WorldWideScience

Sample records for low-enriched uranium-molybdenum fuel

  1. 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

  2. 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

  3. 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

  4. FY16 Status Report for the Uranium-Molybdenum Fuel Concept

    Energy Technology Data Exchange (ETDEWEB)

    Bennett, Wendy D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Doherty, Ann L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Henager, Charles H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lavender, Curt A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Montgomery, Robert O. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Omberg, Ronald P. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Smith, Mark T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Webster, Ryan A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-09-22

    The Fuel Cycle Research and Development program of the Office of Nuclear Energy has implemented a program to develop a Uranium-Molybdenum metal fuel for light water reactors. Uranium-Molybdenum fuel has the potential to provide superior performance based on its thermo-physical properties. With sufficient development, it may be able to provide the Light Water Reactor industry with a melt-resistant, accident-tolerant fuel with improved safety response. The Pacific Northwest National Laboratory has been tasked with extrusion development and performing ex-reactor corrosion testing to characterize the performance of Uranium-Molybdenum fuel in both these areas. This report documents the results of the fiscal year 2016 effort to develop the Uranium-Molybdenum metal fuel concept for light water reactors.

  5. Low-temperature irradiation behavior of uranium-molybdenum alloy dispersion fuel

    Science.gov (United States)

    Meyer, M. K.; Hofman, G. L.; Hayes, S. L.; Clark, C. R.; Wiencek, T. C.; Snelgrove, J. L.; Strain, R. V.; Kim, K.-H.

    2002-08-01

    Irradiation tests have been conducted to evaluate the performance of a series of high-density uranium-molybdenum (U-Mo) alloy, aluminum matrix dispersion fuels. Fuel plates incorporating alloys with molybdenum content in the range of 4-10 wt% were tested. Two irradiation test vehicles were used to irradiate low-enrichment fuels to approximately 40 and 70 at.% 235U burnup in the advanced test reactor at fuel temperatures of approximately 65 °C. The fuel particles used to fabricate dispersion specimens for most of the test were produced by generating filings from a cast rod. In general, fuels with molybdenum contents of 6 wt% or more showed stable in-reactor fission gas behavior, exhibiting a distribution of small, stable gas bubbles. Fuel particle swelling was moderate and decreased with increasing alloy content. Fuel particles with a molybdenum content of 4 wt% performed poorly, exhibiting extensive fuel-matrix interaction and the growth of relatively large fission gas bubbles. Fuel particles with 4 or 6 wt% molybdenum reacted more rapidly with the aluminum matrix than those with higher-alloy content. Fuel particles produced by an atomization process were also included in the test to determine the effect of fuel particle morphology and microstructure on fuel performance for the U-10Mo composition. Both of the U-10Mo fuel particle types exhibited good irradiation performance, but showed visible differences in fission gas bubble nucleation and growth behavior.

  6. 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.

  7. 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.

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

    Science.gov (United States)

    Van Kleeck, Melissa A.

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

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. Effects of heat treatments on the thermal diffusivity of Uranium-Molybdenum alloy

    Science.gov (United States)

    Camarano, D. M.; Mansur, F. A.; Santos, A. M. M.; Ferraz, W. B.; Pedrosa, T. A.

    2016-07-01

    U-Mo alloys are the most investigated nuclear fuel material to be used in research reactors. The addition of molybdenum stabilizes the gamma phase of uranium and increases its melting point. A research program under development at Nuclear Technology Development Center (CDTN) aims the obtaining of uranium-molybdenum alloys to enable the high enriched uranium (HEU) to low enriched uranium (LEU) conversions. U-Mo ingots with 10% by weight were induction melted and heat treated at 300 °C for 72 h, 120 h and 240 h. Thermal diffusivity was determined by the laser flash method and thermal quadrupole method, from room temperature to 300 oC and 400oC. It was observed that the thermal diffusivity tends to increase with increasing temperature.

  14. Development of a high density fuel based on uranium-molybdenum alloys with high compatibility in high temperatures; Desenvolvimento de um combustivel de alta densidade a base das ligas uranio-molibdenio com alta compatibilidade em altas temperaturas

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Fabio Branco Vaz de

    2008-07-01

    This work has as its objective the development of a high density and low enriched nuclear fuel based on the gamma-UMo alloys, for utilization where it is necessary satisfactory behavior in high temperatures, considering its utilization as dispersion. For its accomplishment, it was started from the analysis of the RERTR ('Reduced Enrichment for Research and Test Reactors') results and some theoretical works involving the fabrication of gamma-uranium metastable alloys. A ternary addition is proposed, supported by the properties of binary and ternary uranium alloys studied, having the objectives of the gamma stability enhancement and an ease to its powder fabrication. Alloys of uranium-molybdenum were prepared with 5 to 10% Mo addition, and 1 and 3% of ternary, over a gamma U7Mo binary base alloy. In all the steps of its preparation, the alloys were characterized with the traditional techniques, to the determination of its mechanical and structural properties. To provide a process for the alloys powder obtention, its behavior under hydrogen atmosphere were studied, in thermo analyser-thermo gravimeter equipment. Temperatures varied from the ambient up to 1000 deg C, and times from 15 minutes to 16 hours. The results validation were made in a semi-pilot scale, where 10 to 50 g of powders of some of the alloys studied were prepared, under static hydrogen atmosphere. Compatibility studies were conducted by the exposure of the alloys under oxygen and aluminum, to the verification of possible reactions by means of differential thermal analysis. The alloys were exposed to a constant heat up to 1000 deg C, and their performances were evaluated in terms of their reaction resistance. On the basis of the results, it was observed that ternary additions increases the temperatures of the reaction with aluminum and oxidation, in comparison with the gamma UMo binaries. A set of conditions to the hydration of the alloys were defined, more restrictive in terms of temperature

  15. Irradiation performance of uranium-molybdenum alloy dispersion fuels; Desempenho sob irradiacao de elementos combustiveis do tipo U-Mo

    Energy Technology Data Exchange (ETDEWEB)

    Almeida, Cirila Tacconi de

    2005-07-01

    The U-Mo-Al dispersion fuels of Material Test Reactors (MTR) are analyzed in terms of their irradiation performance. The irradiation performance aspects are associated to the neutronic and thermal hydraulics aspects to propose a new core configuration to the IEA-R1 reactor of IPEN-CNEN/SP using U-Mo-Al fuels. Core configurations using U-10Mo-Al fuels with uranium densities variable from 3 to 8 gU/cm{sup 3} were analyzed with the computational programs Citation and MTRCR-IEA R1. Core configurations for fuels with uranium densities variable from 3 to 5 gU/cm{sup 3} showed to be adequate to use in IEA-R1 reactor e should present a stable in reactor performance even at high burn-up. (author)

  16. Mechanical properties examined by nanoindentation for selected phases relevant to the development of monolithic uranium-molybdenum metallic fuels

    Science.gov (United States)

    Newell, Ryan; Park, Youngjoo; Mehta, Abhishek; Keiser, Dennis; Sohn, Yongho

    2017-04-01

    Nanomechanical properties, specifically the reduced modulus and hardness of several intermetallic and solid solution phases are reported to assist the development of the U-10 wt% Mo (U-10Mo) monolithic fuel system for research and test reactors. Findings from this study and reported values of mechanical properties provide data critical for understanding and predicting the structural behavior of the fuel system during fabrication and irradiation. The phases examined are products of interdiffusion and reaction between (1) the AA6061 cladding and the Zr diffusion barrier, namely (Al,Si)3Zr and Al3Zr, (2) the U-10Mo fuel and the Zr diffusion barrier, namely UZr2, Mo2Zr, and α-U, and (3) the U (or U-10Mo) and Mo, namely a mixture gradient of α- and γ-phases. The UC inclusions observed within the fuel alloy were also examined. Only phases present in thick or continuous microstructure on cross-sectioned fuel plates and diffusion couples were investigated for reduced modulus and hardness. Concentration-dependence of room-temperature reduced modulus in U solid solution with 0-10 wt% Mo was semi-quantitatively modeled based on mixture of α- and γ-phases and solid solutioning within the γ-phase.

  17. Uranium-molybdenum nuclear fuel plates behaviour under heavy ion irradiation: An X-ray diffraction analysis

    Science.gov (United States)

    Palancher, H.; Wieschalla, N.; Martin, P.; Tucoulou, R.; Sabathier, C.; Petry, W.; Berar, J.-F.; Valot, C.; Dubois, S.

    2009-03-01

    Heavy ion irradiation has been proposed for discriminating UMo/Al specimens which are good candidates for research reactor fuels. Two UMo/Al dispersed fuels (U-7 wt%Mo/Al and U-10 wt%Mo/Al) have been irradiated with a 80 MeV 127I beam up to an ion fluence of 2 × 1017 cm-2. Microscopy and mainly X-ray diffraction using large and micrometer sized beams have enabled to characterize the grown interaction layer: UAl3 appears to be the only produced crystallized phase. The presence of an amorphous additional phase can however not be excluded. These results are in good agreement with characterizations performed on in-pile irradiated fuels and encourage new studies with heavy ion irradiation.

  18. 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.

  19. 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)

  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. 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.

  2. 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)

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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

  8. 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

  9. 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.

  10. 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

  11. Procedure for Uranium-Molybdenum Density Measurements and Porosity Determination

    Energy Technology Data Exchange (ETDEWEB)

    Prabhakaran, Ramprashad [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Devaraj, Arun [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Joshi, Vineet V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lavender, Curt A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-08-13

    The purpose of this document is to provide guidelines for preparing uranium-molybdenum (U-Mo) specimens, performing density measurements, and computing sample porosity. Typical specimens (solids) will be sheared to small rectangular foils, disks, or pieces of metal. A mass balance, solid density determination kit, and a liquid of known density will be used to determine the density of U-Mo specimens using the Archimedes principle. A standard test weight of known density would be used to verify proper operation of the system. By measuring the density of a U-Mo sample, it is possible to determine its porosity.

  12. 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.

  13. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Daniel, G. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Rudisill, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Almond, P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); O' Rourke, P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-06-28

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

  15. 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

  16. Development and Validation of Capabilities to Measure Thermal Properties of Layered Monolithic U-Mo Alloy Plate-Type Fuel

    Science.gov (United States)

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

    2014-07-01

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

  17. 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

  18. 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

  19. Aqueous corrosion behavior of uranium-molybdenum alloys

    Science.gov (United States)

    Gardner, Levi D.

    Nuclear fuel characterization requires understanding of the various conditions to which materials are exposed in-reactor. One of these important conditions is corrosion, particularly that of fuel constituents. Therefore, corrosion behavior is of special interest and an essential part of nuclear materials characterization efforts. In support of the Office of Material Management and Minimization's Reactor Conversion Program, monolithic uranium-10 wt% molybdenum alloy (U-Mo) is being investigated as a low enriched uranium alternative to highly enriched uranium dispersion fuel currently used in domestic high performance research reactors. The aqueous corrosion behavior of U-Mo is being examined at Pacific Northwest National Laboratory (PNNL) as part of U-Mo fuel fabrication capability activity. No prior study adequately represents this behavior given the current state of alloy composition and thermomechanical processing methods, and research reactor water chemistry. Two main measurement techniques were employed to evaluate U-Mo corrosion behavior. Low-temperature corrosion rate values were determined by means of U-Mo immersion testing and subsequent mass-loss measurements. The electrochemical behavior of each processing condition was also qualitatively examined using the techniques of corrosion potential and anodic potentiodynamic polarization. Scanning electron microscopy (SEM) and optical metallography (OM) imagery and hardness measurements provided supplemental corrosion analysis in an effort to relate material corrosion behavior to processing. The processing effects investigated as part of this were those of homogenization heat treatment (employed to mitigate the effects of coring in castings) and sub-eutectoid heat treatment, meant to represent additional steps in fabrication (such as hot isostatic pressing) performed at similar temperatures. Immersion mass loss measurements and electrochemical results both showed very little appreciable difference between

  20. 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.

  1. 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.

  2. Fuel Thermo-physical Characterization Project: Evaluation of Models to Calculate Thermal Diffusivity of Layered Composites

    Energy Technology Data Exchange (ETDEWEB)

    Burkes, Douglas [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Casella, Amanda J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Gardner, Levi D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Casella, Andrew M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Huber, Tanja K. [Technische Universität München, Munich (Germany); Breitkreutz, Harald [Technische Universität München, Munich (Germany)

    2015-02-11

    The Office of Material Management and Minimization Fuel Thermo-physical Characterization Project at Pacific Northwest National Laboratory (PNNL) is tasked with using PNNL facilities and processes to receive irradiated low enriched uranium-molybdenum fuel plate samples and perform analyses in support of the Office of Material Management and Minimization Reactor Conversion Program. This work is in support of the Fuel Development Pillar that is managed by Idaho National Laboratory. A key portion of the scope associated with this project was to measure the thermal properties of fuel segments harvested from plates that were irradiated in the Advanced Test Reactor. Thermal diffusivity of samples prepared from the fuel segments was measured using laser flash analysis. Two models, one developed by PNNL and the other developed by the Technische Universität München (TUM), were evaluated to extract the thermal diffusivity of the uranium-molybdenum alloy from measurements made on the irradiated, layered composites. The experimental data of the “TC” irradiated fuel segment was evaluated using both models considering a three-layer and five-layer system. Both models are in acceptable agreement with one another and indicate that the zirconium diffusion barrier has a minimal impact on the overall thermal diffusivity of the monolithic U-Mo fuel.

  3. 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

  4. 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.

  5. 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.

  6. Measurement of Fission Gas Release from Irradiated U-Mo Monolithic Fuel Samples

    Energy Technology Data Exchange (ETDEWEB)

    Burkes, Douglas; Casella, Amanda J.; Casella, Andrew M.; Luscher, Walter G.; Rice, Francine; Pool, Karl N.

    2015-06-01

    The uranium-molybdenum (U-Mo) alloy in a monolithic form has been proposed as one fuel design capable of converting some of the world’s highest power research reactors from the use of high enriched uranium (HEU) to low enriched uranium (LEU). One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An apparatus capable of annealing post-irradiated small-scale samples cut from larger fuel segments according to specified thermal profiles under a controlled atmosphere has been installed into a hot cell. Results show that optimized experimental parameters to investigate fission product release from small samples have been established. Initial measurements conducted on aluminum alloy clad uranium-molybdenum monolithic fuel samples reveal three clear fission gas release events over the temperature range of 30-1050 C. The mechanisms responsible for these events are discussed, and the results have been compared with available information in literature.

  7. 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.

  8. PURIFICATION OF URANIUM FROM URANIUM/MOLYBDENUM ALLOY

    Energy Technology Data Exchange (ETDEWEB)

    Pierce, R; Ann Visser, A; James Laurinat, J

    2007-10-15

    The Savannah River Site will recycle a nuclear fuel comprised of 90% uranium-10% molybdenum by weight. The process flowsheet calls for dissolution of the material in nitric acid to a uranium concentration of 15-20 g/L without the formation of precipitates. The dissolution will be followed by separation of uranium from molybdenum using solvent extraction with 7.5% tributylphosphate in n-paraffin. Testing with the fuel validated dissolution and solubility data reported in the literature. Batch distribution coefficient measurements were performed for the extraction, strip and wash stages with particular focus on the distribution of molybdenum.

  9. 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

  10. Spectrographic analysis of uranium-molybdenum alloys; Analisis espectrografico de aleaciones uranio-molibdeno

    Energy Technology Data Exchange (ETDEWEB)

    Roca, M.

    1967-07-01

    A spectrographic method of analysis has been developed for uranium-molybdenum alloys containing up to 10 % Mo. The carrier distillation technique, with gallium oxide and graphite as carriers, is used for the semiquantitative determination of Al, Cr, Fe, Ni and Si, involving the conversion of the samples into oxides. As a consequence of the study of the influence of the molybdenum on the line intensities, it is useful to prepare only one set of standards with 0,6 % MoO{sub 3}. Total burning excitation is used for calcium, employing two sets of standards with 0,6 and 7.5 MoO{sub 3}. (Author) 5 refs.

  11. 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

  12. 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

  13. 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.

  14. Obtention of uranium-molybdenum alloy ingots technique to avoid carbon contamination

    Energy Technology Data Exchange (ETDEWEB)

    Pedrosa, Tercio A.; Paula, Joao Bosco de; Reis, Sergio C.; Brina, Jose Giovanni M.; Faeda, Kelly Cristina M.; Ferraz, Wilmar B., E-mail: tap@cdtn.b, E-mail: jbp@cdtn.b, E-mail: jgmb@cdtn.b, E-mail: ferrazw@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The replacement of high enriched uranium (U{sup 235} > 85 wt%) by low enriched uranium (U{sup 235} < 20wt%) nuclear fuels in research and test reactors is being implemented as an initiative of the Reduced Enrichment for Research and Test Reactors (RERTR) program, conceived in the USA since mid-70s, in order to avoid nuclear weapons proliferation. Such replacement implies in the use of compounds or alloys with higher uranium densities. Among the several uranium alloys investigated since then, U-Mo presents great application potential due to its physical properties and good behavior during irradiation, which makes it an important option as a nuclear fuel material for the Brazilian Multipurpose Reactor - RMB. The development of the plate-type nuclear fuel based on U-Mo alloy is being performed at the Nuclear Technology Development Centre (CDTN) and also at IPEN. The carbon contamination of the alloy is one of the great concerns during the melting process. It was observed that U-Mo alloy is more critical considering carbon contamination when using graphite crucibles. Alternative melting technique was implemented at CDTN in order to avoid carbon contamination from graphite crucible using Yttria stabilized ZrO{sub 2} crucibles. Ingots with low carbon content and good internal quality were obtained. (author)

  15. Obtention of uranium-molybdenum alloy ingots microstructure and phase characterization

    Energy Technology Data Exchange (ETDEWEB)

    Pedrosa, Tercio A.; Braga, Daniel M.; Paula, Joao Bosco de; Brina, Jose Giovanni M.; Ferraz, Wilmar B., E-mail: tap@cdtn.b, E-mail: bragadm@cdtn.b, E-mail: jbp@cdtn.b, E-mail: jgmb@cdtn.b, E-mail: ferrazw@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The replacement of high enriched uranium (U-{sup 235} > 85 wt%) by low enriched uranium (U-{sup 235} < 20 wt%) nuclear fuels in research and test reactors is being implemented as an initiative of the Reduced Enrichment for Research and Test Reactors (RERTR) program, conceived in the USA since mid-70s, in order to avoid nuclear weapons proliferation. Such replacement implies in the use of compounds or alloys with higher uranium densities. Several uranium alloys that fill this requirement has been investigated since then. Among these alloys, U-Mo presents great application potential due to its physical properties and good behavior during irradiation, which makes it an important option as a nuclear fuel material for the Brazilian Multipurpose Reactor - RMB. The development of the plate-type nuclear fuel based on U-Mo alloys is being performed at the Nuclear Technology Development Centre (CDTN) and also at the Institute of Energetic and Nuclear Research - IPEN. U-{sup 10}Mo ingots were melted in an induction furnace with protective argon atmosphere. The microstructure of the ingots were characterized through optical and scanning electronic microscopy in the as cast and heat treated conditions. Energy Dispersive Spectrometry and X-Ray Diffraction were used as characterization techniques for elemental analysis and phases determination. It was confirmed the presence of metastable gamma-phase in the as cast condition, surrounded by hypereutectoid alpha-phase (uranium-rich phase), as well as a pearlite-like constituent, composed by alternated lamellas of U{sub 2}Mo compound and alpha-phase, in the heat treated condition. (author)

  16. Nuclear fuel alloys or mixtures and method of making thereof

    Science.gov (United States)

    Mariani, Robert Dominick; Porter, Douglas Lloyd

    2016-04-05

    Nuclear fuel alloys or mixtures and methods of making nuclear fuel mixtures are provided. Pseudo-binary actinide-M fuel mixtures form alloys and exhibit: body-centered cubic solid phases at low temperatures; high solidus temperatures; and/or minimal or no reaction or inter-diffusion with steel and other cladding materials. Methods described herein through metallurgical and thermodynamics advancements guide the selection of amounts of fuel mixture components by use of phase diagrams. Weight percentages for components of a metallic additive to an actinide fuel are selected in a solid phase region of an isothermal phase diagram taken at a temperature below an upper temperature limit for the resulting fuel mixture in reactor use. Fuel mixtures include uranium-molybdenum-tungsten, uranium-molybdenum-tantalum, molybdenum-titanium-zirconium, and uranium-molybdenum-titanium systems.

  17. 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.

  18. Aqueous processing of U-10Mo scrap for high performance research reactor fuel

    Science.gov (United States)

    Youker, Amanda J.; Stepinski, Dominique C.; Maggos, Laura E.; Bakel, Allen J.; Vandegrift, George F.

    2012-08-01

    The Global Threat Reduction Initiative (GTRI) Conversion program, which is part of the US government's National Nuclear Security Administration (NNSA), supports the conversion of civilian use of highly enriched uranium (HEU) to low enriched uranium (LEU) for reactor fuel and targets. The reason for conversion is to eliminate the use of any material that may pose a threat to the United States or other foreign countries. High performance research reactors (HPRRs) cannot make the conversion to a standard LEU fuel because they require a more dense fuel to meet their performance requirements. As a result, a more dense fuel consisting of a monolithic uranium-molybdenum alloy containing 10% (w/w) Mo with Al cladding and a Zr bonding-layer is being considered. Significant losses are expected in the fabrication of this fuel, so a means to recycle the scrap pieces is needed. Argonne National Laboratory has developed an aqueous-processing flowsheet for scrap recovery in the fuel fabrication process for high-density LEU-monolithic fuel based on data found in the literature. Experiments have been performed to investigate dissolution conditions for solutions containing approximately 20 g-U/L and 50 g-U/L with and without Fe(NO3)3. HNO3 and HF concentrations have been optimized for timely dissolution of the fuel scrap and prevention of the formation of the U-Zr2 intermetallic, explosive complex, while meeting the requirements needed for further processing.

  19. 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.

  20. Aluminum cladding oxidation of prefilmed in-pile fueled experiments

    Science.gov (United States)

    Marcum, W. R.; Wachs, D. M.; Robinson, A. B.; Lillo, M. A.

    2016-04-01

    A series of fueled irradiation experiments were recently completed within the Advanced Test Reactor Full size plate In center flux trap Position (AFIP) and Gas Test Loop (GTL) campaigns. The conduct of the AFIP experiments supports ongoing efforts within the global threat reduction initiative (GTRI) to qualify a new ultra-high loading density low enriched uranium-molybdenum fuel. This study details the characterization of oxide growth on the fueled AFIP experiments and cross-correlates the empirically measured oxide thickness values to existing oxide growth correlations and convective heat transfer correlations that have traditionally been utilized for such an application. This study adds new and valuable empirical data to the scientific community with respect to oxide growth measurements of highly irradiated experiments, of which there is presently very limited data. Additionally, the predicted oxide thickness values are reconstructed to produce an oxide thickness distribution across the length of each fueled experiment (a new application and presentation of information that has not previously been obtainable in open literature); the predicted distributions are compared against experimental data and in general agree well with the exception of select outliers.

  1. Measurement of fission gas release from irradiated UMo dispersion fuel samples

    Energy Technology Data Exchange (ETDEWEB)

    Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.

    2016-09-01

    The uranium-molybdenum (U-Mo) alloy dispersed in an Al-Si matrix has been proposed as one fuel design capable of converting some of the world’s highest power research reactors from the use of high enriched uranium (HEU) to low enriched uranium (LEU). One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. In this paper, two irradiated samples containing 53.6 vol% U-7wt% Mo fuel particles dispersed in an Al-2wt% Si matrix were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Measurements revealed three distinct fission gas release events for the samples from 400 to 700 oC, as well as a number of minor fission gas releases below and above this temperature range. The mechanisms responsible for these events are discussed, and the results have been compared with available information in the literature with exceptional agreement.

  2. Measurement of fission gas release from irradiated Usbnd Mo dispersion fuel samples

    Science.gov (United States)

    Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.

    2016-09-01

    The uranium-molybdenum (Usbnd Mo) alloy dispersed in an Alsbnd Si matrix has been proposed as one fuel design capable of converting some of the world's highest power research reactors from the use of high enriched uranium (HEU) to low enriched uranium (LEU). One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. In this paper, two irradiated samples containing 53.9 vol% U-7wt% Mo fuel particles dispersed in an Al-2wt% Si matrix were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Measurements revealed three distinct fission gas release events for the samples from 400 to 700 °C, as well as a number of minor fission gas releases below and above this temperature range. The mechanisms responsible for these events are discussed, and the results have been compared with available information in the literature with exceptional agreement.

  3. The influence of cladding on fission gas release from irradiated U-Mo monolithic fuel

    Science.gov (United States)

    Burkes, Douglas E.; Casella, Amanda J.; Casella, Andrew M.

    2017-04-01

    The monolithic uranium-molybdenum (U-Mo) alloy has been proposed as a fuel design capable of converting the world's highest power research reactors from use of high enriched uranium to low enriched uranium. However, a zirconium (Zr) diffusion barrier must be used to eliminate interactions that form between the U-Mo monolith and aluminum alloy 6061 (AA6061) cladding during fabrication and are enhanced during irradiation. One aspect of fuel development and qualification is to demonstrate an appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An exothermic reaction has previously been observed between the AA6061 cladding and Zr diffusion layer. In this paper, two fuel segments with different irradiation history were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Samples from each segment were tested with cladding and without cladding to investigate the effect, if any, that the exothermic reaction has on fission gas release mechanisms. Measurements revealed there is an instantaneous effect of the cladding/Zr exothermic reaction, but not necessarily a cumulative effect above approximately 973 K (700 °C). The mechanisms responsible for fission gas release events are discussed.

  4. Conceptual Design Parameters for HFIR LEU U-Mo Fuel Conversion Experimental Irradiations

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-03-01

    The High Flux Isotope Reactor (HFIR) is a versatile research reactor that is operated at the Oak Ridge National Laboratory (ORNL). The HFIR core is loaded with high-enriched uranium (HEU) and operates at a power level of 85 MW. The primary scientific missions of the HFIR include cold and thermal neutron scattering, materials irradiation, and isotope production. An engineering design study of the conversion of the HFIR from HEU to low-enriched uranium (LEU) fuel is ongoing at the Oak Ridge National Laboratory. The LEU fuel considered is based on a uranium-molybdenum alloy that is 10 percent by weight molybdenum (U-10Mo) with a 235U enrichment of 19.75 wt %. The LEU core design discussed in this report is based on the design documented in ORNL/TM-2010/318. Much of the data reported in Sections 1 and 2 of this document was derived from or taken directly out of ORNL/TM-2010/318. The purpose of this report is to document the design parameters for and the anticipated normal operating conditions of the conceptual HFIR LEU fuel to aid in developing requirements for HFIR irradiation experiments.

  5. Influence of a doping by Al stainless steel on kinetics and character of interaction with the metallic nuclear fuel

    Science.gov (United States)

    Nikitin, S. N.; Shornikov, D. P.; Tarasov, B. A.; Baranov, V. G.

    2016-04-01

    Metallic nuclear fuel is a perspective kind of fuel for fast reactors. In this paper we conducted a study of the interaction between uranium-molybdenum alloy and ferritic- martensitic steels with additions of aluminum at a temperature of 700 ° C for 25 hours. The rate constants of the interaction layer growth at 700 °C is about 2.8.10-14 m2/s. It is established that doping Al stainless steel leads to decrease in interaction with uranium-molybdenum alloys. The phase composition of the interaction layer is determined.

  6. 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.

  7. 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

  8. 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.

  9. 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.

  10. Swelling of U-7Mo/Al-Si dispersion fuel plates under irradiation - Non-destructive analysis of the AFIP-1 fuel plates

    Science.gov (United States)

    Wachs, D. M.; Robinson, A. B.; Rice, F. J.; Kraft, N. C.; Taylor, S. C.; Lillo, M.; Woolstenhulme, N.; Roth, G. A.

    2016-08-01

    Extensive fuel-matrix interactions leading to plate pillowing have proven to be a significant impediment to the development of a suitable high density low-enriched uranium molybdenum alloy (U-Mo) based dispersion fuel for high power applications in research reactors. The addition of silicon to the aluminum matrix was previously demonstrated to reduce interaction layer growth in mini-plate experiments. The AFIP-1 project involved the irradiation, in-canal examination, and post-irradiation examination of two fuel plates. The irradiation of two distinct full size, flat fuel plates (one using an Al-2wt%Si matrix and the other an Al-4043 (∼4.8 wt% Si) matrix) was performed in the INL ATR reactor in 2008-2009. The irradiation conditions were: ∼250 W/cm2 peak Beginning Of Life (BOL) power, with a ∼3.5e21 f/cm3 peak burnup. The plates were successfully irradiated and did not show any pillowing at the end of the irradiation. This paper reports the results and interpretation of the in-canal and post-irradiation non-destructive examinations that were performed on these fuel plates. It further compares additional PIE results obtained on fuel plates irradiated in contemporary campaigns in order to allow a complete comparison with all results obtained under similar conditions. Except for a brief indication of accelerated swelling early in the irradiation of the Al-2Si plate, the fuel swelling is shown to evolve linearly with the fission density through the maximum burnup.

  11. A model to predict failure of irradiated U–Mo dispersion fuel

    Energy Technology Data Exchange (ETDEWEB)

    Burkes, Douglas E.; Senor, David J.; Casella, Andrew M.

    2016-12-01

    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 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. The current paper extends a failure model originally developed for UO2-stainless steel dispersion fuels and used currently available thermal-mechanical property information for the materials of interest in the current proposed design. A number of fabrication and irradiation parameters were investigated to understand the conditions at which failure of the matrix, classified as pore formation in the matrix, might occur. The results compared well with experimental observations published as part of the Reduced Enrichment for Research and Test Reactors (RERTR)-6 and -7 mini-plate experiments. Fission rate, a function of the 235U enrichment, appeared to be the most influential parameter in premature failure, mainly as a result of increased interaction layer formation and operational temperature, which coincidentally decreased the yield strength of the matrix and caused more rapid fission gas production and recoil into the surrounding matrix material. Addition of silicon to the matrix appeared effective at reducing the rate of interaction layer formation and can extend the performance of a fuel plate under a certain set of irradiation conditions, primarily moderate heat flux and burnup. Increasing the dispersed fuel particle diameter may also be effective, but only when combined with other parameters, e.g., lower enrichment and increased Si concentration. The model may serve as a valuable tool in initial experimental design.

  12. 78 FR 75579 - Low Enriched Uranium From France

    Science.gov (United States)

    2013-12-12

    ... COMMISSION Low Enriched Uranium From France Determination On the basis of the record \\1\\ developed in the... antidumping duty order on low enriched uranium from France would be likely to lead to continuation or...), entitled Low Enriched Uranium from France: Investigation No. 731-TA-909 (Second Review). By order of...

  13. 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.

  14. 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.

  15. 31 CFR 540.308 - Low Enriched Uranium (LEU).

    Science.gov (United States)

    2010-07-01

    ... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false Low Enriched Uranium (LEU). 540.308... OF FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.308 Low Enriched Uranium (LEU). The term low...

  16. Aqueous processing of U-10Mo scrap for high performance research reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Youker, Amanda J., E-mail: youker@anl.gov [Chemical Sciences and Engineering Division, Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States); Stepinski, Dominique C.; Maggos, Laura E.; Bakel, Allen J.; Vandegrift, George F. [Chemical Sciences and Engineering Division, Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer GTRI program supports conversion from HEU to LEU. Black-Right-Pointing-Pointer High performance research reactors require a dense LEU fuel such as U-10Mo foils. Black-Right-Pointing-Pointer Dissolution conditions for U-10Mo foils in acidic media have been optimized. Black-Right-Pointing-Pointer Solvent-extraction processing can be used to recover U lost in fuel fabrication. Black-Right-Pointing-Pointer Flowsheets were developed using Argonne-design contactors but other contactors can be used as well. - Abstract: The Global Threat Reduction Initiative (GTRI) Conversion program, which is part of the US government's National Nuclear Security Administration (NNSA), supports the conversion of civilian use of highly enriched uranium (HEU) to low enriched uranium (LEU) for reactor fuel and targets. The reason for conversion is to eliminate the use of any material that may pose a threat to the United States or other foreign countries. High performance research reactors (HPRRs) cannot make the conversion to a standard LEU fuel because they require a more dense fuel to meet their performance requirements. As a result, a more dense fuel consisting of a monolithic uranium-molybdenum alloy containing 10% (w/w) Mo with Al cladding and a Zr bonding-layer is being considered. Significant losses are expected in the fabrication of this fuel, so a means to recycle the scrap pieces is needed. Argonne National Laboratory has developed an aqueous-processing flowsheet for scrap recovery in the fuel fabrication process for high-density LEU-monolithic fuel based on data found in the literature. Experiments have been performed to investigate dissolution conditions for solutions containing approximately 20 g-U/L and 50 g-U/L with and without Fe(NO{sub 3}){sub 3}. HNO{sub 3} and HF concentrations have been optimized for timely dissolution of the fuel scrap and prevention of the formation of the U-Zr{sub 2} intermetallic, explosive complex, while

  17. 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.

  18. Recent Accomplishments in the Irradiation Testing of Engineering-Scale Monolithic Fuel Specimens

    Energy Technology Data Exchange (ETDEWEB)

    N.E. Woolstenhulme; D.M. Wachs; M.K. Meyer; H.W. Glunz; R.B. Nielson

    2012-10-01

    The US fuel development team is focused on qualification and demonstration of the uranium-molybdenum monolithic fuel including irradiation testing of engineering-scale specimens. The team has recently accomplished the successful irradiation of the first monolithic multi-plate fuel element assembly within the AFIP-7 campaign. The AFIP-6 MKII campaign, while somewhat truncated by hardware challenges, exhibited successful irradiation of a large-scale monolithic specimen under extreme irradiation conditions. The channel gap and ultrasonic data are presented for AFIP-7 and AFIP-6 MKII, respectively. Finally, design concepts are summarized for future irradiations such as the base fuel demonstration and design demonstration experiment campaigns.

  19. Complex plasmochemical processing of solid fuel

    Directory of Open Access Journals (Sweden)

    Vladimir Messerle

    2012-12-01

    Full Text Available Technology of complex plasmaochemical processing of solid fuel by Ecibastuz bituminous and Turgay brown coals is presented. Thermodynamic and experimental study of the technology was fulfilled. Use of this technology allows producing of synthesis gas from organic mass of coal and valuable components (technical silicon, ferrosilicon, aluminum and silicon carbide and microelements of rare metals: uranium, molybdenum, vanadium etc. from mineral mass of coal. Produced a high-calorific synthesis gas can be used for methanol synthesis, as high-grade reducing gas instead of coke, as well as energy gas in thermal power plants.

  20. Corrosion report for the U-Mo fuel concept

    Energy Technology Data Exchange (ETDEWEB)

    Henager, Charles H. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Bennett, Wendy D. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Doherty, Ann L. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Fuller, E. S. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Hardy, John S. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Omberg, Ronald P. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States)

    2014-08-28

    The Fuel Cycle Research and Development (FCRD) program of the Office of Nuclear Energy (NE) has implemented a program to develop a Uranium-Molybdenum (U-Mo) metal fuel for Light Water Reactors (LWR)s. Uranium-Molybdenum fuel has the potential to provide superior performance based on its thermo-physical properties, which includes high thermal conductivity for less stored heat energy. With sufficient development, it may be able to provide the Light Water industry with a melt-resistant accident tolerant fuel with improved safety response. However, the corrosion of this fuel in reactor water environments needs to be further explored and optimized by additional alloying. The Pacific Northwest National Laboratory has been tasked with performing ex-reactor corrosion testing to characterize the performance of U-Mo fuel. This report documents the results of the effort to characterize and develop the U-Mo metal fuel concept for LWRs with regard to corrosion testing. The results of a simple screening test in buffered water at 30°C using surface alloyed U-10Mo is documented and discussed. The screening test was used to guide the selection of several potential alloy improvements that were found and are recommended for further testing in autoclaves to simulate PWR water conditions more closely.

  1. The OSU Hydro-Mechanical Fuel Test Facility: Standard Fuel Element Testing

    Energy Technology Data Exchange (ETDEWEB)

    Wade R. Marcum; Brian G. Woods; Ann Marie Phillips; Richard G. Ambrosek; James D. Wiest; Daniel M. Wachs

    2001-10-01

    Oregon State University (OSU) and the Idaho National Laboratory (INL) are currently collaborating on a test program which entails hydro-mechanical testing of a generic plate type fuel element, or standard fuel element (SFE), for the purpose of qualitatively demonstrating mechanical integrity of uranium-molybdenum monolithic plates as compared to that of uranium aluminum dispersion, and aluminum fuel plates due to hydraulic forces. This test program supports ongoing work conducted for/by the fuel development program and will take place at OSU in the Hydro-Mechanical Fuel Test Facility (HMFTF). Discussion of a preliminary test matrix, SFE design, measurement and instrumentation techniques, and facility description are detailed in this paper.

  2. 77 FR 19642 - Low Enriched Uranium From France: Final Results of Antidumping Duty Changed Circumstances Review

    Science.gov (United States)

    2012-04-02

    ... International Trade Administration Low Enriched Uranium From France: Final Results of Antidumping Duty Changed... antidumping duty order on low enriched uranium (LEU) from France on February 10, 2012,\\1\\ in which the... Results. \\1\\ See Low Enriched Uranium from France: Preliminary Results of Antidumping Duty...

  3. 77 FR 7128 - Low Enriched Uranium From France: Preliminary Results of Antidumping Duty Changed Circumstances...

    Science.gov (United States)

    2012-02-10

    ... International Trade Administration Low Enriched Uranium From France: Preliminary Results of Antidumping Duty... review (CCR) of the antidumping duty order of low enriched uranium (LEU) from France.\\1\\ We preliminarily.... \\1\\ See Low Enriched Uranium from France: Initiation of Antidumping Duty Changed Circumstances...

  4. 77 FR 1059 - Low Enriched Uranium From France: Initiation of Antidumping Duty Changed Circumstances Review

    Science.gov (United States)

    2012-01-09

    ... International Trade Administration Low Enriched Uranium From France: Initiation of Antidumping Duty Changed... review of the antidumping duty order on low enriched uranium (LEU) from France with respect to Eurodif S... deadline for re-exporting the LEU entry at issue. \\1\\ See Letter from AREVA, ``Low Enriched Uranium...

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

    Energy Technology Data Exchange (ETDEWEB)

    McMurray, J. W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Silva, C. M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Helmreich, G. W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gerczak, T. J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Dyer, J. A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Collins, J. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hunt, R. D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Lindemer, T. B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, K. A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-06-01

    A large batch of UN microspheres to be used as kernels for TRISO particle fuel was produced using carbothermic reduction and nitriding of a sol-gel feedstock bearing tailored amounts of low-enriched uranium (LEU) oxide and carbon. The process parameters, established in a previous study, produced phasepure NaCl structure UN with dissolved C on the N sublattice. The composition, calculated by refinement of the lattice parameter from X-ray diffraction, was determined to be UC0.27N0.73. The final accepted product weighed 197.4 g. The microspheres had an average diameter of 797±1.35 μm and a composite mean theoretical density of 89.9±0.5% for a solid solution of UC and UN with the same atomic ratio; both values are reported with their corresponding calculated standard error.

  6. 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.

  7. A study of phase transformations processes in 0,5 to 4% mo uranium-molybdenum alloys; Etude des processus des transformations dans les alliages uranium-molybdene de teneur 0,5 a 4% en poids de molybdene

    Energy Technology Data Exchange (ETDEWEB)

    Lehmann, J. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1959-06-15

    Isothermal and continuous cooling transformations process have been established on uranium-molybdenum alloys containing 0,5 to 4 w% Mo. Transformations process of the {beta} and {gamma} solid solutions are described. These processes depend upon molybdenum concentration. Out of the {beta} solid solution phase appears an eutectoid decomposition of {beta} to ({alpha} + {gamma}) or the formation of a martensitic phase {alpha}''. The {gamma} solid solution shows a decomposition of {gamma} to ({alpha} + {gamma}) or ({alpha} + {gamma}'), or a formation of martensitic phases a' or a'{sub b}. The U-Mo equilibrium diagram is discussed, particularly in low concentrations zones. Limits between domains ({alpha} + {gamma}) and ({beta} + {gamma}), ({beta} + {gamma}) and {gamma}, ({beta} + {gamma}) and {beta}, have been determined. (author) [French] Les processus des transformations isothermes, et au cours de refroidissements continus ont ete etablis sur les alliages uranium-molybdene de 0,5 a 4 % en poids de Mo. Ceci a permis de mettre en evidence les processus des transformations de solutions solides {beta} et {gamma}, differents suivant la teneur en molybdene de l'alliage. Dans le premier cas il y a decomposition eutectoide de {beta} en ({alpha} + {gamma}) ou formations d'une phase martensitique {alpha}''. Dans le second cas il y a decomposition de {gamma} soit en ({alpha} + {gamma}) soit en ({alpha} + {gamma}') suivant la temperature, ou bien formation des phases martensitiques {alpha}' ou {alpha}'{sub b}. Le diagramme d'equilibre, uranium-molybdene est sujet a de nombreuses controverses, en particulier dans la zone des faibles concentrations. Les limites entre les domaines ({alpha} + {gamma}) et ({beta} + {gamma}), ({beta} + {gamma}) et {gamma}, ({beta} + {gamma}) et {beta}, ont ete determinees. (auteur)

  8. 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.

  9. 78 FR 52905 - Low Enriched Uranium From France: Initiation of Expedited Changed Circumstances Review, and...

    Science.gov (United States)

    2013-08-27

    ... International Trade Administration Low Enriched Uranium From France: Initiation of Expedited Changed... uranium from France with respect to Eurodif S.A. and AREVA NP Inc. (collectively, AREVA). Moreover, the... published an order on low- enriched uranium from France.\\1\\ The order contains a provision to exclude...

  10. 78 FR 21100 - Low Enriched Uranium From France: Final Results of the Expedited Second Sunset Review of the...

    Science.gov (United States)

    2013-04-09

    ... International Trade Administration Low Enriched Uranium From France: Final Results of the Expedited Second... low enriched uranium (``LEU'') from France. The Department finds that revocation of this antidumping... Determination of Sales at Less Than Fair Value and Antidumping Duty Order: Low Enriched Uranium From France,...

  11. 77 FR 71626 - Low Enriched Uranium From France; Institution of a Five-Year Review Concerning the Antidumping...

    Science.gov (United States)

    2012-12-03

    ... COMMISSION Low Enriched Uranium From France; Institution of a Five-Year Review Concerning the Antidumping Duty Order on Low Enriched Uranium From France AGENCY: United States International Trade Commission... the antidumping duty order on low enriched uranium from France would be likely to lead to...

  12. 78 FR 21416 - Low Enriched Uranium From France; Scheduling of a Full Five-year Review Concerning the...

    Science.gov (United States)

    2013-04-10

    ... COMMISSION Low Enriched Uranium From France; Scheduling of a Full Five-year Review Concerning the Antidumping Duty Order on Low Enriched Uranium from France AGENCY: United States International Trade Commission... revocation of the antidumping duty order on low enriched uranium from France would be likely to lead...

  13. Hot rolling of thick uranium molybdenum alloys

    Science.gov (United States)

    DeMint, Amy L.; Gooch, Jack G.

    2015-11-17

    Disclosed herein are processes for hot rolling billets of uranium that have been alloyed with about ten weight percent molybdenum to produce cold-rollable sheets that are about one hundred mils thick. In certain embodiments, the billets have a thickness of about 7/8 inch or greater. Disclosed processes typically involve a rolling schedule that includes a light rolling pass and at least one medium rolling pass. Processes may also include reheating the rolling stock and using one or more heavy rolling passes, and may include an annealing step.

  14. 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

  15. 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).

  16. Low-enrichment and long-life Scalable LIquid Metal cooled small Modular (SLIMM-1.2) reactor

    Energy Technology Data Exchange (ETDEWEB)

    El-Genk, Mohamed S., E-mail: mgenk@unm.edu [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM (United States); Nuclear Engineering Department, University of New Mexico, Albuquerque, NM (United States); Mechanical Engineering Department, University of New Mexico, Albuquerque, NM (United States); Chemical and Biological Engineering Department, University of New Mexico, Albuquerque, NM (United States); Palomino, Luis M.; Schriener, Timothy M. [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM (United States); Nuclear Engineering Department, University of New Mexico, Albuquerque, NM (United States)

    2017-05-15

    Highlights: • Developed low enrichment and natural circulation cooled SLIMM-1.2 SMR for generating 10–100 MW{sub th}. • Neutronics analyses estimate operation life and temperature reactivity feedback. • At 100 MW{sub th}, SLIMM-1.2 operates for 6.3 FPY without refueling. • SLIMM-1.2 has relatively low power peaking and maximum UN fuel temperature < 1400 K. - Abstract: The Scalable LIquid Metal cooled small Modular (SLIMM-1.0) reactor with uranium nitride fuel enrichment of 17.65% had been developed for generating 10–100 MW{sub th} continuously, without refueling for ∼66 and 5.9 full power years, respectively. Natural circulation of in-vessel liquid sodium (Na) cools the core of this fast energy spectrum reactor during nominal operation and after shutdown, with the aid of a tall chimney and an annular Na/Na heat exchanger (HEX) of concentric helically coiled tubes. The HEX at the top of the downcomer maximizes the static pressure head for natural circulation. In addition to the independent emergency shutdown (RSS) and reactor control (RC), the core negative temperature reactivity feedback safely decreases the reactor thermal power, following modest increases in the temperatures of UN fuel and in-vessel liquid sodium. The decay heat is removed from the core by natural circulation of in-vessel liquid sodium, with aid of the liquid metal heat pipes laid along the reactor vessel wall, and the passive backup cooling system (BCS) using natural circulation of ambient air along the outer surface of the guard vessel wall. This paper investigates modifying the SLIMM-1.0 reactor design to lower the UN fuel enrichment. To arrive at a final reactor design (SLIMM-1.2), the performed neutronics and reactivity depletion analyses examined the effects of various design and material choices on both the cold-clean and the hot-clean excess reactivity, the reactivity shutdown margin, the full power operation life at 100 MW{sub th}, the fissile production and depletion, the

  17. Benchmark calculation for water reflected STACY cores containing low enriched uranyl nitrate solution

    Energy Technology Data Exchange (ETDEWEB)

    Miyoshi, Yoshinori; Yamamoto, Toshihiro; Nakamura, Takemi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-08-01

    In order to validate the availability of criticality calculation codes and related nuclear data library, a series of fundamental benchmark experiments on low enriched uranyl nitrate solution have been performed with a Static Experiment Criticality Facility, STACY in JAERI. The basic core composed of a single tank with water reflector was used for accumulating the systematic data with well-known experimental uncertainties. This paper presents the outline of the core configurations of STACY, the standard calculation model, and calculation results with a Monte Carlo code and JENDL 3.2 nuclear data library. (author)

  18. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-01-01

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

  20. Benchmark critical experiments on low-enriched uranium oxide systems with H/U = 0. 77

    Energy Technology Data Exchange (ETDEWEB)

    Tuck, G; Oh, I

    1979-08-01

    Ten benchmark experiments were performed at the Critical Mass Laboratory at Rockwell International's Rocky Flats Plant, Golden, Colorado, for the US Nuclear Regulatory Commission. They provide accurate criticality data for low-enriched damp uranium oxide (U/sub 3/O/sub 8/) systems. The core studied consisted of 152 mm cubical aluminum cans containing an average of 15,129 g of low-enriched (4.46% /sup 235/U) uranium oxide compacted to a density of 4.68 g/cm/sup 3/ and with an H/U atomic ratio of 0.77. One hundred twenty five (125) of these cans were arranged in an approx. 770 mm cubical array. Since the oxide alone cannot be made critical in an array of this size, an enriched (approx. 93% /sup 235/U) metal or solution driver was used to achieve criticality. Measurements are reported for systems having the least practical reflection and for systems reflected by approx. 254-mm-thick concrete or plastic. Under the three reflection conditions, the mass of the uranium metal driver ranged from 29.87 kg to 33.54 kg for an oxide core of 1864.6 kg. For an oxide core of 1824.9 kg, the weight of the high concentration (351.2 kg U/m/sup 3/) solution driver varied from 14.07 kg to 16.14 kg, and the weight of the low concentration (86.4 kg U/m/sup 3/) solution driver from 12.4 kg to 14.0 kg.

  1. Plasma sprayed and electrospark deposited zirconium metal diffusion barrier coatings

    Energy Technology Data Exchange (ETDEWEB)

    Hollis, Kendall J [Los Alamos National Laboratory; Pena, Maria I [Los Alamos National Laboratory

    2010-01-01

    Zirconium metal coatings applied by plasma spraying and electrospark deposition (ESD) have been investigated for use as diffusion barrier coatings on low enrichment uranium fuel for research nuclear reactors. The coatings have been applied to both stainless steel as a surrogate and to simulated nuclear fuel uranium-molybdenum alloy substrates. Deposition parameter development accompanied by coating characterization has been performed. The structure of the plasma sprayed coating was shown to vary with transferred arc current during deposition. The structure of ESD coatings was shown to vary with the capacitance of the deposition equipment.

  2. Low-enriched uranium high-density target project. Compendium report

    Energy Technology Data Exchange (ETDEWEB)

    Vandegrift, George; Brown, M. Alex; Jerden, James L.; Gelis, Artem V.; Stepinski, Dominique C.; Wiedmeyer, Stanley; Youker, Amanda; Hebden, Andrew; Solbrekken, G; Allen, C; Robertson., D; El-Gizawy, Sherif; Govindarajan, Srisharan; Hoyer, Annemarie; Makarewicz, Philip; Harris, Jacob; Graybill, Brian; Gunn, Andy; Berlin, James; Bryan, Chris; Sherman, Steven; Hobbs, Randy; Griffin, F. P.; Chandler, David; Hurt, C. J.; Williams, Paul; Creasy, John; Tjader, Barak; McFall, Danielle; Longmire, Hollie

    2016-09-01

    At present, most 99Mo is produced in research, test, or isotope production reactors by irradiation of highly enriched uranium targets. To achieve the denser form of uranium needed for switching from high to low enriched uranium (LEU), targets in the form of a metal foil (~125-150 µm thick) are being developed. The LEU High Density Target Project successfully demonstrated several iterations of an LEU-fission-based Mo-99 technology that has the potential to provide the world’s supply of Mo-99, should major producers choose to utilize the technology. Over 50 annular high density targets have been successfully tested, and the assembly and disassembly of targets have been improved and optimized. Two target front-end processes (acidic and electrochemical) have been scaled up and demonstrated to allow for the high-density target technology to mate up to the existing producer technology for target processing. In the event that a new target processing line is started, the chemical processing of the targets is greatly simplified. Extensive modeling and safety analysis has been conducted, and the target has been qualified to be inserted into the High Flux Isotope Reactor, which is considered above and beyond the requirements for the typical use of this target due to high fluence and irradiation duration.

  3. Minimum mass of moderator required for criticality of homogeneous low-enriched uranium systems

    Energy Technology Data Exchange (ETDEWEB)

    Jordan, W.C.; Turner, J.C.

    1992-12-01

    A parametric calculational analysis has been performed in order to estimate the minimum mass of moderator required for criticality of homogeneous low-enriched uranium systems. The analysis was performed using a version of the SCALE-4.0 code system and the 27-group ENDF/B-IV cross-section library. Water-moderated uranyl fluoride (UO{sub 2}F{sub 2} and H{sub 2}O) and hydrofluoric-acid-moderated uranium hexaflouride (UF{sub 6} and HF) systems were considered in the analysis over enrichments of 1.4 to 5 wt % {sup 235}U. Estimates of the minimum critical volume, minimum critical mass of uranium, and the minimum mass of moderator required for criticality are presented. There was significant disagreement between the values generated in this study when compared with a similar undocumented study performed in 1983 using ANISN and the Knight-modified Hansen-Roach cross sections. An investigation into the cause of the disagreement was made, and the results are presented.

  4. Minimum mass of moderator required for criticality of homogeneous low-enriched uranium systems

    Energy Technology Data Exchange (ETDEWEB)

    Jordan, W.C.; Turner, J.C.

    1992-12-01

    A parametric calculational analysis has been performed in order to estimate the minimum mass of moderator required for criticality of homogeneous low-enriched uranium systems. The analysis was performed using a version of the SCALE-4.0 code system and the 27-group ENDF/B-IV cross-section library. Water-moderated uranyl fluoride (UO[sub 2]F[sub 2] and H[sub 2]O) and hydrofluoric-acid-moderated uranium hexaflouride (UF[sub 6] and HF) systems were considered in the analysis over enrichments of 1.4 to 5 wt % [sup 235]U. Estimates of the minimum critical volume, minimum critical mass of uranium, and the minimum mass of moderator required for criticality are presented. There was significant disagreement between the values generated in this study when compared with a similar undocumented study performed in 1983 using ANISN and the Knight-modified Hansen-Roach cross sections. An investigation into the cause of the disagreement was made, and the results are presented.

  5. Development of industrial-scale fission {sup 99}Mo production process using low enriched uranium target

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Seung Kon; Lee, Jun Sig [Radioisotope Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Beyer, Gerd J. [Grunicke Strasse 15, Leipzig (Germany)

    2016-06-15

    Molybdenum-99 ({sup 99}Mo) is the most important isotope because its daughter isotope, technetium-99m ({sup 99}mTc), has been the most widely used medical radioisotope for more than 50 years, accounting for > 80% of total nuclear diagnostics worldwide. In this review, radiochemical routes for the production of {sup 99}Mo, and the aspects for selecting a suitable process strategy are discussed from the historical viewpoint of {sup 99}Mo technology developments. Most of the industrial-scale {sup 99}Mo processes have been based on the fission of {sup 235}U. Recently, important issues have been raised for the conversion of fission {sup 99}Mo targets from highly enriched uranium to low enriched uranium (LEU). The development of new LEU targets with higher density was requested to compensate for the loss of {sup 99}Mo yield, caused by a significant reduction of {sup 235}U enrichment, from the conversion. As the dramatic increment of intermediate level liquid waste is also expected from the conversion, an effective strategy to reduce the waste generation from the fission {sup 99}Mo production is required. The mitigation of radioxenon emission from medical radioisotope production facilities is discussed in relation with the monitoring of nuclear explosions and comprehensive nuclear test ban. Lastly, the {sup 99}Mo production process paired with the Korea Atomic Energy Research Institute's own LEU target is proposed as one of the most suitable processes for the LEU target.

  6. 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....

  7. 78 FR 19311 - Low Enriched Uranium From France; Notice of Commission Determination to Conduct a Full Five-Year...

    Science.gov (United States)

    2013-03-29

    ... From the Federal Register Online via the Government Publishing Office INTERNATIONAL TRADE COMMISSION Low Enriched Uranium From France; Notice of Commission Determination to Conduct a Full Five-Year... in the subject five-year review pursuant to section 751(c)(5) of the Act. The Commission found that...

  8. Accelerator-driven sub-critical research facility with low-enriched fuel in lead matrix: Neutron flux calculation

    Directory of Open Access Journals (Sweden)

    Avramović Ivana

    2007-01-01

    Full Text Available The H5B is a concept of an accelerator-driven sub-critical research facility (ADSRF being developed over the last couple of years at the Vinča Institute of Nuclear Sciences, Belgrade, Serbia. Using well-known computer codes, the MCNPX and MCNP, this paper deals with the results of a tar get study and neutron flux calculations in the sub-critical core. The neutron source is generated by an interaction of a proton or deuteron beam with the target placed inside the sub-critical core. The results of the total neutron flux density escaping the target and calculations of neutron yields for different target materials are also given here. Neutrons escaping the target volume with the group spectra (first step are used to specify a neutron source for further numerical simulations of the neutron flux density in the sub-critical core (second step. The results of the calculations of the neutron effective multiplication factor keff and neutron generation time L for the ADSRF model have also been presented. Neutron spectra calculations for an ADSRF with an uranium tar get (highest values of the neutron yield for the selected sub-critical core cells for both beams have also been presented in this paper.

  9. 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.

  10. 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.

  11. Environmental assessment for the purchase of Russian low enriched uranium derived from the dismantlement of nuclear weapons in the countries of the former Soviet Union

    Energy Technology Data Exchange (ETDEWEB)

    1994-01-01

    The United States is proposing to purchase from the Russian Federation low enriched uranium (LEU) derived from highly enriched uranium (HEU) resulting from the dismantlement of nuclear weapons in the countries of the former Soviet Union. The purchase would be accomplished through a proposed contract requiring the United States to purchase 15,250 metric tons (tonnes) of LEU (or 22,550 tonnes of UF{sub 6}) derived from blending 500 metric tones uranium (MTU) of HEU from nuclear warheads. The LEU would be in the form of uranium hexafluoride (UF{sub 6}) and would be converted from HEU in Russia. The United States Enrichment Corporation (USEC) is the entity proposing to undertake the contract for purchase, sale, and delivery of the LEU from the Russian Federation. The US Department of Energy (DOE) is negotiating the procedure for gaining confidence that the LEU is derived from HEU that is derived from dismantled nuclear weapons (referred to as ``transparency),`` and would administer the transparency measures for the contract. There are six environments that could potentially be affected by the proposed action; marine (ocean); US ports of entry; truck or rail transportation corridors; the Portsmouth GDP; the electric power industry; and the nuclear fuel cycle industry. These environmental impacts are discussed.

  12. SOLVENT EXTRACTION FOR URANIUM MOLYBDENUM ALLOY DISSOLUTION FLOWSHEET

    Energy Technology Data Exchange (ETDEWEB)

    Visser, A; Robert Pierce, R

    2007-06-07

    H-Canyon Engineering requested the Savannah River National Laboratory (SRNL) to perform two solvent extraction experiments using dissolved Super Kukla (SK) material. The SK material is an uranium (U)-molybdenum (Mo) alloy material of 90% U/10% Mo by weight with 20% 235U enrichment. The first series of solvent extraction tests involved a series of batch distribution coefficient measurements with 7.5 vol % tributylphosphate (TBP)/n-paraffin for extraction from 4-5 M nitric acid (HNO{sub 3}), using 4 M HNO{sub 3}-0.02 M ferrous sulfamate (Fe(SO3NH2)2) scrub, 0.01 M HNO3 strip steps with particular emphasis on the distribution of U and Mo in each step. The second set of solvent extraction tests determined whether the 2.5 wt % sodium carbonate (Na2CO3) solvent wash change frequency would need to be modified for the processing of the SK material. The batch distribution coefficient measurements were performed using dissolved SK material diluted to 20 g/L (U + Mo) in 4 M HNO{sub 3} and 5 M HNO{sub 3}. In these experiments, U had a distribution coefficient greater than 2.5 while at least 99% of the nickel (Ni) and greater than 99.9% of the Mo remained in the aqueous phase. After extraction, scrub, and strip steps, the aqueous U product from the strip contains nominally 7.48 {micro}g Mo/g U, significantly less than the maximum allowable limit of 800 {micro}g Mo/g U. Solvent washing experiments were performed to expose a 2.5 wt % Na2CO3 solvent wash solution to the equivalent of 37 solvent wash cycles. The low Mo batch distribution coefficient in this solvent extraction system yields only 0.001-0.005 g/L Mo extracted to the organic. During the solvent washing experiments, the Mo appears to wash from the organic.

  13. A physical description of fission product behavior fuels for advanced power reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Kaganas, G.; Rest, J.; Nuclear Engineering Division; Florida International Univ.

    2007-10-18

    The Global Nuclear Energy Partnership (GNEP) is considering a list of reactors and nuclear fuels as part of its chartered initiative. Because many of the candidate materials have not been explored experimentally under the conditions of interest, and in order to economize on program costs, analytical support in the form of combined first principle and mechanistic modeling is highly desirable. The present work is a compilation of mechanistic models developed in order to describe the fission product behavior of irradiated nuclear fuel. The mechanistic nature of the model development allows for the possibility of describing a range of nuclear fuels under varying operating conditions. Key sources include the FASTGRASS code with an application to UO{sub 2} power reactor fuel and the Dispersion Analysis Research Tool (DART ) with an application to uranium-silicide and uranium-molybdenum research reactor fuel. Described behavior mechanisms are divided into subdivisions treating fundamental materials processes under normal operation as well as the effect of transient heating conditions on these processes. Model topics discussed include intra- and intergranular gas-atom and bubble diffusion, bubble nucleation and growth, gas-atom re-solution, fuel swelling and ?scion gas release. In addition, the effect of an evolving microstructure on these processes (e.g., irradiation-induced recrystallization) is considered. The uranium-alloy fuel, U-xPu-Zr, is investigated and behavior mechanisms are proposed for swelling in the {alpha}-, intermediate- and {gamma}-uranium zones of this fuel. The work reviews the FASTGRASS kinetic/mechanistic description of volatile ?scion products and, separately, the basis for the DART calculation of bubble behavior in amorphous fuels. Development areas and applications for physical nuclear fuel models are identified.

  14. 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)

  15. Advanced research reactor fuel development

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Kyu; Pak, H. D.; Kim, K. H. [and others

    2000-05-01

    -plates will be conducted in the Advanced Test Reactor(ATR). 49 compacts with a uranium density of 8 gU/cc consist of 7 different atomized uranium-molybdenum alloy powders. The tensile strength increased and the elongation decreased with increasing the volume fraction of U-10Mo powders in dispersion fuel. The tensile strength was lower and elongation was larger in dispersion fuel using atomized U-10Mo powders than that using comminuted fuel powders. The green strength of the comminuted powder compacts was about twice as large as that of the atomized powder compacts. It is suggested that the compacting condition required to fabricate the atomized powder compacts is over the 350MPa. The comminuted irregular shaped particles and smaller particle size of fuel powders showed improved homogeneity of powder mixture. The homogeneity of powder mixtures increased to a minimum at approximately 0.10 wt% moisture and then decreased with moisture content.

  16. Low enriched uranium foil targets with different geometries for the production of Molybdenum-99 in the BMR (Brazilian Multipurpose Reactor)

    Energy Technology Data Exchange (ETDEWEB)

    Domingos, Douglas B.; Silva, Antonio T. e; Joao, Thiago G.; Muniz, Rafael O.R.; Coelho, Talita S., E-mail: teixeira@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    A new research reactor is being planned in Brazil to take care of the demand of radiopharmaceuticals in the country and conduct research in various areas. This new reactor, the Brazilian Multipurpose Reactor (RMB), planned for 30 MW, is now in the conception design phase. Two low enriched (<20% {sup 235}U) metallic uranium foil targets (cylinder and plate geometries) are being considered for production of Molybdenum-99 ({sup 99}Mo) by fission. Neutronic and thermal-hydraulics calculations were performed to compare the production of {sup 99}Mo for these targets in the RMB and to determine the temperatures achieved in the targets. For the neutronic calculations were utilized the computer codes HAMMER-TECHNION, CITATION and SCALE and for the thermal-hydraulics calculations were utilized the computer codes MTRCR-IEA-R1 and ANSYS CFX. (author)

  17. Validation of KENO V. a. and two cross-section libraries for criticality calculations of low-enriched uranium systems

    Energy Technology Data Exchange (ETDEWEB)

    Easter, M.E.

    1985-07-01

    The SCALE code system, utilizing the Monte Carlo computer code KENO V.a, was employed to calculate 37 critical experiments. The critical assemblies had /sup 235/U enrichments of 5% or less and cover a variety of geometries and materials. Values of k/sub eff/ were calculated using two different results using either of the cross-section libraries. The 16-energy-group Hansen-Roach and the 27-energy-group ENDF/B-IV cross-section libraries, available in SCALE, were used in this validation study, and both give good results for the experiments considered. It is concluded that the code and cross sections are adequate for low-enriched uranium systems and that reliable criticality safety calculations can be made for such systems provided the limits of validated applicability are not exceeded.

  18. Environmental assessment: Transfer of normal and low-enriched uranium billets to the United Kingdom, Hanford Site, Richland, Washington

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-11-01

    Under the auspices of an agreement between the U.S. and the United Kingdom, the U.S. Department of Energy (DOE) has an opportunity to transfer approximately 710,000 kilograms (1,562,000 pounds) of unneeded normal and low-enriched uranium (LEU) to the United Kingdom; thus, reducing long-term surveillance and maintenance burdens at the Hanford Site. The material, in the form of billets, is controlled by DOE`s Defense Programs, and is presently stored as surplus material in the 300 Area of the Hanford Site. The United Kingdom has expressed a need for the billets. The surplus uranium billets are currently stored in wooden shipping containers in secured facilities in the 300 Area at the Hanford Site (the 303-B and 303-G storage facilities). There are 482 billets at an enrichment level (based on uranium-235 content) of 0.71 weight-percent. This enrichment level is normal uranium; that is, uranium having 0.711 as the percentage by weight of uranium-235 as occurring in nature. There are 3,242 billets at an enrichment level of 0.95 weight-percent (i.e., low-enriched uranium). This inventory represents a total of approximately 532 curies. The facilities are routinely monitored. The dose rate on contact of a uranium billet is approximately 8 millirem per hour. The dose rate on contact of a wooden shipping container containing 4 billets is approximately 4 millirem per hour. The dose rate at the exterior of the storage facilities is indistinguishable from background levels.

  19. Environmental assessment: Transfer of normal and low-enriched uranium billets to the United Kingdom, Hanford Site, Richland, Washington

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-11-01

    Under the auspices of an agreement between the U.S. and the United Kingdom, the U.S. Department of Energy (DOE) has an opportunity to transfer approximately 710,000 kilograms (1,562,000 pounds) of unneeded normal and low-enriched uranium (LEU) to the United Kingdom; thus, reducing long-term surveillance and maintenance burdens at the Hanford Site. The material, in the form of billets, is controlled by DOE`s Defense Programs, and is presently stored as surplus material in the 300 Area of the Hanford Site. The United Kingdom has expressed a need for the billets. The surplus uranium billets are currently stored in wooden shipping containers in secured facilities in the 300 Area at the Hanford Site (the 303-B and 303-G storage facilities). There are 482 billets at an enrichment level (based on uranium-235 content) of 0.71 weight-percent. This enrichment level is normal uranium; that is, uranium having 0.711 as the percentage by weight of uranium-235 as occurring in nature. There are 3,242 billets at an enrichment level of 0.95 weight-percent (i.e., low-enriched uranium). This inventory represents a total of approximately 532 curies. The facilities are routinely monitored. The dose rate on contact of a uranium billet is approximately 8 millirem per hour. The dose rate on contact of a wooden shipping container containing 4 billets is approximately 4 millirem per hour. The dose rate at the exterior of the storage facilities is indistinguishable from background levels.

  20. Preliminary Accident Analyses for Conversion of the Massachusetts Institute of Technology Reactor (MITR) from Highly Enriched to Low Enriched Uranium

    Energy Technology Data Exchange (ETDEWEB)

    Dunn, Floyd E. [Argonne National Lab. (ANL), Argonne, IL (United States); Olson, Arne P. [Argonne National Lab. (ANL), Argonne, IL (United States); Wilson, Erik H. [Argonne National Lab. (ANL), Argonne, IL (United States); Sun, Kaichao S. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Newton, Jr., Thomas H. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Hu, Lin-wen [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

    2013-09-30

    The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. It delivers a neutron flux comparable to current LWR power reactors in a compact 6 MW core using Highly Enriched Uranium (HEU) fuel. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context most research and test reactors, both domestic and international, have started a program of conversion to the use of 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 MITR. This report presents the preliminary accident analyses for MITR cores fueled with LEU monolithic U-Mo alloy fuel with 10 wt% Mo. Preliminary results demonstrate adequate performance, including thermal margin to expected safety limits, for the LEU accident scenarios analyzed.

  1. Neutron and hard X-ray diffraction studies of the isothermal transformation kinetics in the research reactor fuel candidate U-8 wt%Mo.

    Science.gov (United States)

    Säubert, Steffen; Jungwirth, Rainer; Zweifel, Tobias; Hofmann, Michael; Hoelzel, Markus; Petry, Winfried

    2016-06-01

    Exposing uranium-molybdenum alloys (UMo) retained in the γ phase to elevated temperatures leads to transformation reactions during which the γ-UMo phase decomposes into the thermal equilibrium phases, i.e. U2Mo and α-U. Since α-U is not suitable for a nuclear fuel exposed to high burn-up, it is necessary to retain the γ-UMo phase during the production process of the fuel elements for modern high-performance research reactors. The present work deals with the isothermal transformation kinetics in U-8 wt%Mo alloys for temperatures between 673 and 798 K and annealing durations of up to 48 h. Annealed samples were examined at room temperature using either X-ray or neutron diffraction to determine the phase composition after thermal treatment, and in situ annealing studies disclosed the onset of phase decomposition. While for temperatures of 698 and 673 K the start of decomposition is delayed, for higher temperatures the first signs of transformation are already observable within 3 h of annealing. The typical C-shaped curves in a time-temperature-transformation (TTT) diagram for both the start and the end of phase decomposition could be determined in the observed temperature regime. Therefore, a revised TTT diagram for U-8 wt%Mo between 673 and 798 K and annealing durations of up to 48 h is proposed.

  2. 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.

  3. Target-fueled nuclear reactor for medical isotope production

    Science.gov (United States)

    Coats, Richard L.; Parma, Edward J.

    2017-06-27

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

  4. 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).

  5. 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)

  6. Preliminary Assessment of the Impact on Reactor Vessel dpa Rates Due to Installation of a Proposed Low Enriched Uranium (LEU) Core in the High Flux Isotope Reactor (HFIR)

    Energy Technology Data Exchange (ETDEWEB)

    Daily, Charles R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-10-01

    An assessment of the impact on the High Flux Isotope Reactor (HFIR) reactor vessel (RV) displacements-per-atom (dpa) rates due to operations with the proposed low enriched uranium (LEU) core described by Ilas and Primm has been performed and is presented herein. The analyses documented herein support the conclusion that conversion of HFIR to low-enriched uranium (LEU) core operations using the LEU core design of Ilas and Primm will have no negative impact on HFIR RV dpa rates. Since its inception, HFIR has been operated with highly enriched uranium (HEU) cores. As part of an effort sponsored by the National Nuclear Security Administration (NNSA), conversion to LEU cores is being considered for future HFIR operations. The HFIR LEU configurations analyzed are consistent with the LEU core models used by Ilas and Primm and the HEU balance-of-plant models used by Risner and Blakeman in the latest analyses performed to support the HFIR materials surveillance program. The Risner and Blakeman analyses, as well as the studies documented herein, are the first to apply the hybrid transport methods available in the Automated Variance reduction Generator (ADVANTG) code to HFIR RV dpa rate calculations. These calculations have been performed on the Oak Ridge National Laboratory (ORNL) Institutional Cluster (OIC) with version 1.60 of the Monte Carlo N-Particle 5 (MCNP5) computer code.

  7. Fuel flexible fuel injector

    Science.gov (United States)

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

    2015-02-03

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

  8. Assembly and Irradiation Modeling of Residual Stresses in Low-Enriched Uranium Foil-Based Annular Targets for Molybdenum-99 Production

    Directory of Open Access Journals (Sweden)

    Srisharan G. Govindarajan

    2013-01-01

    Full Text Available This paper considers a composite cylindrical structure, with low-enriched uranium (LEU foil enclosed between two aluminum 6061-T6 cylinders. A recess is cut all around the outer circumference of the inner tube to accommodate the LEU foil of open-cross section. To obtain perfect contact at the interfaces of the foil and the tubes, an internal pressure is applied to the inner tube, thereby plastically and elastically deforming it. The residual stresses resulting from the assembly process are used along with a thermal stress model to predict the stress margins in the cladding during irradiation. The whole process was simulated as a steady-state two-dimensional problem using the commercial finite element code Abaqus FEA. The irradiation behavior of the annular target has been presented, and the effect of the assembly residual stresses has been discussed.

  9. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Rasmussen B.

    2010-01-01

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

  11. Facts and issues of direct disposal of spent fuel; Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Parks, P.B.

    1993-10-01

    This report reviews those facts and issues that affect the direct disposal of spent reactor fuels. It is intended as a resource document for those impacted by the current Department of Energy (DOE) guidance that calls for the cessation of fuel reprocessing. It is not intended as a study of the specific impacts (schedules and costs) to the Savannah River Site (SRS) alone. Commercial fuels, other low enriched fuels, highly enriched defense-production, research, and naval reactor fuels are included in this survey, except as prevented by rules on classification.

  12. 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%.

  13. Low enriched uranium UAl{sub X}-Al targets for the production of Molybdenum-99 in the IEA-R1 and RMB reactors

    Energy Technology Data Exchange (ETDEWEB)

    Domingos, Douglas B.; Silva, Antonio T. e; Joao, Thiago G.; Silva, Jose Eduardo R. da, E-mail: teixeira@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Nishiyama, Pedro J.B. de O., E-mail: pedro.julio@ctmsp.mar.mil.b [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), SP (Brazil)

    2011-07-01

    The IEA-R1 reactor of IPEN/CNEN-SP in Brazil is a pool type research reactor cooled and moderated by demineralized water and having Beryllium and Graphite as reflectors. In 1997 the reactor received the operating licensing for 5 MW. A new research reactor is being planned in Brazil to replace the IEA-R1 reactor. This new reactor, the Brazilian Multipurpose Reactor (RMB), planned for 30 MW, is now in the conception design phase. Low enriched uranium (LEU) (<20% {sup 235}U) UAl{sub x} dispersed in Al targets are being considered for production of Molybdenum-99 ({sup 99}Mo) by fission. Neutronic and thermal-hydraulics calculations were performed, respectively, to compare the production of {sup 99}Mo for these targets in IEA-R1 reactor and RMB and to determine the temperatures achieved in the UAl{sub x}-Al targets during irradiation. For the neutronic calculations were utilized the computer codes HAMMER-TECHNION, CITATION and SCALE and for the thermal-hydraulics calculations was utilized the computer code MTRCR-IEAR1. (author)

  14. 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.

  15. Theory of Antineutrino Monitoring of Burning MOX Plutonium Fuels

    CERN Document Server

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

    2011-01-01

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

  16. 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.

  17. Non-destructive Quantitative Phase Analysis and Microstructural Characterization of Zirconium Coated U-10Mo Fuel Foils via Neutron Diffraction

    Energy Technology Data Exchange (ETDEWEB)

    Cummins, Dustin Ray [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Vogel, Sven C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hollis, Kendall Jon [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Brown, Donald William [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dombrowski, David E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-10-18

    This report uses neutron diffraction to investigate the crystal phase composition of uranium-molybdenum alloy foils (U-10Mo) for the CONVERT MP-1 Reactor Conversion Project, and determines the effect on alpha-uranium contamination following the deposition of a Zr metal diffusion layer by various methods: plasma spray deposition of Zr powders at LANL and hot co-rolling with Zr foils at BWXT. In summary, there is minimal decomposition of the gamma phase U-10Mo foil to alpha phase contamination following both plasma spraying and hot co-rolling. The average unit cell volume, i.e. lattice spacing, of the Zr layer can be mathematically extracted from the diffraction data; co-rolled Zr matches well with literature values of bulk Zr, while plasma sprayed Zr shows a slight increase in the lattice spacing, indicative of interstitial oxygen in the lattice. Neutron diffraction is a beneficial alternative to conventional methods of phase composition, i.e. x ray diffraction (XRD) and destructive metallography. XRD has minimal penetration depth in high atomic number materials, particularly uranium, and can only probe the first few microns of the fuel plate; neutrons pass completely through the foil, allowing for bulk analysis of the foil composition and no issues with addition of cladding layers, as in the final, aluminum-clad reactor fuel plates. Destructive metallography requires skilled technicians, cutting of the foil into small sections, hazardous etching conditions, long polishing and microscopy times, etc.; the neutron diffraction system has an automated sample loader and can fit larger foils, so there is minimal analysis preparation; the total spectrum acquisition time is ~ 1 hour per sample. The neutron diffraction results are limited by spectra refinement/calculation times and the availability of the neutron beam source. In the case of LANSCE at Los Alamos, the beam operates ~50% of the year. Following the lessons learned from these preliminary results, optimizations to

  18. Non-destructive Quantitative Phase Analysis and Microstructural Characterization of Zirconium Coated U-10Mo Fuel Foils via Neutron Diffraction

    Energy Technology Data Exchange (ETDEWEB)

    Cummins, Dustin Ray [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Vogel, Sven C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hollis, Kendall Jon [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Brown, Donald William [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dombrowski, David E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-10-18

    This report uses neutron diffraction to investigate the crystal phase composition of uranium-molybdenum alloy foils (U-10Mo) for the CONVERT MP-1 Reactor Conversion Project and determines the effect on alpha uranium contamination following the deposition of a Zr metal diffusion layer by various methods: plasma spray deposition of Zr powders at LANL and hot co-rolling with Zr foils at BWXT. In summary, there is minimal decomposition of the gamma phase U-10Mo foil to alpha phase contamination following both plasma spraying and hot co-rolling. The average unit cell volume, i.e. lattice spacing, of the Zr layer can be mathematically extracted from the diffraction data; co-rolled Zr matches well with literature values of bulk Zr, while plasma sprayed Zr shows a slight increase in the lattice spacing, indicative of interstitial oxygen in the lattice. Neutron diffraction is a beneficial alternative to conventional methods of phase composition, i.e. x ray diffraction (XRD) and destructive metallography. XRD has minimal penetration depth in high atomic number materials, particularly uranium, and can only probe the first few microns of the fuel plate; neutrons pass completely through the foil, allowing for bulk analysis of the foil composition and no issues with addition of cladding layers, as in the final, aluminum clad reactor fuel plates. Destructive metallography requires skilled technicians, cutting of the foil into small sections, hazardous etching conditions, long polishing and microscopy times, etc.; the neutron diffraction system has an automated sample loader and can fit larger foils, so there is minimal analysis preparation; the total spectrum acquisition time is ~ 1 hour per sample. The neutron diffraction results are limited by spectra refinement/calculation times and the availability of the neutron beam source. In the case of LANSCE at Los Alamos, the beam operates ~50% of the year. Following the lessons learned from these preliminary results, optimizations to

  19. 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.

  20. 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.

  1. 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.

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

    Science.gov (United States)

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

    2017-08-01

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

  3. Third session: new projects; Session 3: Les Projets nouveaux

    Energy Technology Data Exchange (ETDEWEB)

    Iracane, D. [CEA Saclay, Dir. de l' Energie Nucleaire (DEN), 91 - Gif sur Yvette (France); Dupuy, J.P. [AREVA-Technicatome, 13 - Aix en Provence (France); Perotto, G. [AREVA-Framatome-ANP, 13 - Aix en Provence (France); Ithurralde, G. [Electricite de France (EDF), 13 - Aix en Provence (France); Cabrillat, J.C. [CEA Cadarache, Dir. de l' Energie Nucleaire (DEN), 13 - Saint-Paul-lez-Durance (France); Becle, D. [FRAMATOME, AREVA-FANP, 69 - Lyon (France); Pivet, S. [CEA Bruyeres-le-Chatel, 91 (France); Minguet, J.L. [AREVA-Technicatome, 75 - Paris (France)

    2005-07-01

    The RJH (Jules Horowitz reactor) is a 100 MWth research reactor, cooled and moderated by water, that has been designed to operate with high density (8 g U/cm{sup 3}) and low enrichment fuel (19.75%) uranium molybdenum fuel. Its main neutron performances are expected to be: -) maximum fast neutron flux (E > 0.907 MeV) in the sample: 5.10{sup 14} n/cm{sup 2}/s, -) maximum thermal neutron flux (E < 0.625 eV): 4.10{sup 14} n/cm{sup 2}/s, and -) maximal neutron damage: 16 dpa/year. RJH that will be built on the Cea's site of Cadarache, is scheduled to enter into service in 2014, it will take over the activities of the Osiris reactor and its operating life is planned to be over 50 years long. The Cabri research reactor is located on the Cadarache site, it has been operating since 1962 and is now undergoing important upgrading operations whose aim is to make the facility fit to study the behaviour of new nuclear fuels in accidental situations. The RES test reactor is a facility whose aim is to qualify nuclear fuels and core configurations of existing propulsion reactors onboard of aircraft carriers or submarines and to propose innovative concepts for future propulsion reactors. (A.C.)

  4. 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)

  5. 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.

  6. 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

  7. 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)

  8. 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%.

  9. The generation of denatured reactor plutonium by different options of the fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Broeders, C.H.M.; Kessler, G. [Inst. for Neutron Physics and Reactor Technology, Research Center Karlsruhe (Germany)

    2006-11-15

    Denatured (proliferation resistant) reactor plutonium can be generated in a number of different fuel cycle options. First denatured reactor plutonium can be obtained if, instead of low enriched U-235 PWR fuel, re-enriched U-235/U-236 from reprocessed uranium is used (fuel type A). Also the envisaged existing 2,500 t of reactor plutonium (being generated world wide up to the year 2010), mostly stored in intermediate fuel storage facilities at present, could be converted during a transition phase into denatured reactor plutonium by the options fuel type B and D. Denatured reactor plutonium could have the same safeguards standard as present low enriched (<20% U-235) LWR fuel. It could be incinerated by recycling once or twice in PWRs and subsequently by multi-recycling in FRs (CAPRA type or IFRs). Once denatured, such reactor plutonium could remain denatured during multiple recycling. In a PWR, e.g., denatured reactor plutonium could be destroyed at a rate of about 250 kg/GWey. While denatured reactor plutonium could be recycled and incinerated under relieved IAEA safeguards, neptunium would still have to be monitored by the IAEA in future for all cases in which considerable amounts of neptunium are produced. (orig.)

  10. 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.

  11. 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.

  12. FLOW TESTING AND ANALYSIS OF THE FSP-1 EXPERIMENT

    Energy Technology Data Exchange (ETDEWEB)

    Hawkes, Grant L.; Jones, Warren F.; Marcum, Wade; Weiss, Aaron; Howard, Trevor

    2017-06-01

    The U.S. High Performance Research Reactor Conversions fuel development team is focused on developing and qualifying the uranium-molybdenum (U-Mo) alloy monolithic fuel to support conversion of domestic research reactors to low enriched uranium. Several previous irradiations have demonstrated the favorable behavior of the monolithic fuel. The Full Scale Plate 1 (FSP-1) fuel plate experiment will be irradiated in the northeast (NE) flux trap of the Advanced Test Reactor (ATR). This fueled experiment contains six aluminum-clad fuel plates consisting of monolithic U-Mo fuel meat. Flow testing experimentation and hydraulic analysis have been performed on the FSP-1 experiment to be irradiated in the ATR at the Idaho National Laboratory (INL). A flow test experiment mockup of the FSP-1 experiment was completed at Oregon State University. Results of several flow test experiments are compared with analyses. This paper reports and shows hydraulic analyses are nearly identical to the flow test results. A water velocity of 14.0 meters per second is targeted between the fuel plates. Comparisons between FSP-1 measurements and this target will be discussed. This flow rate dominates the flow characteristics of the experiment and model. Separate branch flows have minimal effect on the overall experiment. A square flow orifice was placed to control the flowrate through the experiment. Four different orifices were tested. A flow versus delta P curve for each orifice is reported herein. Fuel plates with depleted uranium in the fuel meat zone were used in one of the flow tests. This test was performed to evaluate flow test vibration with actual fuel meat densities and reported herein. Fuel plate deformation tests were also performed and reported.

  13. CERCA LEU fuel assemblies testing in Maria Reactor - safety analysis summary and testing program scope.

    Energy Technology Data Exchange (ETDEWEB)

    Pytel, K.; Mieleszczenko, W.; Lechniak, J.; Moldysz, A.; Andrzejewski, K.; Kulikowska, T.; Marcinkowska, A.; Garner, P. L.; Hanan, N. A.; Nuclear Engineering Division; Institute of Atomic Energy (Poland)

    2010-03-01

    The presented paper contains neutronic and thermal-hydraulic (for steady and unsteady states) calculation results prepared to support annex to Safety Analysis Report for MARIA reactor in order to obtain approval for program of testing low-enriched uranium (LEU) lead test fuel assemblies (LTFA) manufactured by CERCA. This includes presentation of the limits and operational constraints to be in effect during the fuel testing investigations. Also, the scope of testing program (which began in August 2009), including additional measurements and monitoring procedures, is described.

  14. Comparison of low enriched uranium (UAl{sub x}-Al and U-Ni) targets with different geometries for the production of molybdenum-99 in the RMB (Brazilian multipurpose reactor)

    Energy Technology Data Exchange (ETDEWEB)

    Domingos, Douglas B.; Silva, Antonio T. e; Joao, Thiago G.; Silva, Jose Eduardo R. da; Angelo, Gabriel; Fedorenko, Giuliana G., E-mail: teixeira@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Nishiyama, Pedro J.B. de O., E-mail: pedro.julio@ctmsp.mar.mil.b [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil)

    2011-07-01

    The Brazilian Multipurpose Reactor (RMB), now in the conception design phase, is being designed in Brazil to attend the demand of radiopharmaceuticals in the country and conduct researches in various areas. The new reactor, planned for 30 MW, will replace the IEA-R1 reactor of IPEN-CNEN/SP. Low enriched uranium (<20% {sup 235}U) UAl{sub x} dispersed in Al (plate geometry) and metallic uranium foil targets (plate and cylinder geometries) are being considered for production of Molybdenum-99 ({sup 99}Mo) by fission. Neutronic and thermal-hydraulics calculations were performed to compare the production of {sup 99}Mo for these targets in the RMB. For the neutronic calculations were utilized the computer codes Hammer-Technion, Citation and Scale and for the thermal-hydraulics calculations were utilized the computer code MTRCR-IEAR1 and ANSYS CFX. (author)

  15. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Carbajo, J.J.

    2005-05-27

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

  17. 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…

  18. 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…

  19. Fabrication and Quality Inspection of U-10wt.%Zr Fuel Rod for Irradiation Test

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ki Hwan; Song, Hoon; Oh, Seok Jin; Lee, Jung Won; Park, Jeong Yong; Lee, Chan Bock [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    Metallic fuel has advantages such as simple fabrication procedures, good neutron economy, high thermal conductivity, excellent compatibility with a Na coolant and inherent passive safety. Metal fuels such as U-Zr alloy have been considered as a starting driver fuel for a proto-type Gen-IV sodium cooled fast reactor (PGSFR) in Korea. To confirm the design and fabrication technologies of metallic fuels with FMS cladding for the loading of metallic fuel in PGSFR, an irradiation test will be performed in BOR-60 in Russia in 2016. In this study, U-10wt.%Zr fuel rods using low enrichment uranium (LEU) have been fabricated and inspected in quality for the fuel verification of PGSFR. Fuel slugs per melting batch without casting defects were fabricated by development of the advanced casting technology and evaluation tests. The optimal GTAW welding conditions and parameters were also established through lots of experiments. And, the qualification test carried out to prove the weld quality of end plug welding of the metallic fuel rods. The wire wrapping of metallic fuel rods for the irradiation test was successfully accomplished in KAERI. So, PGSFR fuel rods for the irradiation test in BOR-60 have been soundly fabricated in KAERI.

  20. 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.

  1. Thorium utilization in a small long-life HTR. Part I: Th/U MOX fuel blocks

    Energy Technology Data Exchange (ETDEWEB)

    Ding, Ming, E-mail: dingm2005@gmail.com [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB, Delft (Netherlands); Harbin Engineering University, Nantong Street 145, 150001 Harbin (China); Kloosterman, Jan Leen, E-mail: j.l.kloosterman@tudelft.nl [Delft University of Technology, Reactor Institute Delft, Mekelweg 15, 2629 JB, Delft (Netherlands)

    2014-02-15

    Highlights: • We propose thorium MOX (TMOX) fuel blocks for a small block-type HTR. • The TMOX fuel blocks with low-enriched uranium are recommended. • More thorium decreases the reactivity swing of the TMOX fuel blocks. • Thorium reduces the negative temperature coefficient of the TMOX fuel blocks. • Thorium increases the conversion ratio of the TMOX fuel blocks. - Abstract: The U-Battery is a small, long-life and transportable high temperature gas-cooled reactor (HTR). The neutronic features of a typical fuel block with uranium and thorium have been investigated for a application of the U-Battery, by parametrically analyzing the composition and geometric parameters. The type of fuel block is defined as Th/U MOX fuel block because uranium and thorium are assumed to be mixed in each fuel kernel as a form of (Th,U)O{sub 2}. If the initially loaded mass of U-235 is mostly consumed in the early period of the lifetime of Th/U MOX fuel block, low-enriched uranium (LEU) as ignited fuel will not largely reduce the neutronic performance of the Th/U MOX fuel block, compared with high-enriched uranium. The radii of fuel kernels and fuel compacts and packing fraction of TRISO particles determine the atomic ratio of the carbon to heavy metal. When the ratio is smaller than 400, the difference among them due to double heterogeneous effects can be neglected for the Th/U MOX fuel block. In the range between 200 and 400, the reactivity swing of the Th/U MOX fuel block during 10 years is sufficiently small. The magnitude of the negative reactivity temperature coefficients of the Th/U MOX fuel block decreases by 20–45%, which is positive to reduce temperature defect of the Th/U MOX fuel block. The conversion ratio (CR) of the fuel increases from 0.48 (typical CR of the LEU-fueled U-Battery) to 0.78. The larger conversion ratio of the Th/U MOX fuel block reduces the reactivity swing during 10 years for the U-Battery.

  2. Calibration of spent fuel measurement assembly

    Science.gov (United States)

    Koleska, Michal; Viererbl, Ladislav; Marek, Milan

    2014-11-01

    The LVR-15 research reactor (Czech Republic) had been converted from the highly enriched IRT-2M to the low enriched IRT-4M fuel. For the possibility of the independent pre-transport evaluation of IRT-2M burnup, a spectrometric system was developed. This spectrometric system consists of the fuel holder, the collimator and the portable Canberra Big MAC HPGe (High Purity Germanium) detector. In order to have well reproducible and reliable experimental data for modeling of the measurement system, calibration with the 110mAg isotope with known activity was performed. This isotope was chosen for having energies similar to isotopes measured in fuel assemblies. The 110mAg isotope was prepared by irradiating of the silver foil in LVR-15 research reactor; its activity was evaluated in the LVR-15's spectrometric laboratory. From the measured data, an efficiency curve of the spectrometric system has been determined. The experimental data were compared to the calculation results with the MCNPX model of the spectrometric system.

  3. Proliferation resistance of small modular reactors fuels

    Energy Technology Data Exchange (ETDEWEB)

    Polidoro, F.; Parozzi, F. [RSE - Ricerca sul Sistema Energetico,Via Rubattino 54, 20134, Milano (Italy); Fassnacht, F.; Kuett, M.; Englert, M. [IANUS, Darmstadt University of Technology, Alexanderstr. 35, D-64283 Darmstadt (Germany)

    2013-07-01

    In this paper the proliferation resistance of different types of Small Modular Reactors (SMRs) has been examined and classified with criteria available in the literature. In the first part of the study, the level of proliferation attractiveness of traditional low-enriched UO{sub 2} and MOX fuels to be used in SMRs based on pressurized water technology has been analyzed. On the basis of numerical simulations both cores show significant proliferation risks. Although the MOX core is less proliferation prone in comparison to the UO{sub 2} core, it still can be highly attractive for diversion or undeclared production of nuclear material. In the second part of the paper, calculations to assess the proliferation attractiveness of fuel in typical small sodium cooled fast reactor show that proliferation risks from spent fuel cannot be neglected. The core contains a highly attractive plutonium composition during the whole life cycle. Despite some aspects of the design like the sealed core that enables easy detection of unauthorized withdrawal of fissile material and enhances proliferation resistance, in case of open Non-Proliferation Treaty break-out, weapon-grade plutonium in sufficient quantities could be extracted from the reactor core.

  4. Complexes of osmium, uranium, molybdenum, and tungsten with the catechol amines adrenaline, noradrenaline, dopamine, dopa, and isoproterenol

    Energy Technology Data Exchange (ETDEWEB)

    El-Hendawy, A.M.; Griffith, W.P.; Pumphrey, C.A.

    1988-07-01

    New complexes of the form trans-(OsO/sub 2/L/sub 2/)/sup 2-/ and UO/sub 2/Lcenter dotnH/sub 2/O (H/sub 2/L = adrenaline (H/sub 2/ad), noradrenaline (H/sub 2/nad), dopamine (H/sub 2/dpm), dopa (H/sub 2/dp), and isoproterenol (H/sub 2/prot)) are reported, as are cis(MO/sub 2/L/sub 2/)/sup 2-/(L = nad, dp, prot for M = Mo or W, and ad for M = W), (MO/sub 2/(Hdpm)/sub 2/) (M = Mo or W), and (Mo/sub 2/O/sub 5/(Had)/sub 2/). The structures of these species are discussed on the basis of their Raman, infrared, /sup 1/H and /sup 13/C n.m.r. spectra.

  5. 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.

  6. 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....

  7. 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.

  8. Testing of LWR fuel rods to support criticality safety analysis of transport accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Purcell, P.C. [BNFL International Transport, Spent Fuel Services (United Kingdom); Dallongeville, M. [COGEMA Logistics (AREVA Group) (France)

    2004-07-01

    For the transport of low enriched materials, criticality safety may be demonstrated by applying pessimistic modelling assumptions that bound any realistic case. Where Light Water Reactor (LWR) fuel is being transported, enrichment levels are usually too high to permit this approach and more realistic data is needed. This requires a method by which the response of LWR fuel under impact accident conditions can be approximated or bounded. In 2000, BNFL and COGEMA LOGISTICS jointly commenced the Fuel Integrity Project (FIP) whose objective was to develop such methods. COGEMA LOGISTICS were well advanced with a method for determining the impact response of unirradiated fuel, but required further test data before acceptance by the Transport Regulators. The joint project team extensively discussed the required inputs to the FIP, from which it was agreed that BNFL would organise new tests on both unirradiated and irradiated fuel samples and COGEMA LOGISTICS would take major responsibility for evaluating the test results. Tests on unirradiated fuel rod samples involved both dynamic and quasi-static loading on fuel samples. PWR fuel rods loaded with uranium pellets were dropped vertically from 9m onto a rigid target and this was repeated on BWR fuel rods, similar tests on empty fuel rods were also conducted. Quasi-static tests were conducted on 530 mm long PWR and BWR fuel specimens under axial loading. Tests on irradiated fuel samples were conducted on high burn-up fuel rods of both PWR and BWR types. These were believed original to the FIP project and involved applying bending loads to simply supported pressurised rod specimens. In one test the fuel rod was heated to nearly 500oC during loading, all specimens were subject to axial impact before testing. Considerable experience of fuel rod testing and new data was gained from this test programme.

  9. Coated U(Mo) Fuel: As-Fabricated Microstructures

    Energy Technology Data Exchange (ETDEWEB)

    Emmanuel Perez; Dennis D. Keiser, Jr.; Ann Leenaers; Sven Van den Berghe; Tom Wiencek

    2014-04-01

    As part of the development of low-enriched uranium fuels, fuel plates have recently been tested in the BR-2 reactor as part of the SELENIUM experiment. These fuel plates contained fuel particles with either Si or ZrN thin film coating (up to 1 µm thickness) around the U-7Mo fuel particles. In order to best understand irradiation performance, it is important to determine the starting microstructure that can be observed in as-fabricated fuel plates. To this end, detailed microstructural characterization was performed on ZrN and Si-coated U-7Mo powder in samples taken from AA6061-clad fuel plates fabricated at 500°C. Of interest was the condition of the thin film coatings after fabrication at a relatively high temperature. Both scanning electron microscopy and transmission electron microscopy were employed. The ZrN thin film coating was observed to consist of columns comprised of very fine ZrN grains. Relatively large amounts of porosity could be found in some areas of the thin film, along with an enrichment of oxygen around each of the the ZrN columns. In the case of the pure Si thin film coating sample, a (U,Mo,Al,Si) interaction layer was observed around the U-7Mo particles. Apparently, the Si reacted with the U-7Mo and Al matrix during fuel plate fabrication at 500°C to form this layer. The microstructure of the formed layer is very similar to those that form in U-7Mo versus Al-Si alloy diffusion couples annealed at higher temperatures and as-fabricated U-7Mo dispersion fuel plates with Al-Si alloy matrix fabricated at 500°C.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-11-01

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

  11. Evaluation of HFIR LEU Fuel Using the COMSOL Multiphysics Platform

    Energy Technology Data Exchange (ETDEWEB)

    Primm, Trent [ORNL; Ruggles, Arthur [ORNL; Freels, James D [ORNL

    2009-03-01

    A finite element computational approach to simulation of the High Flux Isotope Reactor (HFIR) Core Thermal-Fluid behavior is developed. These models were developed to facilitate design of a low enriched core for the HFIR, which will have different axial and radial flux profiles from the current HEU core and thus will require fuel and poison load optimization. This report outlines a stepwise implementation of this modeling approach using the commercial finite element code, COMSOL, with initial assessment of fuel, poison and clad conduction modeling capability, followed by assessment of mating of the fuel conduction models to a one dimensional fluid model typical of legacy simulation techniques for the HFIR core. The model is then extended to fully couple 2-dimensional conduction in the fuel to a 2-dimensional thermo-fluid model of the coolant for a HFIR core cooling sub-channel with additional assessment of simulation outcomes. Finally, 3-dimensional simulations of a fuel plate and cooling channel are presented.

  12. Production and Characterization of Atomized U-Mo Powder by the Rotating Electrode Process

    Energy Technology Data Exchange (ETDEWEB)

    C.R. Clark; B.R. Muntifering; J.F. Jue

    2007-09-01

    In order to produce feedstock fuel powder for irradiation testing, the Idaho National Laboratory has produced a rotating electrode type atomizer to fabricate uranium-molybdenum alloy fuel. Operating with the appropriate parameters, this laboratory-scale atomizer produces fuel in the desired size range for the RERTR dispersion experiments. Analysis of the powder shows a homogenous, rapidly solidified microstructure with fine equiaxed grains. This powder has been used to produce irradiation experiments to further test adjusted matrix U-Mo dispersion fuel.

  13. 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...... and nuclear fuel-based energy technologies....

  14. 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

  15. 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.

  16. 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

  17. 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

  18. Preliminary plan for the qualification of the LEU/Th fuel cycle for the Fort St. Vrain HTGR

    Energy Technology Data Exchange (ETDEWEB)

    Gulden, T.D.; Gainey, B.W.; Altschwager, C.J. (comps.)

    1980-03-01

    This plan was prepared to ensure that low-enriched uranium/thorium (LEU/Th) would be available as a backup to the highly enriched uranium/thorium (HEU/Th) fuel cycle currently being used in the Fort St. Vrain (FSV) high-temperature gas-cooled reactor (HTGR) in the event that the US nonproliferation policies require it. It describes the program that would be required to develop, qualify, and introduce an LEU/Th fuel cycle into the FSV HTGR on the earliest possible and most optimistic schedule. The results of the study indicate that licensing of the LEU/Th fuel cycle for FSV could be completed and fuel manufacturing could begin about 4.5 years from inception of the program.

  19. Final Report on Two-Stage Fast Spectrum Fuel Cycle Options

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Won Sik [Purdue Univ., West Lafayette, IN (United States); Lin, C. S. [Purdue Univ., West Lafayette, IN (United States); Hader, J. S. [Purdue Univ., West Lafayette, IN (United States); Park, T. K. [Purdue Univ., West Lafayette, IN (United States); Deng, P. [Purdue Univ., West Lafayette, IN (United States); Yang, G. [Purdue Univ., West Lafayette, IN (United States); Jung, Y. S. [Purdue Univ., West Lafayette, IN (United States); Kim, T. K. [Argonne National Lab. (ANL), Argonne, IL (United States); Stauff, N. E. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-01-30

    This report presents the performance characteristics of two “two-stage” fast spectrum fuel cycle options proposed to enhance uranium resource utilization and to reduce nuclear waste generation. One is a two-stage fast spectrum fuel cycle option of continuous recycle of plutonium (Pu) in a fast reactor (FR) and subsequent burning of minor actinides (MAs) in an accelerator-driven system (ADS). The first stage is a sodium-cooled FR fuel cycle starting with low-enriched uranium (LEU) fuel; at the equilibrium cycle, the FR is operated using the recovered Pu and natural uranium without supporting LEU. Pu and uranium (U) are co-extracted from the discharged fuel and recycled in the first stage, and the recovered MAs are sent to the second stage. The second stage is a sodium-cooled ADS in which MAs are burned in an inert matrix fuel form. The discharged fuel of ADS is reprocessed, and all the recovered heavy metals (HMs) are recycled into the ADS. The other is a two-stage FR/ADS fuel cycle option with MA targets loaded in the FR. The recovered MAs are not directly sent to ADS, but partially incinerated in the FR in order to reduce the amount of MAs to be sent to the ADS. This is a heterogeneous recycling option of transuranic (TRU) elements

  20. Interim assessment of the denatured /sup 233/U fuel cycle: feasibility and nonproliferation characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Abbott, L.S.; Bartine, D.E.; Burns, T.J. (eds.)

    1978-12-01

    A fuel cycle that employs /sup 233/U denatured with /sup 238/U and mixed with thorium fertile material is examined with respect to its proliferation-resistance characteristics and its technical and economic feasibility. The rationale for considering the denatured /sup 233/U fuel cycle is presented, and the impact of the denatured fuel on the performance of Light-Water Reactors, Spectral-Shift-Controlled Reactors, Gas-Cooled Reactors, Heavy-Water Reactors, and Fast Breeder Reactors is discussed. The scope of the R, D and D programs to commercialize these reactors and their associated fuel cycles is also summarized and the resource requirements and economics of denatured /sup 233/U cycles are compared to those of the conventional Pu/U cycle. In addition, several nuclear power systems that employ denatured /sup 233/U fuel and are based on the energy center concept are evaluated. Under this concept, dispersed power reactors fueled with denatured or low-enriched uranium fuel are supported by secure energy centers in which sensitive activities of the nuclear cycle are performed. These activities include /sup 233/U production by Pu-fueled transmuters (thermal or fast reactors) and reprocessing. A summary chapter presents the most significant conclusions from the study and recommends areas for future work.

  1. 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

  2. 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.

  3. Calculational Benchmark Problems for VVER-1000 Mixed Oxide Fuel Cycle

    Energy Technology Data Exchange (ETDEWEB)

    Emmett, M.B.

    2000-03-17

    Standard problems were created to test the ability of American and Russian computational methods and data regarding the analysis of the storage and handling of Russian pressurized water reactor (VVER) mixed oxide fuel. Criticality safety and radiation shielding problems were analyzed. Analysis of American and Russian multiplication factors for fresh fuel storage for low-enriched uranium (UOX), weapons- (MOX-W) and reactor-grade (MOX-R) MOX differ by less than 2% for all variations of water density. For shielding calculations for fresh fuel, the ORNL results for the neutron source differ from the Russian results by less than 1% for UOX and MOX-R and by approximately 3% for MOX-W. For shielding calculations for fresh fuel assemblies, neutron dose rates at the surface of the assemblies differ from the Russian results by 5% to 9%; the level of agreement for gamma dose varies depending on the type of fuel, with UOX differing by the largest amount. The use of different gamma group structures and instantaneous versus asymptotic decay assumptions also complicate the comparison. For the calculation of dose rates from spent fuel in a shipping cask, the neutron source for UOX after 3-year cooling is within 1% and for MOX-W within 5% of one of the Russian results while the MOX-R difference is the largest at over 10%. These studies are a portion of the documentation required by the Russian nuclear regulatory authority, GAN, in order to certify Russian programs and data as being acceptably accurate for the analysis of mixed oxide fuels.

  4. Closed fuel cycle with increased fuel burn-up and economy applying of thorium resources

    Science.gov (United States)

    Kulikov, G. G.; Apse, V. A.

    2017-01-01

    The possible role of existing thorium reserves in the Russian Federation on engaging thorium in being currently closed (U-Pu)-fuel cycle of nuclear power of the country is considered. The application efficiency of thermonuclear neutron sources with thorium blanket for the economical use of existing thorium reserves is demonstrated. The aim of the work is to find solutions of such major tasks as the reduction of both front-end and back-end of nuclear fuel cycle and an enhancing its protection against the uncontrolled proliferation of fissile materials by means of the smallest changes in the fuel cycle. During implementation of the work we analyzed the results obtained earlier by the authors, brought new information on the number of thorium available in the Russian Federation and made further assessments. On the basis of proposal on the inclusion of hybrid reactors with Th-blanket into the future nuclear power for the production of light uranium fraction 232+233+234U, and 231Pa, we obtained the following results: 1. The fuel cycle will shift from fissile 235U to 233U which is more attractive for thermal power reactors. 2. The light uranium fraction is the most "protected" in the uranium component of fuel and mixed with regenerated uranium will in addition become a low enriched uranium fuel, that will weaken the problem of uncontrolled proliferation of fissile materials. 3. 231Pa doping into the fuel stabilizes its multiplying properties that will allow us to implement long-term fuel residence time and eventually to increase the export potential of all nuclear power technologies. 4. The thorium reserves being near city Krasnoufimsk (Russia) are large enough for operation of large-scale nuclear power of the Russian Federation of 70 GWe capacity during more than a quarter century under assumption that thorium is loaded into blankets of hybrid TNS only. The general conclusion: the inclusion of a small number of hybrid reactors with Th-blanket into the future nuclear

  5. 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)

  6. 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...

  7. Evaluation of design parameters for TRISO-coated fuel particles to establish manufacturing critical limits using PARFUME

    Science.gov (United States)

    Skerjanc, William F.; Maki, John T.; Collin, Blaise P.; Petti, David A.

    2016-02-01

    The success of modular high temperature gas-cooled reactors is highly dependent on the performance of the tristructural-isotopic (TRISO) coated fuel particle and the quality to which it can be manufactured. During irradiation, TRISO-coated fuel particles act as a pressure vessel to contain fission gas and mitigate the diffusion of fission products to the coolant boundary. The fuel specifications place limits on key attributes to minimize fuel particle failure under irradiation and postulated accident conditions. PARFUME (an integrated mechanistic coated particle fuel performance code developed at the Idaho National Laboratory) was used to calculate fuel particle failure probabilities. By systematically varying key TRISO-coated particle attributes, failure probability functions were developed to understand how each attribute contributes to fuel particle failure. Critical manufacturing limits were calculated for the key attributes of a low enriched TRISO-coated nuclear fuel particle with a kernel diameter of 425 μm. These critical manufacturing limits identify ranges beyond where an increase in fuel particle failure probability is expected to occur.

  8. Transition Analysis of Promising U.S. Future Fuel Cycles Using ORION

    Energy Technology Data Exchange (ETDEWEB)

    Sunny, Eva E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Worrall, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Peterson, Joshua L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gehin, Jess C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gregg, Robert [Preston Laboratory, Springfields Works (United Kingdom)

    2015-01-01

    The US Department of Energy Office of Fuel Cycle Technologies performed an evaluation and screening (E&S) study of nuclear fuel cycle options to help prioritize future research and development decisions. Previous work for this E&S study focused on establishing equilibrium conditions for analysis examples of 40 nuclear fuel cycle evaluation groups (EGs) and evaluating their performance according to a set of 22 standardized metrics. Following the E&S study, additional studies are being conducted to assess transitioning from the current US fuel cycle to future fuel cycle options identified by the E&S study as being most promising. These studies help inform decisions on how to effectively achieve full transition, estimate the length of time needed to undergo transition from the current fuel cycle, and evaluate performance of nuclear systems and facilities in place during the transition. These studies also help identify any barriers to achieve transition. Oak Ridge National Laboratory (ORNL) Fuel Cycle Options Campaign team used ORION to analyze the transition pathway from the existing US nuclear fuel cycle—the once-through use of low-enriched-uranium (LEU) fuel in thermal-spectrum light water reactors (LWRs) —to a new fuel cycle with continuous recycling of plutonium and uranium in sodium fast reactors (SFRs). This paper discusses the analysis of the transition from an LWR to an SFR fleet using ORION, highlights the role of lifetime extensions of existing LWRs to aid transition, and discusses how a slight delay in SFR deployment can actually reduce the time to achieve an equilibrium fuel cycle.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-08-01

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

  10. 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.

  11. Simulated fission product-SiC interaction in Triso-coated LEU or MEU HTGR fuel particles

    Energy Technology Data Exchange (ETDEWEB)

    Pearson, R.L.; Lindemer, T.B.; Beahm, E.C.

    1980-11-01

    Proliferation issues relating to the use of highly enriched uranium (HEU) have led to an evaluation of the fission product-SiC interaction problems that might arise if low enriched uranium (LEU) or medium enriched uranium (MEU) were used as fissile fuel in HTGR systems. Simulated Triso-coated UO/sub 2/, UC/sub 2/, and UO/sub 2//UC/sub 2/ particles mixed with varying amounts of Mo, Ru, Rh, Pd, Ag, and Cd were prepared. These fission products were chosen because, after full burnup, their concentrations are higher in LEU and MEU fuels than in HEU fuel. After the particles were heat treated in the laboratory, their behavior was examined by use of metallography, scanning electron microscopy, and electron microprobe x-ray analysis.

  12. 3D COMSOL Simulations for Thermal Deflection of HFIR Fuel Plate in the "Cheverton-Kelley" Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Jain, Prashant K [ORNL; Freels, James D [ORNL; Cook, David Howard [ORNL

    2012-08-01

    Three dimensional simulation capabilities are currently being developed at Oak Ridge National Laboratory using COMSOL Multiphysics, a finite element modeling software, to investigate thermal expansion of High Flux Isotope Reactor (HFIR) s low enriched uranium fuel plates. To validate simulations, 3D models have also been developed for the experimental setup used by Cheverton and Kelley in 1968 to investigate the buckling and thermal deflections of HFIR s highly enriched uranium fuel plates. Results for several simulations are presented in this report, and comparisons with the experimental data are provided when data are available. A close agreement between the simulation results and experimental findings demonstrates that the COMSOL simulations are able to capture the thermal expansion physics accurately and that COMSOL could be deployed as a predictive tool for more advanced computations at realistic HFIR conditions to study temperature-induced fuel plate deflection behavior.

  13. The Rhode Island Nuclear Science Center conversion from HEU to LEU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Tehan, Terry

    2000-09-27

    The 2-MW Rhode Island Nuclear Science Center (RINSC) open pool reactor was converted from 93% UAL-High Enriched Uranium (HEU) fuel to 20% enrichment U3Si2-AL Low Enriched Uranium (LEU) fuel. The conversion included redesign of the core to a more compact size and the addition of beryllium reflectors and a beryllium flux trap. A significant increase in thermal flux level was achieved due to greater neutron leakage in the new compact core configuration. Following the conversion, a second cooling loop and an emergency core cooling system were installed to permit operation at 5 MW. After re-licensing at 2 MW, a power upgrade request will be submitted to the NRC.

  14. Neutron diffraction measurement of residual stresses, dislocation density and texture in Zr-bonded U-10Mo "mini" fuel foils and plates

    Science.gov (United States)

    Brown, D. W.; Okuniewski, M. A.; Sisneros, T. A.; Clausen, B.; Moore, G. A.; Balogh, L.

    2016-12-01

    Al clad U-10Mo fuel plates are being considered for conversion of several research reactors from high-enriched to low-enriched U fuel. Neutron diffraction measurements of the textures, residual phase stresses, and dislocation densities in the individual phases of the mini-foils throughout several processing steps and following hot-isostatic pressing to the Al cladding, have been completed. Recovery and recrystallization of the bare U-10Mo fuel foil, as indicated by the dislocation density and texture, are observed depending on the state of the material prior to annealing and the duration and temperature of the annealing process. In general, the cladding procedure significantly reduces the dislocation density, but the final state of the clad plate, both texture and dislocation density, depends strongly on the final processing step of the fuel foil. In contrast, the residual stress state of the final plate is dominated by the thermal expansion mismatch of the constituent materials.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-01-01

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

  16. Analysis of LOCA Scenarios in the NIST Research Reactor Before and After Fuel Conversion

    Energy Technology Data Exchange (ETDEWEB)

    Baek, J. S. [Brookhaven National Lab. (BNL), Upton, NY (United States); Cheng, L. Y. [Brookhaven National Lab. (BNL), Upton, NY (United States); Diamond, D. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2015-08-30

    An analysis has been done of hypothetical loss-of-coolant-accidents (LOCAs) in the research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The purpose of the analysis is to determine if the peak clad temperature remains below the Safety Limit, which is the blister temperature for the fuel. The configuration of the NBSR considered in the analysis is that projected for the future when changes will be made so that shutdown pumps do not operate when a LOCA signal is detected. The analysis was done for the present core with high-enriched uranium (HEU) fuel and with the proposed low-enriched uranium (LEU) fuel that would be used when the NBSR is converted from one to the other. The analysis consists of two parts. The first examines how the water would drain from the primary system following a break and the possibility for the loss of coolant from within the fuel element flow channels. This work is performed using the TRACE system thermal-hydraulic code. The second looks at the fuel clad temperature as a function of time given that the water may have drained from many of the flow channels and the water in the vessel is in a quasi-equilibrium state. The temperature behavior is investigated using the three-dimensional heat conduction code HEATING7.3. The results in all scenarios considered for both HEU and LEU fuel show that the peak clad temperature remains below the blister temperature.

  17. 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%.

  18. 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).

  19. Redundancy of Supply in the International Nuclear Fuel Fabrication Market: Are Fabrication Services Assured?

    Energy Technology Data Exchange (ETDEWEB)

    Seward, Amy M.; Toomey, Christopher; Ford, Benjamin E.; Wood, Thomas W.; Perkins, Casey J.

    2011-11-14

    For several years, Pacific Northwest National Laboratory (PNNL) has been assessing the reliability of nuclear fuel supply in support of the U.S. Department of Energy/National Nuclear Security Administration. Three international low enriched uranium reserves, which are intended back up the existing and well-functioning nuclear fuel market, are currently moving toward implementation. These backup reserves are intended to provide countries credible assurance that of the uninterrupted supply of nuclear fuel to operate their nuclear power reactors in the event that their primary fuel supply is disrupted, whether for political or other reasons. The efficacy of these backup reserves, however, may be constrained without redundant fabrication services. This report presents the findings of a recent PNNL study that simulated outages of varying durations at specific nuclear fuel fabrication plants. The modeling specifically enabled prediction and visualization of the reactors affected and the degree of fuel delivery delay. The results thus provide insight on the extent of vulnerability to nuclear fuel supply disruption at the level of individual fabrication plants, reactors, and countries. The simulation studies demonstrate that, when a reasonable set of qualification criteria are applied, existing fabrication plants are technically qualified to provide backup fabrication services to the majority of the world's power reactors. The report concludes with an assessment of the redundancy of fuel supply in the nuclear fuel market, and a description of potential extra-market mechanisms to enhance the security of fuel supply in cases where it may be warranted. This report is an assessment of the ability of the existing market to respond to supply disruptions that occur for technical reasons. A forthcoming report will address political disruption scenarios.

  20. 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.

  1. Development of low enriched uranium target plates by thermo-mechanical processing of UAl{sub 2}–Al matrix for production of {sup 99}Mo in Pakistan

    Energy Technology Data Exchange (ETDEWEB)

    Ali, Kanwar Liaqat; Khan, Akhlaque Ahmad [Pakistan Institute of Nuclear Science and Technology (PINSTECH) P.O Nilore, Islamabad (Pakistan); Mushtaq, Ahmad, E-mail: amushtaq1@hotmail.com [Pakistan Institute of Nuclear Science and Technology (PINSTECH) P.O Nilore, Islamabad (Pakistan); Imtiaz, Farhan; Ziai, Maratab Ali; Gulzar, Amir; Farooq, Muhammad; Hussain, Nazar; Ahmed, Nisar; Pervez, Shahid; Zaidi, Jamshed Hussain [Pakistan Institute of Nuclear Science and Technology (PINSTECH) P.O Nilore, Islamabad (Pakistan)

    2013-02-15

    Uranium aluminide predominated with UAl{sub 2} phase was prepared by arc-melting procedures and comminuted to required particle size. UAl{sub 2} and Al powders were blended and compacted to achieve LEU fuel density of 2.17 g/cm{sup 3}. The picture-frame technique was used to clad the dispersions (UAl{sub 2}–Al) with aluminum. A few target plates were fabricated by thermo-mechanical processing (hot rolling and annealing) of UAl{sub 2}–Al matrix contained in roll billet of Al. The fabricated plates were characterized by destructive and some of non-destructive testing techniques and then annealed to achieve required phase of uranium aluminide for proper dissolution in basic media.

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

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.; Goldmann, Andrew S. (Texas A& M University, College Station, TX); Wagner, Kenneth C.; Powers, Dana Auburn; Ashbaugh, Scott G.; Longmire, Pamela

    2010-04-01

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

  3. 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.

  4. 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...

  5. 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...

  6. 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.

  7. ASSESSMENT OF POSSIBLE CYCLE LENGTHS FOR FULLY-CERAMIC MICRO-ENCAPSULATED FUEL-BASED LIGHT WATER REACTOR CONCEPTS

    Energy Technology Data Exchange (ETDEWEB)

    R. Sonat Sen; Michael A. Pope; Abderrafi M. Ougouag; Kemal Pasamehmetoglu; Francesco Venneri

    2012-04-01

    The use of TRISO-particle-based dispersion fuel within SiC matrix and cladding materials has the potential to allow the design of extremely safe LWRs with failure-proof fuel. This paper examines the feasibility of LWR-like cycle length for such a low enriched uranium fuel with the imposed constraint of strictly retaining the original geometry of the fuel pins and assemblies. The motivation for retaining the original geometry is to provide the ability to incorporate the fuel 'as-is' into existing LWRs while retaining their thermal-hydraulic characteristics. The feasibility of using this fuel is assessed by looking at cycle lengths and fuel failure rates. Other considerations (e.g., safety parameters, etc.) were not considered at this stage of the study. The study includes the examination of different TRISO kernel diameters without changing the coating layer thicknesses. The study shows that a naive use of UO{sub 2} results in cycle lengths too short to be practical for existing LWR designs and operational demands. Increasing fissile inventory within the fuel compacts shows that acceptable cycle lengths can be achieved. In this study, starting with the recognized highest packing fraction practically achievable (44%), higher enrichment, larger fuel kernel sizes, and the use of higher density fuels have been evaluated. The models demonstrate cycle lengths comparable to those of ordinary LWRs. As expected, TRISO particles with extremely large kernels are shown to fail under all considered scenarios. In contrast, the designs that do not depart too drastically from those of the nominal NGNP HTR fuel TRISO particles are shown to perform satisfactorily and display a high rates of survival under all considered scenarios. Finally, it is recognized that relaxing the geometry constraint will result in satisfactory cycle lengths even using UO{sub 2}-loaded TRISO particles-based fuel with enrichment at or below 20 w/o.

  8. 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.

  9. 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...

  10. 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.

  11. Supplemental Reactor Physics Calculations and Analysis of ELF Mk 1A Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Pope, Michael A. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-10-01

    These calculations supplement previous the reactor physics work evaluating the Enhanced Low Enriched Uranium (LEU) Fuel (ELF) Mk 1A element. This includes various additional comparisons between the current Highly Enriched Uranium (HEU) and LEU along with further characterization of the performance of the ELF fuel. The excess reactivity to be held down at BOC for ELF Mk 1A fuel is estimated to be approximately $2.75 greater than with HEU for a typical cycle. This is a combined effect of the absence of burnable poison in the ELF fuel and the reduced neck shim worth in LEU fuel compared to HEU. Burnable poison rods were conceptualized for use in the small B positions containing Gd2O3 absorber. These were shown to provide $2.37 of negative reactivity at BOC and to burn out in less than half of a cycle. The worth of OSCCs is approximately the same between HEU and ELF Mk 1A (LEU) fuels in the representative loading evaluated. This was evaluated by rotating all banks simultaneously. The safety rod worth is relatively unchanged between HEU and ELF Mk 1A (LEU) fuels in the representative loading evaluated. However, this should be reevaluated with different loadings. Neutron flux, both total and fast (>1 MeV), is either the same or reduced upon changing from HEU to ELF Mk 1A (LEU) fuels in the representative loading evaluated. This is consistent with the well-established trend of lower neutron fluxes for a given power in LEU than HEU.The IPT loop void reactivity is approximately the same or less positive with ELF Mk 1A (LEU) fuel than HEU in the representative loading evaluated.

  12. Advanced Gas Reactor (AGR)-5/6/7 Fuel Irradiation Experiments in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    A. Joseph Palmer; David A. Petti; S. Blaine Grover

    2014-04-01

    The United States Department of Energy’s Very High Temperature Reactor (VHTR) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which each consist of at least five separate capsules, are being irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gases also have on-line fission product monitoring the effluent from each capsule to track performance of the fuel during irradiation. The first two experiments (designated AGR-1 and AGR-2), have been completed. The third and fourth experiments have been combined into a single experiment designated AGR-3/4, which started its irradiation in December 2011 and is currently scheduled to be completed in April 2014. The design of the fuel qualification experiment, designated AGR-5/6/7, is well underway and incorporates lessons learned from the three previous experiments. Various design issues will be discussed with particular details related to selection of thermometry.

  13. Status of the NGNP Fuel Experiment AGR-2 Irradiated in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Blaine Grover

    2012-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2), which utilized the same experiment design as well as control and monitoring systems as AGR-1, started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The design of this experiment and support systems will be briefly discussed, followed by the progress and status of the experiment to date.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-03-02

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Luther, Erik; Rooyen, Isabella van; Leckie, Rafael; Papin, Pallas; Nelson, Andrew; Hunter, James

    2015-03-01

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

  16. Measurement of fission gas release from irradiated U–Mo monolithic fuel samples

    Energy Technology Data Exchange (ETDEWEB)

    Burkes, Douglas E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Casella, Amanda J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Casella, Andrew M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Luscher, Walter G. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Rice, Francine J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Pool, Karl N. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-06-01

    The uranium–molybdenum (U-Mo) alloy in a monolithic form has been proposed as one fuel design capable of converting some of the world’s highest power research reactors from the use of high enriched uranium (HEU) to low enriched uranium (LEU). One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An apparatus capable of heating post-irradiated small-scale samples cut from larger fuel segments according to specified thermal profiles under a controlled atmosphere has been installed into a hot cell. Results show that optimized experimental parameters to investigate fission product release from small samples have been established. Initial measurements conducted on aluminum alloy clad uranium–molybdenum monolithic fuel samples reveal three clear fission gas release events over the temperature range of 30-1000 °C. The mechanisms responsible for these events are discussed, and the results have been compared with available information in the literature.

  17. Global Threat Reduction Initiative Fuel Thermo-Physical Characterization Project: Sample Management Plan

    Energy Technology Data Exchange (ETDEWEB)

    Casella, Amanda J.; Pereira, Mario M.; Steen, Franciska H.

    2013-01-01

    This sample management plan provides guidelines for sectioning, preparation, acceptance criteria, analytical path, and end-of-life disposal for the fuel element segments utilized in the Global Threat Reduction Initiative (GTRI), Fuel Thermo-Physical Characterization Project. The Fuel Thermo-Physical Characterization Project is tasked with analysis of irradiated Low Enriched Uranium (LEU) Molybdenum (U-Mo) fuel element samples to support the GTRI conversion program. Sample analysis may include optical microscopy (OM), scanning electron microscopy (SEM) fuel-surface interface analysis, gas pycnometry (density) measurements, laser flash analysis (LFA), differential scanning calorimetry (DSC), thermogravimetry and differential thermal analysis with mass spectroscopy (TG /DTA-MS), Inductively Coupled Plasma Spectrophotometry (ICP), alpha spectroscopy, and Thermal Ionization Mass Spectroscopy (TIMS). The project will utilize existing Radiochemical Processing Laboratory (RPL) operating, technical, and administrative procedures for sample receipt, processing, and analyses. Test instructions (TIs), which are documents used to provide specific details regarding the implementation of an existing RPL approved technical or operational procedure, will also be used to communicate to staff project specific parameters requested by the Principal Investigator (PI). TIs will be developed, reviewed, and issued in accordance with the latest revision of the RPL-PLN-700, RPL Operations Plan. Additionally, the PI must approve all project test instructions and red-line changes to test instructions.

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

    Energy Technology Data Exchange (ETDEWEB)

    Pereira, Mario M.; Slonecker, Bruce D.

    2012-06-01

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

  19. 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

  20. 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

  1. 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

  2. Application of Nondestructive Methods for Qualification of High Density Fuels in the IEA-R1 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Silva, J.E.R.; Silva, A.T.; Domingos, D.B.; Terremoto, L.A.A. [Instituto de Pesquisas Energeticas e Nucleares, Comissao Nacional de Energia Nuclear (IPEN-CNEN/SP), Av.Prof. Lineu Prestes 2242, Cidade Universitaria 05508-000, Sao Paulo, SP (Brazil)

    2011-07-01

    The IEA-R1 reactor of IPEN/CNEN-SP in Brazil is a pool type research reactor cooled and moderated by demineralised water and having Beryllium and Graphite as reflectors. Since 1990, IPEN/CNEN-SP has been fabricating and qualifying its own U{sub 3}O{sub 8}-Al and U{sub 3}Si{sub 2}-Al dispersion fuels. The U{sub 3}O{sub 8}-Al dispersion fuel is qualified to a uranium density of 2.3 gU/cm{sup 3} and the U{sub 3}Si{sub 2}-Al dispersion fuel up to 3.0 gU/cm{sup 3}. The IEA-R1 reactor core is constituted of the fuels above, with low enrichment in U-235 (19.9% of U-235). Nowadays, IPEN/CNEN-SP is interested in qualifying the above dispersion fuels at higher densities. Fuel miniplates of U{sub 3}O{sub 8}-Al and U{sub 3}Si{sub 2}-Al fuels, with densities of 3.0 gU/cm{sup 3} and 4.8 gU/cm{sup 3}, respectively, which are the maximal uranium densities qualified worldwide for these dispersion fuels, were fabricated at IPEN/CNEN-SP. The miniplates were put in an irradiation device, with similar external dimensions of IEA-R1 fuel assemblies, which was placed in a peripheral position of the IEA-R1 reactor core. IPEN/CNEN-SP has no hot cells to provide destructive analysis of the irradiated fuel. As a consequence, non destructive methods are being used to evaluate irradiation performance of the fuel miniplates: i) monitoring the fuel miniplate performance during the IEA-R1 operation for the following parameters: reactor power, time of operation, neutron flux at the position of each fuel assembly, burnup, inlet and outlet water, and radiochemistry analysis of reactor water; ii) periodic underwater visual inspection of fuel miniplates and eventual sipping test for the fuel miniplate suspected of leakage. The miniplates are being periodically visually inspected by an underwater radiation-resistant camera inside the IEA-R1 reactor pool, to verify its integrity and its general plate surface conditions. A new special system was designed for the fuel miniplate swelling evaluation. The

  3. High-quality thorium TRISO fuel performance in HTGRs

    Energy Technology Data Exchange (ETDEWEB)

    Verfondern, Karl [Forschungszentrum Juelich GmbH (Germany); Allelein, Hans-Josef [Forschungszentrum Juelich GmbH (Germany); Technische Hochschule Aachen (Germany); Nabielek, Heinz; Kania, Michael J.

    2013-11-01

    Thorium as a nuclear fuel has received 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 HTGR development employed thorium together with high-enriched uranium (HEU). After 1980, HTGR fuel systems switched to low-enriched uranium (LEU). After completing fuel development for the AVR and the THTR with BISO coated particles, the German program expanded its efforts utilizing thorium and HEU TRISO coated particles in advanced HTGR concepts for process heat applications (PNP) and direct-cycle electricity production (HHT). The combination of a low-temperature isotropic (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 high-temperature isotropic (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 HTGR 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 to 13% FIMA and fast fluences to 8 x 10{sup 25} n/m{sup 2} (E> 16 fJ), the performance results exceed the design limits on manufacturing and operational requirements for the German HTR-Modul concept, which are 6.5 x 10{sup -5} for manufacturing, 2 x 10{sup -4} for normal operating conditions, and 5 x 10{sup -4

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

    Science.gov (United States)

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

    2017-04-01

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

  5. Analysis of the optimal fuel composition for the Indonesian experimental power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Liem, Peng Hong [Nippon Advanced Information Service (NAIS Co., Inc.), Ibaraki (Japan); Sembiring, Tagor Malem [National Nuclear Energy Agency of Indonesia, Banten (Indonesia). Center for Nuclear Reactor Technology and Safety; Arbie, Bakri; Subki, Iyos [PT MOTAB Technology, Jakarta Barat (Indonesia)

    2017-03-15

    The optimal fuel composition of the 10 MWth Experimental Power Reactor (RDE), to be built by the Indonesian National Nuclear Energy Agency (BATAN), is a very important design parameter since it will directly affect the fuel cost, new and spent fuel storage capacity, and other back-end environmental burden. The RDE is a very small sized pebble-bed high temperature gas-cooled reactor (HTGR) with low enriched uranium (LEU) UO{sub 2} TRISO fuel under multipass or once-through-then-out fueling scheme. A scoping study on fuel composition parameters, namely heavy metal (HM) loading per pebble and uranium enrichment is conducted. All burnup, criticality calculations and core equilibrium search are carried out by using BATAN-MPASS, a general in-core fuel management code for pebble bed HTGRs, featured with many automatic equilibrium searching options as well as thermal-hydraulic calculation capability. The RDE User Requirement Document issued by BATAN is used to derive the main core design parameters and constraints. The scoping study is conducted over uranium enrichment in the range of 10 to 20 w/o and HM loading in the range of 4 g to 10 g/pebble. Fissile loading per unit energy generated (kg/GWd) is taken as the objective function for the present scoping study. The analysis results show that the optimal HM loading is around 8 g/pebble. Under the constraint of 80 GWd/t fuel discharge burnup imposed by the technical specification, the uranium enrichment for the optimal HM loading is approximately 13 w/o.

  6. 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)

  7. Molybdenum-99-producing 37-element fuel bundle neutronically and thermal-hydraulically equivalent to a standard CANDU fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Nichita, E., E-mail: Eleodor.Nichita@uoit.ca; Haroon, J., E-mail: Jawad.Haroon@uoit.ca

    2016-10-15

    Highlights: • A 37-element fuel bundle modified for {sup 99}Mo production in CANDU reactors is presented. • The modified bundle is neutronically and thermal-hydraulically equivalent to the standard bundle. • The modified bundle satisfies all safety criteria satisfied by the standard bundle. - Abstract: {sup 99m}Tc, the most commonly used radioisotope in diagnostic nuclear medicine, results from the radioactive decay of {sup 99}Mo which is currently being produced at various research reactors around the globe. In this study, the potential use of CANDU power reactors for the production of {sup 99}Mo is investigated. A modified 37-element fuel bundle, suitable for the production of {sup 99}Mo in existing CANDU-type reactors is proposed. The new bundle is specifically designed to be neutronically and thermal-hydraulically equivalent to the standard 37-element CANDU fuel bundle in normal, steady-state operation and, at the same time, be able to produce significant quantities of {sup 99}Mo when irradiated in a CANDU reactor. The proposed bundle design uses fuel pins consisting of a depleted-uranium centre surrounded by a thin layer of low-enriched uranium. The new molybdenum-producing bundle is analyzed using the lattice transport code DRAGON and the diffusion code DONJON. The proposed design is shown to produce 4081 six-day Curies of {sup 99}Mo activity per bundle when irradiated in the peak-power channel of a CANDU core, while maintaining the necessary reactivity and power rating limits. The calculated {sup 99}Mo yield corresponds to approximately one third of the world weekly demand. A production rate of ∼3 bundles per week can meet the global demand of {sup 99}Mo.

  8. Alpha particle spectroscopy — A useful tool for the investigation of spent nuclear fuel from high temperature gas-cooled reactors

    Science.gov (United States)

    Helmbold, M.

    1984-06-01

    For more than a decade, alpha particle spectrometry of spent nuclear fuel has been used at the Kernforschungsanlage Jülich (KFA) in the field of research for the German high temperature reactor (HTR). Techniques used for the preparation of samples for alpha spectrometry have included deposition from aqueous solutions of spent fuel, annealing of fuel particles in an oven and the evaporation of fuel material by a laser beam. The resulting sources are very thin but of low activity and the alpha spectrometry data obtained from them must be evaluated with sophisticated computer codes to achieve the required accuracy. Measurements have been made on high and low enriched uranium fuel and on a variety of parameters relevant to the fuel cycle. In this paper the source preparation and data evaluation techniques will be discussed together with the results obtained to data, i.e. production of alpha active actinide isotopes, correlations between actinide isotopes and fission products, build up and transmutation of actinides during burn-up of HTR fuel, diffusion coefficients of actinides for fuel particle kernels and coating materials. All these KFA results have helped to establish the basis for the design, licensing and operation of HTR power plants, including reprocessing and waste management.

  9. HTGR Fuel performance basis

    Energy Technology Data Exchange (ETDEWEB)

    Shamasundar, B.I.; Stansfield, O.M.; Jensen, D.D.

    1982-05-01

    The safety characteristics of the high-temperature gas-cooled reactor (HTGR) during normal and accident conditions are determined in part by HTGR fuel performance. During normal operation, less than 0.1% fuel failure occurs, primarily from defective particles. This low fuel failure fraction limits circulating activity to acceptable levels. During severe accidents, the radiological consequence is influenced by high-temperature fuel particle behavior. An empirical fuel failure model, supported by recent experimental data, is presented. The onset of significant fuel particle failure occurs at temperatures in excess of 1600/sup 0/C, and complete fuel failure occurs at 2660/sup 0/C. This indicates that the fuel is more retentive at higher temperatures than previously assumed. The more retentive nature of the fuel coupled with the high thermal capacitance of the core results in slow release of fission products from the core during severe accidents.

  10. Fuels Combustion Research: Supercritical Fuel Pyrolysis

    National Research Council Canada - National Science Library

    Glassman, Irvin

    2001-01-01

    Present and anticipated variation in jet propulsion fuels due to advanced engine compression ratios and airframe cooling requirements necessitate greater understanding of chemical phenomena associated...

  11. Fuels Combustion Research: Supercritical Fuel Pyrolysis

    National Research Council Canada - National Science Library

    Glassman, Irvin

    2000-01-01

    Present and anticipated variation in jet propulsion fuels due to advanced engine compression ratios and airframe cooling requirements necessitate greater understanding of chemical phenomena associated...

  12. 77 FR 699 - Regulation of Fuels and Fuel Additives: Identification of Additional Qualifying Renewable Fuel...

    Science.gov (United States)

    2012-01-05

    ... January 5, 2012 Part V Environmental Protection Agency 40 CFR Part 80 Regulation of Fuels and Fuel Additives: Identification of Additional Qualifying Renewable Fuel Pathways Under the Renewable Fuel Standard... Fuels and Fuel Additives: Identification of Additional Qualifying Renewable Fuel Pathways Under...

  13. Materials for fuel cells

    Directory of Open Access Journals (Sweden)

    Sossina M Haile

    2003-03-01

    Full Text Available Because of their potential to reduce the environmental impact and geopolitical consequences of the use of fossil fuels, fuel cells have emerged as tantalizing alternatives to combustion engines. Like a combustion engine, a fuel cell uses some sort of chemical fuel as its energy source but, like a battery, the chemical energy is directly converted to electrical energy, without an often messy and relatively inefficient combustion step. In addition to high efficiency and low emissions, fuel cells are attractive for their modular and distributed nature, and zero noise pollution. They will also play an essential role in any future hydrogen fuel economy.

  14. Composite nuclear fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Dollard, W.J.; Ferrari, H.M.

    1982-04-27

    An open lattice elongated nuclear fuel assembly including small diameter fuel rods disposed in an array spaced a selected distance above an array of larger diameter fuel rods for use in a nuclear reactor having liquid coolant flowing in an upward direction. Plenums are preferably provided in the upper portion of the upper smaller diameter fuel rods and in the lower portion of the lower larger diameter fuel rods. Lattice grid structures provide lateral support for the fuel rods and preferably the lowest grid about the upper rods is directly and rigidly affixed to the highest grid about the lower rods.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-09-30

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

  16. DUPIC fuel compatibility assessment

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok; Rho, G. H.; Park, J. W. [and others

    2000-03-01

    The purpose of this study is to assess the compatibility of DUPIC(Direct Use of Spent PWR Fuel in CANDU Reactors) fuel with the current CANDU 6 reactor, which is one of the technology being developed to utilize the spent PWR fuel in CANDU reactors. The phase 1 study of this project includes the feasibility analysis on applicability of the current core design method, the feasibility analysis on operation of the DUPIC fuel core, the compatibility analysis on individual reactor system, the sensitivity analysis on the fuel composition, and the economic analysis on DUPIC fuel cycle. The results of the validation calculations have confirmed that the current core analysis system is acceptable for the feasibility study of the DUPIC fuel compatibility analysis. The results of core simulations have shown that both natural uranium and DUPIC fuel cores are almost the same from the viewpoint of the operational performance. For individual reactor system including reactively devices, the functional requirements of each system are satisfied in general. However, because of the pronounced power flattening in the DUPIC core, the radiation damage on the critical components increases, which should be investigated more in the future. The DUPIC fuel composition heterogeneity dose not to impose any serious effect on the reactor operation if the fuel composition is adjusted. The economics analysis has been performed through conceptual design studies on the DUPIC fuel fabrication, fuel handling in a plant, and spent fuel disposal, which has shown that the DUPIC fuel cycle is comparable to the once-trough fuel cycle considering uncertainties associated with unit costs of the fuel cycle components. The results of Phase 1 study have shown that it is feasible to use the DUPIC fuel in CANDU reactors without major changes in hardware. However further studies are required to confirm the safety of the reactor under accident condition.

  17. Toxicity of uranium, molybdenum, nickel, and arsenic to Hyalella azteca and Chironomus dilutus in water-only and spiked-sediment toxicity tests.

    Science.gov (United States)

    Liber, Karsten; Doig, Lorne E; White-Sobey, Suzanne L

    2011-07-01

    A series of laboratory spiked-sediment toxicity tests with the amphipod Hyalella azteca and the midge Chironomus dilutus were undertaken to determine acute and chronic toxicity thresholds for uranium (U), molybdenum (Mo), nickel (Ni), and arsenic (As) based on both whole-sediment (total) and pore water exposure concentrations. Water-only toxicity data were also generated from separate experiments to determine the toxicities of these metals/metalloids under our test conditions and to help evaluate the hypothesis that pore water metal concentrations are better correlated with sediment toxicity to benthic organisms than whole-sediment metal concentrations. The relative toxicity of the four elements tested differed depending on which test species was used and whether whole-sediment or pore water metal concentrations were correlated with effects. Based on measured whole-sediment concentrations, Ni and As were the two most acutely toxic elements to H. azteca with 10-d LC50s of 521 and 532 mg/kg d.w., respectively. Measured pore water concentrations indicated that U and Ni were the two most acutely toxic elements, with 10-d LC50s to H. azteca of 2.15 and 2.05 mg/L, respectively. Based on pore water metal concentrations, the no-observed-effect concentrations (NOECs) for growth were (H. azteca and C. dilutus, respectively) 0.67 and 0.21 mg/L for U, azteca and C. dilutus, respectively) 2.99 and 0.48 mg/L for U, 0.37 and 2.33 mg/L for Ni, and 58.99 and 0.42 mg/L for As. For U and Ni, results from 96-h water-only acute toxicity tests correlated well with pore water metal concentrations in acutely toxic metal-spiked sediment. This was not true for As where metalloid concentrations in overlying water (diffusion from sediment) may have contributed to toxicity. The lowest whole-sediment LOEC reported here for As was 6.6- and 4-fold higher than the Canadian Council of Ministers of the Environment interim sediment quality guideline and the Canadian Nuclear Safety Commission (CNSC) lowest effect level (LEL), respectively. The lowest whole-sediment LOECs reported here for Ni, U and Mo were 4-, 17.5-, and >260-fold higher, respectively, than the CNSC LELs for these metals/metalloids. Data on pore water metal concentrations in toxic sediment would be a useful addition to future Guidelines documents.

  18. EDTA返滴定法测定铀钼合金中钼含量%Determination of Molybdenum in Uranium-Molybdenum Alloy by EDTA Back Titration

    Institute of Scientific and Technical Information of China (English)

    王芸; 肖吉群

    2014-01-01

    介绍了采用EDTA返滴定法测定铀钼合金中钼含量的方法.以王水溶样,用磷酸三丁酯-氢化煤油萃取分离铀,采用盐酸羟胺将MC+还原为Mo5+,用EDTA络合Mo5+,然后用锌标准溶液返滴定过量的EDTA,从而完成铀钼合金中的钼含量测定.优化了方法的实验条件,实验表明,相对标准偏差(RSD)为1.4%.

  19. 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.

  20. Direct hydrocarbon fuel cells

    Science.gov (United States)

    Barnett, Scott A.; Lai, Tammy; Liu, Jiang

    2010-05-04

    The direct electrochemical oxidation of hydrocarbons in solid oxide fuel cells, to generate greater power densities at lower temperatures without carbon deposition. The performance obtained is comparable to that of fuel cells used for hydrogen, and is achieved by using novel anode composites at low operating temperatures. Such solid oxide fuel cells, regardless of fuel source or operation, can be configured advantageously using the structural geometries of this invention.

  1. Navy Fuel Specification Standardization

    Science.gov (United States)

    1992-04-01

    surfaced periodically to convert further to a single-fuel operation, i.e., one fuel for both aircraft and ship propulsion /power systems. This study...lead to the development of a single distillate fuel for ship propulsion , resulting eventually in the MIL-F-16884 Naval Distillate Fuel (NDF) used today...for both aircraft and ship propulsion /power systems. This report summarizes a study to consider this problem in light of current systems and

  2. The STAT7 Code for Statistical Propagation of Uncertainties In Steady-State Thermal Hydraulics Analysis of Plate-Fueled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dunn, Floyd E. [Argonne National Lab. (ANL), Argonne, IL (United States); Hu, Lin-wen [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States). Nuclear Reactor Lab.; Wilson, Erik [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-12-01

    The STAT code was written to automate many of the steady-state thermal hydraulic safety calculations for the MIT research reactor, both for conversion of the reactor from high enrichment uranium fuel to low enrichment uranium fuel and for future fuel re-loads after the conversion. A Monte-Carlo statistical propagation approach is used to treat uncertainties in important parameters in the analysis. These safety calculations are ultimately intended to protect against high fuel plate temperatures due to critical heat flux or departure from nucleate boiling or onset of flow instability; but additional margin is obtained by basing the limiting safety settings on avoiding onset of nucleate boiling. STAT7 can simultaneously analyze all of the axial nodes of all of the fuel plates and all of the coolant channels for one stripe of a fuel element. The stripes run the length of the fuel, from the bottom to the top. Power splits are calculated for each axial node of each plate to determine how much of the power goes out each face of the plate. By running STAT7 multiple times, full core analysis has been performed by analyzing the margin to ONB for each axial node of each stripe of each plate of each element in the core.

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

    Energy Technology Data Exchange (ETDEWEB)

    Sarnoski, Sarah E.; Fast, James E.; Fulsom, Bryan G.; Gilbert, Andrew J.; Jacomb-Hood, Timothy W.; VanDevender, Brent A.; Wood, Lynn S.

    2017-07-17

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

  4. Criticality investigations for the fixed bed nuclear reactor using thorium fuel mixed with plutonium or minor actinides

    Energy Technology Data Exchange (ETDEWEB)

    Sahin, Suemer [Beykoz Lojistik Meslek Yueksekokulu, Beykoz, Istanbul (Turkey)], E-mail: sumer@gazi.edu.tr; Sahin, Haci Mehmet; Acir, Adem [Beykoz Lojistik Meslek Yueksekokulu, Istanbul (Turkey); Al-Kusayer, Tawfik Ahmed [King Saud University, College of Engineering, P.O. Box 800, Riyadh 11421 (Saudi Arabia)

    2009-08-15

    Prospective fuels for a new reactor type, the so called fixed bed nuclear reactor (FBNR) are investigated with respect to reactor criticality. These are (1) low enriched uranium (LEU); (2) weapon grade plutonium + ThO{sub 2}; (3) reactor grade plutonium + ThO{sub 2}; and (4) minor actinides in the spent fuel of light water reactors (LWRs) + ThO{sub 2}. Reactor grade plutonium and minor actinides are considered as highly radio-active and radio-toxic nuclear waste products so that one can expect that they will have negative fuel costs. The criticality calculations are conducted with SCALE5.1 using S{sub 8}-P{sub 3} approximation in 238 neutron energy groups with 90 groups in thermal energy region. The study has shown that the reactor criticality has lower values with uranium fuel and increases passing to minor actinides, reactor grade plutonium and weapon grade plutonium. Using LEU, an enrichment grade of 9% has resulted with k{sub eff} = 1.2744. Mixed fuel with weapon grade plutonium made of 20% PuO{sub 2} + 80% ThO{sub 2} yields k{sub eff} = 1.2864. Whereas a mixed fuel with reactor grade plutonium made of 35% PuO{sub 2} + 65% ThO{sub 2} brings it to k{sub eff} = 1.267. Even the very hazardous nuclear waste of LWRs, namely minor actinides turn out to be high quality nuclear fuel due to the excellent neutron economy of FBNR. A relatively high reactor criticality of k{sub eff} = 1.2673 is achieved by 50% MAO{sub 2} + 50% ThO{sub 2}. The hazardous actinide nuclear waste products can be transmuted and utilized as fuel in situ. A further output of the study is the possibility of using thorium as breeding material in combination with these new alternative fuels.

  5. STATUS OF TRISO FUEL IRRADIATIONS IN THE ADVANCED TEST REACTOR SUPPORTING HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGNS

    Energy Technology Data Exchange (ETDEWEB)

    Davenport, Michael; Petti, D. A.; Palmer, Joe

    2016-11-01

    The United States Department of Energy’s Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and completed in October 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and completed in April 2014. Since the purpose of this experiment was to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment was significantly different from the first two experiments, though the control

  6. Modeling: driving fuel cells

    Directory of Open Access Journals (Sweden)

    Michael Francis

    2002-05-01

    Fuel cells were invented in 1839 by Sir William Grove, a Welsh judge and gentleman scientist, as a result of his experiments on the electrolysis of water. To put it simply, fuel cells are electrochemical devices that take hydrogen gas from fuel, combine it with oxygen from the air, and generate electricity and heat, with water as the only by-product.

  7. Alternate Fuels Combustion Research

    Science.gov (United States)

    1983-10-01

    properties of the other fuels are varied systematically beyond the specification limits imposed on the reference fuels, principally in the direction of...lower hydrogen content- Comparison of fuel nozzles, Figurae ,6.32. shows stronger dependence bet- ween oeiseslona and hydrogen content for airblast and

  8. Vented nuclear fuel element

    Science.gov (United States)

    Grossman, Leonard N.; Kaznoff, Alexis I.

    1979-01-01

    A nuclear fuel cell for use in a thermionic nuclear reactor in which a small conduit extends from the outside surface of the emitter to the center of the fuel mass of the emitter body to permit escape of volatile and gaseous fission products collected in the center thereof by virtue of molecular migration of the gases to the hotter region of the fuel.

  9. Alternative Fuels Data Center

    Energy Technology Data Exchange (ETDEWEB)

    None

    2013-06-01

    Fact sheet describes the Alternative Fuels Data Center, which provides information, data, and tools to help fleets and other transportation decision makers find ways to reduce petroleum consumption through the use of alternative and renewable fuels, advanced vehicles, and other fuel-saving measures.

  10. Fuel cell catalyst degradation

    DEFF Research Database (Denmark)

    Arenz, Matthias; Zana, Alessandro

    2016-01-01

    Fuel cells are an important piece in our quest for a sustainable energy supply. Although there are several different types of fuel cells, the by far most popular is the proton exchange membrane fuel cell (PEMFC). Among its many favorable properties are a short start up time and a high power density...

  11. A Multi-Dimensional Heat Transfer Model of a Tie-Tube and Hexagonal Fuel Element for Nuclear Thermal Propulsion

    Science.gov (United States)

    Gomez, C. F.; Mireles, O. R.; Stewart, E.

    2016-01-01

    The Space Capable Cryogenic Thermal Engine (SCCTE) effort considers a nuclear thermal rocket design based around a Low-Enriched Uranium (LEU) design fission reactor. The reactor core is comprised of bundled hexagonal fuel elements that directly heat hydrogen for expansion in a thrust chamber and hexagonal tie-tubes that house zirconium hydride moderator mass for the purpose of thermalizing fast neutrons resulting from fission events. Created 3D steady state Hex fuel rod model with 1D flow channels. Hand Calculation were used to set up initial conditions for fluid flow. The Hex Fuel rod uses 1D flow paths to model the channels using empirical correlations for heat transfer in a pipe. Created a 2-D axisymmetric transient to steady state model using the CFD turbulent flow and Heat Transfer module in COMSOL. This model was developed to find and understand the hydrogen flow that might effect the thermal gradients axially and at the end of the tie tube where the flow turns and enters an annulus. The Hex fuel rod and Tie tube models were made based on requirements given to us by CSNR and the SCCTE team. The models helped simplify and understand the physics and assumptions. Using pipe correlations reduced the complexity of the 3-D fuel rod model and is numerically more stable and computationally more time-efficient compared to the CFD approach. The 2-D axisymmetric tie tube model can be used as a reference "Virtual test model" for comparing and improving 3-D Models.

  12. Alternative aviation turbine fuels

    Science.gov (United States)

    Grobman, J.

    1977-01-01

    The efficient utilization of fossil fuels by future jet aircraft may necessitate the broadening of current aviation turbine fuel specifications. The most significant changes in specifications would be an increased aromatics content and a higher final boiling point in order to minimize refinery energy consumption and costs. These changes would increase the freezing point and might lower the thermal stability of the fuel and could cause increased pollutant emissions, increased smoke and carbon formation, increased combustor liner temperatures, and poorer ignition characteristics. This paper discusses the effects that broadened specification fuels may have on present-day jet aircraft and engine components and the technology required to use fuels with broadened specifications.

  13. Fuel cells : a viable fossil fuel alternative

    Energy Technology Data Exchange (ETDEWEB)

    Paduada, M.

    2007-02-15

    This article presented a program initiated by Natural Resources Canada (NRCan) to develop proof-of-concept of underground mining vehicles powered by fuel cells in order to eliminate emissions. Recent studies on American and Canadian underground mines provided the basis for estimating the operational cost savings of switching from diesel to fuel cells. For the Canadian mines evaluated, the estimated ventilation system operating cost reductions ranged from 29 per cent to 75 per cent. In order to demonstrate the viability of a fuel cell-powered vehicle, NRCan has designed a modified Caterpillar R1300 loader with a 160 kW hybrid power plant in which 3 stacks of fuel cells deliver up to 90 kW continuously, and a nickel-metal hydride battery provides up to 70 kW. The battery subsystem transiently boosts output to meet peak power requirements and also accommodates regenerative braking. Traction for the loader is provided by a brushless permanent magnet traction motor. The hydraulic pump motor is capable of a 55 kW load continuously. The loader's hydraulic and traction systems are operated independently. Future fuel cell-powered vehicles designed by the program may include a locomotive and a utility vehicle. Future mines running their operations with hydrogen-fueled equipment may also gain advantages by employing fuel cells in the operation of handheld equipment such as radios, flashlights, and headlamps. However, the proton exchange membrane (PEM) fuel cells used in the project are prohibitively expensive. The catalytic content of a fuel cell can add hundreds of dollars per kW of electric output. Production of catalytic precious metals will be strongly connected to the scale of use and acceptance of fuel cells in vehicles. In addition, the efficiency of hydrogen production and delivery is significantly lower than the well-to-tank efficiency of many conventional fuels. It was concluded that an adequate hydrogen infrastructure will be required for the mining industry

  14. Oxy-fuel combustion of solid fuels

    DEFF Research Database (Denmark)

    Toftegaard, Maja Bøg; Brix, Jacob; Jensen, Peter Arendt

    2010-01-01

    Oxy-fuel combustion is suggested as one of the possible, promising technologies for capturing CO2 from power plants. The concept of oxy-fuel combustion is removal of nitrogen from the oxidizer to carry out the combustion process in oxygen and, in most concepts, recycled flue gas to lower the flame...... temperature. The flue gas produced thus consists primarily of carbon dioxide and water. Much research on the different aspects of an oxy-fuel power plant has been performed during the last decade. Focus has mainly been on retrofits of existing pulverized-coal-fired power plant units. Green-field plants which...... provide additional options for improvement of process economics are however likewise investigated. Of particular interest is the change of the combustion process induced by the exchange of carbon dioxide and water vapor for nitrogen as diluent. This paper reviews the published knowledge on the oxy-fuel...

  15. 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.

  16. Fuel safety research 1999

    Energy Technology Data Exchange (ETDEWEB)

    Uetsuka, Hiroshi (ed.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-07-01

    In April 1999, the Fuel Safety Research Laboratory was newly established as a result of reorganization of the Nuclear Safety Research Center, JAERI. The laboratory was organized by combining three laboratories, the Reactivity Accident Laboratory, the Fuel Reliability Laboratory, and a part of the Sever Accident Research Laboratory. Consequently, the Fuel Safety Research Laboratory is now in charge of all the fuel safety research in JAERI. Various types of experimental and analytical researches are conducted in the laboratory by using the unique facilities such as the Nuclear Safety Research Reactor (NSRR), the Japan Material Testing Reactor (JMTR), the Japan Research Reactor 3 (JRR-3) and hot cells in JAERI. The laboratory consists of five research groups corresponding to each research fields. They are; (a) Research group of fuel behavior under the reactivity initiated accident conditions (RIA group). (b) Research group of fuel behavior under the loss-of-coolant accident conditions (LOCA group). (c) Research group of fuel behavior under the normal operation conditions (JMTR/BOCA group). (d) Research group of fuel behavior analysis (FEMAXI group). (e) Research group of FP release/transport behavior from irradiated fuel (VEGA group). This report summarizes the outline of research activities and major outcomes of the research executed in 1999 in the Fuel Safety Research Laboratory. (author)

  17. ATR LEU Monolithic Foil-Type Fuel with Integral Cladding Burnable Absorber – Neutronics Performance Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Gray Chang

    2012-03-01

    The Advanced Test Reactor (ATR), currently operating in the United States, is used for material testing at very high neutron fluxes. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth. Because of the large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting HEU driven reactor cores to low-enriched uranium (LEU) cores. The burnable absorber - 10B, was added in the inner and outer plates to reduce the initial excess reactivity, and to improve the peak ratio of the inner/outer heat flux. The present work investigates the LEU Monolithic foil-type fuel with 10B Integral Cladding Burnable Absorber (ICBA) design and evaluates the subsequent neutronics operating effects of this proposed fuel designs. The proposed LEU fuel specification in this work is directly related to both the RERTR LEU Development Program and the Advanced Test Reactor (ATR) LEU Conversion Project at Idaho National Laboratory (INL).

  18. Fuel related risks; Braenslerisker

    Energy Technology Data Exchange (ETDEWEB)

    Englund, Jessica; Sernhed, Kerstin; Nystroem, Olle; Graveus, Frank (Grontmij AB, (Sweden))

    2012-02-15

    The project, within which this work report was prepared, aimed to complement the Vaermeforsk publication 'Handbook of fuels' on fuel related risks and measures to reduce the risks. The fuels examined in this project where the fuels included in the first version of the handbook from 2005 plus four additional fuels that will be included in the second and next edition of the handbook. Following fuels were included: woodfuels (sawdust, wood chips, powder, briquettes), slash, recycled wood, salix, bark, hardwood, stumps, straw, reed canary grass, hemp, cereal, cereal waste, olive waste, cocoa beans, citrus waste, shea, sludge, forest industrial sludge, manure, Paper Wood Plastic, tyre, leather waste, cardboard rejects, meat and bone meal, liquid animal and vegetable wastes, tall oil pitch, peat, residues from food industry, biomal (including slaughterhouse waste) and lignin. The report includes two main chapters; a general risk chapter and a chapter of fuel specific risks. The first one deals with the general concept of risk, it highlights laws and rules relevant for risk management and it discuss general risks that are related to the different steps of fuel handling, i.e. unloading, storing, processing the fuel, transportation within the facility, combustion and handling of ashes. The information that was used to produce this chapter was gathered through a literature review, site visits, and the project group's experience from risk management. The other main chapter deals with fuel-specific risks and the measures to reduce the risks for the steps of unloading, storing, processing the fuel, internal transportation, combustion and handling of the ashes. Risks and measures were considered for all the biofuels included in the second version in the handbook of fuels. Information about the risks and risk management was gathered through interviews with people working with different kinds of fuels in electricity and heat plants in Sweden. The information from

  19. Direct Fuel Injector Temporal Measurements

    Science.gov (United States)

    2014-10-01

    optimize engine performance and emissions. Fuel injectors contain an actuator, pintle (or needle), and nozzle. The most common actuator is a solenoid ...Introduction Fuel injectors have a long history in metering fuel in modern engines by either port fuel injection (PFI) or direct fuel injection (DFI...Compared with a carburetor, fuel injectors have more accurate fuel delivering capability, thus giving engineers and technicians more flexibility to

  20. 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

  1. Neutron Diffraction Measurement of Residual Stresses, Dislocation Density and Texture in Zr-bonded U-10Mo ''Mini'' Fuel Foils and Plates

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Donald W. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Okuniewski, M. A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sisneros, Thomas A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Clausen, Bjorn [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Moore, G. A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Balogh, L [Queen' s Univ., Kingston, ON (Canada)

    2014-08-07

    Aluminum clad monolithic uranium 10 weight percent molybdenum (U-10Mo) fuel plates are being considered for conversion of several research and test nuclear reactors from high-enriched to low-enriched uranium fuel due to the inherently high density of fissile material. Comprehensive neutron diffraction measurements of the evolution of the textures, residual phase stresses, and dislocation densities in the individual phases of the mini-foils throughout several processing steps and following hot-isostatic pressing to the Al cladding, have been completed. Recovery and recrystallization of the bare U-10Mo fuel foil, as indicated by the dislocation density and texture, are observed depending on the state of the material prior to annealing and the duration and temperature of the annealing process. In general, the HIP procedure significantly reduces the dislocation density, but the final state of the clad plate, both texture and dislocation density, depends strongly on the final processing step of the fuel foil. In contrast, the residual stresses in the clad fuel plate do not depend strongly on the final processing step of the bare foil prior to HIP bonding. Rather, the residual stresses are dominated by the thermal expansion mismatch of the constituent materials of the fuel plate.

  2. Performance assessment of the direct disposal in unsaturated tuff or spent nuclear fuel and high-level waste owned by USDOE: Volume 2, Methodology and results

    Energy Technology Data Exchange (ETDEWEB)

    Rechard, R.P. [ed.

    1995-03-01

    This assessment studied the performance of high-level radioactive waste and spent nuclear fuel in a hypothetical repository in unsaturated tuff. The results of this 10-month study are intended to help guide the Office of Environment Management of the US Department of Energy (DOE) on how to prepare its wastes for eventual permanent disposal. The waste forms comprised spent fuel and high-level waste currently stored at the Idaho National Engineering Laboratory (INEL) and the Hanford reservations. About 700 metric tons heavy metal (MTHM) of the waste under study is stored at INEL, including graphite spent nuclear fuel, highly enriched uranium spent fuel, low enriched uranium spent fuel, and calcined high-level waste. About 2100 MTHM of weapons production fuel, currently stored on the Hanford reservation, was also included. The behavior of the waste was analyzed by waste form and also as a group of waste forms in the hypothetical tuff repository. When the waste forms were studied together, the repository was assumed also to contain about 9200 MTHM high-level waste in borosilicate glass from three DOE sites. The addition of the borosilicate glass, which has already been proposed as a final waste form, brought the total to about 12,000 MTHM.

  3. Fuel cell generator with fuel electrodes that control on-cell fuel reformation

    Science.gov (United States)

    Ruka, Roswell J.; Basel, Richard A.; Zhang, Gong

    2011-10-25

    A fuel cell for a fuel cell generator including a housing including a gas flow path for receiving a fuel from a fuel source and directing the fuel across the fuel cell. The fuel cell includes an elongate member including opposing first and second ends and defining an interior cathode portion and an exterior anode portion. The interior cathode portion includes an electrode in contact with an oxidant flow path. The exterior anode portion includes an electrode in contact with the fuel in the gas flow path. The anode portion includes a catalyst material for effecting fuel reformation along the fuel cell between the opposing ends. A fuel reformation control layer is applied over the catalyst material for reducing a rate of fuel reformation on the fuel cell. The control layer effects a variable reformation rate along the length of the fuel cell.

  4. HTPEM Fuel Cell Impedance

    DEFF Research Database (Denmark)

    Vang, Jakob Rabjerg

    As part of the process to create a fossil free Denmark by 2050, there is a need for the development of new energy technologies with higher efficiencies than the current technologies. Fuel cells, that can generate electricity at higher efficiencies than conventional combustion engines, can...... potentially play an important role in the energy system of the future. One of the fuel cell technologies, that receives much attention from the Danish scientific community is high temperature proton exchange membrane (HTPEM) fuel cells based on polybenzimidazole (PBI) with phosphoric acid as proton conductor....... This type of fuel cell operates at higher temperature than comparable fuel cell types and they distinguish themselves by high CO tolerance. Platinum based catalysts have their efficiency reduced by CO and the effect is more pronounced at low temperature. This Ph.D. Thesis investigates this type of fuel...

  5. Solid electrolytic fuel cell

    Energy Technology Data Exchange (ETDEWEB)

    Sakai, Masayasu; Yamauchi, Yasuhiro; Kamisaka, Mitsuo; Notomi, Kei.

    1989-04-21

    Concerning a solid electrolytic fuel cell with a gas permeable substrate pipe, a fuel electrode installed on this substrate pipe and an air electrode which is laminated on this fuel electrode with the electrolyte in between, the existing fuel cell of this kind uses crystals of CaMnO3, etc. for the material of the air electrode, but its electric resistance is big and in order to avert this, it is necessary to make the film thickness of the air electrode big. However, in such a case, the entry of the air into its inside worsens and the cell performance cannot develop satisfactorily. In view of the above, in order to obtain a high performance solid electrolytic fuel cell which can improve electric conductivity without damaging diffusion rate of the air, this invention proposes with regard to the aforementioned solid electrolytic fuel cell to install a heat resistant and conductive member inside the above air electrode. 6 figs.

  6. Spiral cooled fuel nozzle

    Energy Technology Data Exchange (ETDEWEB)

    Fox, Timothy; Schilp, Reinhard

    2012-09-25

    A fuel nozzle for delivery of fuel to a gas turbine engine. The fuel nozzle includes an outer nozzle wall and a center body located centrally within the nozzle wall. A gap is defined between an inner wall surface of the nozzle wall and an outer body surface of the center body for providing fuel flow in a longitudinal direction from an inlet end to an outlet end of the fuel nozzle. A turbulating feature is defined on at least one of the central body and the inner wall for causing at least a portion of the fuel flow in the gap to flow transverse to the longitudinal direction. The gap is effective to provide a substantially uniform temperature distribution along the nozzle wall in the circumferential direction.

  7. Fuel cycle analysis of once-through nuclear systems.

    Energy Technology Data Exchange (ETDEWEB)

    Kim, T. K.; Taiwo, T. A.; Nuclear Engineering Division

    2010-08-10

    Once-through fuel cycle systems are commercially used for the generation of nuclear power, with little exception. The bulk of these once-through systems have been water-cooled reactors (light-water and heavy water reactors, LWRs and HWRs). Some gas-cooled reactors are used in the United Kingdom. The commercial power systems that are exceptions use limited recycle (currently one recycle) of transuranic elements, primarily plutonium, as done in Europe and nearing deployment in Japan. For most of these once-through fuel cycles, the ultimate storage of the used (spent) nuclear fuel (UNF, SNF) will be in a geologic repository. Besides the commercial nuclear plants, new once-through concepts are being proposed for various objectives under international advanced nuclear fuel cycle studies and by industrial and venture capital groups. Some of the objectives for these systems include: (1) Long life core for remote use or foreign export and to support proliferation risk reduction goals - In these systems the intent is to achieve very long core-life with no refueling and limited or no access to the fuel. Most of these systems are fast spectrum systems and have been designed with the intent to improve plant economics, minimize nuclear waste, enhance system safety, and reduce proliferation risk. Some of these designs are being developed under Generation IV International Forum activities and have generally not used fuel blankets and have limited the fissile content of the fuel to less than 20% for the purpose on meeting international nonproliferation objectives. In general, the systems attempt to use transuranic elements (TRU) produced in current commercial nuclear power plants as this is seen as a way to minimize the amount of the problematic radio-nuclides that have to be stored in a repository. In this case, however, the reprocessing of the commercial LWR UNF to produce the initial fuel will be necessary. For this reason, some of the systems plan to use low enriched uranium

  8. Fuel safety research 2001

    Energy Technology Data Exchange (ETDEWEB)

    Uetsuka, Hiroshi (ed.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-11-01

    The Fuel Safety Research Laboratory is in charge of research activity which covers almost research items related to fuel safety of water reactor in JAERI. Various types of experimental and analytical researches are being conducted by using some unique facilities such as the Nuclear Safety Research Reactor (NSRR), the Japan Material Testing Reactor (JMTR), the Japan Research Reactor 3 (JRR-3) and the Reactor Fuel Examination Facility (RFEF) of JAERI. The research to confirm the safety of high burn-up fuel and MOX fuel under accident conditions is the most important item among them. The laboratory consists of following five research groups corresponding to each research fields; Research group of fuel behavior under the reactivity initiated accident conditions (RIA group). Research group of fuel behavior under the loss-of-coolant accident conditions (LOCA group). Research group of fuel behavior under the normal operation conditions (JMTR/BOCA group). Research group of fuel behavior analysis (FEMAXI group). Research group of radionuclides release and transport behavior from irradiated fuel under severe accident conditions (VEGA group). The research conducted in the year 2001 produced many important data and information. They are, for example, the fuel behavior data under BWR power oscillation conditions in the NSRR, the data on failure-bearing capability of hydrided cladding under LOCA conditions and the FP release data at very high temperature in steam which simulate the reactor core condition during severe accidents. This report summarizes the outline of research activities and major outcomes of the research executed in 2001 in the Fuel Safety Research Laboratory. (author)

  9. Liquid fuel cells.

    Science.gov (United States)

    Soloveichik, Grigorii L

    2014-01-01

    The advantages of liquid fuel cells (LFCs) over conventional hydrogen-oxygen fuel cells include a higher theoretical energy density and efficiency, a more convenient handling of the streams, and enhanced safety. This review focuses on the use of different types of organic fuels as an anode material for LFCs. An overview of the current state of the art and recent trends in the development of LFC and the challenges of their practical implementation are presented.

  10. Portable Fuel Quality Analyzer

    Science.gov (United States)

    2014-01-27

    other transportation industries, such as trucking. The PFQA could also be used in fuel blending operations performed at petroleum, ethanol and biodiesel plants. ...used to identify fuel type and determine performance properties. The Phase I measurements identified the best spectral resolution, spectral region and...identified the best spectral resolution, spectral region and sample path length to differentiate between diesel and jet fuels, as well as to determine

  11. Alternative Fuels (Briefing Charts)

    Science.gov (United States)

    2009-06-19

    feedstock for HRJ, plant cost for F-T) Courtesy AFRL, Dr. Tim Edwards Unclassified • Agricultural crop oils (canola, jatropha, soy, palm , etc...Fuels Focus  Various conversion processes  Upgraded to meet fuel specs Diverse energy sources Petroleum Crude Oil Petroleum based Single Fuel in the...University of North Dakota EERC – UOP – General Electric (GE) – Swedish Biofuels AB • Cellulosic and algal feedstocks that are non- competitive with

  12. Fuel Tank Technology

    Science.gov (United States)

    1989-11-01

    structures b) - Equal thermic inertia c) - Equal fluid volume d) - Equal pressure variation on both wings at the change of the room temperature - This...individual fuel sections. Each fuel section is further ccmpartmentated by metall tank shear walls and tank floors into three individual fuel cells to...plate Dy a stretch forming process, and the metallic tank floors . The air intake segments extend from one bulkhead to the other, thus reducing assembly

  13. Microstructural Characterization of a Mg Matrix U-Mo Dispersion Fuel Plate Irradiated in the Advanced Test Reactor to High Fission Density: SEM Results

    Science.gov (United States)

    Keiser, Dennis D.; Jue, Jan-Fong; Miller, Brandon D.; Gan, Jian; Robinson, Adam B.; Medvedev, Pavel G.; Madden, James W.; Moore, Glenn A.

    2016-06-01

    Low-enriched (U-235 reactors. In most cases, fuel plates with Al or Al-Si alloy matrices have been tested in the Advanced Test Reactor to support this development. In addition, fuel plates with Mg as the matrix have also been tested. The benefit of using Mg as the matrix is that it potentially will not chemically interact with the U-Mo fuel particles during fabrication or irradiation, whereas with Al and Al-Si alloys such interactions will occur. Fuel plate R9R010 is a Mg matrix fuel plate that was aggressively irradiated in ATR. This fuel plate was irradiated as part of the RERTR-8 experiment at high temperature, high fission rate, and high power, up to high fission density. This paper describes the results of the scanning electron microscopy (SEM) analysis of an irradiated fuel plate using polished samples and those produced with a focused ion beam. A follow-up paper will discuss the results of transmission electron microscopy (TEM) analysis. Using SEM, it was observed that even at very aggressive irradiation conditions, negligible chemical interaction occurred between the irradiated U-7Mo fuel particles and Mg matrix; no interconnection of fission gas bubbles from fuel particle to fuel particle was observed; the interconnected fission gas bubbles that were observed in the irradiated U-7Mo particles resulted in some transport of solid fission products to the U-7Mo/Mg interface; the presence of microstructural pathways in some U-9.1 Mo particles that could allow for transport of fission gases did not result in the apparent presence of large porosity at the U-7Mo/Mg interface; and, the Mg-Al interaction layers that were present at the Mg matrix/Al 6061 cladding interface exhibited good radiation stability, i.e. no large pores.

  14. Fuel Cell Demonstration Program

    Energy Technology Data Exchange (ETDEWEB)

    Gerald Brun

    2006-09-15

    In an effort to promote clean energy projects and aid in the commercialization of new fuel cell technologies the Long Island Power Authority (LIPA) initiated a Fuel Cell Demonstration Program in 1999 with six month deployments of Proton Exchange Membrane (PEM) non-commercial Beta model systems at partnering sites throughout Long Island. These projects facilitated significant developments in the technology, providing operating experience that allowed the manufacturer to produce fuel cells that were half the size of the Beta units and suitable for outdoor installations. In 2001, LIPA embarked on a large-scale effort to identify and develop measures that could improve the reliability and performance of future fuel cell technologies for electric utility applications and the concept to establish a fuel cell farm (Farm) of 75 units was developed. By the end of October of 2001, 75 Lorax 2.0 fuel cells had been installed at the West Babylon substation on Long Island, making it the first fuel cell demonstration of its kind and size anywhere in the world at the time. Designed to help LIPA study the feasibility of using fuel cells to operate in parallel with LIPA's electric grid system, the Farm operated 120 fuel cells over its lifetime of over 3 years including 3 generations of Plug Power fuel cells (Lorax 2.0, Lorax 3.0, Lorax 4.5). Of these 120 fuel cells, 20 Lorax 3.0 units operated under this Award from June 2002 to September 2004. In parallel with the operation of the Farm, LIPA recruited government and commercial/industrial customers to demonstrate fuel cells as on-site distributed generation. From December 2002 to February 2005, 17 fuel cells were tested and monitored at various customer sites throughout Long Island. The 37 fuel cells operated under this Award produced a total of 712,635 kWh. As fuel cell technology became more mature, performance improvements included a 1% increase in system efficiency. Including equipment, design, fuel, maintenance

  15. Fuels Processing Laboratory

    Data.gov (United States)

    Federal Laboratory Consortium — NETL’s Fuels Processing Laboratory in Morgantown, WV, provides researchers with the equipment they need to thoroughly explore the catalytic issues associated with...

  16. Fuel assembly reconstitution

    Energy Technology Data Exchange (ETDEWEB)

    Morgado, Mario M.; Oliveira, Monica G.N.; Ferreira Junior, Decio B.M.; Santos, Barbara O. dos; Santos, Jorge E. dos, E-mail: mongeor@eletronuclear.gov.b [ELETROBRAS Termonuclear S.A. - ELETRONUCLEAR, Angra dos Reis, RJ (Brazil)

    2009-07-01

    Fuel failures have been happened in Nuclear Power Plants worldwide, without lost of integrity and safety, mainly for the public, environment and power plants workers. The most common causes of these events are corrosion (CRUD), fretting and pellet cladding interaction. These failures are identified by increasing the activity of fission products, verified by chemical analyses of reactor coolant. Through these analyses, during the fourth operation cycle of Angra 2 Nuclear Power Plant, was possible to observe fuel failure indication. This indication was confirmed in the end of the cycle during the unloading of reactor core through leakage tests of fuel assembly, using the equipment called 'In Mast Sipping' and 'Box Sipping'. After confirmed, the fuel assembly reconstitution was scheduled, and happened in April, 2007, where was identified the cause and the fuel rod failure, which was substitute by dummy rods (zircaloy). The cause was fretting by 'debris'. The actions to avoid and prevent fuel assemblies failures are important. The goals of this work are to describe the methodology of fuel assembly reconstitution using the FARE (Fuel Assembly Reconstitution Equipment) system, to describe the results of this task in economic and security factors of the company and show how the fuel assembly failures are identified during operation and during the outage. (author)

  17. Rethinking nuclear fuel recycling.

    Science.gov (United States)

    von Hippel, Frank N

    2008-05-01

    Spent nuclear fuel contains plutonium which can be extracted and used in new fuel. To reduce the amount of long-lived radioactive waste, the U.S. Department of Energy has proposed reprocessing spent fuel in this way and then "burning" the plutonium in special reactors. But reprocesssing is very expensive. Also, spent fuel emits lethal radiation, whereas separated plutonium can be handled easily. So reprocessing invites the possibility that terrorists might steal plutonium and construct an atom bormb. The authors argue against reprocessing and for storing the waste in casks until an underground repository is ready.

  18. Fuel Cells: Reshaping the Future

    Science.gov (United States)

    Toay, Leo

    2004-01-01

    In conjunction with the FreedomCAR (Cooperative Automotive Research) and Fuel Initiative, President George W. Bush has pledged nearly two billion dollars for fuel cell research. Chrysler, Ford, and General Motors have unveiled fuel cell demonstration vehicles, and all three of these companies have invested heavily in fuel cell research. Fuel cell…

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-09-01

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

  20. Framing car fuel efficiency : linearity heuristic for fuel consumption and fuel-efficiency ratings

    NARCIS (Netherlands)

    Schouten, T.M.; Bolderdijk, J.W.; Steg, L.

    2014-01-01

    People are sensitive to the way information on fuel efficiency is conveyed. When the fuel efficiency of cars is framed in terms of fuel per distance (FPD; e.g. l/100 km), instead of distance per units of fuel (DPF; e.g. km/l), people have a more accurate perception of potential fuel savings. People

  1. 77 FR 13009 - Regulation of Fuels and Fuel Additives: Identification of Additional Qualifying Renewable Fuel...

    Science.gov (United States)

    2012-03-05

    ... AGENCY 40 CFR Part 80 RIN 2060-AR07 Regulation of Fuels and Fuel Additives: Identification of Additional Qualifying Renewable Fuel Pathways Under the Renewable Fuel Standard Program AGENCY: Environmental Protection... January 5, 2012 to amend the Renewable Fuel Standard program regulations. Because EPA received...

  2. www.FuelEconomy.gov

    Data.gov (United States)

    U.S. Environmental Protection Agency — FuelEconomy.gov provides comprehensive information about vehicles' fuel economy. The official U.S. government site for fuel economy information, it is operated by...

  3. Improved hybrid rocket fuel

    Science.gov (United States)

    Dean, David L.

    1995-01-01

    McDonnell Douglas Aerospace, as part of its Independent R&D, has initiated development of a clean burning, high performance hybrid fuel for consideration as an alternative to the solid rocket thrust augmentation currently utilized by American space launch systems including Atlas, Delta, Pegasus, Space Shuttle, and Titan. It could also be used in single stage to orbit or as the only propulsion system in a new launch vehicle. Compared to solid propellants based on aluminum and ammonium perchlorate, this fuel is more environmentally benign in that it totally eliminates hydrogen chloride and aluminum oxide by products, producing only water, hydrogen, nitrogen, carbon oxides, and trace amounts of nitrogen oxides. Compared to other hybrid fuel formulations under development, this fuel is cheaper, denser, and faster burning. The specific impulse of this fuel is comparable to other hybrid fuels and is between that of solids and liquids. The fuel also requires less oxygen than similar hybrid fuels to produce maximum specific impulse, thus reducing oxygen delivery system requirements.

  4. Durable fuel electrode

    DEFF Research Database (Denmark)

    2017-01-01

    the composite. The invention also relates to the use of the composite as a fuel electrode, solid oxide fuel cell, and/or solid oxide electrolyser. The invention discloses a composite for an electrode, comprising a three-dimensional network of dispersed metal particles, stabilised zirconia particles and pores...

  5. Toward sustainable fuel cells

    DEFF Research Database (Denmark)

    Stephens, Ifan; Rossmeisl, Jan; Chorkendorff, Ib

    2016-01-01

    to a regular gasoline car. However, current fuel cells require 0.25 g of platinum (Pt) per kilowatt of power (2) as catalysts to drive the electrode reactions. If the entire global annual production of Pt were devoted to fuel cell vehicles, fewer than 10 million vehicles could be produced each year, a mere 10...

  6. MICROBIAL FUEL CELL

    DEFF Research Database (Denmark)

    2008-01-01

    A novel microbial fuel cell construction for the generation of electrical energy. The microbial fuel cell comprises: (i) an anode electrode, (ii) a cathode chamber, said cathode chamber comprising an in let through which an influent enters the cathode chamber, an outlet through which an effluent...

  7. Solar Fuel Generator

    Science.gov (United States)

    Lewis, Nathan S. (Inventor); West, William C. (Inventor)

    2017-01-01

    The disclosure provides conductive membranes for water splitting and solar fuel generation. The membranes comprise an embedded semiconductive/photoactive material and an oxygen or hydrogen evolution catalyst. Also provided are chassis and cassettes containing the membranes for use in fuel generation.

  8. Bioethanol: fuel or feedstock?

    DEFF Research Database (Denmark)

    Rass-Hansen, Jeppe; Falsig, Hanne; Jørgensen, Betina

    2007-01-01

    Increasing amounts of bioethanol are being produced from fermentation of biomass, mainly to counteract the continuing depletion of fossil resources and the consequential escalation of oil prices. Today, bioethanol is mainly utilized as a fuel or fuel additive in motor vehicles, but it could also...

  9. Nanofluidic fuel cell

    Science.gov (United States)

    Lee, Jin Wook; Kjeang, Erik

    2013-11-01

    Fuel cells are gaining momentum as a critical component in the renewable energy mix for stationary, transportation, and portable power applications. State-of-the-art fuel cell technology benefits greatly from nanotechnology applied to nanostructured membranes, catalysts, and electrodes. However, the potential of utilizing nanofluidics for fuel cells has not yet been explored, despite the significant opportunity of harnessing rapid nanoscale reactant transport in close proximity to the reactive sites. In the present article, a nanofluidic fuel cell that utilizes fluid flow through nanoporous media is conceptualized and demonstrated for the first time. This transformative concept captures the advantages of recently developed membraneless and catalyst-free fuel cell architectures paired with the enhanced interfacial contact area enabled by nanofluidics. When compared to previously reported microfluidic fuel cells, the prototype nanofluidic fuel cell demonstrates increased surface area, reduced activation overpotential, superior kinetic characteristics, and moderately enhanced fuel cell performance in the high cell voltage regime with up to 14% higher power density. However, the expected mass transport benefits in the high current density regime were constrained by high ohmic cell resistance, which could likely be resolved through future optimization studies.

  10. Are Solar Fuels Sustainable?

    NARCIS (Netherlands)

    Meuwese, Anne

    2012-01-01

    Summary The combined problems of too little fossil fuels to supply the world’s future energy needs and the possible negative environmental effects of carbon dioxide emissions which are coupled to their usage has led to the development of fuels based on s

  11. Fuel cells: Operating flexibly

    Science.gov (United States)

    Lee, Young Moo

    2016-09-01

    Fuel cells typically function well only in rather limited temperature and humidity ranges. Now, a proton exchange membrane consisting of ion pair complexes is shown to enable improved fuel cell performance under a wide range of conditions that are unattainable with conventional approaches.

  12. Are Solar Fuels Sustainable?

    NARCIS (Netherlands)

    Meuwese, Anne

    2012-01-01

    Summary The combined problems of too little fossil fuels to supply the world’s future energy needs and the possible negative environmental effects of carbon dioxide emissions which are coupled to their usage has led to the development of fuels based on s

  13. Solar fuel generator

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, Nathan S.; West, William C.

    2017-01-17

    The disclosure provides conductive membranes for water splitting and solar fuel generation. The membranes comprise an embedded semiconductive/photoactive material and an oxygen or hydrogen evolution catalyst. Also provided are chassis and cassettes containing the membranes for use in fuel generation.

  14. CO2-Neutral Fuels

    NARCIS (Netherlands)

    Goede, A.; van de Sanden, M. C. M.

    2016-01-01

    Mimicking the biogeochemical cycle of System Earth, synthetic hydrocarbon fuels are produced from recycled CO2 and H2O powered by renewable energy. Recapturing CO2 after use closes the carbon cycle, rendering the fuel cycle CO2 neutral. Non-equilibrium molecular CO2 vibrations are key to high energy

  15. CO2-Neutral Fuels

    Science.gov (United States)

    Goede, Adelbert; van de Sanden, Richard

    2016-06-01

    Mimicking the biogeochemical cycle of System Earth, synthetic hydrocarbon fuels are produced from recycled CO2 and H2O powered by renewable energy. Recapturing CO2 after use closes the carbon cycle, rendering the fuel cycle CO2 neutral. Non-equilibrium molecular CO2 vibrations are key to high energy efficiency.

  16. CO2-Neutral Fuels

    NARCIS (Netherlands)

    Goede, A.; van de Sanden, M. C. M.

    2016-01-01

    Mimicking the biogeochemical cycle of System Earth, synthetic hydrocarbon fuels are produced from recycled CO2 and H2O powered by renewable energy. Recapturing CO2 after use closes the carbon cycle, rendering the fuel cycle CO2 neutral. Non-equilibrium molecular CO2 vibrations are key to high energy

  17. PLATINUM AND FUEL CELLS

    Science.gov (United States)

    Platinum requirements for fuel cell vehicles (FCVS) have been identified as a concern and possible problem with FCV market penetration. Platinum is a necessary component of the electrodes of fuel cell engines that power the vehicles. The platinum is deposited on porous electrodes...

  18. Hydrogen Fuel Quality

    Energy Technology Data Exchange (ETDEWEB)

    Rockward, Tommy [Los Alamos National Laboratory

    2012-07-16

    For the past 6 years, open discussions and/or meetings have been held and are still on-going with OEM, Hydrogen Suppliers, other test facilities from the North America Team and International collaborators regarding experimental results, fuel clean-up cost, modeling, and analytical techniques to help determine levels of constituents for the development of an international standard for hydrogen fuel quality (ISO TC197 WG-12). Significant progress has been made. The process for the fuel standard is entering final stages as a result of the technical accomplishments. The objectives are to: (1) Determine the allowable levels of hydrogen fuel contaminants in support of the development of science-based international standards for hydrogen fuel quality (ISO TC197 WG-12); and (2) Validate the ASTM test method for determining low levels of non-hydrogen constituents.

  19. Assessment of automotive fuels

    Science.gov (United States)

    Isenberg, G.

    Energy demand all over the world increases steadily and, within the next decades, is almost completely met by fossil fuels. This poses increasing pressure on oil supply and reserves. Concomitant is the concern about environmental pollution, especially by carbon dioxide from fossil fuel combustion, with the risk of global warming. Environmental well-being requires a modified mix of energy sources to emit less carbon dioxide, starting with a move to natural gas and ending with the market penetration of renewable energies. Efforts should focus on advanced oil and gas production and processing technologies and on regeneratively produced fuels like hydrogen or bio-fuels as well. Within the framework of an industrial initiative in Germany, a process of defining one or two alternative fuels was started, to bring them into the market within the next years.

  20. Fuels and Combustion

    KAUST Repository

    Johansson, Bengt

    2016-08-17

    This chapter discusses the combustion processes and the link to the fuel properties that are suitable for them. It describes the basic three concepts, including spark ignition (SI) and compression ignition (CI), and homogeneous charge compression ignition (HCCI). The fuel used in a CI engine is vastly different from that in an SI engine. In an SI engine, the fuel should sustain high pressure and temperature without autoignition. Apart from the dominating SI and CI engines, it is also possible to operate with a type of combustion: autoignition. With HCCI, the fuel and air are fully premixed before combustion as in the SI engine, but combustion is started by the increased pressure and temperature during the compression stroke. Apart from the three combustion processes, there are also a few combined or intermediate concepts, such as Spark-Assisted Compression Ignition (SACI). Those concepts are discussed in terms of the requirements of fuel properties.

  1. Rejuvenation of automotive fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yu Seung; Langlois, David A.

    2016-08-23

    A process for rejuvenating fuel cells has been demonstrated to improve the performance of polymer exchange membrane fuel cells with platinum/ionomer electrodes. The process involves dehydrating a fuel cell and exposing at least the cathode of the fuel cell to dry gas (nitrogen, for example) at a temperature higher than the operating temperature of the fuel cell. The process may be used to prolong the operating lifetime of an automotive fuel cell.

  2. Status of the Combined Third and Fourth NGNP Fuel Irradiations In the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover; David A. Petti; Michael E. Davenport

    2013-07-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and is currently scheduled to be completed in September 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and is currently scheduled to be completed in April 2014. Since the purpose of this combined experiment is to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment is

  3. A model to predict failure of irradiated U–Mo dispersion fuel

    Energy Technology Data Exchange (ETDEWEB)

    Burkes, Douglas E., E-mail: Douglas.Burkes@pnnl.gov; Senor, David J.; Casella, Andrew M.

    2016-12-15

    Highlights: • Simple model to predict failure of dispersion fuel meat designs. • Evaluated as a function of fabrication parameters and irradiation conditions. • Predictions compare well with experimental measurements of miniature fuel plates. • Interaction layer formation reduces matrix strength and increases temperature. • Si additions to the matrix appear effective only at moderate heat flux and burnup. - Abstract: 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 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. The current paper extends a failure model originally developed for UO{sub 2}-stainless steel dispersion fuels and uses currently available thermal–mechanical property information for the materials of interest in the currently proposed design. A number of fabrication and irradiation parameters were investigated to understand the conditions at which failure of the matrix, classified as onset of pore formation in the matrix, might occur. The results compared well with experimental observations published as part of the Reduced Enrichment for Research and Test Reactors (RERTR)-6 and -7 mini-plate experiments. Fission rate, a function of the {sup 235}U enrichment, appeared to be the most influential parameter in premature failure, mainly as a result of increased interaction layer formation and operational temperature, which coincidentally decreased the strength of the matrix and caused more rapid fission gas production and recoil into the surrounding matrix material. Addition of silicon to the matrix appeared effective at reducing the rate of

  4. A risk-informed evaluation of MOX fuel loading in PWRS

    Energy Technology Data Exchange (ETDEWEB)

    Lyman, E.S. [Nuclear Control Institute, Washington, DC (United States)

    2001-07-01

    The full text follows: The U.S. Department of Energy (DOE) has signed a contract with Duke Cogema Stone and Webster (DCS) for fabrication of mixed-oxide (MOX) fuel and irradiation of the MOX fuel at the Catawba and McGuire pressurized-water reactors (PWRs), operated by Duke Power. The first load of MOX fuel is scheduled for 2007. In order to use MOX in these plants, Duke Power will have to apply to the Nuclear Regulatory Commission (NRC) for amendments to their operating licenses. Until recently, there have been no numerical guidelines for determining the acceptability of license amendment requests. However, such guidelines are now at hand with the adoption in 1998 of NRC Regulatory Guide 1.174, which defines a maximum value for the permissible increase in risk to the public resulting from a proposed change to a nuclear plant's licensing basis (LB). The substitution of MOX fuel for low-enriched uranium (LEU) fuel in LWRs will have an impact on risk to the public that will require regulatory evaluation. One of the major differences is that use of MOX will increase the inventories of plutonium and minor actinides in the reactor core, thereby increasing the source term for certain severe accidents, such as a core melt with early containment failure or a spent fuel pool drain-down. The goal of this paper is to quantitatively evaluate the increase in risk associated with the greater actinide source term in MOX-fueled reactors, and to compare this increase with RG 1.174 guidelines. Standard computer programs (SCALE and MACCS2) are used to estimate the increase in severe accident risk to the public associated with the DCS plan to use 40% cores of weapons-grade MOX fuel. These values are then compared to the RG 1.174 acceptance criteria, using publicly available risk information. Since RG 1.174 guidelines are based on the assumption that severe accident source terms are not affected by LB changes, the RG 1.174 formalism must be modified for this case. A similar

  5. Alternate-Fueled Flight: Halophytes, Algae, Bio-, and Synthetic Fuels

    Science.gov (United States)

    Hendricks, R. C.

    2012-01-01

    Synthetic and biomass fueling are now considered to be near-term aviation alternate fueling. The major impediment is a secure sustainable supply of these fuels at reasonable cost. However, biomass fueling raises major concerns related to uses of common food crops and grasses (some also called "weeds") for processing into aviation fuels. These issues are addressed, and then halophytes and algae are shown to be better suited as sources of aerospace fuels and transportation fueling in general. Some of the history related to alternate fuels use is provided as a guideline for current and planned alternate fuels testing (ground and flight) with emphasis on biofuel blends. It is also noted that lessons learned from terrestrial fueling are applicable to space missions. These materials represent an update (to 2009) and additions to the Workshop on Alternate Fueling Sustainable Supply and Halophyte Summit at Twinsburg, Ohio, October 17 to 18, 2007.

  6. Low contaminant formic acid fuel for direct liquid fuel cell

    Science.gov (United States)

    Masel, Richard I.; Zhu, Yimin; Kahn, Zakia; Man, Malcolm

    2009-11-17

    A low contaminant formic acid fuel is especially suited toward use in a direct organic liquid fuel cell. A fuel of the invention provides high power output that is maintained for a substantial time and the fuel is substantially non-flammable. Specific contaminants and contaminant levels have been identified as being deleterious to the performance of a formic acid fuel in a fuel cell, and embodiments of the invention provide low contaminant fuels that have improved performance compared to known commercial bulk grade and commercial purified grade formic acid fuels. Preferred embodiment fuels (and fuel cells containing such fuels) including low levels of a combination of key contaminants, including acetic acid, methyl formate, and methanol.

  7. Direct Methanol Fuel Cell, DMFC

    Directory of Open Access Journals (Sweden)

    Amornpitoksuk, P.

    2003-09-01

    Full Text Available Direct Methanol Fuel Cell, DMFC is a kind of fuel cell using methanol as a fuel for electric producing. Methanol is low cost chemical substance and it is less harmful than that of hydrogen fuel. From these reasons it can be commercial product. The electrocatalytic reaction of methanol fuel uses Pt-Ru metals as the most efficient catalyst. In addition, the property of membrane and system designation are also effect to the fuel cell efficient. Because of low power of methanol fuel cell therefore, direct methanol fuel cell is proper to use for the energy source of small electrical devices and vehicles etc.

  8. Preliminary Results of an On-Line, Multi-Spectrometer Fission Product Monitoring System to Support Advanced Gas Reactor Fuel Testing and Qualification in the Advanced Test Reactor at the Idaho National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Dawn M. Scates; John K. Hartwell; John B. Walter; Mark W. Drigert

    2007-10-01

    The Advanced Gas Reactor -1 (AGR-1) experiment is the first experiment in a series of eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotropic (TRISO) particle fuel (in compact form) experiments scheduled for placement in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The experiment began irradiation in the ATR with a cycle that reached full power on December 26, 2006 and will continue irradiation for about 2.5 years. During this time six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The goals of the irradiation experiment is to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. This paper presents the preliminary test details of the fuel performance, as measured by the control and acquisition software.

  9. Halden Gd fuel experiments IFA-515, IFA-636 and IFA-681

    Energy Technology Data Exchange (ETDEWEB)

    Tolonen, Pekka; Tverberg, Terje

    2004-07-01

    The nuclear power industry has used gadolinia (Gd) as a burnable poison widely in Boiling Water Reactors (BWR) since the 1970's, and its use in Pressurised Water Reactors (PWR) as well has become common since the 1990's. Given the poorer thermal conductivity of Gd bearing fuels in comparison with UO{sub 2} fuels, the Gd fuel power and discharge burnup are generally limited by lower enrichment to compensate for the relatively higher fuel temperature. The global tendency of increasing discharge burnup and thus operating powers to improve fuel economy brings about a demand to improve the knowledge on the thermal-mechanical performance characteristics of Gd bearing fuels and hence, reduce the reactivity penalty due to large margins in fuel design and low enrichment. Aiming to respond to the demand for more profound knowledge on Gd fuel performance, the OECD Halden Reactor Project is conducting a series of Gd experiments, which started with the special Gd fuel experiment IFA-515.10 from 1994 to 2000. The experiment was dedicated to the thermal conductivity study of non absorbing low diameter Gd fuel pellets up to an average burnup of 76 MWd/kgUO{sub 2}. Since 1998 it has been followed by an comparative integral Gd/UO{sub 2} fuel performance experiment IFA-636.1 loaded with production line pellets fabricated by ENUSA. This experiment has generated data on key thermal-mechanical performance issues, such as fuel densification and swelling and fission gas release. However, two instrumentation failures resulted in partial loss of data from the experiment and the low initial operating power in Gd rods left doubts about the initial fuel densification results. The experiment has accumulated maximum burnup of approx25 MWd/kgUO{sub 2} in Gd fuel, and will be discharged in 2004. Some of the rods will be subjected to post irradiation examinations (PIE) at the Kjeller hot cell laboratory, while options exist to load selected rods into other experiment rigs for further

  10. AFIP-6 Irradiation Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Danielle M Perez; M. A. Lillo; G. S. Chang; G. A. Roth; N. E. Woolstenhulme; D. M. Wachs

    2011-09-01

    The Advanced Test Reactor (ATR) Full size plate In center flux trap Position (AFIP) experiment AFIP-6 was designed to evaluate the performance of monolithic uranium-molybdenum (U-Mo) fuels at a length prototypic to that of the ATR fuel plates (45 inches in length). The AFIP-6 test was the first test with plates in a swaged condition with longer fuel zones of approximately 22.5 inches in length1,2. The following report summarizes the life of the AFIP-6 experiment through end of irradiation, including a brief description of the safety analysis, as-run neutronic analysis results, hydraulic testing results, and thermal analysis results.

  11. Fuel cycle analysis of once-through nuclear systems.

    Energy Technology Data Exchange (ETDEWEB)

    Kim, T. K.; Taiwo, T. A.; Nuclear Engineering Division

    2010-08-10

    Once-through fuel cycle systems are commercially used for the generation of nuclear power, with little exception. The bulk of these once-through systems have been water-cooled reactors (light-water and heavy water reactors, LWRs and HWRs). Some gas-cooled reactors are used in the United Kingdom. The commercial power systems that are exceptions use limited recycle (currently one recycle) of transuranic elements, primarily plutonium, as done in Europe and nearing deployment in Japan. For most of these once-through fuel cycles, the ultimate storage of the used (spent) nuclear fuel (UNF, SNF) will be in a geologic repository. Besides the commercial nuclear plants, new once-through concepts are being proposed for various objectives under international advanced nuclear fuel cycle studies and by industrial and venture capital groups. Some of the objectives for these systems include: (1) Long life core for remote use or foreign export and to support proliferation risk reduction goals - In these systems the intent is to achieve very long core-life with no refueling and limited or no access to the fuel. Most of these systems are fast spectrum systems and have been designed with the intent to improve plant economics, minimize nuclear waste, enhance system safety, and reduce proliferation risk. Some of these designs are being developed under Generation IV International Forum activities and have generally not used fuel blankets and have limited the fissile content of the fuel to less than 20% for the purpose on meeting international nonproliferation objectives. In general, the systems attempt to use transuranic elements (TRU) produced in current commercial nuclear power plants as this is seen as a way to minimize the amount of the problematic radio-nuclides that have to be stored in a repository. In this case, however, the reprocessing of the commercial LWR UNF to produce the initial fuel will be necessary. For this reason, some of the systems plan to use low enriched uranium

  12. Behaviour of fission gas in the rim region of high burn-up UO 2 fuel pellets with particular reference to results from an XRF investigation

    Science.gov (United States)

    Mogensen, M.; Pearce, J. H.; Walker, C. T.

    1999-01-01

    XRF and EPMA results for retained xenon from Battelle's high burn-up effects program are re-evaluated. The data reviewed are from commercial low enriched BWR fuel with burn-ups of 44.8-54.9 GWd/tU and high enriched PWR fuel with burn-ups from 62.5 to 83.1 GWd/tU. It is found that the high burn-up structure penetrated much deeper than initially reported. The local burn-up threshold for the formation of the high burn-up structure in those fuels with grain sizes in the normal range lay between 60 and 75 GWd/tU. The high burn-up structure was not detected by EPMA in a fuel that had a grain size of 78 μm although the local burn-up at the pellet rim had exceeded 80 GWd/tU. It is concluded that fission gas had been released from the high burn-up structure in three PWR fuel sections with burn-ups of 70.4, 72.2 and 83.1 GWd/tU. In the rim region of the last two sections at the locations where XRF indicated gas release the local burn-up was higher than 75 GWd/tU.

  13. Metallic fuels for advanced reactors

    Science.gov (United States)

    Carmack, W. J.; Porter, D. L.; Chang, Y. I.; Hayes, S. L.; Meyer, M. K.; Burkes, D. E.; Lee, C. B.; Mizuno, T.; Delage, F.; Somers, J.

    2009-07-01

    In the framework of the Generation IV Sodium Fast Reactor Program, the Advanced Fuel Project has conducted an evaluation of the available fuel systems supporting future sodium cooled fast reactors. This paper presents an evaluation of metallic alloy fuels. Early US fast reactor developers originally favored metal alloy fuel due to its high fissile density and compatibility with sodium. The goal of fast reactor fuel development programs is to develop and qualify a nuclear fuel system that performs all of the functions of a conventional fast spectrum nuclear fuel while destroying recycled actinides. This will provide a mechanism for closure of the nuclear fuel cycle. Metal fuels are candidates for this application, based on documented performance of metallic fast reactor fuels and the early results of tests currently being conducted in US and international transmutation fuel development programs.

  14. 2009 Fuel Cell Market Report

    Energy Technology Data Exchange (ETDEWEB)

    Vincent, Bill [Breakthrough Technologies Inst., Washington, DC (United States); Gangi, Jennifer [Breakthrough Technologies Inst., Washington, DC (United States); Curtin, Sandra [Breakthrough Technologies Inst., Washington, DC (United States); Delmont, Elizabeth [Breakthrough Technologies Inst., Washington, DC (United States)

    2010-11-01

    Fuel cells are electrochemical devices that combine hydrogen and oxygen to produce electricity, water, and heat. Unlike batteries, fuel cells continuously generate electricity, as long as a source of fuel is supplied. Moreover, fuel cells do not burn fuel, making the process quiet, pollution-free and two to three times more efficient than combustion. Fuel cell systems can be a truly zero-emission source of electricity, if the hydrogen is produced from non-polluting sources. Global concerns about climate change, energy security, and air pollution are driving demand for fuel cell technology. More than 630 companies and laboratories in the United States are investing $1 billion a year in fuel cells or fuel cell component technologies. This report provides an overview of trends in the fuel cell industry and markets, including product shipments, market development, and corporate performance. It also provides snapshots of select fuel cell companies, including general.

  15. Alkaline fuel cells applications

    Science.gov (United States)

    Kordesch, Karl; Hacker, Viktor; Gsellmann, Josef; Cifrain, Martin; Faleschini, Gottfried; Enzinger, Peter; Fankhauser, Robert; Ortner, Markus; Muhr, Michael; Aronson, Robert R.

    On the world-wide automobile market technical developments are increasingly determined by the dramatic restriction on emissions as well as the regimentation of fuel consumption by legislation. Therefore there is an increasing chance of a completely new technology breakthrough if it offers new opportunities, meeting the requirements of resource preservation and emission restrictions. Fuel cell technology offers the possibility to excel in today's motive power techniques in terms of environmental compatibility, consumer's profit, costs of maintenance and efficiency. The key question is economy. This will be decided by the costs of fuel cell systems if they are to be used as power generators for future electric vehicles. The alkaline hydrogen-air fuel cell system with circulating KOH electrolyte and low-cost catalysed carbon electrodes could be a promising alternative. Based on the experiences of Kordesch [K. Kordesch, Brennstoffbatterien, Springer, Wien, 1984, ISBN 3-387-81819-7; K. Kordesch, City car with H 2-air fuel cell and lead-battery, SAE Paper No. 719015, 6th IECEC, 1971], who operated a city car hybrid vehicle on public roads for 3 years in the early 1970s, improved air electrodes plus new variations of the bipolar stack assembly developed in Graz are investigated. Primary fuel choice will be a major issue until such time as cost-effective, on-board hydrogen storage is developed. Ammonia is an interesting option. The whole system, ammonia dissociator plus alkaline fuel cell (AFC), is characterised by a simple design and high efficiency.

  16. Hydrogen vehicle fueling station

    Energy Technology Data Exchange (ETDEWEB)

    Daney, D.E.; Edeskuty, F.J.; Daugherty, M.A. [Los Alamos National Lab., NM (United States)] [and others

    1995-09-01

    Hydrogen fueling stations are an essential element in the practical application of hydrogen as a vehicle fuel, and a number of issues such as safety, efficiency, design, and operating procedures can only be accurately addressed by a practical demonstration. Regardless of whether the vehicle is powered by an internal combustion engine or fuel cell, or whether the vehicle has a liquid or gaseous fuel tank, the fueling station is a critical technology which is the link between the local storage facility and the vehicle. Because most merchant hydrogen delivered in the US today (and in the near future) is in liquid form due to the overall economics of production and delivery, we believe a practical refueling station should be designed to receive liquid. Systems studies confirm this assumption for stations fueling up to about 300 vehicles. Our fueling station, aimed at refueling fleet vehicles, will receive hydrogen as a liquid and dispense it as either liquid, high pressure gas, or low pressure gas. Thus, it can refuel any of the three types of tanks proposed for hydrogen-powered vehicles -- liquid, gaseous, or hydride. The paper discusses the fueling station design. Results of a numerical model of liquid hydrogen vehicle tank filling, with emphasis on no vent filling, are presented to illustrate the usefulness of the model as a design tool. Results of our vehicle performance model illustrate our thesis that it is too early to judge what the preferred method of on-board vehicle fuel storage will be in practice -- thus our decision to accommodate all three methods.

  17. Design package for fuel retrieval system fuel handling tool modification

    Energy Technology Data Exchange (ETDEWEB)

    TEDESCHI, D.J.

    1999-03-17

    This is a design package that contains the details for a modification to a tool used for moving fuel elements during loading of MCO Fuel Baskets for the Fuel Retrieval System. The tool is called the fuel handling tool (or stinger). This document contains requirements, development design information, tests, and test reports.

  18. Design package for fuel retrieval system fuel handling tool modification

    Energy Technology Data Exchange (ETDEWEB)

    TEDESCHI, D.J.

    1998-11-09

    This is a design package that contains the details for a modification to a tool used for moving fuel elements during loading of MCO Fuel Baskets for the Fuel Retrieval System. The tool is called the fuel handling tool (or stinger). This document contains requirements, development design information, tests, and test reports.

  19. FUEL3-D: A Spatially Explicit Fractal Fuel Distribution Model

    Science.gov (United States)

    Russell A. Parsons

    2006-01-01

    Efforts to quantitatively evaluate the effectiveness of fuels treatments are hampered by inconsistencies between the spatial scale at which fuel treatments are implemented and the spatial scale, and detail, with which we model fire and fuel interactions. Central to this scale inconsistency is the resolution at which variability within the fuel bed is considered. Crown...

  20. Heating subsurface formations by oxidizing fuel on a fuel carrier

    Energy Technology Data Exchange (ETDEWEB)

    Costello, Michael; Vinegar, Harold J.

    2012-10-02

    A method of heating a portion of a subsurface formation includes drawing fuel on a fuel carrier through an opening formed in the formation. Oxidant is supplied to the fuel at one or more locations in the opening. The fuel is combusted with the oxidant to provide heat to the formation.

  1. Fuel nozzle tube retention

    Energy Technology Data Exchange (ETDEWEB)

    Cihlar, David William; Melton, Patrick Benedict

    2017-02-28

    A system for retaining a fuel nozzle premix tube includes a retention plate and a premix tube which extends downstream from an outlet of a premix passage defined along an aft side of a fuel plenum body. The premix tube includes an inlet end and a spring support feature which is disposed proximate to the inlet end. The premix tube extends through the retention plate. The spring retention feature is disposed between an aft side of the fuel plenum and the retention plate. The system further includes a spring which extends between the spring retention feature and the retention plate.

  2. Molten carbonate fuel cell

    Science.gov (United States)

    Kaun, T.D.; Smith, J.L.

    1986-07-08

    A molten electrolyte fuel cell is disclosed with an array of stacked cells and cell enclosures isolating each cell except for access to gas manifolds for the supply of fuel or oxidant gas or the removal of waste gas. The cell enclosures collectively provide an enclosure for the array and effectively avoid the problems of electrolyte migration and the previous need for compression of stack components. The fuel cell further includes an inner housing about and in cooperation with the array enclosure to provide a manifold system with isolated chambers for the supply and removal of gases. An external insulated housing about the inner housing provides thermal isolation to the cell components.

  3. Ammonia as a Suitable Fuel for Fuel Cells

    OpenAIRE

    Lan, Rong; Tao, Shanwen

    2014-01-01

    Ammonia, an important basic chemical, is produced at a scale of 150 million tons per year. Half of hydrogen produced in chemical industry is used for ammonia production. Ammonia containing 17.5 wt% hydrogen is an ideal carbon-free fuel for fuel cells. Compared to hydrogen, ammonia has many advantages. In this mini-review, the suitability of ammonia as fuel for fuel cells, the development of different types of fuel cells using ammonia as the fuel and the potential applications of ammonia fuel ...

  4. Ammonia as a suitable fuel for fuel cells

    Directory of Open Access Journals (Sweden)

    Rong eLan

    2014-08-01

    Full Text Available Ammonia, an important basic chemical, is produced at a scale of 150 million tons per year. Half of hydrogen produced in chemical industry is used for ammonia production. Ammonia containing 17.5wt% hydrogen is an ideal carbon-free fuel for fuel cells. Compared to hydrogen, ammonia has many advantages. In this mini-review, the suitability of ammonia as fuel for fuel cells, the development of different types of fuel cells using ammonia as the fuel and the potential applications of ammonia fuel cells are briefly reviewed.

  5. Organic fuel cells and fuel cell conducting sheets

    Science.gov (United States)

    Masel, Richard I.; Ha, Su; Adams, Brian

    2007-10-16

    A passive direct organic fuel cell includes an organic fuel solution and is operative to produce at least 15 mW/cm.sup.2 when operating at room temperature. In additional aspects of the invention, fuel cells can include a gas remover configured to promote circulation of an organic fuel solution when gas passes through the solution, a modified carbon cloth, one or more sealants, and a replaceable fuel cartridge.

  6. Fuel Cell Power Plants Renewable and Waste Fuels

    Science.gov (United States)

    2011-01-13

    Fuel Cell Power Plants Renewable and Waste Fuels DOE-DOD Workshop Washington, DC. January 13, 2011 reliable, efficient, ultra-clean Report...2011 2. REPORT TYPE 3. DATES COVERED 00-00-2011 to 00-00-2011 4. TITLE AND SUBTITLE Fuel Cell Power Plants Renewable and Waste Fuels 5a. CONTRACT...Approved for public release; distribution unlimited 13. SUPPLEMENTARY NOTES presented at the DOE-DOD Waste-to-Energy using Fuel Cells Workshop held

  7. Analysis of Fresh Fuel Critical Experiments Appropriate for Burnup Credit Validation

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.

    1995-01-01

    The ANS/ANS-8.1 standard requires that calculational methods used in determining criticality safety limits for applications outside reactors be validated by comparison with appropriate critical experiments. This report provides a detailed description of 34 fresh fuel critical experiments and their analyses using the SCALE-4.2 code system and the 27-group ENDF/B-IV cross-section library. The 34 critical experiments were selected based on geometry, material, and neutron interaction characteristics that are applicable to a transportation cask loaded with pressurized-water-reactor spent fuel. These 34 experiments are a representative subset of a much larger data base of low-enriched uranium and mixed-oxide critical experiments. A statistical approach is described and used to obtain an estimate of the bias and uncertainty in the calculational methods and to predict a confidence limit for a calculated neutron multiplication factor. The SCALE-4.2 results for a superset of approximately 100 criticals are included in uncertainty analyses, but descriptions of the individual criticals are not included.

  8. Analysis of fresh fuel critical experiments appropriate for burnup credit validation

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.; Bowman, S.M.

    1995-10-01

    The ANS/ANS-8.1 standard requires that calculational methods used in determining criticality safety limits for applications outside reactors be validated by comparison with appropriate critical experiments. This report provides a detailed description of 34 fresh fuel critical experiments and their analyses using the SCALE-4.2 code system and the 27-group ENDF/B-IV cross-section library. The 34 critical experiments were selected based on geometry, material, and neutron interaction characteristics that are applicable to a transportation cask loaded with pressurized-water-reactor spent fuel. These 34 experiments are a representative subset of a much larger data base of low-enriched uranium and mixed-oxide critical experiments. A statistical approach is described and used to obtain an estimate of the bias and uncertainty in the calculational methods and to predict a confidence limit for a calculated neutron multiplication factor. The SCALE-4.2 results for a superset of approximately 100 criticals are included in uncertainty analyses, but descriptions of the individual criticals are not included.

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

    Energy Technology Data Exchange (ETDEWEB)

    Croft, Stephen [Los Alamos National Laboratory; Favalli, Andrea [Los Alamos National Laboratory; Swinhoe, Martyn T. [Los Alamos National Laboratory

    2012-06-19

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

  10. Advanced Fuels Campaign 2012 Accomplishments

    Energy Technology Data Exchange (ETDEWEB)

    Not Listed

    2012-11-01

    The Advanced Fuels Campaign (AFC) under the Fuel Cycle Research and Development (FCRD) program is responsible for developing fuels technologies to support the various fuel cycle options defined in the DOE Nuclear Energy Research and Development Roadmap, Report to Congress, April 2010. The fiscal year 2012 (FY 2012) accomplishments are highlighted below. Kemal Pasamehmetoglu is the National Technical Director for AFC.

  11. Seventh Edition Fuel Cell Handbook

    Energy Technology Data Exchange (ETDEWEB)

    NETL

    2004-11-01

    Provides an overview of fuel cell technology and research projects. Discusses the basic workings of fuel cells and their system components, main fuel cell types, their characteristics, and their development status, as well as a discussion of potential fuel cell applications.

  12. BIOCHEMICAL FUEL CELLS.

    Science.gov (United States)

    used to evaluate kinetics of alcoholic fermentation . Evaluation of results indicated that 1% ethanol can be generated in 1 hour. One per cent ethanol is the minimum fuel concentration required for this system. (Author)

  13. Fuels and Lubricants Facility

    Data.gov (United States)

    Federal Laboratory Consortium — Modern naval aircraft and turbine-powered craft require reliable and high-quality fuels and lubricants to satisfy the demands imposed upon them for top performance...

  14. Nuclear Fuel Reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Michael F. Simpson; Jack D. Law

    2010-02-01

    This is an a submission for the Encyclopedia of Sustainable Technology on the subject of Reprocessing Spent Nuclear Fuel. No formal abstract was required for the article. The full article will be attached.

  15. Renewable jet fuel.

    Science.gov (United States)

    Kallio, Pauli; Pásztor, András; Akhtar, M Kalim; Jones, Patrik R

    2014-04-01

    Novel strategies for sustainable replacement of finite fossil fuels are intensely pursued in fundamental research, applied science and industry. In the case of jet fuels used in gas-turbine engine aircrafts, the production and use of synthetic bio-derived kerosenes are advancing rapidly. Microbial biotechnology could potentially also be used to complement the renewable production of jet fuel, as demonstrated by the production of bioethanol and biodiesel for piston engine vehicles. Engineered microbial biosynthesis of medium chain length alkanes, which constitute the major fraction of petroleum-based jet fuels, was recently demonstrated. Although efficiencies currently are far from that needed for commercial application, this discovery has spurred research towards future production platforms using both fermentative and direct photobiological routes. Copyright © 2013 Elsevier Ltd. All rights reserved.

  16. Alternative fuel information sources

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

    This short document contains a list of more than 200 US sources of information (Name, address, phone number, and sometimes contact) related to the use of alternative fuels in automobiles and trucks. Electric-powered cars are also included.

  17. Hydrogen as a fuel

    Energy Technology Data Exchange (ETDEWEB)

    1979-01-01

    A panel of the Committee on Advanced Energy Storage Systems of the Assembly of Engineering has examined the status and problems of hydrogen manufacturing methods, hydrogen transmission and distribution networks, and hydrogen storage systems. This examination, culminating at a time when rapidly changing conditions are having noticeable impact on fuel and energy availability and prices, was undertaken with a view to determining suitable criteria for establishing the pace, timing, and technical content of appropriate federally sponsored hydrogen R and D programs. The increasing urgency to develop new sources and forms of fuel and energy may well impact on the scale and timing of potential future hydrogen uses. The findings of the panel are presented. Chapters are devoted to hydrogen sources, hydrogen as a feedstock, hydrogen transport and storage, hydrogen as a heating fuel, automotive uses of hydrogen, aircraft use of hydrogen, the fuel cell in hydrogen energy systems, hydrogen research and development evaluation, and international hydrogen programs.

  18. Very High Temperature Reactor (VHTR) Deep Burn Core and Fuel Analysis -- Complete Design Selection for the Pebble Bed Reactor

    Energy Technology Data Exchange (ETDEWEB)

    B. Boer; A. M. Ougouag

    2010-09-01

    The Deep-Burn (DB) concept focuses on the destruction of transuranic nuclides from used light water reactor fuel. These transuranic nuclides are incorporated into TRISO coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400). Although it has been shown in the previous Fiscal Year (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup, while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239-Pu, 240-Pu and 241-Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a ”standard,” UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. The effort of this task in FY 2010 has focused on the optimization of the core to maximize the pebble discharge

  19. Analysis of maximum allowable fragment heights during dissolution of high flux isotope reactor fuel in an h-canyon dissolver

    Energy Technology Data Exchange (ETDEWEB)

    Daniel, G. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Rudisill, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-07-17

    As part of the Spent Nuclear Fuel (SNF) processing campaign, H-Canyon is planning to begin dissolving High Flux Isotope Reactor (HFIR) fuel in late FY17 or early FY18. Each HFIR fuel core contains inner and outer fuel elements which were fabricated from uranium oxide (U3O8) dispersed in a continuous Al phase using traditional powder metallurgy techniques. Fuels fabricated in this manner, like other SNF’s processed in H-Canyon, dissolve by the same general mechanisms with similar gas generation rates and the production of H2. The HFIR fuel cores will be dissolved using a flowsheet developed by the Savannah River National Laboratory (SRNL) in either the 6.4D or 6.1D dissolver using a unique insert. Multiple cores will be charged to the same dissolver solution maximizing the concentration of dissolved Al. The recovered U will be down-blended into low-enriched U for subsequent use as commercial reactor fuel. During the development of the HFIR fuel dissolution flowsheet, the cycle time for the initial core was estimated at 28 to 40 h. Once the cycle is complete, H-Canyon personnel will open the dissolver and probe the HFIR insert wells to determine the height of any fuel fragments which did not dissolve. Before the next core can be charged to the dissolver, an analysis of the potential for H2 gas generation must show that the combined surface area of the fuel fragments and the subsequent core will not generate H2 concentrations in the dissolver offgas which exceeds 60% of the lower flammability limit (LFL) of H2 at 200 °C. The objective of this study is to identify the maximum fuel fragment height as a function of the Al concentration in the dissolving solution which will provide criteria for charging successive HFIR cores to an H-Canyon dissolver.

  20. Fuel nitrogen conversion in solid fuel fired systems

    Energy Technology Data Exchange (ETDEWEB)

    P. Glarborg; A.D. Jensen; J.E. Johnsson [Technical University of Denmark, Lyngby (Denmark). Department of Chemical Engineering

    2003-07-01

    Understanding of the chemical and physical processes that govern formation and destruction of nitrogen oxides (NOx) in combustion of solid fuels continues to be a challenge. There are still unresolved issues that may limit the potential of primary measures for NOx control. In most solid fuel fired systems oxidation of fuel-bound nitrogen constitutes the dominating source of nitrogen oxides. The paper reviews some fundamental aspects of fuel nitrogen conversion in these systems, emphasizing combustion of coal since most previous work deal with this fuel. Results on biomass combustion are also discussed. Homogeneous and heterogeneous pathways in fuel NO formation and destruction are discussed and the effect of fuel characteristics, devolatilization conditions and combustion mode on the oxidation selectivity towards NO and N{sub 2} is evaluated. Results indicate that even under idealized conditions, such as a laminar pulverized-fuel flame, the governing mechanisms for fuel nitrogen conversion are not completely understood. Light gases, tar, char and soot may all be important vehicles for fuel-N conversion, with their relative importance depending on fuel rank and reaction conditions. Oxygen availability and fuel-nitrogen level are major parameters determining the oxidation selectivity of fuel-N towards NO and N{sub 2}, but also the ability of char and soot to reduce NO is potentially important. The impact of fuel/oxidizer mixing pattern on NO formation appears to be less important in solid-fuel flames than in homogeneous flames. 247 refs., 14 figs., 2 tabs.

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

    Energy Technology Data Exchange (ETDEWEB)

    Hyder, M L; Perkins, W C; Thompson, M C; Burney, G A; Russell, E R; Holcomb, H P; Landon, L F

    1979-04-01

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

  2. Vibrating fuel grapple. [LMFBR

    Science.gov (United States)

    Chertock, A.J.; Fox, J.N.; Weissinger, R.B.

    A reactor refueling method is described which utilizes a vibrating fuel grapple for removing spent fuel assemblies from a reactor core. It incorporates a pneumatic vibrator in the grapple head which allows additional withdrawal capability without exceeding the allowable axial force limit. The only moving part in the vibrator is a steel ball, pneumatically driven by a gas, such as argon, around a track, with centrifugal force created by the ball being transmitted through the grapple to the assembly handling socket.

  3. Solar fuels generator

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, Nathan S.; Spurgeon, Joshua M.

    2016-10-25

    The solar fuels generator includes an ionically conductive separator between a gaseous first phase and a second phase. A photoanode uses one or more components of the first phase to generate cations during operation of the solar fuels generator. A cation conduit is positioned provides a pathway along which the cations travel from the photoanode to the separator. The separator conducts the cations. A second solid cation conduit conducts the cations from the separator to a photocathode.

  4. Environmentally safe aviation fuels

    Science.gov (United States)

    Liberio, Patricia D.

    1995-01-01

    In response to the Air Force directive to remove Ozone Depleting Chemicals (ODC's) from military specifications and Defense Logistics Agency's Hazardous Waste Minimization Program, we are faced with how to ensure a quality aviation fuel without using such chemicals. Many of these chemicals are found throughout the fuel and fuel related military specifications and are part of test methods that help qualify the properties and quality of the fuels before they are procured. Many years ago there was a directive for military specifications to use commercially standard test methods in order to provide standard testing in private industry and government. As a result the test methods used in military specifications are governed by the American Society of Testing and Materials (ASTM). The Air Force has been very proactive in the removal or replacement of the ODC's and hazardous materials in these test methods. For example, ASTM D3703 (Standard Test Method for Peroxide Number of Aviation Turbine Fuels), requires the use of Freon 113, a known ODC. A new rapid, portable hydroperoxide test for jet fuels similar to ASTM D3703 that does not require the use of ODC's has been developed. This test has proved, in limited testing, to be a viable substitute method for ASTM D3703. The Air Force is currently conducting a round robin to allow the method to be accepted by ASTM and therefore replace the current method. This paper will describe the Air Force's initiatives to remove ODC's and hazardous materials from the fuel and fuel related military specifications that the Air Force Wright Laboratory.

  5. TARDEC Assured Fuels Initiative

    Science.gov (United States)

    2008-05-07

    Objectives • Develop and build two microchannel reactors − FT reactor : convert CO + H2 to long-chain hydrocarbons − Hydrocracker: upgrade FT wax to...present depending on the FT reactor conditions and catalyst used and subsequent upgrade-to-finished fuel processing • Lack of aromatics impacts fuel... Microchannel Processing Technology (MPT) The Technology • Stacks of closely spaced thin plates form microchannels • Process fluids pass through

  6. Nuclear fuel element

    Science.gov (United States)

    Meadowcroft, Ronald Ross; Bain, Alastair Stewart

    1977-01-01

    A nuclear fuel element wherein a tubular cladding of zirconium or a zirconium alloy has a fission gas plenum chamber which is held against collapse by the loops of a spacer in the form of a tube which has been deformed inwardly at three equally spaced, circumferential positions to provide three loops. A heat resistant disc of, say, graphite separates nuclear fuel pellets within the cladding from the plenum chamber. The spacer is of zirconium or a zirconium alloy.

  7. EPRI fuel cladding integrity program

    Energy Technology Data Exchange (ETDEWEB)

    Yang, R. [Electric Power Research Institute, Palo Alto, CA (United States)

    1997-01-01

    The objectives of the EPRI fuel program is to supplement the fuel vendor research to assure that utility economic and operational interests are met. To accomplish such objectives, EPRI has conducted research and development efforts to (1) reduce fuel failure rates and mitigate the impact of fuel failures on plant operation, (2) provide technology to extend burnup and reduce fuel cycle cost. The scope of R&D includes fuel and cladding. In this paper, only R&D related to cladding integrity will be covered. Specific areas aimed at improving fuel cladding integrity include: (1) Fuel Reliability Data Base; (2) Operational Guidance for Defective Fuel; (3) Impact of Water Chemistry on Cladding Integrity; (4) Cladding Corrosion Data and Model; (5) Cladding Mechanical Properties; and (6) Transient Fuel Cladding Response.

  8. Hydrogen fuel - Universal energy

    Science.gov (United States)

    Prince, A. G.; Burg, J. A.

    The technology for the production, storage, transmission, and consumption of hydrogen as a fuel is surveyed, with the physical and chemical properties of hydrogen examined as they affect its use as a fuel. Sources of hydrogen production are described including synthesis from coal or natural gas, biomass conversion, thermochemical decomposition of water, and electrolysis of water, of these only electrolysis is considered economicially and technologically feasible in the near future. Methods of production of the large quantities of electricity required for the electrolysis of sea water are explored: fossil fuels, hydroelectric plants, nuclear fission, solar energy, wind power, geothermal energy, tidal power, wave motion, electrochemical concentration cells, and finally ocean thermal energy conversion (OTEC). The wind power and OTEC are considered in detail as the most feasible approaches. Techniques for transmission (by railcar or pipeline), storage (as liquid in underwater or underground tanks, as granular metal hydride, or as cryogenic liquid), and consumption (in fuel cells in conventional power plants, for home usage, for industrial furnaces, and for cars and aircraft) are analyzed. The safety problems of hydrogen as a universal fuel are discussed, noting that they are no greater than those for conventional fuels.

  9. Alternative Fuels: Research Progress

    Directory of Open Access Journals (Sweden)

    Maher A.R. Sadiq Al-Baghdadi

    2013-01-01

    Full Text Available Chapter 1: Pollutant Emissions and Combustion Characteristics of Biofuels and Biofuel/Diesel Blends in Laminar and Turbulent Gas Jet Flames. R. N. Parthasarathy, S. R. Gollahalli Chapter 2: Sustainable Routes for The Production of Oxygenated High-Energy Density Biofuels from Lignocellulosic Biomass. Juan A. Melero, Jose Iglesias, Gabriel Morales, Marta Paniagua Chapter 3: Optical Investigations of Alternative-Fuel Combustion in an HSDI Diesel Engine. T. Huelser, M. Jakob, G. Gruenefeld, P. Adomeit, S. Pischinger Chapter 4: An Insight into Biodiesel Physico-Chemical Properties and Exhaust Emissions Based on Statistical Elaboration of Experimental Data. Evangelos G. Giakoumis Chapter 5: Biodiesel: A Promising Alternative Energy Resource. A.E. Atabani Chapter 6: Alternative Fuels for Internal Combustion Engines: An Overview of the Current Research. Ahmed A. Taha, Tarek M. Abdel-Salam, Madhu Vellakal Chapter 7: Investigating the Hydrogen-Natural Gas Blends as a Fuel in Internal Combustion Engine. ?lker YILMAZ Chapter 8: Conversion of Bus Diesel Engine into LPG Gaseous Engine; Method and Experiments Validation. M. A. Jemni , G. Kantchev , Z. Driss , R. Saaidia , M. S. Abid Chapter 9: Predicting the Combustion Performance of Different Vegetable Oils-Derived Biodiesel Fuels. Qing Shu, ChangLin Yu Chapter 10: Production of Gasoline, Naphtha, Kerosene, Diesel, and Fuel Oil Range Fuels from Polypropylene and Polystyrene Waste Plastics Mixture by Two-Stage Catalytic Degradation using ZnO. Moinuddin Sarker, Mohammad Mamunor Rashid

  10. Nalco Fuel Tech

    Energy Technology Data Exchange (ETDEWEB)

    Michalak, S.

    1995-12-31

    The Nalco Fuel Tech with its seat at Naperville (near Chicago), Illinois, is an engineering company working in the field of technology and equipment for environmental protection. A major portion of NALCO products constitute chemical materials and additives used in environmental protection technologies (waste-water treatment plants, water treatment, fuel modifiers, etc.). Basing in part on the experience, laboratories and RD potential of the mother company, the Nalco Fuel Tech Company developed and implemented in the power industry a series of technologies aimed at the reduction of environment-polluting products of fuel combustion. The engineering solution of Nalco Fuel Tech belong to a new generation of environmental protection techniques developed in the USA. They consist in actions focused on the sources of pollutants, i.e., in upgrading the combustion chambers of power engineering plants, e.g., boilers or communal and/or industrial waste combustion units. The Nalco Fuel Tech development and research group cooperates with leading US investigation and research institutes.

  11. Fuel Class Higher Alcohols

    KAUST Repository

    Sarathy, Mani

    2016-08-17

    This chapter focuses on the production and combustion of alcohol fuels with four or more carbon atoms, which we classify as higher alcohols. It assesses the feasibility of utilizing various C4-C8 alcohols as fuels for internal combustion engines. Utilizing higher-molecular-weight alcohols as fuels requires careful analysis of their fuel properties. ASTM standards provide fuel property requirements for spark-ignition (SI) and compression-ignition (CI) engines such as the stability, lubricity, viscosity, and cold filter plugging point (CFPP) properties of blends of higher alcohols. Important combustion properties that are studied include laminar and turbulent flame speeds, flame blowout/extinction limits, ignition delay under various mixing conditions, and gas-phase and particulate emissions. The chapter focuses on the combustion of higher alcohols in reciprocating SI and CI engines and discusses higher alcohol performance in SI and CI engines. Finally, the chapter identifies the sources, production pathways, and technologies currently being pursued for production of some fuels, including n-butanol, iso-butanol, and n-octanol.

  12. Motor Fuel Excise Taxes

    Energy Technology Data Exchange (ETDEWEB)

    2015-09-01

    A new report from the National Renewable Energy Laboratory (NREL) explores the role of alternative fuels and energy efficient vehicles in motor fuel taxes. Throughout the United States, it is common practice for federal, state, and local governments to tax motor fuels on a per gallon basis to fund construction and maintenance of our transportation infrastructure. In recent years, however, expenses have outpaced revenues creating substantial funding shortfalls that have required supplemental funding sources. While rising infrastructure costs and the decreasing purchasing power of the gas tax are significant factors contributing to the shortfall, the increased use of alternative fuels and more stringent fuel economy standards are also exacerbating revenue shortfalls. The current dynamic places vehicle efficiency and petroleum use reduction polices at direct odds with policies promoting robust transportation infrastructure. Understanding the energy, transportation, and environmental tradeoffs of motor fuel tax policies can be complicated, but recent experiences at the state level are helping policymakers align their energy and environmental priorities with highway funding requirements.

  13. Hot Isostatic Press Can Optimization for Aluminum Cladding of U-10Mo Reactor Fuel Plates: FY12 Final Report and FY13 Update

    Energy Technology Data Exchange (ETDEWEB)

    Clarke, Kester D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Crapps, Justin M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Scott, Jeffrey E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Aikin, Beverly [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Vargas, Victor D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dvornak, Matthew J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Duffield, Andrew N. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Weinberg, Richard Y. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Alexander, David J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Montalvo, Joel D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hudson, Richard W. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mihaila, Bogdan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Liu, Cheng [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Lovato, Manuel L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dombrowski, David E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2013-08-26

    Currently, the proposed processing path for low enriched uranium – 10 wt. pct. molybdenum alloy (LEU-10Mo) monolithic fuel plates for high power research and test reactors includes hot isostatic pressing (HIP) to bond the aluminum cladding that encapsulates the fuel foil. Initial HIP experiments were performed at Idaho National Laboratory (INL) on approximately ¼ scale “mini” fuel plate samples using a HIP can design intended for these smaller experimental trials. These experiments showed that, with the addition of a co-rolled zirconium diffusion barrier on the LEU-10Mo alloy fuel foil, the HIP bonding process is a viable method for producing monolithic fuel plates. Further experimental trials at Los Alamos National Laboratory (LANL) effectively scaled-up the “mini” can design to produce full-size fuel prototypic plates. This report summarizes current efforts at LANL to produce a HIP can design that is further optimized for higher volume production runs. The production-optimized HIP can design goals were determined by LANL and Babcock & Wilcox (B&W) to include maintaining or improving the quality of the fuel plates produced with the baseline scaled-up mini can design, while minimizing material usage, improving dimensional stability, easing assembly and disassembly, eliminating machining, and significantly reducing welding. The initial small-scale experiments described in this report show that a formed-can approach can achieve the goals described above. Future work includes scaling the formed-can approach to full-size fuel plates, and current progress toward this goal is also summarized here.

  14. Automotive Fuel Processor Development and Demonstration with Fuel Cell Systems

    Energy Technology Data Exchange (ETDEWEB)

    Nuvera Fuel Cells

    2005-04-15

    The potential for fuel cell systems to improve energy efficiency and reduce emissions over conventional power systems has generated significant interest in fuel cell technologies. While fuel cells are being investigated for use in many applications such as stationary power generation and small portable devices, transportation applications present some unique challenges for fuel cell technology. Due to their lower operating temperature and non-brittle materials, most transportation work is focusing on fuel cells using proton exchange membrane (PEM) technology. Since PEM fuel cells are fueled by hydrogen, major obstacles to their widespread use are the lack of an available hydrogen fueling infrastructure and hydrogen's relatively low energy storage density, which leads to a much lower driving range than conventional vehicles. One potential solution to the hydrogen infrastructure and storage density issues is to convert a conventional fuel such as gasoline into hydrogen onboard the vehicle using a fuel processor. Figure 2 shows that gasoline stores roughly 7 times more energy per volume than pressurized hydrogen gas at 700 bar and 4 times more than liquid hydrogen. If integrated properly, the fuel processor/fuel cell system would also be more efficient than traditional engines and would give a fuel economy benefit while hydrogen storage and distribution issues are being investigated. Widespread implementation of fuel processor/fuel cell systems requires improvements in several aspects of the technology, including size, startup time, transient response time, and cost. In addition, the ability to operate on a number of hydrocarbon fuels that are available through the existing infrastructure is a key enabler for commercializing these systems. In this program, Nuvera Fuel Cells collaborated with the Department of Energy (DOE) to develop efficient, low-emission, multi-fuel processors for transportation applications. Nuvera's focus was on (1) developing fuel

  15. Outlook for alternative transportation fuels

    Energy Technology Data Exchange (ETDEWEB)

    Gushee, D.E. [Univ. of Illinois, Chicago, IL (United States)

    1996-12-31

    This presentation provides a brief review of regulatory issues and Federal programs regarding alternative fuel use in automobiles. A number of U.S. DOE initiatives and studies aimed at increasing alternative fuels are outlined, and tax incentives in effect at the state and Federal levels are discussed. Data on alternative fuel consumption and alternative fuel vehicle use are also presented. Despite mandates, tax incentives, and programs, it is concluded alternative fuels will have minimal market penetration. 7 refs., 5 tabs.

  16. Deep desulfurization of hydrocarbon fuels

    Science.gov (United States)

    Song, Chunshan [State College, PA; Ma, Xiaoliang [State College, PA; Sprague, Michael J [Calgary, CA; Subramani, Velu [State College, PA

    2012-04-17

    The invention relates to processes for reducing the sulfur content in hydrocarbon fuels such as gasoline, diesel fuel and jet fuel. The invention provides a method and materials for producing ultra low sulfur content transportation fuels for motor vehicles as well as for applications such as fuel cells. The materials and method of the invention may be used at ambient or elevated temperatures and at ambient or elevated pressures without the need for hydrogen.

  17. Analysis of partial and total flow blockage of a single fuel assembly of an MTR research reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Adorni, Martina [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Facolta di Ingegneria, Universita di Pisa, Via Diotisalvi, 2-56100 Pisa (Italy)]. E-mail: martina_adorni@tin.it; Bousbia-Salah, Anis [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Facolta di Ingegneria, Universita di Pisa, Via Diotisalvi, 2-56100 Pisa (Italy); Hamidouche, Tewfik [Commissariat a l' Energie Atomique, Centre de Recherche Nucleaire d' Alger-Algeria, 02 Boulevard Frantz fanon, BP 399 Alger-gare (Algeria); Maro, Beniamino Di [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Facolta di Ingegneria, Universita di Pisa, Via Diotisalvi, 2-56100 Pisa (Italy); Pierro, Franco [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Facolta di Ingegneria, Universita di Pisa, Via Diotisalvi, 2-56100 Pisa (Italy); D' Auria, Francesco [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Facolta di Ingegneria, Universita di Pisa, Via Diotisalvi, 2-56100 Pisa (Italy)

    2005-10-15

    The main aim of the following study is to perform a safety analysis of the IAEA 10 MW MTR Pool type Research Reactor [IAEA-TECDOC-233, 1980. IAEA Research Reactor Core Conversion from the use of high-enriched uranium to the use of low enriched uranium fuels Guidebook] under flow blockage of a single Fuel Assembly (FA) conditions. Such event was rarely investigated in the open literature notwithstanding the fat that it constitutes a severe accident that may lead to local dryout and eventually to loss of the FA integrity. The transients herein considered are related to partial and total obstruction of the cooling channel of a single Fuel Assembly of the reactor core. This study constitutes the first step of a larger work, which consists in performing a 3D simulation using the Best Estimate coupled code technique. However, as a first approach the instantaneous reactor power is derived through the point kinetic approach of the used thermal-hydraulic system code.

  18. 78 FR 14190 - Regulation of Fuels and Fuel Additives: Identification of Additional Qualifying Renewable Fuel...

    Science.gov (United States)

    2013-03-05

    ... AGENCY 40 CFR Part 80 RIN 2060-AR07 Regulation of Fuels and Fuel Additives: Identification of Additional Qualifying Renewable Fuel Pathways Under the Renewable Fuel Standard Program AGENCY: Environmental Protection Agency (EPA). ACTION: Final rule. SUMMARY: EPA is issuing a final rule identifying additional...

  19. Apparatus and method for grounding compressed fuel fueling operator

    Science.gov (United States)

    Cohen, Joseph Perry; Farese, David John; Xu, Jianguo

    2002-06-11

    A safety system for grounding an operator at a fueling station prior to removing a fuel fill nozzle from a fuel tank upon completion of a fuel filling operation is provided which includes a fuel tank port in communication with the fuel tank for receiving and retaining the nozzle during the fuel filling operation and a grounding device adjacent to the fuel tank port which includes a grounding switch having a contact member that receives physical contact by the operator and where physical contact of the contact member activates the grounding switch. A releasable interlock is included that provides a lock position wherein the nozzle is locked into the port upon insertion of the nozzle into the port and a release position wherein the nozzle is releasable from the port upon completion of the fuel filling operation and after physical contact of the contact member is accomplished.

  20. Progress of the DUPIC Fuel Compatibility Analysis (IV) - Fuel Performance

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok; Ryu, Ho Jin; Roh, Gyu Hong; Jeong, Chang Joon; Park, Chang Je; Song, Kee Chan; Lee, Jung Won

    2005-10-15

    This study describes the mechanical compatibility of the direct use of spent pressurized water reactor (PWR) fuel in Canada deuterium uranium (CANDU) reactors (DUPIC) fuel, when it is loaded into a CANDU reactor. The mechanical compatibility can be assessed for the fuel management, primary heat transport system, fuel channel, and the fuel handling system in the reactor core by both the experimental and analytic methods. Because the physical dimensions of the DUPIC fuel bundle adopt the CANDU flexible (CANFLEX) fuel bundle design which has already been demonstrated for a commercial use in CANDU reactors, the experimental compatibility analyses focused on the generation of material property data and the irradiation tests of the DUPIC fuel, which are used for the computational analysis. The intermediate results of the mechanical compatibility analysis have shown that the integrity of the DUPIC fuel is mostly maintained under the high power and high burnup conditions even though some material properties like the thermal conductivity is a little lower compared to the uranium fuel. However it is required to slightly change the current DUPIC fuel design to accommodate the high internal pressure of the fuel element. It is also strongly recommended to perform more irradiation tests of the DUPIC fuel to accumulate a database for the demonstration of the DUPIC fuel performance in the CANDU reactor.

  1. NASA Alternative Aviation Fuel Research

    Science.gov (United States)

    Anderson, B. E.; Beyersdorf, A. J.; Thornhill, K. L., II; Moore, R.; Shook, M.; Winstead, E.; Ziemba, L. D.; Crumeyrolle, S.

    2015-12-01

    We present an overview of research conducted by NASA Aeronautics Research Mission Directorate to evaluate the performance and emissions of "drop-in" alternative jet fuels, highlighting experiment design and results from the Alternative Aviation Fuel Experiments (AAFEX-I & -II) and Alternative Fuel-Effects on Contrails and Cruise Emissions flight series (ACCESS-I & II). These projects included almost 100 hours of sampling exhaust emissions from the NASA DC-8 aircraft in both ground and airborne operation and at idle to takeoff thrust settings. Tested fuels included Fischer-Tropsch (FT) synthetic kerosenes manufactured from coal and natural-gas feedstocks; Hydro-treated Esters and Fatty-Acids (HEFA) fuels made from beef-tallow and camelina-plant oil; and 50:50 blends of these alternative fuels with Jet A. Experiments were also conducted with FT and Jet A fuels doped with tetrahydrothiophene to examine the effects of fuel sulfur on volatile aerosol and contrail formation and microphysical properties. Results indicate that although the absence of aromatic compounds in the alternative fuels caused DC-8 fuel-system leaks, the fuels did not compromise engine performance or combustion efficiency. And whereas the alternative fuels produced only slightly different gas-phase emissions, dramatic reductions in non-volatile particulate matter (nvPM) emissions were observed when burning the pure alternative fuels, particularly at low thrust settings where particle number and mass emissions were an order of magnitude lower than measured from standard jet fuel combustion; 50:50 blends of Jet A and alternative fuels typically reduced nvPM emissions by ~50% across all thrust settings. Alternative fuels with the highest hydrogen content produced the greatest nvPM reductions. For Jet A and fuel blends, nvPM emissions were positively correlated with fuel aromatic and naphthalene content. Fuel sulfur content regulated nucleation mode aerosol number and mass concentrations within aging

  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. Development of PEM fuel cell technology at international fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Wheeler, D.J.

    1996-04-01

    The PEM technology has not developed to the level of phosphoric acid fuel cells. Several factors have held the technology development back such as high membrane cost, sensitivity of PEM fuel cells to low level of carbon monoxide impurities, the requirement to maintain full humidification of the cell, and the need to pressurize the fuel cell in order to achieve the performance targets. International Fuel Cells has identified a hydrogen fueled PEM fuel cell concept that leverages recent research advances to overcome major economic and technical obstacles.

  4. Final Report on Utilization of TRU TRISO Fuel as Applied to HTR Systems Part I: Pebble Bed Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Brian Boer; Abderrafi M. Ougouag

    2011-03-01

    The Deep-Burn (DB) concept [ ] focuses on the destruction of transuranic nuclides from used light water reactor (LWR) fuel. These transuranic nuclides are incorporated into tri-isotopic (TRISO) coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400) [ ]. Although it has been shown in the previous Fiscal Year (FY) (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking, and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239Pu, 240Pu, and 241Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a standard, UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. Regarding the coated particle performance, the FY 2009 investigations showed that no

  5. Dry Process Fuel Performance Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Myung Seung; Song, K. C.; Moon, J. S. and others

    2005-04-15

    The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase II R and D. In order to fulfil this objectives, irradiation test of DUPIC fuel was carried out in HANARO using the non-instrumented and SPND-instrumented rig. Also, the analysis on the in-reactor behavior analysis of DUPIC fuel, out-pile test using simulated DUPIC fuel as well as performance and integrity assessment in a commercial reactor were performed during this Phase. The R and D results of the Phase II are summarized as follows : - Performance evaluation of DUPIC fuel via irradiation test in HANARO - Post irradiation examination of irradiated fuel and performance analysis - Development of DUPIC fuel performance code (modified ELESTRES) considering material properties of DUPIC fuel - Irradiation behavior and integrity assessment under the design power envelope of DUPIC fuel - Foundamental technology development of thermal/mechanical performance evaluation using ANSYS (FEM package)

  6. Pulverized fuel-oxygen burner

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, Curtis; Patterson, Brad; Perdue, Jayson

    2017-09-05

    A burner assembly combines oxygen and fuel to produce a flame. The burner assembly includes an oxygen supply tube adapted to receive a stream of oxygen and a solid fuel conduit arranged to extend through the oxygen tube to convey a stream of fluidized, pulverized, solid fuel into a flame chamber. Oxygen flowing through the oxygen supply tube passes generally tangentially through a first set of oxygen-injection holes formed in the solid fuel conduit and off-tangentially from a second set of oxygen-injection holes formed in the solid fuel conduit and then mixes with fluidized, pulverized, solid fuel passing through the solid fuel conduit to create an oxygen-fuel mixture in a downstream portion of the solid fuel conduit. This mixture is discharged into a flame chamber and ignited in the flame chamber to produce a flame.

  7. 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

  8. Comparison of fresh fuel experimental measurements to MCNPX calculations using self-interrogation neutron resonance densitometry

    Energy Technology Data Exchange (ETDEWEB)

    LaFleur, Adrienne M., E-mail: alafleur@lanl.gov [Nuclear Nonproliferation Division, Los Alamos National Laboratory, P.O. Box 1663 MS E540, Los Alamos, NM 87545 (United States); Charlton, William S., E-mail: wcharlton@tamu.edu [Nuclear Security Science and Policy Institute, Texas A and M University, 3473 TAMU, College Station, TX 77843 (United States); Menlove, Howard O., E-mail: hmenlove@lanl.gov [Nuclear Nonproliferation Division, Los Alamos National Laboratory, P.O. Box 1663 MS E540, Los Alamos, NM 87545 (United States); Swinhoe, Martyn T., E-mail: swinhoe@lanl.gov [Nuclear Nonproliferation Division, Los Alamos National Laboratory, P.O. Box 1663 MS E540, Los Alamos, NM 87545 (United States)

    2012-07-11

    A new non-destructive assay technique called Self-Interrogation Neutron Resonance Densitometry (SINRD) is currently being developed at Los Alamos National Laboratory (LANL) to improve existing nuclear safeguards measurements for Light Water Reactor (LWR) fuel assemblies. SINRD consists of four {sup 235}U fission chambers (FCs): bare FC, boron carbide shielded FC, Gd covered FC, and Cd covered FC. Ratios of different FCs are used to determine the amount of resonance absorption from {sup 235}U in the fuel assembly. The sensitivity of this technique is based on using the same fissile materials in the FCs as are present in the fuel because the effect of resonance absorption lines in the transmitted flux is amplified by the corresponding (n,f) reaction peaks in the fission chamber. In this work, experimental measurements were performed in air with SINRD using a reference Pressurized Water Reactor (PWR) 15 Multiplication-Sign 15 low enriched uranium (LEU) fresh fuel assembly at LANL. The purpose of this experiment was to assess the following capabilities of SINRD: (1) ability to measure the effective {sup 235}U enrichment of the PWR fresh LEU fuel assembly and (2) sensitivity and penetrability to the removal of fuel pins from an assembly. These measurements were compared to Monte Carlo N-Particle eXtended transport code (MCNPX) simulations to verify the accuracy of the MCNPX model of SINRD. The reproducibility of experimental measurements via MCNPX simulations is essential to validating the results and conclusions obtained from the simulations of SINRD for LWR spent fuel assemblies. - Highlights: Black-Right-Pointing-Pointer Development of new measurement technique called SINRD to improve LWR safeguards. Black-Right-Pointing-Pointer Performed SINRD experiment to measure {sup 235}U and pin diversions in PWR fresh assembly. Black-Right-Pointing-Pointer Excellent agreement of MCNPX and measured results confirmed accuracy of SINRD model. Black-Right-Pointing-Pointer SINRD

  9. HTGR spent fuel composition and fuel element block flow

    Energy Technology Data Exchange (ETDEWEB)

    Hamilton, C.J.; Holder, N.D.; Pierce, V.H.; Robertson, M.W.

    1976-07-01

    The High-Temperature Gas-Cooled Reactor (HTGR) utilizes the thorium-uranium fuel cycle. Fully enriched uranium fissile material and thorium fertile material are used in the initial reactor core and for makeup fuel in the recycle core loadings. Bred /sup 233/U and unburned /sup 235/U fissile materials are recovered from spent fuel elements, refabricated into recycle fuel elements, and used as part of the recycle core loading along with the makeup fuel elements. A typical HTGR employs a 4-yr fuel cycle with approximately one-fourth of the core discharged and reloaded annually. The fuel element composition, including heavy metals, impurity nuclides, fission products, and activation products, has been calculated for discharged spent fuel elements and for reload fresh fuel and recycle fuel elements for each cycle over the life of a typical HTGR. Fuel element compositions are presented for the conditions of equilibrium recycle. Data describing compositions for individual reloads throughout the reactor life are available in a detailed volume upon request. Fuel element block flow data have been compiled based on a forecast HTGR market. Annual block flows are presented for each type of fuel element discharged from the reactors for reprocessing and for refabrication.

  10. Loss-of-Flow and Loss-of-Pressure Simulations of the BR2 Research Reactor with HEU and LEU Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Licht, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Sikik, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Van den Branden, G. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Koonen, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium)

    2016-01-01

    Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The reactor core of BR2 is located inside a pressure vessel that contains 79 channels in a hyperboloid configuration. The core configuration is highly variable as each channel can contain a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Because of this variability, a representative core configuration, based on current reactor use, has been defined for the fuel conversion analyses. The code RELAP5/Mod 3.3 was used to perform the transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. The input model has been modernized relative to that historically used at BR2 taking into account the best modeling practices developed by Argonne National Laboratory (ANL) and BR2 engineers.

  11. Loss-of-Flow and Loss-of-Pressure Simulations of the BR2 Research Reactor with HEU and LEU Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Licht, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Sikik, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Van den Branden, G. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Koonen, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium)

    2016-01-01

    Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The reactor core of BR2 is located inside a pressure vessel that contains 79 channels in a hyperboloid configuration (Figure 1). The core configuration is highly variable as each channel can contain a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Because of this variability, a representative core configuration (Figure 2), based on current reactor use, has been defined for the fuel conversion analyses [1]. The code RELAP5/Mod 3.3 [2] was used to perform the transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. The input model has been modernized relative to that historically used at BR2 taking into account the best modeling practices developed by Argonne National Laboratory (ANL) and BR2 engineers.

  12. Selection of Nuclear Fuel for TREAT: UO2 vs U3O8

    Energy Technology Data Exchange (ETDEWEB)

    Glazoff, Michael Vasily [Idaho National Lab. (INL), Idaho Falls, ID (United States); Van Rooyen, Isabella Johanna [Idaho National Lab. (INL), Idaho Falls, ID (United States); Coryell, Benjamin David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Parga, Clemente Jose [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-04-01

    The Transient Reactor Test (TREAT) that resides at the Materials and Fuels Complex (MFC) at Idaho National Laboratory (INL), first achieved criticality in 1959, and successfully performed many transient tests on nuclear fuel until 1994 when its operations were suspended. Resumption of operations at TREAT was approved in February 2014 to meet the U.S. Department of Energy (DOE) Office of Nuclear Energy’s objectives in transient testing of nuclear fuels. The National Nuclear Security Administration’s is converting TREAT from its existing highly enriched uranium (HEU) core to a new core containing low enriched uranium (LEU) (i.e., U-235< 20% by weight). The TREAT Conversion project is currently progressing with conceptual design phase activities. Dimensional stability of the fuel element assemblies, predictable fuel can oxidation and sufficient heat conductivity by the fuel blocks are some of the critical performance requirements of the new LEU fuel. Furthermore, to enable the design team to design fuel block and can specifications, it is amongst the objectives to evaluate TREAT LEU fuel and cladding material’s chemical interaction. This information is important to understand the viability of Zr-based alloys and fuel characteristics for the fabrication of the TREAT LEU fuel and cladding. Also, it is very important to make the right decision on what type of nuclear fuel will be used at TREAT. In particular, one has to consider different oxides of uranium, and most importantly, UO2 vs U3O8. In this report, the results are documented pertaining to the choice mentioned above (UO2 vs U3O8). The conclusion in favor of using UO2 was made based on the analysis of historical data, up-to-date literature, and self-consistent calculations of phase equilibria and thermodynamic properties in the U-O and U-O-C systems. The report is organized as follows. First, the criteria that were used to make the

  13. Proton exchange membrane fuel cells

    CERN Document Server

    Qi, Zhigang

    2013-01-01

    Preface Proton Exchange Membrane Fuel CellsFuel CellsTypes of Fuel CellsAdvantages of Fuel CellsProton Exchange Membrane Fuel CellsMembraneCatalystCatalyst LayerGas Diffusion MediumMicroporous LayerMembrane Electrode AssemblyPlateSingle CellStackSystemCell Voltage Monitoring Module (CVM)Fuel Supply Module (FSM)Air Supply Module (ASM)Exhaust Management Module (EMM)Heat Management Module (HMM)Water Management Module (WMM)Internal Power Supply Module (IPM)Power Conditioning Module (PCM)Communications Module (COM)Controls Module (CM)SummaryThermodynamics and KineticsTheoretical EfficiencyVoltagePo

  14. Fuel Element Technical Manual

    Energy Technology Data Exchange (ETDEWEB)

    Burley, H.H. [ed.

    1956-08-01

    It is the purpose of the Fuel Element Technical Manual to Provide a single document describing the fabrication processes used in the manufacture of the fuel element as well as the technical bases for these processes. The manual will be instrumental in the indoctrination of personnel new to the field and will provide a single data reference for all personnel involved in the design or manufacture of the fuel element. The material contained in this manual was assembled by members of the Engineering Department and the Manufacturing Department at the Hanford Atomic Products Operation between the dates October, 1955 and June, 1956. Arrangement of the manual. The manual is divided into six parts: Part I--introduction; Part II--technical bases; Part III--process; Part IV--plant and equipment; Part V--process control and improvement; and VI--safety.

  15. Household fuel demand analysis

    Energy Technology Data Exchange (ETDEWEB)

    Cohn, S.; Hirst, E.; Jackson, J.

    1976-01-01

    This study develops econometric models of residential demands for electricity, natural gas, and petroleum products. Fuel demands per household are estimated as functions of fuel prices, per capita income, heating degree days, and mean July temperature. Cross-sectional models are developed using a large data base containing observations for each state and year from 1951 through 1974. Long-run own-price elasticities for all three fuels are greater than unity with natural gas showing the greatest sensitivity to own-price changes. Cross-price elasticities are all less than unity except for the elasticity of demand for oil with respect to the price of gas (which is even larger than the own-price elasticity of demand for oil). The models show considerable stabiity with respect to own-price elasticities but much instability with respect to the cross-price and income elasticities.

  16. Fuel cells in transportation

    Energy Technology Data Exchange (ETDEWEB)

    Erdmann, G. [Technische Univ., Berlin (Germany); Hoehlein, B. [Research Center Juelich (Germany)

    1996-12-01

    A promising new power source for electric drive systems is the fuel cell technology with hydrogen as energy input. The worldwide fuel cell development concentrates on basic research efforts aiming at improving this new technology and at developing applications that might reach market maturity in the very near future. Due to the progress achieved, the interest is now steadily turning to the development of overall systems such as demonstration plants for different purposes: electricity generation, drive systems for road vehicles, ships and railroads. This paper does not present results concerning the market potential of fuel cells in transportation but rather addresses some questions and reflections that are subject to further research of both engineers and economists. Some joint effort of this research will be conducted under the umbrella of the IEA Implementing Agreement 026 - Annex X, but there is a lot more to be done in this challenging but also promising fields. (EG) 18 refs.

  17. Emergency fuels utilization guidebook. Alternative Fuels Utilization Program

    Energy Technology Data Exchange (ETDEWEB)

    1980-08-01

    The basic concept of an emergency fuel is to safely and effectively use blends of specification fuels and hydrocarbon liquids which are free in the sense that they have been commandeered or volunteered from lower priority uses to provide critical transportation services for short-duration emergencies on the order of weeks, or perhaps months. A wide variety of liquid hydrocarbons not normally used as fuels for internal combustion engines have been categorized generically, including limited information on physical characteristics and chemical composition which might prove useful and instructive to fleet operators. Fuels covered are: gasoline and diesel fuel; alcohols; solvents; jet fuels; kerosene; heating oils; residual fuels; crude oils; vegetable oils; gaseous fuels.

  18. Fuel cell engineering

    CERN Document Server

    Sundmacher

    2012-01-01

    Fuel cells are attractive electrochemical energy converters featuring potentially very high thermodynamic efficiency factors. The focus of this volume of Advances in Chemical Engineering is on quantitative approaches, particularly based on chemical engineering principles, to analyze, control and optimize the steady state and dynamic behavior of low and high temperature fuel cells (PEMFC, DMFC, SOFC) to be applied in mobile and stationary systems. * Updates and informs the reader on the latest research findings using original reviews * Written by leading industry experts and scholars * Review

  19. FUELS IN TOBACCO PRODUCTION

    Directory of Open Access Journals (Sweden)

    M. Čavlek

    2008-09-01

    Full Text Available Energy production from biomass can reduce „greenhouse effect” and contribute to solving energy security especially in the agricultural households which rely on energy from fossil fuels. In Croatia fuel-cured tobacco is produced on about 5000 ha. Gross income for the whole production is about 180 000 000 kn/year. Flue-cured tobacco is a high energy consuming crop. There are two parts of energy consumption, for mechanization used for the field production (11% and, energy for bulk-curing (89%. In each case, presently used fuels of fossil origin need to be substituted by an alternative energy source of organic origin. Hereafter attention is paid to finding a more economic and ecologically acceptable fuel for curing tobacco. Curing flue-cured tobacco is done by heated air in curing burns. Various sources of heat have been used; wood, coal, oil and gas. In each case different burning facilities of different efficiency have been used. This has had an impact on curing costs and ecology. Recently, mostly used fuel has been natural gas. However, gas is getting expensive. Consequently, an alternative fuel for curing tobacco is sought for. According to literature, agricultural crops suitable for the latter purpose could be wheat, barley, maize, sorghum, sugar beet and some other annual and perennial plant species. Wooden pellets (by-products are suitable for combustion too. Ligno-cellulose fuels have been used for heating since long time. However, not sufficient research has been done from an applied point of view (Venturi and Venturi, 2003. Fuel combustion is getting more efficient with developing technological innovations. The curing barn manufacturers are offering technology for combusting wooden pellets (by-products for curing tobacco. The pellets are available on domestic market. The same technology can be used for combustion of maize grain. Within “Hrvatski duhani” research on suitability of using wooden pellets and maize grain and whole

  20. FUELS IN TOBACCO PRODUCTION

    OpenAIRE

    Čavlek, M.; Boić, M.; Kristina Gršić; V. Kozumplik

    2008-01-01

    Energy production from biomass can reduce „greenhouse effect” and contribute to solving energy security especially in the agricultural households which rely on energy from fossil fuels. In Croatia fuel-cured tobacco is produced on about 5000 ha. Gross income for the whole production is about 180 000 000 kn/year. Flue-cured tobacco is a high energy consuming crop. There are two parts of energy consumption, for mechanization used for the field production (11%) and, energy for bulk-curing (89%)....

  1. 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

  2. Fuel and fuel blending components from biomass derived pyrolysis oil

    Science.gov (United States)

    McCall, Michael J.; Brandvold, Timothy A.; Elliott, Douglas C.

    2012-12-11

    A process for the conversion of biomass derived pyrolysis oil to liquid fuel components is presented. The process includes the production of diesel, aviation, and naphtha boiling point range fuels or fuel blending components by two-stage deoxygenation of the pyrolysis oil and separation of the products.

  3. Fuel Cell Electrodes for Hydrogen-Air Fuel Cell Assemblies.

    Science.gov (United States)

    The report describes the design and evaluation of a hydrogen-air fuel cell module for use in a portable hydrid fuel cell -battery system. The fuel ... cell module consists of a stack of 20 single assemblies. Each assembly contains 2 electrically independent cells with a common electrolyte compartment

  4. NEUTRONICS ANALYSIS ON MINI TEST FUEL IN THE RSG-GAS CORE

    Directory of Open Access Journals (Sweden)

    Tukiran Surbakti

    2016-03-01

    Full Text Available Abstract NEUTRONICS ANALYSIS ON MINI TEST FUEL IN THE RSG-GAS CORE. Research of UMo fuel for research reactor has been developing  right now. The fuel of  research reactor used is uranium low enrichment with high density. For supporting the development of fuel, an assessment of mini fuel in the RSG-GAS core was performed. The mini fuel are U7Mo-Al and U6Zr-Al with densitis of 7.0gU/cc and 5.2 gU/cc, respectively. The size of both fuel are the same namely 630x70.75x1.30 mm were inserted to the 3 plates of dummy fuel. Before being irradiated in the core, a calculation for safety analysis  from neutronics and thermohydrolics aspects were required. However, in this paper will discuss safety analysis of the U7Mo-Al and U6Zr-Al mini fuels from neutronic point of view.  The calculation was done using WIMSD-5B and Batan-3DIFF code. The result showed that both of the mini fuels could be irradiated in the RSG-GAS core with burn up less than 70 % within 12 cycles of operation without over limiting the safety margin. Power density of U7Mo-Al mini fuel bigger than U6Zr-Al fuel.   Key words: mini fuel, neutronics analysis, reactor core, safety analysis   Abstrak ANALISIS NEUTRONIK ELEMEN BAKAR UJI MINI DI TERAS RSG-GAS. Penelitian tentang bahan bakar UMo untuk reaktor riset terus berkembang saat ini. Bahan bakar reaktor riset yang digunakan adalah uranium pengkayaan rendah namun densitas tinggi.  Untuk mendukung pengembangan bahan bakar dilakukan uji elemen bakar mini di teras reakror RSG-GAS dengan tujuan menentukan jumlah siklus di dalam teras sehingga tercapai fraksi bakar maksimum. Bahan bakar yang diuji adalah U7Mo-Al dengan densitas 7,0 gU/cc dan U6Zr-Al densitas 5,2 gU/cc. Ukuran kedua bahan bakar uji tersebut adalah sama 630x70,75x1,30 mm dimasukkan masing masing kedalam 3 pelat dummy bahan bakar. Sebelum diiradiasi ke dalam teras reaktor maka perlu dilakukan perhitungan keselamatan baik secara neutronik maupun termohidrolik. Dalam makalah ini

  5. Mass transfer in fuel cells

    Science.gov (United States)

    Walker, R. D., Jr.

    1973-01-01

    Developments in the following areas are reported: surface area and pore size distribution in electrolyte matrices, electron microscopy of electrolyte matrices, surface tension of KOH solutions, water transport in fuel cells, and effectiveness factors for fuel cell components.

  6. Fuel cell system with interconnect

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Zhien; Goettler, Richard; Delaforce, Philip Mark

    2016-03-08

    The present invention includes a fuel cell system having an interconnect that reduces or eliminates diffusion (leakage) of fuel and oxidant by providing an increased densification, by forming the interconnect as a ceramic/metal composite.

  7. Fuel Cell Technical Team Roadmap

    Energy Technology Data Exchange (ETDEWEB)

    None

    2013-06-01

    The Fuel Cell Technical Team promotes the development of a fuel cell power system for an automotive powertrain that meets the U.S. DRIVE Partnership (United States Driving Research and Innovation for Vehicle efficiency and Energy sustainability) goals.

  8. Design and Status of the NGNP Fuel Experiment AGR-3/4 Irradiated in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Blaine Grover

    2012-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The third and fourth experiments have been combined into a single experiment designated AGR-3/4, which started its irradiation in December 2011 and is currently scheduled to be completed in November 2013. Since the purpose of this experiment is to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment is

  9. Supplemental Thermal-Hydraulic Transient Analyses of BR2 in Support of Conversion to LEU Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Licht, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Sikik, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Van den Branden, G. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Koonen, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium)

    2016-01-01

    Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The RELAP5/Mod 3.3 code has been used to perform transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. A RELAP5 model of BR2 has been validated against select transient BR2 reactor experiments performed in 1963 by showing agreement with measured cladding temperatures. Following the validation, the RELAP5 model was then updated to represent the current use of the reactor; taking into account core configuration, neutronic parameters, trip settings, component changes, etc. Simulations of the 1963 experiments were repeated with this updated model to re-evaluate the boiling risks associated with the currently allowed maximum heat flux limit of 470 W/cm2 and temporary heat flux limit of 600 W/cm2. This document provides analysis of additional transient simulations that are required as part of a modern BR2 safety analysis report (SAR). The additional simulations included in this report are effect of pool temperature, reduced steady-state flow rate, in-pool loss of coolant accidents, and loss of external cooling. The simulations described in this document have been performed for both an HEU- and LEU-fueled core.

  10. LIQUID HYDROCARBON FUEL CELL DEVELOPMENT.

    Science.gov (United States)

    A compound anode consists of a reforming catalyst bed in direct contact with a palladium-silver fuel cell anode. The objective of this study was to...prove the feasibility of operating a compound anode fuel cell on a liquid hydrocarbon and to define the important parameters that influence cell...performance. Both reformer and fuel cell tests were conducted with various liquid hydrocarbon fuels. Included in this report is a description of the

  11. Residential fuel quality

    Energy Technology Data Exchange (ETDEWEB)

    Santa, T. [Santa Fuel, Inc., Bridgeport, CT (United States)

    1997-09-01

    This report details progress made in improving the performance of No. 2 heating oil in residential applications. Previous research in this area is documented in papers published in the Brookhaven Oil Heat Technology Conference Proceedings in 1993, 1994 and 1996. By way of review we have investigated a number of variables in the search for improved fuel system performance. These include the effect of various additives designed to address stability, dispersion, biotics, corrosion and reaction with metals. We have also investigated delivery methods, filtration, piping arrangements and the influence of storage tank size and location. As a result of this work Santa Fuel Inc. in conjunction with Mobile Oil Corporation have identified an additive package which shows strong evidence of dramatically reducing the occurrence of fuel system failures in residential oil burners. In a broad market roll-out of the additized product we have experienced a 29% reduction in fuel related service calls when comparing the 5 months ending January 1997 to the same period ending January 1996.

  12. Minimum fuel mode evaluation

    Science.gov (United States)

    Orme, John S.; Nobbs, Steven G.

    1995-01-01

    The minimum fuel mode of the NASA F-15 research aircraft is designed to minimize fuel flow while maintaining constant net propulsive force (FNP), effectively reducing thrust specific fuel consumption (TSFC), during cruise flight conditions. The test maneuvers were at stabilized flight conditions. The aircraft test engine was allowed to stabilize at the cruise conditions before data collection initiated; data were then recorded with performance seeking control (PSC) not-engaged, then data were recorded with the PSC system engaged. The maneuvers were flown back-to-back to allow for direct comparisons by minimizing the effects of variations in the test day conditions. The minimum fuel mode was evaluated at subsonic and supersonic Mach numbers and focused on three altitudes: 15,000; 30,000; and 45,000 feet. Flight data were collected for part, military, partial, and maximum afterburning power conditions. The TSFC savings at supersonic Mach numbers, ranging from approximately 4% to nearly 10%, are in general much larger than at subsonic Mach numbers because of PSC trims to the afterburner.

  13. Fields of Fuel

    Science.gov (United States)

    Russ, Rosemary S.; Wangen, Steve; Nye, D. Leith; Shapiro, R. Benjamin; Strinz, Will; Ferris, Michael

    2015-01-01

    To help teachers engage students in discussions about sustainability, the authors designed Fields of Fuel, a multiplayer, web-based simulation game that allows players to explore the environmental and economic trade-offs of a realistic sustainable system. Computer-based simulations of real-world phenomena engage students and have been shown to…

  14. Solid Oxide Fuel Cell

    DEFF Research Database (Denmark)

    2010-01-01

    The solid oxide fuel cell comprising a metallic support material, an active anode layer consisting of a good hydrocarbon cracking catalyst, an electrolyte layer, an active cathode layer, and a transition layer consisting of preferably a mixture of LSM and a ferrite to the cathode current collector...

  15. Fire and fuels

    Science.gov (United States)

    Brandon Collins; Carl Skinner

    2014-01-01

    Recent studies of historical fire regimes indicate that fires occurring prior to Euro-American settlement were characterized by a high degree of spatial complexity that was driven by heterogeneity in vegetation/fuels and topography and influenced by variability in climate, which mediated the timing, effects, and extents of fires over time. Although there are many...

  16. Future fuel production

    NARCIS (Netherlands)

    Jacobs, E.

    2011-01-01

    This paper is written for the TIDO-course AR0532 Smart & Bioclimatic Design Theory. The objective of this paper is to see where the possibilities are in future fuel production in a sustainable way, and the integration of it in the built environment. Technologies are compared and evaluated, and the

  17. Solid Oxide Fuel Cell

    DEFF Research Database (Denmark)

    2010-01-01

    The solid oxide fuel cell comprising a metallic support material, an active anode layer consisting of a good hydrocarbon cracking catalyst, an electrolyte layer, an active cathode layer, and a transition layer consisting of preferably a mixture of LSM and a ferrite to the cathode current collector...

  18. Fuel nitrogen conversion in solid fuel fired systems

    Energy Technology Data Exchange (ETDEWEB)

    Glarborg, P.; Jensen, A.D.; Johnsson, J.E. [Technical University of Denmark, Lyngby (Denmark). Department of Chemical Engineering

    2003-07-01

    Understanding of the chemical and physical processes that govern formation and destruction of nitrogen oxides (NO{sub x}) in combustion of solid fuels continues to be a challenge. Even though this area has been the subject of extensive research over the last three decades, there are still unresolved issues that may limit the potential of primary measures for NO{sub x} control. In most solid fuel fired systems oxidation of fuel-bound nitrogen constitutes the dominating source of nitrogen oxides. The present paper reviews some fundamental aspects of fuel nitrogen conversion in these systems, emphasizing mostly combustion of coal since most previous work deal with this fuel. However, also results on biomass combustion is discussed. Homogeneous and heterogeneous pathways in fuel NO formation and destruction are discussed and the effect of fuel characteristics, devolatilization conditions and combustion mode on the oxidation selectivity towards NO and N{sub 2} is evaluated. Results indicate that even under idealized conditions, such as a laminar pulverized-fuel flame, the governing mechanisms for fuel nitrogen conversion are not completely understood. Light gases, tar, char and soot may all be important vehicles for fuel-N conversion, with their relative importance depending on fuel rank and reaction conditions. Oxygen availability and fuel-nitrogen level are major parameters determining the oxidation selectivity of fuel-N towards NO and N{sub 2}, but also the ability of char and soot to reduce NO is potentially important. The impact of fuel/oxidizer mixing pattern on NO formation appears to be less important in solid-fuel flames than in homogeneous flames. (author)

  19. Fuel Handbook[Wood and other renewable fuels

    Energy Technology Data Exchange (ETDEWEB)

    Stroemberg, Birgitta [TPS Termiska Processer AB, Nykoeping (SE)] (ed.)

    2006-03-15

    This handbook on renewable fuels is intended for power and heat producers in Sweden. This fuel handbook provides, from a plant owner's perspective, a method to evaluate different fuels on the market. The fuel handbook concerns renewable fuels (but does not include household waste) that are available on the Swedish market today or fuels that have potential to be available within the next ten years. The handbook covers 26 different fuels. Analysis data, special properties, operating experiences and literature references are outlined for each fuel. [Special properties, operating experiences and literature references are not included in this English version] The handbook also contains: A proposed methodology for introduction of new fuels. A recommendation of analyses and tests to perform in order to reduce the risk of problems is presented. [The recommendation of analyses and tests is not included in the English version] A summary of relevant laws and taxes for energy production, with references to relevant documentation. [Only laws and taxes regarding EU are included] Theory and background to evaluate a fuel with respect to combustion, ash and corrosion properties and methods that can be used for such evaluations. Summary of standards, databases and handbooks on biomass fuels and other solid fuels, and links to web sites where further information about the fuels can be found. The appendices includes: A methodology for trial firing of fuels. Calculations procedures for, amongst others, heating value, flue gas composition, key number and free fall velocity [Free fall velocity is not included in the English version]. In addition, conversion routines between different units for a number of different applications are provided. Fuel analyses are presented in the appendix. (The report is a translation of parts of the report VARMEFORSK--911 published in 2005)

  20. Nuclear and radiological safety in the substitution process of the fuel HEU to LEU 30/20 in the Reactor TRIGA Mark III of the ININ; Seguridad nuclear y radiologica en el proceso de sustitucion del combustible HEU a LEU 30/20 en el Reactor TRIGA Mark III del Instituto Nacional de Investigaciones Nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez G, J., E-mail: jaime.hernandez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-10-15

    Inside the safety initiative in the international ambit, with the purpose of reducing the risks associated with the use of high enrichment nuclear fuels (HEU) for different proposes to the peaceful uses of the nuclear energy, Mexico contributes by means of the substitution of the high enrichment fuel HEU for low enrichment fuel LEU 30/20 in the TRIGA Mark III Reactor, belonging to Instituto Nacional de Investigaciones Nucleares (ININ). The conversion process was carried out by means of the following activities: analysis of the proposed core, reception and inspection of the fuel LEU 30/20, the discharge of the fuels of the mixed reactor core, shipment of the fuels HEU fresh and irradiated to the origin country, reload activities with the fuels LEU 30/20 and parameters measurement of the core operation. In order to maintaining the personnel's integrity and infrastructure associated to the Reactor, during the whole process the measurements of nuclear and radiological safety were controlled to detail, in execution with the license requirements of the installation. This work describes the covering activities and radiological inspections more relevant, as well as the measurements of radiological control implemented with base in the estimate of the equivalent dose of the substitution process. (Author)

  1. Nuclear fuel elements design, fabrication and performance

    CERN Document Server

    Frost, Brian R T

    1982-01-01

    Nuclear Fuel Elements: Design, Fabrication and Performance is concerned with the design, fabrication, and performance of nuclear fuel elements, with emphasis on fast reactor fuel elements. Topics range from fuel types and the irradiation behavior of fuels to cladding and duct materials, fuel element design and modeling, fuel element performance testing and qualification, and the performance of water reactor fuels. Fast reactor fuel elements, research and test reactor fuel elements, and unconventional fuel elements are also covered. This volume consists of 12 chapters and begins with an overvie

  2. Fuel cell report to congress

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2003-02-28

    This report describes the status of fuel cells for Congressional committees. It focuses on the technical and economic barriers to the use of fuel cells in transportation, portable power, stationary, and distributed power generation applications, and describes the need for public-private cooperative programs to demonstrate the use of fuel cells in commercial-scale applications by 2012. (Department of Energy, February 2003).

  3. Commercialization of fuel-cells

    Energy Technology Data Exchange (ETDEWEB)

    Penner, S.S.; Appleby, A.J.; Baker, B.S.; Bates, J.L.; Buss, L.B.; Dollard, W.J.; Farris, P.J.; Gillis, E.A.; Gunsher, J.A.; Khandkar, A.; Krumpelt, M.; O' Sullivan, J.B.; Runte, G.; Savinell, R.F.; Selman, J.R.; Shores, D.A.; Tarman, P.

    1995-03-01

    This report is an abbreviated version of the ''Report of the DOE Advanced Fuel Cell Commercialization Working Group (AFC2WG),'' released January 1995. We describe fuel-cell commercialization for stationary power applications of phosphoric acid, molten carbonate, solid oxide, and polymer electrolyte membrane fuel cells.

  4. Fuel cell sub-assembly

    Science.gov (United States)

    Chi, Chang V.

    1983-01-01

    A fuel cell sub-assembly comprising a plurality of fuel cells, a first section of a cooling means disposed at an end of the assembly and means for connecting the fuel cells and first section together to form a unitary structure.

  5. Foaming-electrolyte fuel cell

    Science.gov (United States)

    Nanis, L.; Saunders, A. P.

    1970-01-01

    Foam structure feeds fuel gas solution into electrolyte. Fuel gas reacts at static, three-phase interface between fuel gas, electrolyte, and electrode material. The foam forms an electrical contact between main body of electrolyte and the electrode, and aids in removal of by-products of the chemical reaction.

  6. VIRTUAL FUEL-PUMP DESIGN

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

    Some concepts of virtual product are discussed. The key technologies of virtual fuel-pump development are in detail analysed, which include virtual fuel-pump product modeling, intelligent simulation, distributed design environment, and virtual assembly. The virtual fuel-pump development prototype system considers requirement analysis, concept design, injection preferment analysis, detailed design, and assembly analysis.

  7. Boiler design for fuel economy

    Energy Technology Data Exchange (ETDEWEB)

    Ramaswamy, M.P.; Sastry, C.V.R.L.; Tharakraj, M.

    1980-03-01

    In view of the limited fuel resources and ever increasing demand, Bharat Heavy Electricals, Ltd. (BHEL), as the leading boiler manufacturer, always endeavours to effect fuel economy in all possible avenues, leaving no stone unturned in this effort. This paper outlines some of the major efforts of BHEL in the area of boiler design to effect fuel economy.

  8. FUEL CELL MANPACK POWER SOURCE.

    Science.gov (United States)

    battery provides required power density and instantly available power while the fuel cell efficiently converts a primary fuel to electrical power at a...field supply, afford an extremely high energy density making the hybrid fuel cell system competitive on cost per kilowatt hour with standard military zinc-carbon primary batteries. (Author)

  9. Fire-safe hydrocarbon fuels

    Energy Technology Data Exchange (ETDEWEB)

    Fodor, G.E.; Weatherford, W.D. Jr.; Wright, B.R.

    1979-11-06

    A stabilized, fire-safe, aqueous hydrocarbon fuel emulsion prepared by mixing: a diesel fuel; an emulsifier (consisting of oleyl diethanolamide, diethanolamine, and diethanolamine soap of oleic acid) which has been treated with about 0 to 7 1/2 of oleic acid. A modified version of this fuel also contains 0 to 0.5% of an antimisting agent, and water.

  10. Carbon fuel particles used in direct carbon conversion fuel cells

    Science.gov (United States)

    Cooper, John F.; Cherepy, Nerine

    2012-10-09

    A system for preparing particulate carbon fuel and using the particulate carbon fuel in a fuel cell. Carbon particles are finely divided. The finely dividing carbon particles are introduced into the fuel cell. A gas containing oxygen is introduced into the fuel cell. The finely divided carbon particles are exposed to carbonate salts, or to molten NaOH or KOH or LiOH or mixtures of NaOH or KOH or LiOH, or to mixed hydroxides, or to alkali and alkaline earth nitrates.

  11. Microbial fuel cell treatment of fuel process wastewater

    Science.gov (United States)

    Borole, Abhijeet P; Tsouris, Constantino

    2013-12-03

    The present invention is directed to a method for cleansing fuel processing effluent containing carbonaceous compounds and inorganic salts, the method comprising contacting the fuel processing effluent with an anode of a microbial fuel ell, the anode containing microbes thereon which oxidatively degrade one or more of the carbonaceous compounds while producing electrical energy from the oxidative degradation, and directing the produced electrical energy to drive an electrosorption mechanism that operates to reduce the concentration of one or more inorganic salts in the fuel processing effluent, wherein the anode is in electrical communication with a cathode of the microbial fuel cell. The invention is also directed to an apparatus for practicing the method.

  12. LMFBR fuel assembly design for HCDA fuel dispersal

    Science.gov (United States)

    Lacko, Robert E.; Tilbrook, Roger W.

    1984-01-01

    A fuel assembly for a liquid metal fast breeder reactor having an upper axial blanket region disposed in a plurality of zones within the fuel assembly. The characterization of a zone is dependent on the height of the axial blanket region with respect to the active fuel region. The net effect of having a plurality of zones is to establish a dispersal flow path for the molten materials resulting during a core meltdown accident. Upward flowing molten material can escape from the core region and/or fuel assembly without solidifying on the surface of fuel rods due to the heat sink represented by blanket region pellets.

  13. 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.

  14. A solid fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sonetaka, K.; Iketani, Y.; Nisino, A.; Takeuti, Y.

    1983-04-15

    A solid fuel is proposed which consists of cylindrical or prismatic pieces, whose surfaces (Pv) are equipped with greater than or equal to one channel, while the pieces themselves have greater than or equal to 1 through opening; the ratio of the total surface to the surface of the channels is within 95 to 5 to 60 to 40. The presence of the channels and the through openings facilitates the feeding of air to the surface of the fuel, providing in such a way, better ignition, the propagation of the flames and the completeness of combustion of the solid fuels based on carbonaceous materials. The optimal composition of the proposed fuel is: 70 to 95 percent carbonaceous base (coal, coke, charcoal, graphite, petroleum coke and so on), 1 to 25 percent desulfurizing additive (carbonate, hydroxide or nitrate of sodium, potassium, calcium or magnesium, 0 to 15 percent combustion accelerator (oxidizers: nitrates of potassium and barium, potassium perchlorate, oxides of magnesium, iron or manganese, aluminum powder and so on), 0 to 10 percent forming additive (bentonite, talc, kaolin and so on) and 0.5 to 5 percent binder (pitch, resin, pulp, cement and so on). The optimal characteristics of the combustion process are reached at a weight of a single piece of fuel of approximately 20 grams, a seeming density of less than or equal to 1.3 grams per cubic centimeter, a cross section surface area of 25 to 1,600 square millimeters and a filling density with combustion of less than or equal to 0.8 grams per cubic centimeter.

  15. Dissolution Flowsheet for High Flux Isotope Reactor Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Daniel, W. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Rudisill, T. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); O' Rourke, P. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Karay, N. S [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-09-27

    As part of the Spent Nuclear Fuel (SNF) processing campaign, H-Canyon is planning to begin dissolving High Flux Isotope Reactor (HFIR) fuel in late FY17 or early FY18. Each HFIR fuel core contains inner and outer fuel elements which were fabricated from uranium oxide (U3O8) dispersed in a continuous Al phase using traditional powder metallurgy techniques. Fuels fabricated in this manner, like other SNF’s processed in H-Canyon, dissolve by the same general mechanisms with similar gas generation rates and the production of H2. The HFIR fuel cores will be dissolved and the recovered U will be down-blended into low-enriched U. HFIR fuel was previously processed in H-Canyon using a unique insert in both the 6.1D and 6.4D dissolvers. Multiple cores will be charged to the same dissolver solution maximizing the concentration of dissolved Al. The objective of this study was to identify flowsheet conditions through literature review and laboratory experimentation to safely and efficiently dissolve the HFIR fuel in H-Canyon. Laboratory-scale experiments were performed to evaluate the dissolution of HFIR fuel using both Al 1100 and Al 6061 T6 alloy coupons. The Al 1100 alloy was considered a representative surrogate which provided an upper bound on the generation of flammable (i.e., H2) gas during the dissolution process. The dissolution of the Al 6061 T6 alloy proceeded at a slower rate than the Al 1100 alloy and was used to verify that the target Al concentration in solution could be achieved for the selected Hg concentration. Mass spectrometry and Raman spectroscopy were used to provide continuous monitoring of the concentration of H2 and other permanent gases in the dissolution offgas allowing the development of H2 generation rate profiles. The H2 generation rates were subsequently used to evaluate if a full HFIR core could be dissolved in an H-Canyon dissolver without exceeding 60% of the calculated lower

  16. Dissolution Flowsheet for High Flux Isotope Reactor Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Daniel, W. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Rudisill, T. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); O' Rourke, P. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Karay, N. S [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-09-27

    As part of the Spent Nuclear Fuel (SNF) processing campaign, H-Canyon is planning to begin dissolving High Flux Isotope Reactor (HFIR) fuel in late FY17 or early FY18. Each HFIR fuel core contains inner and outer fuel elements which were fabricated from uranium oxide (U3O8) dispersed in a continuous Al phase using traditional powder metallurgy techniques. Fuels fabricated in this manner, like other SNF’s processed in H-Canyon, dissolve by the same general mechanisms with similar gas generation rates and the production of H2. The HFIR fuel cores will be dissolved and the recovered U will be down-blended into low-enriched U. HFIR fuel was previously processed in H-Canyon using a unique insert in both the 6.1D and 6.4D dissolvers. Multiple cores will be charged to the same dissolver solution maximizing the concentration of dissolved Al. The objective of this study was to identify flowsheet conditions through literature review and laboratory experimentation to safely and efficiently dissolve the HFIR fuel in H-Canyon. Laboratory-scale experiments were performed to evaluate the dissolution of HFIR fuel using both Al 1100 and Al 6061 T6 alloy coupons. The Al 1100 alloy was considered a representative surrogate which provided an upper bound on the generation of flammable (i.e., H2) gas during the dissolution process. The dissolution of the Al 6061 T6 alloy proceeded at a slower rate than the Al 1100 alloy, and was used to verify that the target Al concentration in solution could be achieved for the selected Hg concentration. Mass spectrometry and Raman spectroscopy were used to provide continuous monitoring of the concentration of H2 and other permanent gases in the dissolution offgas, allowing the development of H2 generation rate profiles. The H2 generation rates were subsequently used to evaluate if a full HFIR core could be dissolved in an H-Canyon dissolver without exceeding 60% of the

  17. Fuel Cell Handbook, Fifth Edition

    Energy Technology Data Exchange (ETDEWEB)

    Energy and Environmental Solutions

    2000-10-31

    Progress continues in fuel cell technology since the previous edition of the Fuel Cell Handbook was published in November 1998. Uppermost, polymer electrolyte fuel cells, molten carbonate fuel cells, and solid oxide fuel cells have been demonstrated at commercial size in power plants. The previously demonstrated phosphoric acid fuel cells have entered the marketplace with more than 220 power plants delivered. Highlighting this commercial entry, the phosphoric acid power plant fleet has demonstrated 95+% availability and several units have passed 40,000 hours of operation. One unit has operated over 49,000 hours. Early expectations of very low emissions and relatively high efficiencies have been met in power plants with each type of fuel cell. Fuel flexibility has been demonstrated using natural gas, propane, landfill gas, anaerobic digester gas, military logistic fuels, and coal gas, greatly expanding market opportunities. Transportation markets worldwide have shown remarkable interest in fuel cells; nearly every major vehicle manufacturer in the U.S., Europe, and the Far East is supporting development. This Handbook provides a foundation in fuel cells for persons wanting a better understanding of the technology, its benefits, and the systems issues that influence its application. Trends in technology are discussed, including next-generation concepts that promise ultrahigh efficiency and low cost, while providing exceptionally clean power plant systems. Section 1 summarizes fuel cell progress since the last edition and includes existing power plant nameplate data. Section 2 addresses the thermodynamics of fuel cells to provide an understanding of fuel cell operation at two levels (basic and advanced). Sections 3 through 8 describe the six major fuel cell types and their performance based on cell operating conditions. Alkaline and intermediate solid state fuel cells were added to this edition of the Handbook. New information indicates that manufacturers have stayed

  18. Development of DUPIC fuel cycle technology - Assessment of Wolsong NPP fuel handling system for DUPIC fuel

    Energy Technology Data Exchange (ETDEWEB)

    Na, Bok Gyun; Nam, Gung Ihn [Korea Power Engineering Company, Taejon (Korea)

    2000-04-01

    The DUPIC fuel loading and discharge path of Wolsong NPP is studied assuming that DUPIC fuel is used at Wolsong NPP. Spent DUPIC fuel discharge path is irrelevant, since it uses the same spent fuel discharge path. Number of factors such as safety, economics of design change, radiation exposure to operators, easy of operation and maintenance, etc, are considered in the evaluation of path. A more detailed analysis of cost estimation of the selected path is also carried out. The study shows that DUPIC fuel loading path following through Spent Fuel Storage Bay and Spent Fuel Discharge Port in reverse direction will minimize the design change and additional equipment and radiation exposure to operators. The estimated total cost of using DUPIC fuel in Wolsong NPP based on price index of year 2000 is around 4.5 billion won. 4 refs., 30 figs., 13 tabs. (Author)

  19. Solid fuels as engine fuels. Kiinteiden polttoaineiden kaeyttoemahdollisuudet moottoripolttoaineena

    Energy Technology Data Exchange (ETDEWEB)

    Vakkilainen, A.; Nylund, N.-O.

    1986-07-01

    The use of solid fuels as engine fuels is discussed in this literature study. The present liquid fuel engines require extensive and expensive changes to overcome difficulties due to solid fuels. The solid particles result in increasing wear in the engine, in the fuel feed system and everywhere, where the particles come into touch with moving surfaces. The rate of wear has been as high as 100-fold compared to that caused by liquid fuels. Large medium-fast or slow diesel engines seem to meet best the requirements set by solid fuels. The experiment carried out by the Swiss Sulzer Engine Works are the most promising engine experiments carried out so far. In Sulzer's experiments, coal-water slurries containing 50-70 wt-% coal have been used as the fuel. Burning has been rather complete, but wear and the high price of the coal-water slurry seem to be unsolvable problems. The development work on enines is still at an early stage and a solid fuel engine will not be manufactured in series in the near future. The pulverous fuel of the future will be a mixture of some liquid and of some solid fuel powder, and hence the handling-technical problems will be considerably smaller than using powder only. Powder manufacture with the present techniques is energy-wasting. Most engine experiments have been carried out on on coal slurries with < 30 ..mu..m particle size. It is not economic to produce such powders at present.

  20. Fuel Cell Handbook, Fourth Edition

    Energy Technology Data Exchange (ETDEWEB)

    Stauffer, D.B; Hirschenhofer, J.H.; Klett, M.G.; Engleman, R.R.

    1998-11-01

    Robust progress has been made in fuel cell technology since the previous edition of the Fuel Cell Handbook was published in January 1994. This Handbook provides a foundation in fuel cells for persons wanting a better understanding of the technology, its benefits, and the systems issues that influence its application. Trends in technology are discussed, including next-generation concepts that promise ultra high efficiency and low cost, while providing exceptionally clean power plant systems. Section 1 summarizes fuel cell progress since the last edition and includes existing power plant nameplate data. Section 2 addresses the thermodynamics of fuel cells to provide an understanding of fuel cell operation at two levels (basic and advanced). Sections 3 through 6 describe the four major fuel cell types and their performance based on cell operating conditions. The section on polymer electrolyte membrane fuel cells has been added to reflect their emergence as a significant fuel cell technology. Phosphoric acid, molten carbonate, and solid oxide fuel cell technology description sections have been updated from the previous edition. New information indicates that manufacturers have stayed with proven cell designs, focusing instead on advancing the system surrounding the fuel cell to lower life cycle costs. Section 7, Fuel Cell Systems, has been significantly revised to characterize near-term and next-generation fuel cell power plant systems at a conceptual level of detail. Section 8 provides examples of practical fuel cell system calculations. A list of fuel cell URLs is included in the Appendix. A new index assists the reader in locating specific information quickly.

  1. Sipping fuel and saving lives: increasing fuel economy withoutsacrificing safety

    Energy Technology Data Exchange (ETDEWEB)

    Gordon, Deborah; Greene, David L.; Ross, Marc H.; Wenzel, Tom P.

    2007-06-11

    The public, automakers, and policymakers have long worried about trade-offs between increased fuel economy in motor vehicles and reduced safety. The conclusion of a broad group of experts on safety and fuel economy in the auto sector is that no trade-off is required. There are a wide variety of technologies and approaches available to advance vehicle fuel economy that have no effect on vehicle safety. Conversely, there are many technologies and approaches available to advance vehicle safety that are not detrimental to vehicle fuel economy. Congress is considering new policies to increase the fuel economy of new automobiles in order to reduce oil dependence and reduce greenhouse gas emissions. The findings reported here offer reassurance on an important dimension of that work: It is possible to significantly increase the fuel economy of motor vehicles without compromising their safety. Automobiles on the road today demonstrate that higher fuel economy and greater safety can co-exist. Some of the safest vehicles have higher fuel economy, while some of the least safe vehicles driven today--heavy, large trucks and SUVs--have the lowest fuel economy. At an October 3, 2006 workshop, leading researchers from national laboratories, academia, auto manufacturers, insurance research industry, consumer and environmental groups, material supply industries, and the federal government agreed that vehicles could be designed to simultaneously improve safety and fuel economy. The real question is not whether we can realize this goal, but the best path to get there. The experts' studies reveal important new conclusions about fuel economy and safety, including: (1) Vehicle fuel economy can be increased without affecting safety, and vice versa; (2) Reducing the weight and height of the heaviest SUVs and pickup trucks will simultaneously increase both their fuel economy and overall safety; and (3) Advanced materials can decouple size from mass, creating important new possibilities

  2. Checklist for transition to new highway fuel(s).

    Energy Technology Data Exchange (ETDEWEB)

    Risch, C.; Santini, D.J. (Energy Systems)

    2011-12-15

    Transportation is vital to the U.S. economy and society. As such, U.S. Presidents have repeatedly stated that the nation needs to reduce dependence on petroleum, especially for the highway transportation sector. Throughout history, highway transportation fuel transitions have been completed successfully both in United States and abroad. Other attempts have failed, as described in Appendix A: Historical Highway Fuel Transitions. Planning for a transition is critical because the changes can affect our nation's ability to compete in the world market. A transition will take many years to complete. While it is tempting to make quick decisions about the new fuel(s) of choice, it is preferable and necessary to analyze all the pertinent criteria to ensure that correct decisions are made. Doing so will reduce the number of changes in highway fuel(s). Obviously, changes may become necessary because of occurrences such as significant technology breakthroughs or major world events. With any and all of the possible transitions to new fuel(s), the total replacement of gasoline and diesel fuels is not expected. These conventional fuels are envisioned to coexist with the new fuel(s) for decades, while the revised fuel and vehicle infrastructures are implemented. The transition process must analyze the needs of the primary 'players,' which consist of the customers, the government, the fuel industry, and the automotive industry. To maximize the probability of future successes, the prime considerations of these groups must be addressed. Section 2 presents a succinct outline of the Checklist. Section 3 provides a brief discussion about the groupings on the Checklist.

  3. Spent-fuel-storage alternatives

    Energy Technology Data Exchange (ETDEWEB)

    1980-01-01

    The Spent Fuel Storage Alternatives meeting was a technical forum in which 37 experts from 12 states discussed storage alternatives that are available or are under development. The subject matter was divided into the following five areas: techniques for increasing fuel storage density; dry storage of spent fuel; fuel characterization and conditioning; fuel storage operating experience; and storage and transport economics. Nineteen of the 21 papers which were presented at this meeting are included in this Proceedings. These have been abstracted and indexed. (ATT)

  4. Research on aviation fuel instability

    Science.gov (United States)

    Baker, C. E.; Bittker, D. A.; Cohen, S. M.; Seng, G. T.

    1984-01-01

    The problems associated with aircraft fuel instability are discussed. What is currently known about the problem is reviewed and a research program to identify those areas where more research is needed is discussed. The term fuel instability generally refers to the gums, sediments, or deposits which can form as a result of a set of complex chemical reactions when a fuel is stored for a long period at ambient conditions or when the fuel is thermally stressed inside the fuel system of an aircraft.

  5. Solar Fuels: Vision and Concepts

    OpenAIRE

    Styring, Stenbjörn

    2012-01-01

    The world needs new, environmentally friendly and renewable fuels to allow an exchange from fossil fuels. The fuel must be made from cheap and ‘endless’ resources that are available everywhere. The new research area on solar fuels, which are made from solar energy and water, aims to meet this demand. The paper discusses why we need a solar fuel and why electricity is not enough; it proposes solar energy as the major renewable energy source to feed from. The present research strategies, involv...

  6. Spent-fuel-storage alternatives

    Energy Technology Data Exchange (ETDEWEB)

    1980-01-01

    The Spent Fuel Storage Alternatives meeting was a technical forum in which 37 experts from 12 states discussed storage alternatives that are available or are under development. The subject matter was divided into the following five areas: techniques for increasing fuel storage density; dry storage of spent fuel; fuel characterization and conditioning; fuel storage operating experience; and storage and transport economics. Nineteen of the 21 papers which were presented at this meeting are included in this Proceedings. These have been abstracted and indexed. (ATT)

  7. Sensor system for fuel transport vehicle

    Science.gov (United States)

    Earl, Dennis Duncan; McIntyre, Timothy J.; West, David L.

    2016-03-22

    An exemplary sensor system for a fuel transport vehicle can comprise a fuel marker sensor positioned between a fuel storage chamber of the vehicle and an access valve for the fuel storage chamber of the vehicle. The fuel marker sensor can be configured to measure one or more characteristics of one or more fuel markers present in the fuel adjacent the sensor, such as when the marked fuel is unloaded at a retail station. The one or more characteristics can comprise concentration and/or identity of the one or more fuel markers in the fuel. Based on the measured characteristics of the one or more fuel markers, the sensor system can identify the fuel and/or can determine whether the fuel has been adulterated after the marked fuel was last measured, such as when the marked fuel was loaded into the vehicle.

  8. Metallic fuel design development

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Woan; Kang, H. Y.; Lee, B. O. and others

    1999-04-01

    This report describes the R and D results of the ''Metallic Fuel Design Development'' project that performed as a part of 'Nuclear Research and Development Program' during the '97 - '98 project years. The objectives of this project are to perform the analysis of thermo-mechanical and irradiation behaviors, and preliminary conceptual design for the fuel system of the KALIMER liquid metal reactor. The following are the major results that obtained through the project. The preliminary design requirements and design criteria which are necessary in conceptual design stage, are set up. In the field of fuel pin design, the pin behavior analysis, failure probability prediction, and sensitivity analysis are performed under the operation conditions of steady-state and transient accidents. In the area of assembly duct analysis; 1) KAFACON-2D program is developed to calculate an array configuration of inner shape of assembly duct, 2) Stress-strain analysis are performed for the components of assembly such as, handling socket, mounting rail and wire wrap, 3) The BDI program is developed to analyze mechanical interaction between pin bundle and duct, 4) a vibration analysis is performed to understand flow-induced vibration of assembly duct, 5) The NUBOW-2D, which is bowing and deformation analysis code for assembly duct, is modified to be operated in KALIMER circumstance, and integrity evaluation of KALIMER core assembly is carried out using the modified NUBOW-2D and the CRAMP code in U.K., and 6) The KALIMER assembly duct is manufactured to be used in flow test. In the area of non-fuel assembly, such as control, reflector, shielding, GEM and USS, the states-of-the-arts and the major considerations in designing are evaluated, and the design concepts are derived. The preliminary design description and their design drawing of KALIMER fuel system are prepared based upon the above mentioned evaluation and analysis. The achievement of conceptual

  9. Nuclear fuels - Present and future

    Science.gov (United States)

    Olander, D.

    2009-06-01

    The important developments in nuclear fuels and their problems are reviewed and compared with the status of present light-water reactor fuels. The limitations of LWR fuels are reviewed with respect to important recent concerns, namely provision of outlet coolant temperatures high enough for use in H 2 production, destruction of plutonium to eliminate proliferation concerns, and burning of the minor actinides to reduce the waste repository heat load and long-term radiation hazard. In addition to current oxide-based fuel rod designs, the hydride fuel with liquid-metal thermal bonding of the fuel-cladding gap is covered. Finally, two of the most promising Generation IV reactor concepts, the very high temperature reactor and the sodium fast reactor, and the accompanying reprocessing technologies, aqueous-based UREX+1a and pyrometallurgical, are summarized. In all of the topics covered, the thermodynamics involved in the fuel's behavior under irradiation and in the reprocessing schemes are emphasized.

  10. Alternative Fuels in Cement Production

    DEFF Research Database (Denmark)

    Larsen, Morten Boberg

    The substitution of alternative for fossil fuels in cement production has increased significantly in the last decade. Of these new alternative fuels, solid state fuels presently account for the largest part, and in particular, meat and bone meal, plastics and tyre derived fuels (TDF) accounted...... of the fuel heating value. In addition, the devolatilization time of alternative fuels cannot be neglected in kiln system process analyses, as these fuels are typically in the cm-size with devolatilization times in the order of minutes. The devolatilization characteristics of large particles of tyre rubber...... time, where increased particle size increased the devolatilization time. Model analyses demonstrated that the overall devolatilization kinetics of large particles of tyre rubber is mainly controlled by heat transfer and intrinsic pyrolysis kinetics, whereas mass transfer has negligible influence...

  11. Fuel services; Servicios de combustible

    Energy Technology Data Exchange (ETDEWEB)

    Marta, H.; Alvarez, P.; Jimenez, J.

    2006-07-01

    Refuelling outages comprise a number of maintenance tasks scheduled long in advance to assure a reliable operation throughout the next cycle and, in the long run, a safer and more efficient plant. Most of these tasks are routine service of mechanical and electrical system and likewise fuel an be considered a critical component as to handling, inspection, cleaning and repair. ENUSA-ENWESA AIE has been working in this area since 1995 growing from fuel repair to a more integrated service that includes new and spent fuel handling, inserts, failed fuel rod detection systems, ultrasonic fuel cleaning,fuel repair and a comprehensive array of inspection and tests related to the reliability of the mechanical components in the fuel assembly, all this, performed in compliance with quality, safety, health physics and any other nuclear standard. (Author)

  12. Initial performance assessment of the disposal of spent nuclear fuel and high-level waste stored at Idaho National Engineering Laboratory. Volume 2: Appendices

    Energy Technology Data Exchange (ETDEWEB)

    Rechard, R.P. [ed.

    1993-12-01

    This performance assessment characterized plausible treatment options conceived by the Idaho National Engineering Laboratory (INEL) for its spent fuel and high-level radioactive waste and then modeled the performance of the resulting waste forms in two hypothetical, deep, geologic repositories: one in bedded salt and the other in granite. The results of the performance assessment are intended to help guide INEL in its study of how to prepare wastes and spent fuel for eventual permanent disposal. This assessment was part of the Waste Management Technology Development Program designed to help the US Department of Energy develop and demonstrate the capability to dispose of its nuclear waste, as mandated by the Nuclear Waste Policy Act of 1982. The waste forms comprised about 700 metric tons of initial heavy metal (or equivalent units) stored at the INEL: graphite spent fuel, experimental low enriched and highly enriched spent fuel, and high-level waste generated during reprocessing of some spent fuel. Five different waste treatment options were studied; in the analysis, the options and resulting waste forms were analyzed separately and in combination as five waste disposal groups. When the waste forms were studied in combination, the repository was assumed to also contain vitrified high-level waste from three DOE sites for a common basis of comparison and to simulate the impact of the INEL waste forms on a moderate-sized repository, The performance of the waste form was assessed within the context of a whole disposal system, using the U.S. Environmental Protection Agency`s Environmental Radiation Protection Standards for Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes, 40 CFR 191, promulgated in 1985. Though the waste form behavior depended upon the repository type, all current and proposed waste forms provided acceptable behavior in the salt and granite repositories.

  13. Operando fuel cell spectroscopy

    Science.gov (United States)

    Kendrick, Ian Michael

    The active state of a catalyst only exists during catalysis (1) provided the motivation for developing operando spectroscopic techniques. A polymer electrolyte membrane fuel cell (PEMFC) was designed to interface with commercially available instruments for acquisition of infrared spectra of the catalytic surface of the membrane electrode assembly (MEA) during normal operation. This technique has provided insight of the complex processes occurring at the electrode surface. Nafion, the solid electrolyte used in most modern-day polymer electrolyte membrane fuel cells (PEMFC), serves many purposes in fuel cell operation. However, there is little known of the interface between Nafion and the electrode surface. Previous studies of complex Stark tuning curves of carbon monoxide on the surface of a platinum electrode were attributed the co-adsorption of bisulfite ions originating from the 0.5M H2SO4 electrolyte used in the study(2). Similar tuning curves obtained on a fuel cell MEA despite the absence of supplemental electrolytes suggest the adsorption of Nafion onto platinum (3). The correlation of spectra obtained using attenuated total reflectance spectroscopy (ATR) and polarization modulated IR reflection-absorption spectroscopy (PM-IRRAS) to a theoretical spectrum generated using density functional theory (DFT) lead to development of a model of Nafion and platinum interaction which identified participation of the SO3- and CF3 groups in Nafion adsorption. The use of ethanol as a fuel stream in proton exchange membrane fuel cells provides a promising alternative to methanol. Relative to methanol, ethanol has a greater energy density, lower toxicity and can be made from the fermentation of biomass(4). Operando IR spectroscopy was used to study the oxidation pathway of ethanol and Stark tuning behavior of carbon monoxide on Pt, Ru, and PtRu electrodes. Potential dependent products such as acetaldehyde, acetic acid and carbon monoxide are identified as well as previously

  14. Fuel pin cladding

    Science.gov (United States)

    Vaidyanathan, S.; Adamson, M.G.

    1986-01-28

    Disclosed is an improved fuel pin cladding, particularly adapted for use in breeder reactors, consisting of composite tubing with austenitic steel on the outer portion of the thickness of the tube wall and with nickel and/or ferritic material on the inner portion of the thickness of the tube wall. The nickel forms a sacrificial barrier as it reacts with certain fission products thereby reducing fission product activity at the austenitic steel interface. The ferritic material forms a preventive barrier for the austenitic steel as it is immune to liquid metal embrittlement. The improved cladding permits the use of high density fuel which in turn leads to a better breeding ratio in breeder reactors, and will increase the threshold at which failure occurs during temperature transients. 2 figs.

  15. 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.

  16. Fuel pin cladding

    Science.gov (United States)

    Vaidyanathan, Swaminathan; Adamson, Martyn G.

    1986-01-01

    An improved fuel pin cladding, particularly adapted for use in breeder reactors, consisting of composite tubing with austenitic steel on the outer portion of the thickness of the tube wall and with nickel and/or ferritic material on the inner portion of the thickness of the tube wall. The nickel forms a sacrificial barrier as it reacts with certain fission products thereby reducing fission product activity at the austenitic steel interface. The ferritic material forms a preventive barrier for the austenitic steel as it is immune to liquid metal embrittlement. The improved cladding permits the use of high density fuel which in turn leads to a better breeding ratio in breeder reactors, and will increase the threshold at which failure occurs during temperature transients.

  17. LIFE Materials: Overview of Fuels and Structural Materials Issues Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, J

    2008-09-08

    The National Ignition Facility (NIF) project, a laser-based Inertial Confinement Fusion (ICF) experiment designed to achieve thermonuclear fusion ignition and burn in the laboratory, is under construction at the Lawrence Livermore National Laboratory (LLNL) and will be completed in April of 2009. Experiments designed to accomplish the NIF's goal will commence in late FY2010 utilizing laser energies of 1 to 1.3 MJ. Fusion yields of the order of 10 to 20 MJ are expected soon thereafter. Laser initiated fusion-fission (LIFE) engines have now been designed to produce nuclear power from natural or depleted uranium without isotopic enrichment, and from spent nuclear fuel from light water reactors without chemical separation into weapons-attractive actinide streams. A point-source of high-energy neutrons produced by laser-generated, thermonuclear fusion within a target is used to achieve ultra-deep burn-up of the fertile or fissile fuel in a sub-critical fission blanket. Fertile fuels including depleted uranium (DU), natural uranium (NatU), spent nuclear fuel (SNF), and thorium (Th) can be used. Fissile fuels such as low-enrichment uranium (LEU), excess weapons plutonium (WG-Pu), and excess highly-enriched uranium (HEU) may be used as well. Based upon preliminary analyses, it is believed that LIFE could help meet worldwide electricity needs in a safe and sustainable manner, while drastically shrinking the nation's and world's stockpile of spent nuclear fuel and excess weapons materials. LIFE takes advantage of the significant advances in laser-based inertial confinement fusion that are taking place at the NIF at LLNL where it is expected that thermonuclear ignition will be achieved in the 2010-2011 timeframe. Starting from as little as 300 to 500 MW of fusion power, a single LIFE engine will be able to generate 2000 to 3000 MWt in steady state for periods of years to decades, depending on the nuclear fuel and engine configuration. Because the fission

  18. Water reactive hydrogen fuel cell power system

    Science.gov (United States)

    Wallace, Andrew P; Melack, John M; Lefenfeld, Michael

    2014-01-21

    A water reactive hydrogen fueled power system includes devices and methods to combine reactant fuel materials and aqueous solutions to generate hydrogen. The generated hydrogen is converted in a fuel cell to provide electricity. The water reactive hydrogen fueled power system includes a fuel cell, a water feed tray, and a fuel cartridge to generate power for portable power electronics. The removable fuel cartridge is encompassed by the water feed tray and fuel cell. The water feed tray is refillable with water by a user. The water is then transferred from the water feed tray into a fuel cartridge to generate hydrogen for the fuel cell which then produces power for the user.

  19. Water reactive hydrogen fuel cell power system

    Science.gov (United States)

    Wallace, Andrew P; Melack, John M; Lefenfeld, Michael

    2014-11-25

    A water reactive hydrogen fueled power system includes devices and methods to combine reactant fuel materials and aqueous solutions to generate hydrogen. The generated hydrogen is converted in a fuel cell to provide electricity. The water reactive hydrogen fueled power system includes a fuel cell, a water feed tray, and a fuel cartridge to generate power for portable power electronics. The removable fuel cartridge is encompassed by the water feed tray and fuel cell. The water feed tray is refillable with water by a user. The water is then transferred from the water feed tray into the fuel cartridge to generate hydrogen for the fuel cell which then produces power for the user.

  20. Fuel storage systems

    Energy Technology Data Exchange (ETDEWEB)

    Donakowski, T.D.; Tison, R.R.

    1979-08-01

    Storage technologies are characterized for solid, liquid, and gaseous fuels. Emphasis is placed on storage methods applicable to Integrated Community Energy Systems based on coal. Items discussed here include standard practice, materials and energy losses, environmental effects, operating requirements, maintenance and reliability, and cost considerations. All storage systems were found to be well-developed and to represent mature technologies; an exception may exist for low-Btu gas storage, which could have materials incompatability.

  1. Refractoriless liquid fuel burner

    Energy Technology Data Exchange (ETDEWEB)

    Musil, J.E.

    1986-07-15

    A liquid fuel burner head is described which consists of: A. a generally annular burner head housing spacedly enveloping a generally cylindrical primary air assembly, the head and assembly each having corresponding forward and rearward ends, (a) the primary air assembly having a plurality of internal primary air supply passage means extending in a generally forwardly direction in the assembly and emerging through annularly disposed primary air port means at the forward end of the primary air assembly, (b) means effective to produce a swirl of primary air in one direction about the axis of the primary air assembly as the air emerges from the primary air port means, (c) means associated with the primary air port means for adjusting the location of flame origin forward of and relative to the primary air port means, (d) the primary air assembly including a liquid fuel supply passage and a nozzle, the nozzle being centrally disposed at the forward end of the primary air assembly and encompassed by the primary air port means, the liquid fuel nozzle being effective to discharge a substantially fan-like spray of liquid fuel just forward of and across the primary air port means, (e) the primary air assembly and the nozzle together being axially movable relative to the housing between forwardmost and rearwardmost positions respectively responsive to change in burner firing rate between minimum and maximum; B. secondary air supply passage means disposed in the space between the housing and the primary air assembly; C. means rearwardly of the secondary air directional means and port means effective to meter the amount of secondary air supplied air port means from a lesser quantity when the primary air assembly and nozzle are in their forwardmost position to a greater quantity when the primary air assembly and nozzle are in their rearwardmost position.

  2. Fuel Cell Stacks

    Science.gov (United States)

    1975-04-01

    AD-A009 587 FUEL CELL STACKS Bernard S. Baker Energy Research Corporation Prepared for: Army Mobility Equipment Research and Development Center April... Mobility Equipment Research and Development Center Unclassified For- Belvoir, Virginia 22060 [15. DE.CLASSIFICATION/L.TWNOGRADING SCREOUJLE 16...the majority of effort has been directed at translating technoilogy for small comn- ponent manufacture on a laboratory scale into large size components

  3. Industrial Fuel Flexibility Workshop

    Energy Technology Data Exchange (ETDEWEB)

    none,

    2006-09-01

    On September 28, 2006, in Washington, DC, ITP and Booz Allen Hamilton conducted a fuel flexibility workshop with attendance from various stakeholder groups. Workshop participants included representatives from the petrochemical, refining, food and beverage, steel and metals, pulp and paper, cement and glass manufacturing industries; as well as representatives from industrial boiler manufacturers, technology providers, energy and waste service providers, the federal government and national laboratories, and developers and financiers.

  4. Protected Nuclear Fuel Element

    Science.gov (United States)

    Kittel, J. H.; Schumar, J. F.

    1962-12-01

    A stainless steel-clad actinide metal fuel rod for use in fast reactors is reported. In order to prevert cladding failures due to alloy formation between the actinide metal and the stainless steel, a mesh-like sleeve of expanded metal is interposed between them, the sleeve metal being of niobium, tantalum, molybdenum, tungsten, zirconium, or vanadium. Liquid alkali metal is added as a heat transfer agent. (AEC)

  5. Fuels from microalgae

    Energy Technology Data Exchange (ETDEWEB)

    1989-06-01

    Many species of aquatic plants can provide a source of renewable energy. Some species of microalgae, in particular, produce lipids -- oils that can be extracted and converted to a diesel fuel substitute or to gasoline. Since 1979 the Aquatic Species Program element of the Biofuels Program, has supported fundamental and applied research to develop the technology for using this renewable energy resource. This document, produced by the Solar Technical Information Program, provides an overview of the DOE/SERI Aquatic Species Program element. Chapter 1 is an introduction to the program and to the microalgae. Chapter 2 is an overview of the general principles involved in making fuels from microalgae. It also outlines the technical challenges to producing economic, high-energy transportation fuels. Chapter 3 provides an overview of the Algal Production and Economic Model (APEM). This model was developed by researchers within the program to identify aspects of the process critical to performance with the greatest potential to reduce costs. The analysis using this model has helped direct research sponsored by the program. Finally, Chapter 4 provides an overview of the Aquatic Species Program and describes current research. 28 refs., 17 figs.

  6. 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

  7. Sustainability of Fossil Fuels

    Science.gov (United States)

    Lackner, K. S.

    2002-05-01

    For a sustainable world economy, energy is a bottleneck. Energy is at the basis of a modern, technological society, but unlike materials it cannot be recycled. Energy or more precisely "negentropy" (the opposite of entropy) is always consumed. Thus, one either accepts the use of large but finite resources or must stay within the limits imposed by dilute but self-renewing resources like sunlight. The challenge of sustainable energy is exacerbated by likely growth in world energy demand due to increased population and increased wealth. Most of the world still has to undergo the transition to a wealthy, stable society with the near zero population growth that characterizes a modern industrial society. This represents a huge unmet demand. If ten billion people were to consume energy like North Americans do today, world energy demand would be ten times higher. In addition, technological advances while often improving energy efficiency tend to raise energy demand by offering more opportunity for consumption. Energy consumption still increases at close to the 2.3% per year that would lead to a tenfold increase over the course of the next century. Meeting future energy demands while phasing out fossil fuels appears extremely difficult. Instead, the world needs sustainable or nearly sustainable fossil fuels. I propose the following definition of sustainable under which fossil fuels would well qualify: The use of a technology or resource is sustainable if the intended and unintended consequences will not force its abandonment within a reasonable planning horizon. Of course sustainable technologies must not be limited by resource depletion but this is only one of many concerns. Environmental impacts, excessive land use, and other constraints can equally limit the use of a technology and thus render it unsustainable. In the foreseeable future, fossil fuels are not limited by resource depletion. However, environmental concerns based on climate change and other environmental

  8. Palm oil and derivatives: fuels or potential fuels?

    OpenAIRE

    Pioch Daniel; Vaitilingom Gilles

    2005-01-01

    Scientific and technical information including field trials about uses of palm oil as fuel has been available for more than half a century now. Several ways were investigated, from the simple mixture with petroleum Diesel fuel, to more sophisticated solutions. The quality of vegetable oils in natura as fuel is difficult to assess because of interferences between properties of the triacylglycerols – the main components – and those of the many minor components, their content varying significant...

  9. Molten Carbonate Fuel Cell Operation With Dual Fuel Flexibility

    Science.gov (United States)

    2007-10-01

    oxygen PAFC Phosphoric Acid Fuel Cell PEMFC Polymer Electrolyte Membrane Fuel Cell PDS Propane Desulfurization System ppm parts per million psig...range of power outputs. In addition , instantaneous and on-load fuel switching from natural gas to propane and back was demonstrated without loss of...issues that required additional investigation included identifying the number and volume of propane tanks needed and a vaporization sys- tem to

  10. Fuels Performance: Navigating the Intersection of Fuels and Combustion (Brochure)

    Energy Technology Data Exchange (ETDEWEB)

    2014-12-01

    Researchers at the National Renewable Energy Laboratory (NREL), the only national laboratory dedicated 100% to renewable energy and energy efficiency, recognize that engine and infrastructure compatibility can make or break the impact of even the most promising fuel. NREL and its industry partners navigate the intersection of fuel chemistry, ignition kinetics, combustion, and emissions, with innovative approaches to engines and fuels that meet drivers' expectations, while minimizing petroleum use and GHGs.

  11. 1986 fuel cell seminar: Program and abstracts

    Energy Technology Data Exchange (ETDEWEB)

    None

    1986-10-01

    Ninety nine brief papers are arranged under the following session headings: gas industry's 40 kw program, solid oxide fuel cell technology, phosphoric acid fuel cell technology, molten carbonate fuel cell technology, phosphoric acid fuel cell systems, power plants technology, fuel cell power plant designs, unconventional fuels, fuel cell application and economic assessments, and plans for commerical development. The papers are processed separately for the data base. (DLC)

  12. MICRO/NANO-STRUCTURAL EXAMINATION AND FISSION PRODUCT IDENTIFICATION IN NEUTRON IRRADIATED AGR-1 TRISO FUEL

    Energy Technology Data Exchange (ETDEWEB)

    van Rooyen, I. J.; Lillo, T. M.; Wen, H. M.; Hill, C. M.; Holesinger, T. G.; Wu, Y. Q.; Aguiara, J. A.

    2016-11-01

    Advanced microscopic and microanalysis techniques were developed and applied to study irradiation effects and fission product behavior in selected low-enriched uranium oxide/uranium carbide TRISO-coated particles from fuel compacts in six capsules irradiated to burnups of 11.2 to 19.6% FIMA. Although no TRISO coating failures were detected during the irradiation, the fraction of Ag-110m retained in individual particles often varied considerably within a single compact and at the capsule level. At the capsule level Ag-110m release fractions ranged from 1.2 to 38% and within a single compact, silver release from individual particles often spanned a range that extended from 100% retention to nearly 100% release. In this paper, selected irradiated particles from Baseline, Variant 1 and Variant 3 type fueled TRISO coated particles were examined using Scanning Electron Microscopy, Atom Probe Tomography; Electron Energy Loss Spectroscopy; Precession Electron Diffraction, Transmission Electron Microscopy, Scanning Transmission Electron Microscopy (STEM), High Resolution Electron Microscopy (HRTEM) examinations and Electron Probe Micro-Analyzer. Particle selection in this study allowed for comparison of the fission product distribution with Ag retention, fuel type and irradiation level. Nano sized Ag-containing features were predominantly identified in SiC grain boundaries and/or triple points in contrast with only two sitings of Ag inside a SiC grain in two different compacts (Baseline and Variant 3 fueled compacts). STEM and HRTEM analysis showed evidence of Ag and Pd co-existence in some cases and it was found that fission product precipitates can consist of multiple or single phases. STEM analysis also showed differences in precipitate compositions between Baseline and Variant 3 fuels. A higher density of fission product precipitate clusters were identified in the SiC layer in particles from the Variant 3 compact compared with the Variant 1 compact. Trend analysis shows

  13. Performance of AGR-1 high-temperature reactor fuel during post-irradiation heating tests

    Energy Technology Data Exchange (ETDEWEB)

    Morris, Robert N., E-mail: morrisrn@ornl.gov [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6093 (United States); Baldwin, Charles A. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6093 (United States); Demkowicz, Paul A. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415 (United States); Hunn, John D. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6093 (United States); Reber, Edward L. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415 (United States)

    2016-09-15

    Highlights: • High-temperature safety tests were performed on 14 irradiated HTGR fuel compacts. • Significant krypton release was detected in only one of the safety tests. • Cesium retention by intact SiC was excellent, even up to 1800 °C. • Release of Ag, Eu, and Sr was dominated by previous release during irradiation. • Silver exhibited the highest fractional release. - Abstract: The fission product retention of irradiated low-enriched uranium oxide/uranium carbide tri-structural isotropic (TRISO) fuel compacts from the Advanced Gas-Cooled Reactor 1 (AGR-1) experiment has been evaluated at temperatures of 1600–1800 °C during post-irradiation safety tests. Fourteen compacts (a total of ∼58,000 particles) with a burnup ranging from 13.4% to 19.1% fissions per initial metal atom (FIMA) have been tested using dedicated furnace systems at Idaho National Laboratory and Oak Ridge National Laboratory. The release of fission products {sup 110m}Ag, {sup 134}Cs, {sup 137}Cs, {sup 154}Eu, {sup 155}Eu, {sup 90}Sr, and {sup 85}Kr was monitored while heating the fuel specimens in flowing helium. The behavior of silver, europium, and strontium appears to be dominated by inventory that was originally released through intact SiC coating layers during irradiation, but was retained in the compact at the end of irradiation and subsequently released during the safety tests. However, at a test temperature of 1800 °C, the data suggest that release of these elements through intact coatings may become significant after ∼100 h. Cesium was very well retained by intact SiC layers, with a fractional release <5 × 10{sup −6} after 300 h at 1600 °C or 100 h at 1800 °C. However, it was rapidly released from individual particles if the SiC layer failed, and therefore the overall cesium release fraction was dominated by the SiC defect and failure fractions in the fuel compacts. No complete TRISO coating layer failures were observed after 300 h at 1600 or 1700 °C, and {sup

  14. Fuel Thermo-physical Characterization Project. Fiscal Year 2014 Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Burkes, Douglas [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Casella, Andrew M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Buck, Edgar C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Casella, Amanda J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Edwards, Matthew K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); MacFarlan, Paul J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pool, Karl N. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Slonecker, Bruce D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Smith, Frances N. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Steen, Franciska H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-03-15

    The Office of Material Management and Minimization (M3) Reactor Conversion Fuel Thermo-Physical Characterization Project at Pacific Northwest National Laboratory (PNNL) was tasked with using PNNL facilities and processes to receive irradiated low enriched uranium–molybdenum (LEU-Mo) fuel plate samples and perform analysis in support of the M3 Reactor Conversion Program. This work is in support of the M3 Reactor Conversion Fuel Development Pillar that is managed by Idaho National Laboratory. The primary research scope was to determine the thermo-physical properties as a function of temperature and burnup. Work conducted in Fiscal Year (FY) 2014 complemented measurements performed in FY 2013 on four additional irradiated LEU-Mo fuel plate samples. Specifically, the work in FY 2014 investigated the influence of different processing methods on thermal property behavior, the absence of aluminum alloy cladding on thermal property behavior for additional model validation, and the influence of higher operating surface heat flux / more aggressive irradiation conditions on thermal property behavior. The model developed in FY 2013 and refined in FY 2014 to extract thermal properties of the U-Mo alloy from the measurements conducted on an integral fuel plate sample (i.e., U-Mo alloy with a thin Zr coating and clad in AA6061) continues to perform very well. Measurements conducted in FY 2014 on samples irradiated under similar conditions compare well to measurements performed in FY 2013. In general, there is no gross influence of fabrication method on thermal property behavior, although the difference in LEU-Mo foil microstructure does have a noticeable influence on recrystallization of grains during irradiation. Samples irradiated under more aggressive irradiation conditions, e.g., higher surface heat flux, revealed lower thermal conductivity when compared to samples irradiated at moderate surface heat fluxes, with the exception of one sample. This report documents thermal

  15. RERTR-12 Insertion 2 Irradiation Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    D. M. Perez; G. S. Chang; D. M. Wachs; G. A. Roth; N. E. Woolstenhulme

    2012-09-01

    The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-12 was designed to provide comprehensive information on the performance of uranium-molybdenum (U-Mo) based monolithic fuels for research reactor applications.1 RERTR-12 insertion 2 includes the capsules irradiated during the last three irradiation cycles. These capsules include Z, Y1, Y2 and Y3 type capsules. The following report summarizes the life of the RERTR-12 insertion 2 experiment through end of irradiation, including as-run neutronic analysis results, thermal analysis results and hydraulic testing results.

  16. Use of alternative fuels in solid oxide fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2007-05-15

    A future sustainable energy system will certainly be based on a variety of environmentally benign energy production technologies. Fuel cells can be a key element in this scenario. One of the fuel cells types the solid oxide fuel cell (SOFC) has a number of advantages that places them in a favorable position: high efficiency, parallel production of electricity and high value heat, prevention of NOx emission, flexibility regarding usable fuels, and certain tolerance towards impurities. It is thus a natural option, to combine such a highly efficient energy conversion tool with a sustainable fuel supply. In the present contribution, the use of alternative compared to conventional fuels in SOFCs was evaluated. Regarding carbon containing, biomass derived fuels, SOFCs showed excellent power output and stability behavior during long-term testing under technologically relevant conditions. Moreover, ammonia can be used directly as fuel. The chemical and structural properties of the SOFC anode makes it even possible, to combine a chemical conversion of the fuel, for example methane into synthesis gas via steam reforming and decomposition of ammonia into hydrogen and nitrogen, with the electrochemical production of electricity in one step. (au)

  17. Fuel economy of hybrid fuel-cell vehicles

    Science.gov (United States)

    Ahluwalia, Rajesh K.; Wang, X.; Rousseau, A.

    The potential improvement in fuel economy of a mid-size fuel-cell vehicle by combining it with an energy storage system has been assessed. An energy management strategy is developed and used to operate the direct hydrogen, pressurized fuel-cell system in a load-following mode and the energy storage system in a charge-sustaining mode. The strategy places highest priority on maintaining the energy storage system in a state where it can supply unanticipated boost power when the fuel-cell system alone cannot meet the power demand. It is found that downsizing a fuel-cell system decreases its efficiency on a drive cycle which is compensated by partial regenerative capture of braking energy. On a highway cycle with limited braking energy the increase in fuel economy with hybridization is small but on the stop-and-go urban cycle the fuel economy can improve by 27%. On the combined highway and urban drive cycles the fuel economy of the fuel-cell vehicle is estimated to increase by up to 15% by hybridizing it with an energy storage system.

  18. Durability of solid oxide fuel cells using sulfur containing fuels

    DEFF Research Database (Denmark)

    Hagen, Anke; Rasmussen, Jens Foldager Bregnballe; Thydén, Karl Tor Sune

    2011-01-01

    The usability of hydrogen and also carbon containing fuels is one of the important advantages of solid oxide fuel cells (SOFCs), which opens the possibility to use fuels derived from conventional sources such as natural gas and from renewable sources such as biogas. Impurities like sulfur compounds...... are critical in this respect. State-of-the-art Ni/YSZ SOFC anodes suffer from being rather sensitive towards sulfur impurities. In the current study, anode supported SOFCs with Ni/YSZ or Ni/ScYSZ anodes were exposed to H2S in the ppm range both for short periods of 24h and for a few hundred hours. In a fuel...

  19. Fuel quality issues in stationary fuel cell systems.

    Energy Technology Data Exchange (ETDEWEB)

    Papadias, D.; Ahmed, S.; Kumar, R. (Chemical Sciences and Engineering Division)

    2012-02-07

    Fuel cell systems are being deployed in stationary applications for the generation of electricity, heat, and hydrogen. These systems use a variety of fuel cell types, ranging from the low temperature polymer electrolyte fuel cell (PEFC) to the high temperature solid oxide fuel cell (SOFC). Depending on the application and location, these systems are being designed to operate on reformate or syngas produced from various fuels that include natural gas, biogas, coal gas, etc. All of these fuels contain species that can potentially damage the fuel cell anode or other unit operations and processes that precede the fuel cell stack. These detrimental effects include loss in performance or durability, and attenuating these effects requires additional components to reduce the impurity concentrations to tolerable levels, if not eliminate the impurity entirely. These impurity management components increase the complexity of the fuel cell system, and they add to the system's capital and operating costs (such as regeneration, replacement and disposal of spent material and maintenance). This project reviewed the public domain information available on the impurities encountered in stationary fuel cell systems, and the effects of the impurities on the fuel cells. A database has been set up that classifies the impurities, especially in renewable fuels, such as landfill gas and anaerobic digester gas. It documents the known deleterious effects on fuel cells, and the maximum allowable concentrations of select impurities suggested by manufacturers and researchers. The literature review helped to identify the impurity removal strategies that are available, and their effectiveness, capacity, and cost. A generic model of a stationary fuel-cell based power plant operating on digester and landfill gas has been developed; it includes a gas processing unit, followed by a fuel cell system. The model includes the key impurity removal steps to enable predictions of impurity breakthrough

  20. Fuel economy and range estimates for fuel cell powered automobiles

    Energy Technology Data Exchange (ETDEWEB)

    Steinbugler, M.; Ogden, J. [Princeton Univ., NJ (United States)

    1996-12-31

    While a number of automotive fuel cell applications have been demonstrated, including a golf cart, buses, and a van, these systems and others that have been proposed have utilized differing configurations ranging from direct hydrogen fuel cell-only power plants to fuel cell/battery hybrids operating on reformed methanol. To date there is no clear consensus on which configuration, from among the possible combinations of fuel cell, peaking device, and fuel type, is the most likely to be successfully commercialized. System simplicity favors direct hydrogen fuel cell vehicles, but infrastructure is lacking. Infrastructure favors a system using a liquid fuel with a fuel processor, but system integration and performance issues remain. A number of studies have analyzed particular configurations on either a system or vehicle scale. The objective of this work is to estimate, within a consistent framework, fuel economies and ranges for a variety of configurations using flexible models with the goal of identifying the most promising configurations and the most important areas for further research and development.

  1. Updated NGNP Fuel Acquisition Strategy

    Energy Technology Data Exchange (ETDEWEB)

    David Petti; Tim Abram; Richard Hobbins; Jim Kendall

    2010-12-01

    A Next Generation Nuclear Plant (NGNP) fuel acquisition strategy was first established in 2007. In that report, a detailed technical assessment of potential fuel vendors for the first core of NGNP was conducted by an independent group of international experts based on input from the three major reactor vendor teams. Part of the assessment included an evaluation of the credibility of each option, along with a cost and schedule to implement each strategy compared with the schedule and throughput needs of the NGNP project. While credible options were identified based on the conditions in place at the time, many changes in the assumptions underlying the strategy and in externalities that have occurred in the interim requiring that the options be re-evaluated. This document presents an update to that strategy based on current capabilities for fuel fabrication as well as fuel performance and qualification testing worldwide. In light of the recent Pebble Bed Modular Reactor (PBMR) project closure, the Advanced Gas Reactor (AGR) fuel development and qualification program needs to support both pebble and prismatic options under the NGNP project. A number of assumptions were established that formed a context for the evaluation. Of these, the most important are: • Based on logistics associated with the on-going engineering design activities, vendor teams would start preliminary design in October 2012 and complete in May 2014. A decision on reactor type will be made following preliminary design, with the decision process assumed to be completed in January 2015. Thus, no fuel decision (pebble or prismatic) will be made in the near term. • Activities necessary for both pebble and prismatic fuel qualification will be conducted in parallel until a fuel form selection is made. As such, process development, fuel fabrication, irradiation, and testing for pebble and prismatic options should not negatively influence each other during the period prior to a decision on reactor type

  2. Fuel alcohol opportunities for Indiana

    Energy Technology Data Exchange (ETDEWEB)

    Greenglass, Bert

    1980-08-01

    Prepared at the request of US Senator Birch Bayh, Chairman of the National Alcohol Fuels Commission, this study may be best utilized as a guidebook and resource manual to foster the development of a statewide fuel alcohol plan. It examines sectors in Indiana which will impact or be impacted upon by the fuel alcohol industry. The study describes fuel alcohol technologies that could be pertinent to Indiana and also looks closely at how such a fuel alcohol industry may affect the economic and policy development of the State. Finally, the study presents options for Indiana, taking into account the national context of the developing fuel alcohol industry which, unlike many others, will be highly decentralized and more under the control of the lifeblood of our society - the agricultural community.

  3. Hybrid Fuel Cell Technology Overview

    Energy Technology Data Exchange (ETDEWEB)

    None available

    2001-05-31

    For the purpose of this STI product and unless otherwise stated, hybrid fuel cell systems are power generation systems in which a high temperature fuel cell is combined with another power generating technology. The resulting system exhibits a synergism in which the combination performs with an efficiency far greater than can be provided by either system alone. Hybrid fuel cell designs under development include fuel cell with gas turbine, fuel cell with reciprocating (piston) engine, and designs that combine different fuel cell technologies. Hybrid systems have been extensively analyzed and studied over the past five years by the Department of Energy (DOE), industry, and others. These efforts have revealed that this combination is capable of providing remarkably high efficiencies. This attribute, combined with an inherent low level of pollutant emission, suggests that hybrid systems are likely to serve as the next generation of advanced power generation systems.

  4. Towards Multi Fuel SOFC Plant

    DEFF Research Database (Denmark)

    Rokni, Masoud; Clausen, Lasse Røngaard; Bang-Møller, Christian

    2011-01-01

    Complete Solid Oxide Fuel Cell (SOFC) plants fed by several different fuels are suggested and analyzed. The plants sizes are about 10 kW which is suitable for single family house with needs for both electricity and heat. Alternative fuels such as, methanol, DME (Di-Methyl Ether) and ethanol...... are also considered and the results will be compared with the base plant fed by Natural Gas (NG). A single plant design will be suggested that can be fed with methanol, DME and ethanol whenever these fuels are available. It will be shown that the plant fed by ethanol will have slightly higher electrical...... efficiency compared with other fuels. A methanator will be suggested to be included into the plants design in order to produce methane from the fuel before entering the anode side of the SOFC stacks. Increasing methane content will decrease the needed compressor effect and thereby increase the plant power....

  5. Direct-fuelled fuel cells

    Science.gov (United States)

    Waidhas, M.; Drenckhahn, W.; Preidel, W.; Landes, H.

    Fuel supply is one important problem to be solved for commercial application of fuel cell technology. Conventional fuel-cell types require hydrogen as the fuel, which has to be free from impurities when operated at temperatures below 100 °C. The storage and distribution of this explosive and extremely fugitive gas is one of the open questions in the context of a customer-oriented broad commercial market. The direct-fuelled fuel cells (DMFCs) overcome the hydrogen specific restrictions. They are capable of directly using natural gas or fuels which are liquid under ambient conditions. In this paper the different options from direct-fuelled systems are described and their general aspects discussed. The state-of-the-art at Siemens in this field, and also the remaining technical questions are outlined as a basis for assessing future applications.

  6. Monitoring of bunker fuel consumption

    Energy Technology Data Exchange (ETDEWEB)

    Faber, J.; Nelissen, D.; Smit, M.

    2013-03-15

    Monitoring of fuel consumption and greenhouse gas emissions from international shipping is currently under discussion at the EU level as well as at the IMO (International Maritime Organization). There are several approaches to monitoring, each with different characteristics. Based on a survey of the literature and information from equipment suppliers, this report analyses the four main methods for monitoring emissions: (1) Bunker delivery notes (i.e. a note provided by the bunker fuel supplier specifying, inter alia, the amount of fuel bunkered); (2) Tank sounding (i.e. systems for measuring the amount of fuel in the fuel tanks); (3) Fuel flow meters (i.e. systems for measuring the amount of fuel supplied to the engines, generators or boilers); and (4) Direct emissions monitoring (i.e. measuring the exhaust emissions in the stack). The report finds that bunker delivery notes and tank soundings have the lowest investment cost. However, unless tank sounding is automated, these systems have higher operational costs than fuel flow meters or direct emissions monitoring because manual readings have to be entered in monitoring systems. Fuel flow meters have the highest potential accuracy. Depending on the technology selected, their accuracy can be an order of magnitude better than the other systems, which typically have errors of a few percent. By providing real-time feed-back on fuel use or emissions, fuel flow meters and direct emissions monitoring provide ship operators with the means to train their crew to adopt fuel-efficient sailing methods and to optimise their maintenance and hull cleaning schedules. Except for bunker delivery notes, all systems allow for both time-based and route-based (or otherwise geographically delineated) systems.

  7. Zeolites Remove Sulfur From Fuels

    Science.gov (United States)

    Voecks, Gerald E.; Sharma, Pramod K.

    1991-01-01

    Zeolites remove substantial amounts of sulfur compounds from diesel fuel under relatively mild conditions - atmospheric pressure below 300 degrees C. Extracts up to 60 percent of sulfur content of high-sulfur fuel. Applicable to petroleum refineries, natural-gas processors, electric powerplants, and chemical-processing plants. Method simpler and uses considerably lower pressure than current industrial method, hydro-desulfurization. Yields cleaner emissions from combustion of petroleum fuels, and protects catalysts from poisoning by sulfur.

  8. Fire Resistant Fuel Program Update

    Science.gov (United States)

    2011-09-13

    38°F Flashpoint • Objectives – developing new emulsified fuel formulations; – investigating anti-mist additives to diminish the fuel fireball ...unreacted diethanolamine. UNCLASSIFIED UNCLASSIFIED Current FRF Formulation – Mist Control Additive (MCA) – (125-250 ppm) reduces initial fireball ...size which results in reduced fireball size UNCLASSIFIED UNCLASSIFIED Current FRF Formulation – Water reduces or extinguishes pool fuel fires (10 vol

  9. Monolithic Fuel Fabrication Process Development

    Energy Technology Data Exchange (ETDEWEB)

    C. R. Clark; N. P. Hallinan; J. F. Jue; D. D. Keiser; J. M. Wight

    2006-05-01

    The pursuit of a high uranium density research reactor fuel plate has led to monolithic fuel, which possesses the greatest possible uranium density in the fuel region. Process developments in fabrication development include friction stir welding tool geometry and cooling improvements and a reduction in the length of time required to complete the transient liquid phase bonding process. Annealing effects on the microstructures of the U-10Mo foil and friction stir welded aluminum 6061 cladding are also examined.

  10. Hanford spent fuel inventory baseline

    Energy Technology Data Exchange (ETDEWEB)

    Bergsman, K.H.

    1994-07-15

    This document compiles technical data on irradiated fuel stored at the Hanford Site in support of the Hanford SNF Management Environmental Impact Statement. Fuel included is from the Defense Production Reactors (N Reactor and the single-pass reactors; B, C, D, DR, F, H, KE and KW), the Hanford Fast Flux Test Facility Reactor, the Shipping port Pressurized Water Reactor, and small amounts of miscellaneous fuel from several commercial, research, and experimental reactors.

  11. SOME ASPECTS OF FUEL CELLS

    OpenAIRE

    2012-01-01

    This work provides literature data to improve solid oxide fuel cells by a direct methane fuel cell and electrode settings of uninterrupted space. The possibility of electrochemical generators SOFC as synthesis gas from natural gas. We describe progress in the creation of new nanomaterials for components SOFC and modern technologies for their manufacture. Briefly described features of the operation and use molten carbonate fuel cells and their accessories and SOFC in cogeneration system (three...

  12. Lignite Fuel Enhancement

    Energy Technology Data Exchange (ETDEWEB)

    Charles Bullinger; Nenad Sarunac

    2010-03-31

    Pulverized coal power plants which fire lignites and other low-rank high-moisture coals generally operate with reduced efficiencies and increased stack emissions due to the impacts of high fuel moisture on stack heat loss and pulverizer and fan power. A process that uses plant waste heat sources to evaporate a portion of the fuel moisture from the lignite feedstock in a moving bed fluidized bed dryer (FBD) was developed in the U.S. by a team led by Great River Energy (GRE). The demonstration was conducted with Department of Energy (DOE) funding under DOE Award Number DE-FC26-04NT41763. The objectives of GRE's Lignite Fuel Enhancement project were to demonstrate reduction in lignite moisture content by using heat rejected from the power plant, apply technology at full scale at Coal Creek Station (CCS), and commercialize it. The Coal Creek Project has involved several stages, beginning with lignite drying tests in a laboratory-scale FBD at the Energy Research Center (ERC) and development of theoretical models for predicting dryer performance. Using results from these early stage research efforts, GRE built a 2 ton/hour pilot-scale dryer, and a 75 ton/hour prototype drying system at Coal Creek Station. Operated over a range of drying conditions, the results from the pilot-scale and prototype-scale dryers confirmed the performance of the basic dryer design concept and provided the knowledge base needed to scale the process up to commercial size. Phase 2 of the GRE's Lignite Fuel Enhancement project included design, construction and integration of a full-scale commercial coal drying system (four FBDs per unit) with Coal Creek Units 1 and 2 heat sources and coal handling system. Two series of controlled tests were conducted at Coal Creek Unit 1 with wet and dried lignite to determine effect of dried lignite on unit performance and emissions. Wet lignite was fired during the first, wet baseline, test series conducted in September 2009. The second test series was

  13. Electricity as Transportation ``Fuel''

    Science.gov (United States)

    Tamor, Michael

    2013-04-01

    The personal automobile is a surprisingly efficient device, but its place in a sustainable transportation future hinges on its ability use a sustainable fuel. While electricity is widely expected to be such a ``fuel,'' the viability of electric vehicles rests on the validity of three assumptions. First, that the emissions from generation will be significantly lower than those from competing chemical fuels whether `renewable' or fossil. Second, that advances in battery technology will deliver adequate range and durability at an affordable cost. Third, that most customers will accept any functional limitations intrinsic to electrochemical energy storage. While the first two are subjects of active research and vigorous policy debate, the third is treated virtually as a given. Popular statements to the effect that ``because 70% of all daily travel is accomplished in less than 100 miles, mass deployment of 100 mile EVs will electrify 70% of all travel'' are based on collections of one-day travel reports such as the National Household Travel Survey, and so effectively ignore the complexities of individual needs. We have analyzed the day-to-day variations of individual vehicle usage in multiple regions and draw very different conclusions. Most significant is that limited EV range results in a level of inconvenience that is likely to be unacceptable to the vast majority of vehicle owners, and for those who would accept that inconvenience, battery costs must be absurdly low to achieve any economic payback. In contrast, the plug-in hybrid (PHEV) does not suffer range limitations and delivers economic payback for most users at realistic battery costs. More importantly, these findings appear to be universal in developed nations, with labor market population density being a powerful predictor of personal vehicle usage. This ``scalable city'' hypothesis may prove to a powerful predictor of the evolution of transportation in the large cities of the developing world.

  14. Fuel Flexible, Low Emission Catalytic Combustor for Opportunity Fuel Applications

    Energy Technology Data Exchange (ETDEWEB)

    Eteman, Shahrokh

    2013-06-30

    Limited fuel resources, increasing energy demand and stringent emission regulations are drivers to evaluate process off-gases or process waste streams as fuels for power generation. Often these process waste streams have low energy content and/or highly reactive components. Operability of low energy content fuels in gas turbines leads to issues such as unstable and incomplete combustion. On the other hand, fuels containing higher-order hydrocarbons lead to flashback and auto-ignition issues. Due to above reasons, these fuels cannot be used directly without modifications or efficiency penalties in gas turbine engines. To enable the use of these wide variety of fuels in gas turbine engines a rich catalytic lean burn (RCL®) combustion system was developed and tested in a subscale high pressure (10 atm.) rig. The RCL® injector provided stability and extended turndown to low Btu fuels due to catalytic pre-reaction. Previous work has shown promise with fuels such as blast furnace gas (BFG) with LHV of 85 Btu/ft3 successfully combusted. This program extends on this work by further modifying the combustor to achieve greater catalytic stability enhancement. Fuels containing low energy content such as weak natural gas with a Lower Heating Value (LHV) of 6.5 MJ/m3 (180 Btu/ft3 to natural gas fuels containing higher hydrocarbon (e.g ethane) with LHV of 37.6 MJ/m3 (1010 Btu/ft3) were demonstrated with improved combustion stability; an extended turndown (defined as the difference between catalytic and non-catalytic lean blow out) of greater than 250oF was achieved with CO and NOx emissions lower than 5 ppm corrected to 15% O2. In addition, for highly reactive fuels the catalytic region preferentially pre-reacted the higher order hydrocarbons with no events of flashback or auto-ignition allowing a stable and safe operation with low NOx and CO emissions.

  15. Materials for high-temperature fuel cells

    CERN Document Server

    Jiang, San Ping; Lu, Max

    2013-01-01

    There are a large number of books available on fuel cells; however, the majority are on specific types of fuel cells such as solid oxide fuel cells, proton exchange membrane fuel cells, or on specific technical aspects of fuel cells, e.g., the system or stack engineering. Thus, there is a need for a book focused on materials requirements in fuel cells. Key Materials in High-Temperature Fuel Cells is a concise source of the most important and key materials and catalysts in high-temperature fuel cells with emphasis on the most important solid oxide fuel cells. A related book will cover key mater

  16. Aviation turbine fuels: An assessment of alternatives

    Science.gov (United States)

    1982-01-01

    The general outlook for aviation turbine fuels, the effect that broadening permissible aviation turbine fuel properties could have on the overall availability of such fuels, the fuel properties most likely to be affected by use of lower grade petroleum crudes, and the research and technology required to ensure that aviation turbine fuels and engines can function satisfactorily with fuels having a range of fuel properties differing from those of current specification fuel are assessed. Views of industry representatives on alternative aviation turbine fuels are presented.

  17. Materials for low-temperature fuel cells

    CERN Document Server

    Ladewig, Bradley; Yan, Yushan; Lu, Max

    2014-01-01

    There are a large number of books available on fuel cells; however, the majority are on specific types of fuel cells such as solid oxide fuel cells, proton exchange membrane fuel cells, or on specific technical aspects of fuel cells, e.g., the system or stack engineering. Thus, there is a need for a book focused on materials requirements in fuel cells. Key Materials in Low-Temperature Fuel Cells is a concise source of the most important and key materials and catalysts in low-temperature fuel cells. A related book will cover key materials in high-temperature fuel cells. The two books form part

  18. Measurements in liquid fuel sprays

    Science.gov (United States)

    Chigier, N.

    1984-01-01

    Techniques for studying the events directly preceding combustion in the liquid fuel sprays are being used to provide information as a function of space and time on droplet size, shape, number density, position, angle of flight and velocity. Spray chambers were designed and constructed for: (1) air-assist liquid fuel research sprays; (2) high pressure and temperature chamber for pulsed diesel fuel sprays; and (3) coal-water slurry sprays. Recent results utilizing photography, cinematography, and calibration of the Malvern particle sizer are reported. Systems for simultaneous measurement of velocity and particle size distributions using laser Doppler anemometry interferometry and the application of holography in liquid fuel sprays are being calibrated.

  19. Fuel Cell Powered Lift Truck

    Energy Technology Data Exchange (ETDEWEB)

    Moulden, Steve [Sysco Food Service, Houston, TX (United States)

    2015-08-20

    This project, entitled “Recovery Act: Fuel Cell-Powered Lift Truck Sysco (Houston) Fleet Deployment”, was in response to DOE funding opportunity announcement DE-PS36-08GO98009, Topic 7B, which promotes the deployment of fuel cell powered material handling equipment in large, multi-shift distribution centers. This project promoted large-volume commercialdeployments and helped to create a market pull for material handling equipment (MHE) powered fuel cell systems. Specific outcomes and benefits involved the proliferation of fuel cell systems in 5-to 20-kW lift trucks at a high-profile, real-world site that demonstrated the benefits of fuel cell technology and served as a focal point for other nascent customers. The project allowed for the creation of expertise in providing service and support for MHE fuel cell powered systems, growth of existing product manufacturing expertise, and promoted existing fuel cell system and component companies. The project also stimulated other MHE fleet conversions helping to speed the adoption of fuel cell systems and hydrogen fueling technology. This document also contains the lessons learned during the project in order to communicate the successes and difficulties experienced, which could potentially assist others planning similar projects.

  20. Fuel cells problems and solutions

    CERN Document Server

    Bagotsky, Vladimir S

    2012-01-01

    The comprehensive, accessible introduction to fuel cells, their applications, and the challenges they pose Fuel cells-electrochemical energy devices that produce electricity and heat-present a significant opportunity for cleaner, easier, and more practical energy. However, the excitement over fuel cells within the research community has led to such rapid innovation and development that it can be difficult for those not intimately familiar with the science involved to figure out exactly how this new technology can be used. Fuel Cells: Problems and Solutions, Second Edition addresses this i

  1. Special Issue: Aviation Alternative Fuels

    Directory of Open Access Journals (Sweden)

    Yang Zhang

    2014-12-01

    Full Text Available The investigation of aviation alternative fuels has increased significantly in recent years in an effort to reduce the environment and climate impact by aviation industry. Special requirements have to be met for qualifying as a suitable aviation fuel. The fuel has to be high in energy content per unit of mass and volume, thermally stable and avoiding freezing at low temperatures. There are also many other special requirements on viscosity, ignition properties and compatibility with the typical aviation materials. There are quite a few contending alternative fuels which can be derived from coal, natural gas and biomass.[...

  2. Seismic behaviour of fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Song, Heuy Gap; Jhung, Myung Jo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1993-11-01

    A general approach for the dynamic time-history analysis of the reactor core is presented in this paper as a part of the fuel assembly qualification program. Several detailed core models are set up to reflect the placement of the fuel assemblies within the core shroud. Peak horizontal responses are obtained for each model for the motions induced from earthquake. The dynamic responses such as fuel assembly shear force, bending moment and displacement, and spacer grid impact loads are carefully investigated. Also, the sensitivity responses are obtained for the earthquake motions and the fuel assembly non-linear response characteristics are discussed. (Author) 9 refs., 24 figs., 1 tab.

  3. Fuel analyzer; Analisador de combustiveis

    Energy Technology Data Exchange (ETDEWEB)

    Cozzolino, Roberval [RS Motors, Indaiatuba, SP (Brazil)

    2008-07-01

    The current technology 'COMBUSTIMETRO' aims to examine the fuel through performance of the engine, as the role of the fuel is to produce energy for the combustion engine in the form of which is directly proportional to the quality and type of fuel. The 'COMBUSTIMETRO' has an engine that always keeps the same entry of air, fuel and fixed point of ignition. His operation is monitored by sensors (Sonda Lambda, RPM and Gases Analyzer) connected to a processor that performs calculations and records the information, generate reports and graphs. (author)

  4. Used Fuel Testing Transportation Model

    Energy Technology Data Exchange (ETDEWEB)

    Ross, Steven B. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Best, Ralph E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Maheras, Steven J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jensen, Philip J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); England, Jeffery L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); LeDuc, Dan [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2014-09-25

    This report identifies shipping packages/casks that might be used by the Used Nuclear Fuel Disposition Campaign Program (UFDC) to ship fuel rods and pieces of fuel rods taken from high-burnup used nuclear fuel (UNF) assemblies to and between research facilities for purposes of evaluation and testing. Also identified are the actions that would need to be taken, if any, to obtain U.S. Nuclear Regulatory (NRC) or other regulatory authority approval to use each of the packages and/or shipping casks for this purpose.

  5. Ecodesign of Liquid Fuel Tanks

    Science.gov (United States)

    Gicevska, Jana; Bazbauers, Gatis; Repele, Mara

    2011-01-01

    The subject of the study is a 10 litre liquid fuel tank made of metal and used for fuel storage and transportation. The study dealt with separate life cycle stages of this product, compared environmental impacts of similar fuel tanks made of metal and plastic, as well as analysed the product's end-of-life cycle stage, studying the waste treatment and disposal scenarios. The aim of this study was to find opportunities for improvement and to develop proposals for the ecodesign of 10 litre liquid fuel tank.

  6. Hydrogen Generation Via Fuel Reforming

    Science.gov (United States)

    Krebs, John F.

    2003-07-01

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

  7. Used Fuel Testing Transportation Model

    Energy Technology Data Exchange (ETDEWEB)

    Ross, Steven B.; Best, Ralph E.; Maheras, Steven J.; Jensen, Philip J.; England, Jeffery L.; LeDuc, Dan

    2014-09-24

    This report identifies shipping packages/casks that might be used by the Used Nuclear Fuel Disposition Campaign Program (UFDC) to ship fuel rods and pieces of fuel rods taken from high-burnup used nuclear fuel (UNF) assemblies to and between research facilities for purposes of evaluation and testing. Also identified are the actions that would need to be taken, if any, to obtain U.S. Nuclear Regulatory (NRC) or other regulatory authority approval to use each of the packages and/or shipping casks for this purpose.

  8. Fuel reprocessing tank

    Energy Technology Data Exchange (ETDEWEB)

    Gonda, Sumitora

    1998-10-09

    A tank of the present invention for spent fuels comprises a stainless steel tank main body for storing a highly corrosive dissolving solution, a steam jet pump disposed to the inside of the tank main body for transferring the dissolving solution to the outside of the tank main body and pipelines connecting them. With such a constitution, abnormal abrasion and drag of mechanical parts are less caused. In addition, a cleaning nozzle and a cleaning liquid pipeline which eliminates clogging of a sucking port of the steam jet pump if clogging is caused by sludges are disposed thereby enabling to avoid possibility of clogging. (T.M.)

  9. Fuels for Thought!

    Directory of Open Access Journals (Sweden)

    Clifford Louime

    2009-07-01

    Full Text Available When it comes to the marketing of the bioenergy brand, one of the catchiest slogans out these days is “25 by ‘25”. Adopted and supported by industries, academia and government agencies alike, this organization simply aims to supply 25 percent of our energy from renewable resources by the year 2025. By focusing its future efforts on wind, solar and biomass resources, the “25 by 25” initiative is expected to create new jobs, develop novel technologies, help mitigate the effects of global warming and reduce our dependence on fossil fuels. [...

  10. Biodesulfurization of fossil fuels.

    Science.gov (United States)

    Gray, Kevin A; Mrachko, Gregory T; Squires, Charles H

    2003-06-01

    Biotechnological techniques enabling the specific removal of sulfur from fossil fuels have been developed. In the past three years there have been important advances in the elucidation of the mechanisms of biodesulfurization; some of the most significant relate to the role of a flavin reductase, DszD, in the enzymology of desulfurization, and to the use of new tools that enable enzyme enhancement via DNA manipulation to influence both the rate and the substrate range of Dsz. Also, a clearer understanding of the unique desulfinase step in the pathway has begun to emerge.

  11. Producing a solid fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kimura, K.; Nisino, A.; Sonedaka, V.; Takeuti, Y.

    1982-08-21

    Coal, semicoke, coke, graphite and other hydrocarbon fuels are formed into cellular briquets with the addition as a binder of 3 percent by weight of alumina or Portland cement, which operates as a sulfur removing agent. To provide for high thermal resistance and resistance to oxidation, cement which consists of 30 to 60 percent lime, 40 to 60 percent alumina and 2 to 10 percent iron oxide is recommended. Ca(OH)/sub 2/, CaCO/sub 3/, dolomite, Na/sub 2/CO/sub 3/, MgO and ZnO are added to promote the removal of the sulfur.

  12. 14 CFR 25.991 - Fuel pumps.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Fuel pumps. 25.991 Section 25.991... STANDARDS: TRANSPORT CATEGORY AIRPLANES Powerplant Fuel System Components § 25.991 Fuel pumps. (a) Main pumps. Each fuel pump required for proper engine operation, or required to meet the fuel...

  13. 49 CFR 238.423 - Fuel tanks.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 4 2010-10-01 2010-10-01 false Fuel tanks. 238.423 Section 238.423 Transportation....423 Fuel tanks. (a) External fuel tanks. Each type of external fuel tank must be approved by FRA's Associate Administrator for Safety upon a showing that the fuel tank provides a level of safety at...

  14. China Gradually Deregulates Aviation Fuels Market

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

    @@ China will gradually deregulate the aviation fuels market to allow the oil and petrochemical enterprises to become shareholders of China Aviation Fuels Corporation (CAFC) so that the aviation fuels suppliers can operate at a lower cost. Deregulation of the air fuels market aims at reduction of aviation fuels price to spur development of China's air transportation industry.

  15. The DUPIC fuel development program in KAERI

    Energy Technology Data Exchange (ETDEWEB)

    Yang, M. S.; Park, H. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-07-01

    This study describes the DUPIC fuel development program in KAERI as follows; Burning spent PWR fuel again in CANDU by DUPIC, Compatibility with existing CANDU system, Feasibility of DUPIC fuel fabrication, Waste reduction, Safeguard ability, Economics of DUPIC fuel cycle, The DUPIC fuel development program, and International prospective. 5 refs., 10 figs.

  16. 46 CFR 28.335 - Fuel systems.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 1 2010-10-01 2010-10-01 false Fuel systems. 28.335 Section 28.335 Shipping COAST GUARD... Than 16 Individuals on Board § 28.335 Fuel systems. (a) Applicability. Except for the components of an...) Portable fuel systems. Portable fuel systems including portable tanks and related fuel lines and...

  17. Microstructural evolution of a uranium-10 wt.% molybdenum alloy for nuclear reactor fuels

    Science.gov (United States)

    Clarke, A. J.; Clarke, K. D.; McCabe, R. J.; Necker, C. T.; Papin, P. A.; Field, R. D.; Kelly, A. M.; Tucker, T. J.; Forsyth, R. T.; Dickerson, P. O.; Foley, J. C.; Swenson, H.; Aikin, R. M.; Dombrowski, D. E.

    2015-10-01

    Low-enriched uranium-10 wt.% molybdenum (LEU-10wt.%Mo) is of interest for the fabrication of monolithic fuels to replace highly-enriched uranium (HEU) dispersion fuels in high performance research and test reactors around the world. In this work, depleted uranium-10 wt.%Mo (DU-10wt.%Mo) is used to simulate the solidification and microstructural evolution of LEU-10wt.%Mo. Electron backscatter diffraction (EBSD) and complementary electron probe microanalysis (EPMA) reveal significant microsegregation present in the metastable γ-phase after solidification. Homogenization is performed at 800 and 1000 °C for times ranging from 1 to 32 h to explore the time-temperature combinations that will reduce the extent of microsegregation, as regions of higher and lower Mo content may influence local mechanical properties and provide preferred regions for γ-phase decomposition. We show for the first time that EBSD can be used to qualitatively assess microstructural evolution in DU-10wt.%Mo after homogenization treatments. Complementary EPMA is used to quantitatively confirm this finding. Homogenization at 1000 °C for 2-4 h may the regions that contain 8 wt.% Mo or lower, whereas homogenization at 1000 °C for longer than 8 h effectively saturates Mo chemical homogeneity, but results in substantial grain growth. The appropriate homogenization time will depend upon additional microstructural considerations, such as grain growth and intended subsequent processing. Higher carbon LEU-10wt.%Mo generally contains more inclusions within the grains and at grain boundaries after solidification. The effect of these inclusions on microstructural evolution (e.g. grain growth) during homogenization and as potential γ-phase decomposition nucleation sites is unclear, but likely requires additional study.

  18. Solid Oxide Fuel Cells Operating on Alternative and Renewable Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Xiaoxing; Quan, Wenying; Xiao, Jing; Peduzzi, Emanuela; Fujii, Mamoru; Sun, Funxia; Shalaby, Cigdem; Li, Yan; Xie, Chao; Ma, Xiaoliang; Johnson, David; Lee, Jeong; Fedkin, Mark; LaBarbera, Mark; Das, Debanjan; Thompson, David; Lvov, Serguei; Song, Chunshan

    2014-09-30

    This DOE project at the Pennsylvania State University (Penn State) initially involved Siemens Energy, Inc. to (1) develop new fuel processing approaches for using selected alternative and renewable fuels – anaerobic digester gas (ADG) and commercial diesel fuel (with 15 ppm sulfur) – in solid oxide fuel cell (SOFC) power generation systems; and (2) conduct integrated fuel processor – SOFC system tests to evaluate the performance of the fuel processors and overall systems. Siemens Energy Inc. was to provide SOFC system to Penn State for testing. The Siemens work was carried out at Siemens Energy Inc. in Pittsburgh, PA. The unexpected restructuring in Siemens organization, however, led to the elimination of the Siemens Stationary Fuel Cell Division within the company. Unfortunately, this led to the Siemens subcontract with Penn State ending on September 23rd, 2010. SOFC system was never delivered to Penn State. With the assistance of NETL project manager, the Penn State team has since developed a collaborative research with Delphi as the new subcontractor and this work involved the testing of a stack of planar solid oxide fuel cells from Delphi.

  19. Response of a direct methanol fuel cell to fuel change

    Energy Technology Data Exchange (ETDEWEB)

    Leo, T.J. [Dpto de Sistemas Oceanicos y Navales- ETSI Navales, Univ. Politecnica de Madrid, Avda Arco de la Victoria s/n, 28040 Madrid (Spain); Raso, M.A.; de la Blanca, E. Sanchez [Dpto de Quimica Fisica I- Fac. CC. Quimicas, Univ. Complutense de Madrid, Avda Complutense s/n, 28040 Madrid (Spain); Navarro, E.; Villanueva, M. [Dpto de Motopropulsion y Termofluidodinamica, ETSI Aeronauticos, Univ. Politecnica de Madrid, Pza Cardenal Cisneros 3, 28040 Madrid (Spain); Moreno, B. [Instituto de Ceramica y Vidrio, Consejo Superior de Investigaciones Cientificas, C/Kelsen 5, Campus de la UAM, 28049 Cantoblanco, Madrid (Spain)

    2010-10-15

    Methanol and ethanol have recently received much attention as liquid fuels particularly as alternative 'energy-vectors' for the future. In this sense, to find a direct alcohol fuel cell that able to interchange the fuel without losing performances in an appreciable way would represent an evident advantage in the field of portable applications. In this work, the response of a in-house direct methanol fuel cell (DMFC) to the change of fuel from methanol to ethanol and its behaviour at different ambient temperature values have been investigated. A corrosion study on materials suitable to fabricate the bipolar plates has been carried out and either 316- or 2205-duplex stainless steels have proved to be adequate for using in direct alcohol fuel cells. Polarization curves have been measured at different ambient temperature values, controlled by an experimental setup devised for this purpose. Data have been fitted to a model taking into account the temperature effect. For both fuels, methanol and ethanol, a linear dependence of adjustable parameters with temperature is obtained. Fuel cell performance comparison in terms of open circuit voltage, kinetic and resistance is established. (author)

  20. Tanker Fuel Consolidation: Impact of Fuel Efficiency on ATO Resiliency

    Science.gov (United States)

    2011-06-01

    Count 23 101 Appendix E. Blue Dart Major Scott Linck, Student, ASAM DSN 650-7752, scott.linck@us.af.mil word count: 748 Tanker Fuel...Mobility ( ASAM ) Class. His next assignment is to Stuttgart, Germany where he will serve as the USTC LNO to AFRICOM.” Keywords: fuel consolidation