering solutions to make the fuel cycle less useful or more difficult for producing weapons-usable material. By their nature, these schemes require modifications to existing fuel cycles. As such, the concomitants of these modifications require engagement from the nuclear reactor and fuel-design community to fully assess their effects. Unfortunately, active pursuit of any scheme that could further complicate the spread of domestic nuclear power will probably be understandably unpopular. Nevertheless, the nonproliferation and counterterrorism issues are paramount

  6. Enhanced Control of PWR Primary Coolant Water Chemistry Using Selective Separation Systems for Recovery and Recycle of Enriched Boric Acid

    Energy Technology Data Exchange (ETDEWEB)

    Ken Czerwinski; Charels Yeamans; Don Olander; Kenneth Raymond; Norman Schroeder; Thomas Robison; Bryan Carlson; Barbara Smit; Pat Robinson

    2006-02-28

    The objective of this project is to develop systems that will allow for increased nuclear energy production through the use of enriched fuels. The developed systems will allow for the efficient and selective recover of selected isotopes that are additives to power water reactors' primary coolant chemistry for suppression of corrosion attack on reactor materials.

  7. Optimum nuclear design of target fuel rod for Mo-99 production in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Myung Hyun [Kyung Hee University, Seoul (Korea)

    1998-04-01

    Nuclear target design for Mo-99 production in HANARO was performed, KAERI proposed target design was analyzed and its feasibility was shown. Three commercial target designs of Cintichem, ANL and KAERI were tested for the HANARO irradiation an d they all satisfied with design specification. A parametric study was done for target design options and Mo-99 yields ratio and surface heat flux were compared. Tested parameters were target fuel thickness, irradiation location, target axial length, packing density of powder fuel, size of target radius, target geometry, fuel enrichment, fuel composition, and cladding material. Optimized target fuel was designed for both LEU and HEU options. (author). 17 refs., 33 figs., 42 tabs.

  8. Temperature coefficients in the Dragon low-enriched power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hansen, U.

    1972-05-15

    The temperature coefficient of the fuel and of the moderator have been evaluated for the Dragon HTR design for different stages in reactor life, initial core, end of no-refuelling period and equilibrium conditions. The investigation has shown the low-enriched HTR to have a strong, positive moderator coefficient. In some cases and for special operating conditions, even leading to a positive total temperature coefficient. This does not imply, however, that the HTR is an unsafe reactor system. By adequate design of the control system, safe and reliable operating characteristics can be achieved. This has already been proved satisfactory through many years of operation of other graphite moderated systems, such as the Magnox stations.

  9. Winter fuels report

    Energy Technology Data Exchange (ETDEWEB)

    1995-01-13

    The Winter Fuels Report is intended to provide concise, timely information to the industry, the press, policymakers, consumers, analysts, and State and local governments on the following topics: distillate fuel oil net production, imports and stocks on a US level and for all Petroleum Administration for Defense Districts (PADD) and product supplied on a US level; propane net production, imports and stocks on a US level and for PADD`s I, II, and III; natural gas supply and disposition and underground storage for the US and consumption for all PADD`s, as well as selected National average prices; residential and wholesale pricing data for heating oil and propane for those States participating in the joint Energy Information Administration (EIA)/State Heating Oil and Propane Program; crude oil and petroleum price comparisons for the US and selected cities; and a 6-10 day, 30-Day, and 90-Day outlook for temperature and precipitation and US total heating degree-days by city.

  10. Enriching an effect calculus with linear types

    DEFF Research Database (Denmark)

    Egger, Jeff; Møgelberg, Rasmus Ejlers; Simpson, Alex

    2009-01-01

    We define an ``enriched effect calculus'' by conservatively extending  a type theory for computational effects with primitives from linear logic. By doing so, we obtain a generalisation of linear type theory, intended as a formalism for expressing linear aspects of effects. As a worked example, we...... formulate  linearly-used continuations in the enriched effect calculus. These are captured by a fundamental translation of the enriched effect calculus into itself, which extends existing call-by-value and call-by-name linearly-used CPS translations. We show that our translation is involutive. Full...... completeness results for the various linearly-used CPS translations  follow. Our main results, the conservativity of enriching the effect calculus with linear primitives, and the involution property of the fundamental translation, are proved using a category-theoretic semantics for the enriched effect calculus...

  11. Determination of IRT-2M fuel burnup by gamma spectrometry.

    Science.gov (United States)

    Koleška, Michal; Viererbl, Ladislav; Marek, Milan; Ernest, Jaroslav; Šunka, Michal; Vinš, Miroslav

    2016-01-01

    A spectrometric system was developed for evaluating spent fuel in the LVR-15 research reactor, which employs highly enriched (36%) IRT-2M-type fuel. Such system allows the measurement of detailed fission product profiles. Within these measurements, nuclides such as (137)Cs, (134)Cs, (144)Ce, (106)Ru and (154)Eu may be detected in fuel assemblies with different cooling times varying between 1.67 and 7.53 years. Burnup calculations using the MCNPX Monte Carlo code data showed good agreement with measurements, though some discrepancies were observed in certain regions. These discrepancies are attributed to the evaluation of irradiation history, reactor regulation pattern and buildup schemes.

  12. Detailed investigation of a vaporising fuel spray. Part 1: Experimental investigation of time averaged spray

    Science.gov (United States)

    Yule, A. J.; Seng, C. A.; Boulderstone, R.; Ungut, A.; Felton, P. G.; Chigier, N. A.

    1980-01-01

    A laser tomographic light scattering technique provides rapid and accurate high resolution measurements of droplet sizes, concentrations, and vaporization. Measurements using a computer interfaced thermocouple are presented and it is found that the potential exists for separating gas and liquid temperature measurements and diagnosing local spray density by in situ analysis of the response characteristics of the thermocouple. The thermocouple technique provides a convenient means for measuring mean gas velocity in both hot and cold two phase flows. The experimental spray is axisymmetric and has carefully controlled initial and boundary conditions. The flow is designed to give relatively insignificant transfer of momentum and mass from spray to air flow. The effects of (1) size-dependent droplet dispersion by the turbulence, (2) the initial spatial segregation of droplet sizes during atomization, and (3) the interaction between droplets and coherent large eddies are diagnosed.

  13. Experimental study on improving cement quality with oxygen- enriched combustion technology

    Science.gov (United States)

    Liu, Y. Q.; Zhang, A. M.; Qing, S.; Li, F. S.; Yang, S. P.; Yang, Z. F.

    2015-12-01

    With the intensification of the global energy crisis, the production cost of enterprises is continuously increasing because of the rising fuel prices and high requirements for environmental protection. As result, energy savings and environmental protection are vital considerations for a variety of enterprises. As a practical energy-saving technology, oxygen- enriched combustion has played a major role in energy saving and emissions reduction as its application in industrial furnaces has been popularized in recent years. This experiment was conducted in a cement rotary kiln with a capacity of 4000 t/d in a factory in China. Based on measured data in the oxygen-enriched combustion experiment, we determined the patterns of variation in the main parameters of the cement rotary kiln under oxygen-enriched production conditions. The results provide important theoretical and practical base for the cement building materials industry in energy saving and emissions reduction.

  14. Recycling as an option of used nuclear fuel management strategy

    Energy Technology Data Exchange (ETDEWEB)

    Zagar, Tomaz, E-mail: tomaz.zagar@gen-energija.s [GEN energija, d.o.o., Cesta 4. julija 42, 8270 Krsko (Slovenia); Institute Jozef Stefan, Jamova 39, 1000 Ljubljana (Slovenia); Bursic, Ales; Spiler, Joze [GEN energija, d.o.o., Cesta 4. julija 42, 8270 Krsko (Slovenia); Kim, Dana; Chiguer, Mustapha; David, Gilles; Gillet, Philippe [AREVA, 33 rue La Fayette, 75009 Paris (France)

    2011-04-15

    The paper presents recycling as an option of used nuclear fuel management strategy with specific focus on the Slovenia. GEN energija is an independent supplier of integral and competitive electricity for Slovenia. In response to growing energy needs, GEN has conducted several feasibility and installation studies of a new nuclear power plant in Slovenia. With sustainable development, the environment, and public acceptance in mind, GEN conducted a study with AREVA concerning the options for the management of its' new plant's used nuclear fuel. After a brief reminder of global political and economic context, solutions for used nuclear fuel management using current technologies are presented in the study as well as an economic assessment of a closed nuclear fuel cycle. The paper evaluates and proposes practical solutions for mid-term issues on used nuclear fuel management strategies. Different scenarios for used nuclear fuel management are presented, where used nuclear fuel recycling (as MOX, for mixed oxide fuel, and ERU, for enriched reprocessed uranium) are considered. The study concludes that closing the nuclear fuel cycle will allow Slovenia to have a supplementary fuel supply for its new reactor via recycling, while reducing the radiotoxicity, thermal output, and volume of its wastes for final disposal, reducing uncertainties, gaining public acceptance, and allowing time for capitalization on investments for final disposal.

  15. Recent trends in aviation turbine fuel properties

    Science.gov (United States)

    Friedman, R.

    1982-01-01

    Plots and tables, compiled from Department of Energy (and predecessor agency) inspection reports from 1969 to 1980, present ranges, averages, extremes, and trends for most of the 22 properties of Jet A aviation turbine fuel. In recent years, average values of aromatics content, mercaptan sulfur content, distillation temperature of 10 percent recovered, smoke point, and freezing point show small but recognizable trends toward their specification limits. About 80 percent of the fuel samples had at least one property near specification, defined as within a standard band about the specification limit. By far the most common near-specification properties were aromatics content, smoke point, and freezing point.

  16. Enrichment of diluted cell populations from large sample volumes using 3D carbon-electrode dielectrophoresis.

    Science.gov (United States)

    Islam, Monsur; Natu, Rucha; Larraga-Martinez, Maria Fernanda; Martinez-Duarte, Rodrigo

    2016-05-01

    Here, we report on an enrichment protocol using carbon electrode dielectrophoresis to isolate and purify a targeted cell population from sample volumes up to 4 ml. We aim at trapping, washing, and recovering an enriched cell fraction that will facilitate downstream analysis. We used an increasingly diluted sample of yeast, 10(6)-10(2) cells/ml, to demonstrate the isolation and enrichment of few cells at increasing flow rates. A maximum average enrichment of 154.2 ± 23.7 times was achieved when the sample flow rate was 10 μl/min and yeast cells were suspended in low electrically conductive media that maximizes dielectrophoresis trapping. A COMSOL Multiphysics model allowed for the comparison between experimental and simulation results. Discussion is conducted on the discrepancies between such results and how the model can be further improved.

  17. LMFBR operation in the nuclear cycle without fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Toshinsky, S.I. [Institute of Physics and Power Engineering, Kaluga (Russian Federation)

    1997-12-01

    Substantiation is given to expediency of investigation of nuclear power (NP) development with fast reactors cooled by lead-bismuth alloy operating during extended time in the open nuclear fuel cycle with slightly enriched or depleted uranium make-up. 9 refs., 1 fig., 6 tabs.

  18. Axial blanket for 16NGF Angra 1 fuel type

    Energy Technology Data Exchange (ETDEWEB)

    Sadde, Luciano Martins; Faria, Eduardo Fernandes [Industrias Nucleares do Brasil (INB), Resende, RJ (Brazil)]. E-mails: sadde@inb.gov.br; faria@inb.gov.br; Sang-Keun You [Korea Nuclear Fuel Co. Ltd. (KNFC), Taejon (Korea, Republic of)]. E-mail: skyou@knfc.co.kr

    2007-07-01

    Angra-1, Kori-2 and Krsko are nuclear power plants with the same design. However, the fuel assemblies have some differences in design due to the countries strategies and the differences in the fabrication process. The 16NGF (16x16 Next Generation Fuel) was developed by INB, KNFC and Westinghouse in order to be used in these three nuclear power plants and the 'Axial Blanket' is one of the new features for the 16NGF design. The main purpose of the Axial Blanket Optimization study is to determine which axial blanket enrichment and length would provide the better fuel cycle cost benefit. All of the calculations were performed using Gadolinium as Burnable Absorber and solid pellets type for Axial Blanket. The results indicate 1.8 w/o U235 enrichment and 8 inches length as the best option of Axial Blanket from the fuel cycle cost benefit standpoint. The economy is about 1.8%. The difference in the reload cost in the range between 1.5 and 2.6 w/o U235 enrichment and for the 6 and 8 inches length is not so significant. Due that, from the Fq limit standpoint and also for longer cycle length requirements, a higher axial blanket enrichment (2.6 w/o) and shorter length (6 inches) is recommended. (author)

  19. Experimental Study of Coal and Gas Outbursts Related to Gas-Enriched Areas

    Science.gov (United States)

    Tu, Qingyi; Cheng, Yuanping; Guo, Pinkun; Jiang, Jingyu; Wang, Liang; Zhang, Rong

    2016-09-01

    A coal and gas outburst can lead to a catastrophic failure in a coal mine. These outbursts usually occur where the distribution of coal seam gas is abnormal, commonly in tectonic belts. To study the effects of the abnormal distribution of this gas on outbursts, an experimental apparatus to collect data on simulated coal seam outbursts was constructed. Experiments on specimens containing discrete gas-enriched areas were run to induce artificial gas outbursts and further study of these outbursts using data from the experiment was conducted. The results suggest that more gas and outburst energy are contained in gas-enriched areas and this permits these areas to cause an outburst easily, even though the gas pressure in them is lower. During mining, the disappearance of the sealing effect of a coal pillar establishes the occurrence conditions for an outburst. When the enriched gas and outburst energy in the gas-enriched area is released suddenly, a reverse unloading wave and a high gas pressure gradient are formed, which have tension effects on the coal. Under these effects, the fragmentation degree of the coal intensifies and the intensity of the outburst increases. Because a high gas pressure gradient is maintained near the exposed surface and the enriched energy release reduces the coal strength, the existence of a gas-enriched area in coal leads to a faster outburst and the average thickness of the spall is smaller than where is no gas-enriched area.

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

    Science.gov (United States)

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

    2009-08-01

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

  1. Non-destructive Testing Dummy Nuclear Fuel Rods by Neutron Radiography

    Institute of Scientific and Technical Information of China (English)

    WEI; Guo-hai; HAN; Song-bai; HE; Lin-feng; WANG; Yu; WANG; Hong-li; LIU; Yun-tao; CHEN; Dong-feng

    2013-01-01

    As a unique non-destructive testing technique,neutron radiography can be used to measure nuclear fuel rods with radioactivity.The images of the dummy nuclear fuel rods were obtained at the CARR.Through imaging analysis methods,the structure defections,the hydrogen accumulation in the cladding and the 235U enrichment of the pellet were studied and analyzed.Experiences for non-destructive testing real PWR nuclear fuel rods by NR

  2. Deuterium enrichment of interstellar dusts

    Science.gov (United States)

    Das, Ankan; Chakrabarti, Sandip Kumar; Majumdar, Liton; Sahu, Dipen

    2016-07-01

    High abundance of some abundant and simple interstellar species could be explained by considering the chemistry that occurs on interstellar dusts. Because of its simplicity, the rate equation method is widely used to study the surface chemistry. However, because the recombination efficiency for the formation of any surface species is highly dependent on various physical and chemical parameters, the Monte Carlo method is best suited for addressing the randomness of the processes. We carry out Monte-Carlo simulation to study deuterium enrichment of interstellar grain mantle under various physical conditions. Based on the physical properties, various types of clouds are considered. We find that in diffuse cloud regions, very strong radiation fields persists and hardly a few layers of surface species are formed. In translucent cloud regions with a moderate radiation field, significant number of layers would be produced and surface coverage is mainly dominated by photo-dissociation products such as, C, CH_3, CH_2D, OH and OD. In the intermediate dense cloud regions (having number density of total hydrogen nuclei in all forms ˜2 × 10^4 cm^{-3}), water and methanol along with their deuterated derivatives are efficiently formed. For much higher density regions (˜10^6 cm^{-3}), water and methanol productions are suppressed but surface coverage of CO, CO_2, O_2, O_3 are dramatically increased. We find a very high degree of fractionation of water and methanol. Observational results support a high fractionation of methanol but surprisingly water fractionation is found to be low. This is in contradiction with our model results indicating alternative routes for de-fractionation of water.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-09-01

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

  4. Averaged Null Energy Condition from Causality

    CERN Document Server

    Hartman, Thomas; Tajdini, Amirhossein

    2016-01-01

    Unitary, Lorentz-invariant quantum field theories in flat spacetime obey microcausality: commutators vanish at spacelike separation. For interacting theories in more than two dimensions, we show that this implies that the averaged null energy, $\\int du T_{uu}$, must be positive. This non-local operator appears in the operator product expansion of local operators in the lightcone limit, and therefore contributes to $n$-point functions. We derive a sum rule that isolates this contribution and is manifestly positive. The argument also applies to certain higher spin operators other than the stress tensor, generating an infinite family of new constraints of the form $\\int du X_{uuu\\cdots u} \\geq 0$. These lead to new inequalities for the coupling constants of spinning operators in conformal field theory, which include as special cases (but are generally stronger than) the existing constraints from the lightcone bootstrap, deep inelastic scattering, conformal collider methods, and relative entropy. We also comment ...

  5. Geographic Gossip: Efficient Averaging for Sensor Networks

    CERN Document Server

    Dimakis, Alexandros G; Wainwright, Martin J

    2007-01-01

    Gossip algorithms for distributed computation are attractive due to their simplicity, distributed nature, and robustness in noisy and uncertain environments. However, using standard gossip algorithms can lead to a significant waste in energy by repeatedly recirculating redundant information. For realistic sensor network model topologies like grids and random geometric graphs, the inefficiency of gossip schemes is related to the slow mixing times of random walks on the communication graph. We propose and analyze an alternative gossiping scheme that exploits geographic information. By utilizing geographic routing combined with a simple resampling method, we demonstrate substantial gains over previously proposed gossip protocols. For regular graphs such as the ring or grid, our algorithm improves standard gossip by factors of $n$ and $\\sqrt{n}$ respectively. For the more challenging case of random geometric graphs, our algorithm computes the true average to accuracy $\\epsilon$ using $O(\\frac{n^{1.5}}{\\sqrt{\\log ...

  6. Bivariate phase-rectified signal averaging

    CERN Document Server

    Schumann, Aicko Y; Bauer, Axel; Schmidt, Georg

    2008-01-01

    Phase-Rectified Signal Averaging (PRSA) was shown to be a powerful tool for the study of quasi-periodic oscillations and nonlinear effects in non-stationary signals. Here we present a bivariate PRSA technique for the study of the inter-relationship between two simultaneous data recordings. Its performance is compared with traditional cross-correlation analysis, which, however, does not work well for non-stationary data and cannot distinguish the coupling directions in complex nonlinear situations. We show that bivariate PRSA allows the analysis of events in one signal at times where the other signal is in a certain phase or state; it is stable in the presence of noise and impassible to non-stationarities.

  7. Hedge algorithm and Dual Averaging schemes

    CERN Document Server

    Baes, Michel

    2011-01-01

    We show that the Hedge algorithm, a method that is widely used in Machine Learning, can be interpreted as a particular instance of Dual Averaging schemes, which have recently been introduced by Nesterov for regret minimization. Based on this interpretation, we establish three alternative methods of the Hedge algorithm: one in the form of the original method, but with optimal parameters, one that requires less a priori information, and one that is better adapted to the context of the Hedge algorithm. All our modified methods have convergence results that are better or at least as good as the performance guarantees of the vanilla method. In numerical experiments, our methods significantly outperform the original scheme.

  8. Asymmetric network connectivity using weighted harmonic averages

    Science.gov (United States)

    Morrison, Greg; Mahadevan, L.

    2011-02-01

    We propose a non-metric measure of the "closeness" felt between two nodes in an undirected, weighted graph using a simple weighted harmonic average of connectivity, that is a real-valued Generalized Erdös Number (GEN). While our measure is developed with a collaborative network in mind, the approach can be of use in a variety of artificial and real-world networks. We are able to distinguish between network topologies that standard distance metrics view as identical, and use our measure to study some simple analytically tractable networks. We show how this might be used to look at asymmetry in authorship networks such as those that inspired the integer Erdös numbers in mathematical coauthorships. We also show the utility of our approach to devise a ratings scheme that we apply to the data from the NetFlix prize, and find a significant improvement using our method over a baseline.

  9. Average Gait Differential Image Based Human Recognition

    Directory of Open Access Journals (Sweden)

    Jinyan Chen

    2014-01-01

    Full Text Available The difference between adjacent frames of human walking contains useful information for human gait identification. Based on the previous idea a silhouettes difference based human gait recognition method named as average gait differential image (AGDI is proposed in this paper. The AGDI is generated by the accumulation of the silhouettes difference between adjacent frames. The advantage of this method lies in that as a feature image it can preserve both the kinetic and static information of walking. Comparing to gait energy image (GEI, AGDI is more fit to representation the variation of silhouettes during walking. Two-dimensional principal component analysis (2DPCA is used to extract features from the AGDI. Experiments on CASIA dataset show that AGDI has better identification and verification performance than GEI. Comparing to PCA, 2DPCA is a more efficient and less memory storage consumption feature extraction method in gait based recognition.

  10. Measurement system analysis (MSA) of the isotopic ratio for uranium isotope enrichment process control

    Energy Technology Data Exchange (ETDEWEB)

    Medeiros, Josue C. de; Barbosa, Rodrigo A.; Carnaval, Joao Paulo R., E-mail: josue@inb.gov.br, E-mail: rodrigobarbosa@inb.gov.br, E-mail: joaocarnaval@inb.gov.br [Industrias Nucleares do Brasil (INB), Rezende, RJ (Brazil)

    2013-07-01

    Currently, one of the stages in nuclear fuel cycle development is the process of uranium isotope enrichment, which will provide the amount of low enriched uranium for the nuclear fuel production to supply 100% Angra 1 and 20% Angra 2 demands. Determination of isotopic ration n({sup 235}U)/n({sup 238}U) in uranium hexafluoride (UF{sub 6} - used as process gas) is essential in order to control of enrichment process of isotopic separation by gaseous centrifugation cascades. The uranium hexafluoride process is performed by gas continuous feeding in separation unit which uses the centrifuge force principle, establishing a density gradient in a gas containing components of different molecular weights. The elemental separation effect occurs in a single ultracentrifuge that results in a partial separation of the feed in two fractions: an enriched on (product) and another depleted (waste) in the desired isotope ({sup 235}UF{sub 6}). Industrias Nucleares do Brasil (INB) has used quadrupole mass spectrometry (QMS) by electron impact (EI) to perform isotopic ratio n({sup 235}U)/n({sup 238}U) analysis in the process. The decision of adjustments and change te input variables are based on the results presented in these analysis. A study of stability, bias and linearity determination has been performed in order to evaluate the applied method, variations and systematic errors in the measurement system. The software used to analyze the techniques above was the Minitab 15. (author)

  11. Numerical analysis of fuel regression rate distribution characteristics in hybrid rocket motors with different fuel types

    Institute of Scientific and Technical Information of China (English)

    LI; XinTian; TIAN; Hui; CAI; GuoBiao

    2013-01-01

    This paper presents three-dimensional numerical simulations of the hybrid rocket motor with hydrogen peroxide (HP) and hy-droxyl terminated polybutadiene (HTPB) propellant combination and investigates the fuel regression rate distribution charac-teristics of different fuel types. The numerical models are established to couple the Navier-Stokes equations with turbulence,chemical reactions, solid fuel pyrolysis and solid-gas interfacial boundary conditions. Simulation results including the temper-ature contours and fuel regression rate distributions are presented for the tube, star and wagon wheel grains. The results demonstrate that the changing trends of the regression rate along the axis are similar for all kinds of fuel types, which decrease sharply near the leading edges of the fuels and then gradually increase with increasing axial locations. The regression rates of the star and wagon wheel grains show apparent three-dimensional characteristics, and they are higher in the regions of fuel surfaces near the central core oxidizer flow. The average regression rates increase as the oxidizer mass fluxes rise for all of the fuel types. However, under same oxidizer mass flux, the average regression rates of the star and wagon wheel grains are much larger than that of the tube grain due to their lower hydraulic diameters.

  12. Radiative capture on $^{242}$Pu for MOX fuel reactors

    CERN Multimedia

    The use of MOX fuel (mixed-oxide fuel made of UO$_{2}$ and PuO$_{2}$) in nuclear reactors allows substituting a large fraction of the enriched Uranium by Plutonium reprocessed from spent fuel. Indeed around 66% of the plutonium from spent fuel is made of $^{239}$Pu and $^{241}$Pu, which are fissile in thermal reactors. A typical reactor of this type uses a fuel with 7% reprocessed Pu and 93% depleted U, thus profiting from both the spent fuel and the remaining $^{238}$U following the $^{235}$U enrichment. With the use of such new fuel compositions rich in Pu the better knowledge of the capture and fission cross sections of the Pu isotopes becomes very important. This is clearly stated in the recent OECD NEA’s “High Priority Request List” and in the WPEC-26 “Uncertainty and target accuracy assessment for innovative systems using recent covariance data evaluations” report. In particular, a new series of cross section evaluations have been recently carried out jointly by the European (JEFF) and United ...

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

    Science.gov (United States)

    Coble, Jamie; Orton, Christopher; Schwantes, Jon

    2017-04-01

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

  14. LISSAT Analysis of a Generic Centrifuge Enrichment Plant

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-05-31

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

  15. Fuel composition optimization in a 78-element fuel bundle for use in a pressure tube type supercritical water-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hummel, D.W.; Novog, D.R. [McMaster Univ., Hamilton, Ontario (Canada)

    2012-07-01

    A 78-element fuel bundle containing a plutonium-thorium fuel mixture has been proposed for a Generation IV pressure tube type supercritical water-cooled reactor. In this work, using a lattice cell model created with the code DRAGON,the lattice pitch, fuel composition (fraction of PuO{sub 2} in ThO{sub 2}) and radial enrichment profile of the 78-element bundle is optimized using a merit function and a metaheuristic search algorithm.The merit function is designed such that the optimal fuel maximizes fuel utilization while minimizing peak element ratings and coolant void reactivity. A radial enrichment profile of 10 wt%, 11 wt% and 20 wt% PuO{sub 2} (inner to outer ring) with a lattice pitch of 25.0 cm was found to provide the optimal merit score based on the aforementioned criteria. (author)

  16. Solar enriched methane production: Assessment of plant potentialities and applications

    Directory of Open Access Journals (Sweden)

    Vincenzo Piemonte

    2010-04-01

    Full Text Available The crucial environmental issue due to fossil fuel use in our society and industries and more and more perceived by the communities is stimulating the development of innovative technologies with the scope of reducing GHGs and pollutants emissions, improving plants efficiency and exploiting renewable energy sources. The idea proposed in the present work links this context: a novel hybrid plant for the production of a mixture of methane and hydrogen (20%vol, called enriched-methane, from a steam reforming reactor whose heat duty is supplied by a concentrating solar power (CSP plant by means of a molten salt stream is here conceived, modelled and assessed. The enriched-methane mixture can be applied in methane internal combustion engines (ICE reducing CO, CO2, unburned emissions and improving engine efficiency. Moreover, the residual sensible heat of solar-heated molten salt stream can be used to generate medium-pressure steam and to produce electricity by a steam-turbine. Therefore, the plant proposed is co-generative, producing both hydrogen and electricity from a solar source. The behaviour of methane steam reforming reactor is simulated by means of a 2D mathematical model and the design of a cogenerative solar plant is proposed, evaluating its potentialities in terms of MWh of electricity produced and number of vehicles fed by enriched-methane. A single CSP module (surface requirement = 1.5 hectares coupled with a 4-tubes-and-shell shaped reactor is able to produce 686 tons/year of hydrogen, equivalent to 3.430 tons/year of 20%vol H2-CH4 mixture and 3.097 MWh/year of clean electricity.

  17. Boiling water reactors with uranium-plutonium mixed oxide fuel. Report 5: Analysis of the reactivity coefficients and the stability of a BWR loaded with MOx fuel

    Energy Technology Data Exchange (ETDEWEB)

    Demaziere, C. [CEA Centre d' Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Direction des Reacteurs Nucleaires

    2000-01-01

    This report is a part of the project titled 'Boiling Water Reactors With Uranium-Plutonium Mixed Oxide (MOx) Fuel'. The aim of this study is to model the impact of a core loading pattern containing MOx bundles upon the main characteristics of a BWR (reactivity coefficients, stability, etc.). For this purpose, the Core Management System (CMS) codes of Studsvik Scandpower are used. This package is constituted by CASMO-4/TABLES-3/SIMULATE-3. It has been shown in previous reports that these codes are able to accurately represent and model MOx bundles. This report is thus devoted to the study of BWR cores loaded (partially or totally) with MOx bundles. The plutonium quality used is the Pu type 2016 (mostly Pu-239, 56 %, and Pu-240, 26 %), but a variation of the plutonium isotopic vector was also investigated, in case of a partial MOx loading. One notices that the reactivity coefficients do not present significant changes in comparison with a full UOx loading. Nevertheless, two main problems arise: the shutdown margin at BOC is lower than 1 % and the stability to in-phase oscillations is slightly decreased. (The SIMULATE-3 version used for this study does not contain the latest MOx enhancements described in literature, since these code developments have not been provided to the department. Nevertheless, as the nominal average enrichment of the MOx bundles is 5.41 % (total amount of plutonium), which can still be considered as a relatively low enrichment, the accuracy of the CMS codes is acceptable without the use of the MOx improvements for this level of Pu enrichment.

  18. Uranium enrichment reduction in the Prototype Gen-IV sodium-cooled fast reactor (PGSFR) with PBO reflector

    Energy Technology Data Exchange (ETDEWEB)

    Hartanto, Donny; Kim, Chi Hyung; Kim, Yong Hee [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology (KAIST), Daejeon (Korea, Republic of)

    2016-04-15

    The Korean Prototype Gen-IV sodium-cooled fast reactor (PGSFR) is supposed to be loaded with a relatively-costly low-enriched U fuel, while its envisaged transuranic fuels are not available for transmutation. In this work, the U-enrichment reduction by improving the neutron economy is pursued to save the fuel cost. To improve the neutron economy of the core, a new reflector material, PbO, has been introduced to replace the conventional HT9 reflector in the current PGSFR core. Two types of PbO reflectors are considered: one is the conventional pin-type and the other one is an inverted configuration. The inverted PbO reflector design is intended to maximize the PbO volume fraction in the reflector assembly. In addition, the core radial configuration is also modified to maximize the performance of the PbO reflector. For the baseline PGSFR core with several reflector options, the U enrichment requirement has been analyzed and the fuel depletion analysis is performed to derive the equilibrium cycle parameters. The linear reactivity model is used to determine the equilibrium cycle performances of the core. Impacts of the new PbO reflectors are characterized in terms of the cycle length, neutron leakage, radial power distribution, and operational fuel cost.

  19. Uranium Enrichment Reduction in the Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR with PBO Reflector

    Directory of Open Access Journals (Sweden)

    Chihyung Kim

    2016-04-01

    Full Text Available The Korean Prototype Gen-IV sodium-cooled fast reactor (PGSFR is supposed to be loaded with a relatively-costly low-enriched U fuel, while its envisaged transuranic fuels are not available for transmutation. In this work, the U-enrichment reduction by improving the neutron economy is pursued to save the fuel cost. To improve the neutron economy of the core, a new reflector material, PbO, has been introduced to replace the conventional HT9 reflector in the current PGSFR core. Two types of PbO reflectors are considered: one is the conventional pin-type and the other one is an inverted configuration. The inverted PbO reflector design is intended to maximize the PbO volume fraction in the reflector assembly. In addition, the core radial configuration is also modified to maximize the performance of the PbO reflector. For the baseline PGSFR core with several reflector options, the U enrichment requirement has been analyzed and the fuel depletion analysis is performed to derive the equilibrium cycle parameters. The linear reactivity model is used to determine the equilibrium cycle performances of the core. Impacts of the new PbO reflectors are characterized in terms of the cycle length, neutron leakage, radial power distribution, and operational fuel cost.

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

    Science.gov (United States)

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

    2016-10-25

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

  1. Spent nuclear fuel discharges from US reactors 1993

    Energy Technology Data Exchange (ETDEWEB)

    1995-02-01

    The Energy Information Administration (EIA) of the U.S. Department of Energy (DOE) administers the Nuclear Fuel Data Survey, Form RW-859. This form is used to collect data on fuel assemblies irradiated at commercial nuclear reactors operating in the United States, and the current inventories and storage capacities of those reactors. These data are important to the design and operation of the equipment and facilities that DOE will use for the future acceptance, transportation, and disposal of spent fuels. The data collected and presented identifies trends in burnup, enrichment, and spent nuclear fuel discharged form commercial light-water reactor as of December 31, 1993. The document covers not only spent nuclear fuel discharges; but also site capacities and inventories; canisters and nonfuel components; and assembly type characteristics.

  2. Direct reuse of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Mohamed, Nader M.A., E-mail: mnader73@yahoo.com

    2014-10-15

    Highlights: • A new design for the PWR assemblies for direct use of spent fuel was proposed. • The PWR spent fuel will be transferred directly (after a certain cooling time) to CANDU reactors. • The proposed assembly has four zircaloy-4 tubes contains a number of CANDU fuel bundles (7 or 8 bundles per tube) stacked end to end. • MCNPX is used for the calculations that showed that the burnup can be increased by about 25%. • Acceptable linear heat generation rate in hot rods and improved Pu proliferation resistance. - Abstract: In this paper we proposed a new design for the PWR fuel assembly for direct use of the PWR spent fuel without processing. The PWR spent fuel will be transferred directly (after a certain cooling time) to CANDU reactors which preferably built in the same site to avoid the problem of transportations. The proposed assembly has four zircaloy-4 tubes contains a number of CANDU fuel bundles (7 or 8 bundles per tube) stacked end to end. Each tube has the same inner diameter of that of CANDU pressure tube. The spaces between the tubes contain low enriched UO{sub 2} fuel rods and guide tubes. MCNPX code is used for the simulation and calculation of the burnup of the proposed assembly. The bundles after the discharge from the PWR with their materials inventories are burned in a CANDU cell after a certain decay time. The results were compared with reference results and the impact of this new design on the uranium utilization improvement and on the proliferation resistance of plutonium is discussed. The effect of this new design on the power peaking, moderator temperature coefficient of reactivity and CANDU coolant void reactivity are discussed as well.

  3. A new approach for Bayesian model averaging

    Institute of Scientific and Technical Information of China (English)

    TIAN XiangJun; XIE ZhengHui; WANG AiHui; YANG XiaoChun

    2012-01-01

    Bayesian model averaging (BMA) is a recently proposed statistical method for calibrating forecast ensembles from numerical weather models.However,successful implementation of BMA requires accurate estimates of the weights and variances of the individual competing models in the ensemble.Two methods,namely the Expectation-Maximization (EM) and the Markov Chain Monte Carlo (MCMC) algorithms,are widely used for BMA model training.Both methods have their own respective strengths and weaknesses.In this paper,we first modify the BMA log-likelihood function with the aim of removing the additional limitation that requires that the BMA weights add to one,and then use a limited memory quasi-Newtonian algorithm for solving the nonlinear optimization problem,thereby formulating a new approach for BMA (referred to as BMA-BFGS).Several groups of multi-model soil moisture simulation experiments from three land surface models show that the performance of BMA-BFGS is similar to the MCMC method in terms of simulation accuracy,and that both are superior to the EM algorithm.On the other hand,the computational cost of the BMA-BFGS algorithm is substantially less than for MCMC and is almost equivalent to that for EM.

  4. Calculating Free Energies Using Average Force

    Science.gov (United States)

    Darve, Eric; Pohorille, Andrew; DeVincenzi, Donald L. (Technical Monitor)

    2001-01-01

    A new, general formula that connects the derivatives of the free energy along the selected, generalized coordinates of the system with the instantaneous force acting on these coordinates is derived. The instantaneous force is defined as the force acting on the coordinate of interest so that when it is subtracted from the equations of motion the acceleration along this coordinate is zero. The formula applies to simulations in which the selected coordinates are either unconstrained or constrained to fixed values. It is shown that in the latter case the formula reduces to the expression previously derived by den Otter and Briels. If simulations are carried out without constraining the coordinates of interest, the formula leads to a new method for calculating the free energy changes along these coordinates. This method is tested in two examples - rotation around the C-C bond of 1,2-dichloroethane immersed in water and transfer of fluoromethane across the water-hexane interface. The calculated free energies are compared with those obtained by two commonly used methods. One of them relies on determining the probability density function of finding the system at different values of the selected coordinate and the other requires calculating the average force at discrete locations along this coordinate in a series of constrained simulations. The free energies calculated by these three methods are in excellent agreement. The relative advantages of each method are discussed.

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

    Science.gov (United States)

    Rouf; Su'ud, Zaki

    2016-08-01

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

  6. ATR PDQ and MCWO Fuel Burnup Analysis Codes Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    G.S. Chang; P. A. Roth; M. A. Lillo

    2009-11-01

    The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) is being studied to determine the feasibility of converting it from the highly enriched Uranium (HEU) fuel that is currently uses to low enriched Uranium (LEU) fuel. In order to achieve this goal, it would be best to qualify some different computational methods than those that have been used at ATR for the past 40 years. This paper discusses two methods of calculating the burnup of ATR fuel elements. The existing method, that uses the PDQ code, is compared to a modern method that uses A General Monte Carlo N-Particle Transport Code (MCNP) combined with the Origen2.2 code. This modern method, MCNP with ORIGEN2.2 (MCWO), is found to give excellent agreement with the existing method (PDQ). Both of MCWO and PDQ are also in a very good agreement to the 235U burnup data generated by an analytical method.

  7. Fuel Cycle Performance of Thermal Spectrum Small Modular Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Worrall, Andrew [ORNL; Todosow, Michael [Brookhaven National Laboratory (BNL)

    2016-01-01

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

  8. The whole-core LEU silicide fuel demonstration in the JMTR

    Energy Technology Data Exchange (ETDEWEB)

    Aso, Tomokazu; Akashi, Kazutomo; Nagao, Yoshiharu [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)] [and others

    1997-08-01

    The JMTR was fully converted to LEU silicide (U{sub 3}Si{sub 2}) fuel with cadmium wires as burnable absorber in January, 1994. The reduced enrichment program for the JMTR was initiated in 1979, and the conversion to MEU (enrichment ; 45%) aluminide fuel was carried out in 1986 as the first step of the program. The final goal of the program was terminated by the present LEU conversion. This paper describes the results of core physics measurement through the conversion phase from MEU fuel core to LEU fuel core. Measured excess reactivities of the LEU fuel cores are mostly in good agreement with predicted values. Reactivity effect and burnup of cadmium wires, therefore, were proved to be well predicted. Control rod worth in the LEU fuel core is mostly less than that in the MEU fuel core. Shutdown margin was verified to be within the safety limit. There is no significant difference in temperature coefficient of reactivity between the MEU and LEU fuel cores. These results verified that the JMTR was successfully and safely converted to LEU fuel. Extension of the operating cycle period was achieved and reduction of spend fuel elements is expected by using the fuel with high uranium density.

  9. Landing on empty: estimating the benefits from reducing fuel uplift in US Civil Aviation

    Science.gov (United States)

    Ryerson, Megan S.; Hansen, Mark; Hao, Lu; Seelhorst, Michael

    2015-09-01

    Airlines and Air Navigation Service Providers are united in their goal to reduce fuel consumption. While changes to flight operations and technology investments are the focus of a number of studies, our study is among the first to investigate an untapped source of aviation fuel consumption: excess contingency fuel loading. Given the downside risk of fuel exhaustion of diverting to an alternate airport, airline dispatchers may load excess fuel onto an aircraft. Such conservatism comes at a cost of consuming excess fuel, as fuel consumed is a function of, among other factors, aircraft weight. The aim of this paper is to quantify, on a per-flight basis, the fuel burned due to carrying fuel beyond what is needed for foreseeable contingencies, and thereby motivate research, federal guidance, and investments that allow airline dispatchers to reduce fuel uplift while maintaining near zero risks of fuel exhaustion. We merge large publicly available aviation and weather databases with a detailed dataset from a major US airline. Upon estimating factors that capture the quantity fuel consumed due to carrying a pound of weight for a range of aircraft types, we calculate the cost and greenhouse gas emissions from carrying unused fuel on arrival and additional contingency fuel above a conservative buffer for foreseeable contingencies. We establish that the major US carrier does indeed load fuel conservatively. We find that 4.48% of the fuel consumed by an average flight is due to carrying unused fuel and 1.04% of the fuel consumed by an average flight is due to carrying additional contingency fuel above a reasonable buffer. We find that simple changes in flight dispatching that maintain a statistically minimal risk of fuel exhaustion could result in yearly savings of 338 million lbs of CO2, the equivalent to the fuel consumed from 4760 flights on midsized commercial aircraft. Moreover, policy changes regarding maximum fuel loads or investments that reduce uncertainty or increase

  10. Quantification of histone modification ChIP-seq enrichment for data mining and machine learning applications

    Directory of Open Access Journals (Sweden)

    Bekiranov Stefan

    2011-08-01

    Full Text Available Abstract Background The advent of ChIP-seq technology has made the investigation of epigenetic regulatory networks a computationally tractable problem. Several groups have applied statistical computing methods to ChIP-seq datasets to gain insight into the epigenetic regulation of transcription. However, methods for estimating enrichment levels in ChIP-seq data for these computational studies are understudied and variable. Since the conclusions drawn from these data mining and machine learning applications strongly depend on the enrichment level inputs, a comparison of estimation methods with respect to the performance of statistical models should be made. Results Various methods were used to estimate the gene-wise ChIP-seq enrichment levels for 20 histone methylations and the histone variant H2A.Z. The Multivariate Adaptive Regression Splines (MARS algorithm was applied for each estimation method using the estimation of enrichment levels as predictors and gene expression levels as responses. The methods used to estimate enrichment levels included tag counting and model-based methods that were applied to whole genes and specific gene regions. These methods were also applied to various sizes of estimation windows. The MARS model performance was assessed with the Generalized Cross-Validation Score (GCV. We determined that model-based methods of enrichment estimation that spatially weight enrichment based on average patterns provided an improvement over tag counting methods. Also, methods that included information across the entire gene body provided improvement over methods that focus on a specific sub-region of the gene (e.g., the 5' or 3' region. Conclusion The performance of data mining and machine learning methods when applied to histone modification ChIP-seq data can be improved by using data across the entire gene body, and incorporating the spatial distribution of enrichment. Refinement of enrichment estimation ultimately improved accuracy

  11. Alternative Fuels (Briefing Charts)

    Science.gov (United States)

    2009-06-19

    SASOL Jet Fuel Qatar GTL Syntroleum Jet fuel in B-52 Nigeria GTL China Coal GTL Shell Bintulu GTL Cellulose ethanol for ground use Ocean Bio...Bintulu GTL Ocean Bio-fuel Factories Bio-butanol for ground use Future Energy Source Resurgence in Nuclear Power Bio-jet tests done ASTM Spec...High energy deoxygenated bio-jet fuel from algae PSU coal derived JP-8 B-52 emissions Scoping study HBR TF emissions New bio- fuel impacts Adv

  12. Analysis of fuel management in the KIPT neutron source facility

    Energy Technology Data Exchange (ETDEWEB)

    Zhong Zhaopeng, E-mail: zzhong@anl.gov [Nuclear Engineering Division, Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Gohar, Yousry; Talamo, Alberto [Nuclear Engineering Division, Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

    2011-05-15

    Research highlights: > Fuel management of KIPT ADS was analyzed. > Core arrangement was shuffled in stage wise. > New fuel assemblies was added into core periodically. > Beryllium reflector could also be utilized to increase the fuel life. - Abstract: Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the conceptual design development of an experimental neutron source facility consisting of an electron accelerator driven sub-critical assembly. The neutron source driving the sub-critical assembly is generated from the interaction of 100 KW electron beam with a natural uranium target. The sub-critical assembly surrounding the target is fueled with low enriched WWR-M2 type hexagonal fuel assemblies. The U-235 enrichment of the fuel material is <20%. The facility will be utilized for basic and applied research, producing medical isotopes, and training young specialists. With the 100 KW electron beam power, the total thermal power of the facility is {approx}360 kW including the fission power of {approx}260 kW. The burnup of the fissile materials and the buildup of fission products continuously reduce the system reactivity during the operation, decrease the neutron flux level, and consequently impact the facility performance. To preserve the neutron flux level during the operation, the fuel assemblies should be added and shuffled for compensating the lost reactivity caused by burnup. Beryllium reflector could also be utilized to increase the fuel life time in the sub-critical core. This paper studies the fuel cycles and shuffling schemes of the fuel assemblies of the sub-critical assembly to preserve the system reactivity and the neutron flux level during the operation.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-05-12

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

  14. Vehicle Fuel-Efficiency Choices, Emission Externalities, and Urban Sprawl

    DEFF Research Database (Denmark)

    Kim, Jinwon

    of excessive sprawl arising from emission externalities is the uses of larger and less-fuel efficient vehicles by suburban residents, which is different from that of congestion externalities. We also analyze the effect of the Corporate Average Fuel Efficiency (CAFE) regulation on the urban spatial structure....

  15. Algorithm of axial fuel optimization based in progressive steps of turned search; Algoritmo de optimizacion axial de combustible basado en etapas progresivas de busqueda de entorno

    Energy Technology Data Exchange (ETDEWEB)

    Martin del Campo, C.; Francois, J.L. [Laboratorio de Analisis en Ingenieria de Reactores Nucleares, FI-UNAM, Paseo Cuauhnahuac 8532, Jiutepec, Morelos (Mexico)

    2003-07-01

    The development of an algorithm for the axial optimization of fuel of boiling water reactors (BWR) is presented. The algorithm is based in a serial optimizations process in the one that the best solution in each stage is the starting point of the following stage. The objective function of each stage adapts to orient the search toward better values of one or two parameters leaving the rest like restrictions. Conform to it advances in those optimization stages, it is increased the fineness of the evaluation of the investigated designs. The algorithm is based on three stages, in the first one are used Genetic algorithms and in the two following Tabu Search. The objective function of the first stage it looks for to minimize the average enrichment of the one it assembles and to fulfill with the generation of specified energy for the operation cycle besides not violating none of the limits of the design base. In the following stages the objective function looks for to minimize the power factor peak (PPF) and to maximize the margin of shutdown (SDM), having as restrictions the one average enrichment obtained for the best design in the first stage and those other restrictions. The third stage, very similar to the previous one, it begins with the design of the previous stage but it carries out a search of the margin of shutdown to different exhibition steps with calculations in three dimensions (3D). An application to the case of the design of the fresh assemble for the fourth fuel reload of the Unit 1 reactor of the Laguna Verde power plant (U1-CLV) is presented. The obtained results show an advance in the handling of optimization methods and in the construction of the objective functions that should be used for the different design stages of the fuel assemblies. (Author)

  16. 77 FR 13367 - General Electric-Hitachi Global Laser Enrichment, LLC, Proposed Laser-Based Uranium Enrichment...

    Science.gov (United States)

    2012-03-06

    ... COMMISSION General Electric-Hitachi Global Laser Enrichment, LLC, Proposed Laser-Based Uranium Enrichment...- Hitachi Global Laser Enrichment, LLC (GLE) Uranium Enrichment Facility. On June 26, 2009, GLE submitted a... uranium enrichment facility (the ``proposed action''). The GLE proposes to locate the facility on...

  17. Management's Ecstasy and Disparity Over Job Enrichment

    Science.gov (United States)

    King, Albert S.

    1976-01-01

    A case study analyzing job enrichment schemes and manager expectations of increased productivity is presented. It was found that it was the managers' expectations of increased productivity, not the reorganization of work, that led to higher productivity. (EC)

  18. Moving toward multilateral mechanisms for the fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-04-17

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

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

    Science.gov (United States)

    McCoy, Kevin; Mays, Claude

    2008-04-01

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

  20. Production of LEU Fully Ceramic Microencapsulated Fuel for Irradiation Testing

    Energy Technology Data Exchange (ETDEWEB)

    Terrani, Kurt A [ORNL; Kiggans Jr, James O [ORNL; McMurray, Jake W [ORNL; Jolly, Brian C [ORNL; Hunt, Rodney Dale [ORNL; Trammell, Michael P [ORNL; Snead, Lance Lewis [ORNL

    2016-01-01

    Fully Ceramic Microencapsulated (FCM) fuel consists of tristructural isotropic (TRISO) fuel particles embedded inside a SiC matrix. This fuel inherently possesses multiple barriers to fission product release, namely the various coating layers in the TRISO fuel particle as well as the dense SiC matrix that hosts these particles. This coupled with the excellent oxidation resistance of the SiC matrix and the SiC coating layer in the TRISO particle designate this concept as an accident tolerant fuel (ATF). The FCM fuel takes advantage of uranium nitride kernels instead of oxide or oxide-carbide kernels used in high temperature gas reactors to enhance heavy metal loading in the highly moderated LWRs. Production of these kernels with appropriate density, coating layer development to produce UN TRISO particles, and consolidation of these particles inside a SiC matrix have been codified thanks to significant R&D supported by US DOE Fuel Cycle R&D program. Also, surrogate FCM pellets (pellets with zirconia instead of uranium-bearing kernels) have been neutron irradiated and the stability of the matrix and coating layer under LWR irradiation conditions have been established. Currently the focus is on production of LEU (7.3% U-235 enrichment) FCM pellets to be utilized for irradiation testing. The irradiation is planned at INL s Advanced Test Reactor (ATR). This is a critical step in development of this fuel concept to establish the ability of this fuel to retain fission products under prototypical irradiation conditions.