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Sample records for generation fuel irradiation

  1. Gas-Cooled Reactor Programs annual progress report for period ending December 31, 1973. [HTGR fuel reprocessing, fuel fabrication, fuel irradiation, core materials, and fission product distribution; GCFR fuel irradiation and steam generator modeling

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Coobs, J.H.; Lotts, A.L.

    1976-04-01

    Progress is summarized in studies relating to HTGR fuel reprocessing, refabrication, and recycle; HTGR fuel materials development and performance testing; HTGR PCRV development; HTGR materials investigations; HTGR fuel chemistry; HTGR safety studies; and GCFR irradiation experiments and steam generator modeling.

  2. Next generation fuel irradiation capability in the High Flux Reactor Petten

    Energy Technology Data Exchange (ETDEWEB)

    Fuetterer, Michael A., E-mail: michael.fuetterer@jrc.n [European Commission, Joint Research Centre, Institute for Energy (JRC-IE), P.O. Box 2, NL-1755 ZG Petten (Netherlands); D' Agata, Elio; Laurie, Mathias; Marmier, Alain; Scaffidi-Argentina, Francesco [European Commission, Joint Research Centre, Institute for Energy (JRC-IE), P.O. Box 2, NL-1755 ZG Petten (Netherlands); Raison, Philippe [European Commission, Joint Research Centre, Institute for Transuranium Elements (JRC-ITU), D-76334 Eggenstein-Leopoldshafen (Germany); Bakker, Klaas; Groot, Sander de; Klaassen, Frodo [Nuclear Research and consultancy Group (NRG), P.O. Box 25, NL-1755 ZG Petten (Netherlands)

    2009-07-15

    This paper describes selected equipment and expertise on fuel irradiation testing at the High Flux Reactor (HFR) in Petten, The Netherlands. The reactor went critical in 1961 and holds an operating license up to at least 2015. While HFR has initially focused on Light Water Reactor fuel and materials, it also played a decisive role since the 1970s in the German High Temperature Reactor (HTR) development program. A variety of tests related to fast reactor development in Europe were carried out for next generation fuel and materials, in particular for Very High Temperature Reactor (V/HTR) fuel, fuel for closed fuel cycles (U-Pu and Th-U fuel cycle) and transmutation, as well as for other innovative fuel types. The HFR constitutes a significant European infrastructure tool for the development of next generation reactors. Experimental facilities addressed include V/HTR fuel tests, a coated particle irradiation rig, and tests on fast reactor, transmutation and thorium fuel. The rationales for these tests are given, results are provided and further work is outlined.

  3. Post irradiation test report of irradiated DUPIC simulated fuel

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Myung Seung; Jung, I. H.; Moon, J. S. and others

    2001-12-01

    The post-irradiation examination of irradiated DUPIC (Direct Use of Spent PWR Fuel in CANDU Reactors) simulated fuel in HANARO was performed at IMEF (Irradiated Material Examination Facility) in KAERI during 6 months from October 1999 to March 2000. The objectives of this post-irradiation test are i) the integrity of the capsule to be used for DUPIC fuel, ii) ensuring the irradiation requirements of DUPIC fuel at HANARO, iii) performance verification in-core behavior at HANARO of DUPIC simulated fuel, iv) establishing and improvement the data base for DUPIC fuel performance verification codes, and v) establishing the irradiation procedure in HANARO for DUPIC fuel. The post-irradiation examination performed are {gamma}-scanning, profilometry, density, hardness, observation the microstructure and fission product distribution by optical microscope and electron probe microanalyser (EPMA)

  4. PLUTON: Three-group neutronic code for burnup analysis of isotope generation and depletion in highly irradiated LWR fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Lemehov, Sergei E; Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-08-01

    PLUTON is a three-group neutronic code analyzing, as functions of time and burnup, the change of radial profiles, together with average values, of power density, burnup, concentration of trans-uranium elements, plutonium buildup, depletion of fissile elements, and fission product generation in water reactor fuel rod with standard UO{sub 2}, UO{sub 2}-Gd{sub 2}O{sub 3}, inhomogeneous MOX, and UO{sub 2}-ThO{sub 2}. The PLUTON code, which has been designed to be run on Windows PC, has adopted a theoretical shape function of neutron attenuation in pellet, which enables users to perform a very fast and accurate calculation easily. The present code includes the irradiation conditions of the Halden Reactor which gives verification data for the code. The total list of trans-uranium elements included in the calculations consists of {sub 92}U{sup 233-239}, {sub 93}Np{sup 237-239}, {sub 94}Pu{sup 238-243}, {sub 95}Am{sup 241-244} (including isomers), and {sub 96}Cm{sup 242-245}. Poisoning fission products are represented by {sub 54}Xe{sup 131,133,135}, {sub 48}Cd{sup 113}, {sub 62}Sm{sup 149,151,152}, {sub 64}Gd{sup 154-160}, {sub 63}Eu{sup 153,155}, {sub 36}Kr{sup 83,85}, {sub 42}Mo{sup 95}, {sub 43}Tc{sup 99}, {sub 45}Rh{sup 103}, {sub 47}Ag{sup 109}, {sub 53}I{sup 127,129,131}, {sub 55}Cs{sup 133}, {sub 57}La{sup 139}, {sub 59}Pr{sup 141}, {sub 60}Nd{sup 143-150}, {sub 61}Pm{sup 147}. Fission gases and volatiles included in the code are {sub 36}Kr{sup 83-86}, {sub 54}Xe{sup 129-136}, {sub 52}Te{sup 125-130}, {sub 53}I{sup 127-131}, {sub 55}Cs{sup 133-137}, and {sub 56}Ba{sup 135-140}. Verification has been performed up to 83 GWd/tU, and a satisfactory agreement has been obtained. (author)

  5. 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.

  6. 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.

  7. 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.

  8. Gamma-ray spectroscopy on irradiated fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Terremoto, Luis Antonio Albiac [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear], e-mail: laaterre@ipen.br

    2009-07-01

    The recording of gamma-ray spectra along an irradiated fuel rod allows the fission products to be qualitatively and quantitatively examined. Among all nondestructive examinations performed on irradiated fuel rods by gamma-ray spectroscopy, the most comprehensive one is the average burnup measurement, which is quantitative. Moreover, burnup measurements by means of gamma-ray spectroscopy are less time-consuming and waste-generating than burnup measurements by radiochemical, destructive methods. This work presents the theoretical foundations and experimental techniques necessary to measure, using nondestructive gamma-ray spectroscopy, the average burnup of irradiated fuel rods in a laboratory equipped with hot cells. (author)

  9. Post irradiation examination of thermal reactor fuels

    Science.gov (United States)

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

    2008-12-01

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

  10. Irradiation and performance evaluation of DUPIC fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Ki Kwang; Yang, M. S.; Song, K. C. [and others

    2000-05-01

    The objectives of the project is to establish the performance evaluation system for the experimental verification of DUPIC fuel. The scope and content for successful accomplishment of the phase 1 objectives is established as follows : irradiation test of DUPIC fuel at HANARO using a noninstrument capsule, study on the characteristics of DUPIC pellets, development of the analysis technology on the thermal behaviour of DUPIC fuel, basic design of a instrument capsule. The R and D results of the phase 1 are summarized as follows : - Performance analysis technology development of DUPIC fuel by model development for DUPIC fuel, review on the extendability of code(FEMAXI-IV, FRAPCON-3, ELESTRESS). - Study on physical properties of DUPIC fuel by design and fabrication of the equipment for measuring the thermal property. - HANARO irradiation test of simulated DUPIC fuel by the noninstrument capsule development. - PIE and result analysis.

  11. Thermal Analysis of KAERI TRISO Fuel Irradiation at HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Moon-Sung; Kim, B. G.; Yang, S. W. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The TRISO(Tri-structural Isotropic)-coated fuel particle for a VHTR has a diameter of around 1 mm, and is composed of a nuclear fuel kernel and four different outer coating layers. These coating layers consist of a buffer PyC (pyrolytic carbon) layer, an inner PyC layer, a SiC layer, and an outer PyC layer. The fuel kernel is a source of a heat generation by the nuclear fission of fissile uranium. The role of each of the four coating layers is different in view of retaining the generated fission products and other interactions during in-reactor service. KAERI has been developing a TRISO-coated particle fuel technology as a part of the Korean VHTR (Very High Temperature modular gas cooled Reactor) project, which started in 2004, and completed its first irradiation test of TRISO fuels in its research reactor, HANARO for an evaluation and prediction of the irradiation behavior of the fuel. The test was started in August 4, 2013 and finished in March 31, 2014 completing its 5 cycle irradiation of 132.2 EFPD. In this paper, thermal performance of TRISO fuels was evaluated for its five cycle irradiation at HANARO which had been carried out in the absence of the fuel temperature monitoring. A COMSOL based FE (finite element) model was utilized in this analysis. Thermal performance of TRISO fuels was evaluated for its five cycle irradiation at HANARO which had been carried out in the absence of the fuel temperature monitoring. A maximum peak temperature of 1,083 .deg. C was obtained in the rod 1 at 25.06 EFPD and the temperatures decreased as the cycle progresses.

  12. Irradiation behavior of metallic fast reactor fuels

    Energy Technology Data Exchange (ETDEWEB)

    Pahl, R.G.; Porter, D.L.; Crawford, D.C.; Walters, L.C.

    1991-01-01

    Metallic fuels were the first fuels chosen for liquid metal cooled fast reactors (LMR's). In the late 1960's world-wide interest turned toward ceramic LMR fuels before the full potential of metallic fuel was realized. However, during the 1970's the performance limitations of metallic fuel were resolved in order to achieve a high plant factor at the Argonne National Laboratory's Experimental Breeder Reactor II. The 1980's spawned renewed interest in metallic fuel when the Integral Fast Reactor (IFR) concept emerged at Argonne National Laboratory. A fuel performance demonstration program was put into place to obtain the data needed for the eventual licensing of metallic fuel. This paper will summarize the results of the irradiation program carried out since 1985.

  13. Post-irradiation examination and R and D programs using irradiated fuels at KAERI

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Yong Bum; Min, Duck Kee; Kim, Eun Ka and others

    2000-12-01

    This report describes the Post-Irradiation Examination(PIE) and R and D programs using irradiated fuels at KAERI. The objectives of post-irradiation examination (PIE) for the PWR irradiated fuels, CANDU fuels, HANARO fuels and test fuel materials are to verify the irradiation performance and their integrity as well as to construct a fuel performance data base. The comprehensive utilization program of the KAERI's post-irradiation examination related nuclear facilities such as Post-Irradiation Examination Facility (PIEF), Irradiated Materials Examination Facility (IMEF) and HANARO is described.

  14. Tensile Hoop Behavior of Irradiated Zircaloy-4 Nuclear Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Jaramillo, Roger A [ORNL; Hendrich, WILLIAM R [ORNL; Packan, Nicolas H [ORNL

    2007-03-01

    A method for evaluating the room temperature ductility behavior of irradiated Zircaloy-4 nuclear fuel cladding has been developed and applied to evaluate tensile hoop strength of material irradiated to different levels. The test utilizes a polyurethane plug fitted within a tubular cladding specimen. A cylindrical punch is used to compress the plug axially, which generates a radial displacement that acts upon the inner diameter of the specimen. Position sensors track the radial displacement of the specimen outer diameter as the compression proceeds. These measurements coupled with ram force data provide a load-displacement characterization of the cladding response to internal pressurization. The development of this simple, cost-effective, highly reproducible test for evaluating tensile hoop strain as a function of internal pressure for irradiated specimens represents a significant advance in the mechanical characterization of irradiated cladding. In this project, nuclear fuel rod assemblies using Zircaloy-4 cladding and two types of mixed uranium-plutonium oxide (MOX) fuel pellets were irradiated to varying levels of burnup. Fuel pellets were manufactured with and without thermally induced gallium removal (TIGR) processing. Fuel pellets manufactured by both methods were contained in fuel rod assemblies and irradiated to burnup levels of 9, 21, 30, 40, and 50 GWd/MT. These levels of fuel burnup correspond to fast (E > 1 MeV) fluences of 0.27, 0.68, 0.98, 1.4 and 1.7 1021 neutrons/cm2, respectively. Following irradiation, fuel rod assemblies were disassembled; fuel pellets were removed from the cladding; and the inner diameter of cladding was cleaned to remove residue materials. Tensile hoop strength of this cladding material was tested using the newly developed method. Unirradiated Zircaloy-4 cladding was also tested. With the goal of determining the effect of the two fuel types and different neutron fluences on clad ductility, tensile hoop strength tests were

  15. Fuel Accident Condition Simulator (FACS) Furnace for Post-Irradiation Heating Tests of VHTR Fuel Compacts

    Energy Technology Data Exchange (ETDEWEB)

    Paul A Demkowicz; Paul Demkowicz; David V Laug

    2010-10-01

    Abstract –Fuel irradiation testing and post-irradiation examination are currently in progress as part of the Next Generation Nuclear Plant Fuels Development and Qualification Program. The PIE campaign will include extensive accident testing of irradiated very high temperature reactor fuel compacts to verify fission product retention characteristics at high temperatures. This work will be carried out at both the Idaho National Laboratory (INL) and the Oak Ridge National Laboratory, beginning with accident tests on irradiated fuel from the AGR-1 experiment in 2010. A new furnace system has been designed, built, and tested at INL to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000°C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, Eu, and I) and fission gases (Kr, Xe). Fission gases released from the fuel to the sweep gas are monitored in real time using dual cryogenic traps fitted with high purity germanium detectors. Condensable fission products are collected on a plate attached to a water-cooled cold finger that can be exchanged periodically without interrupting the test. Analysis of fission products on the condensation plates involves dry gamma counting followed by chemical analysis of selected isotopes. This paper will describe design and operational details of the Fuel Accident Condition Simulator (FACS) furnace system, as well as preliminary system calibration results.

  16. Advanced Reactor Fuels Irradiation Experiment Design Objectives

    Energy Technology Data Exchange (ETDEWEB)

    Chichester, Heather Jean MacLean [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hayes, Steven Lowe [Idaho National Lab. (INL), Idaho Falls, ID (United States); Dempsey, Douglas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    This report summarizes the objectives of the current irradiation testing activities being undertaken by the Advanced Fuels Campaign relative to supporting the development and demonstration of innovative design features for metallic fuels in order to realize reliable performance to ultra-high burnups. The AFC-3 and AFC-4 test series are nearing completion; the experiments in this test series that have been completed or are in progress are reviewed and the objectives and test matrices for the final experiments in these two series are defined. The objectives, testing strategy, and test parameters associated with a future AFC test series, AFC-5, are documented. Finally, the future intersections and/or synergies of the AFC irradiation testing program with those of the TREAT transient testing program, emerging needs of proposed Versatile Test Reactor concepts, and the Joint Fuel Cycle Study program’s Integrated Recycle Test are discussed.

  17. Pulse irradiation tests of rock-like oxide fuel

    Science.gov (United States)

    Okonogi, K.; Nakamura, T.; Yoshinaga, M.; Ishijima, K.; Akie, H.; Takano, H.

    1999-08-01

    Pulse irradiation tests of special oxide fuel designed for plutonium disposal, called rock-like oxide (ROX), have been conducted in the Nuclear Safety Research Reactor (NSRR) to investigate the transient behavior of ROX fuel under reactivity initiated accident (RIA) conditions. An uranium free ROX, (Zr,Y)O 2-MgAl 2O 4-PuO 2, is proposed for once-through use of Pu in light water reactors. However, because of smaller negative Doppler and void reactivity coefficients in the ROX fuel, higher peak fuel enthalpies are expected under RIAs than for UO 2 fuel. Thus, the tests of simulated ROX, in which Pu was replaced by U for easier realization, were conducted to a peak fuel enthalpy of 0.96 kJ g -1 (230 cal g -1), which is above current Japanese safety limits for UO 2. The transient behavior of the simulated ROX fuel was quite different from that of UO 2, because of its different thermo-physical properties. Fuel failure was associated with fuel melting at peak fuel enthalpies of 1.63 kJ g -1 (390 cal g -1) to 2.22 kJ g -1 (530 cal g -1). Significant mechanical energy generation, the reason for the limit, however, was not observed.

  18. Nanowire mesh solar fuels generator

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Peidong; Chan, Candace; Sun, Jianwei; Liu, Bin

    2016-05-24

    This disclosure provides systems, methods, and apparatus related to a nanowire mesh solar fuels generator. In one aspect, a nanowire mesh solar fuels generator includes (1) a photoanode configured to perform water oxidation and (2) a photocathode configured to perform water reduction. The photocathode is in electrical contact with the photoanode. The photoanode may include a high surface area network of photoanode nanowires. The photocathode may include a high surface area network of photocathode nanowires. In some embodiments, the nanowire mesh solar fuels generator may include an ion conductive polymer infiltrating the photoanode and the photocathode in the region where the photocathode is in electrical contact with the photoanode.

  19. 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.

  20. The 3rd irradiation test plan of DUPIC fuel

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Myung Seung; Song, K. C.; Park, J. H. and others

    2001-05-01

    The objective of the 3rd irradiation test of DUPIC fuel at the HANARO is to estimate the in-core behaviour of a DUPIC pellet that is irradiated up to more than average burnup of CANDU fuel. The irradiation of DUPIC fuel is planned to start at May 21, 2001, and will be continued at least for 8 months. The burnup of DUPIC fuel through this irradiation test is thought to be more than 7,000 MWd/tHE. The DUPIC irradiation rig instrumented with three SPN detectors will be used to accumulate the experience for the instrumented irradiation and to estimate the burnup of irradiated DUPIC fuel more accurately. Under normal operating condition, the maximum linear power of DUPIC fuel was estimated as 55.06 kW/m, and the centerline temperature of a pellet was calculated as 2510 deg C. In order to assess the integrity of DUPIC fuel under the accident condition postulated at the HANARO, safety analyses on the locked rotor and reactivity insertion accidents were carried out. The maximum centerline temperature of DUPIC fuel was estimated 2590 deg C and 2094 deg C for each accident, respectively. From the results of the safety analysis, the integrity of DUPIC fuel during the HANARO irradiation test will be secured. The irradiated DUPIC fuel will be transported to the IMEF. The post-irradiation examinations are planned to be performed at the PIEF and IMEF.

  1. 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.

  2. 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.

  3. Maritime Fuel Cell Generator Project.

    Energy Technology Data Exchange (ETDEWEB)

    Pratt, Joseph William [Sandia National Lab. (SNL-CA), Livermore, CA (United States)

    2017-07-01

    Fuel costs and emissions in maritime ports are an opportunity for transportation energy efficiency improvement and emissions reduction efforts. Ocean-going vessels, harbor craft, and cargo handling equipment are still major contributors to air pollution in and around ports. Diesel engine costs continually increase as tighter criteria pollutant regulations come into effect and will continue to do so with expected introduction of carbon emission regulations. Diesel fuel costs will also continue to rise as requirements for cleaner fuels are imposed. Both aspects will increase the cost of diesel-based power generation on the vessel and on shore. Although fuel cells have been used in many successful applications, they have not been technically or commercially validated in the port environment. One opportunity to do so was identified in Honolulu Harbor at the Young Brothers Ltd. wharf. At this facility, barges sail regularly to and from neighbor islands and containerized diesel generators provide power for the reefers while on the dock and on the barge during transport, nearly always at part load. Due to inherent efficiency characteristics of fuel cells and diesel generators, switching to a hydrogen fuel cell power generator was found to have potential emissions and cost savings.

  4. Project on New Domestic Zirconium Alloy Fuel Assembly Irradiation

    Institute of Scientific and Technical Information of China (English)

    ZHANG; Pei-sheng; ZHANG; Ai-min

    2012-01-01

    <正>The objectives of the project is to conduct irradiation at research reactor for small fuel assembly with domestic new zirconium alloy, and then to carry out post irradiation examination, and finally to acquire

  5. Post-irradiation examination of AlFeNi cladded U 3Si 2 fuel plates irradiated under severe conditions

    Science.gov (United States)

    Leenaers, A.; Koonen, E.; Parthoens, Y.; Lemoine, P.; Van den Berghe, S.

    2008-04-01

    Three full size AlFeNi cladded U 3Si 2 fuel plates were irradiated in the BR2 reactor of the Belgian Nuclear Research Centre (SCK·CEN) under relatively severe, but well defined conditions. The irradiation was part of the qualification campaign for the fuel to be used in the future Jules Horowitz reactor in Cadarache, France. After the irradiation, the fuel plates were submitted to an extensive post-irradiation campaign in the hot cell laboratory of SCK·CEN. The PIE shows that the fuel plates withstood the irradiation successfully, as no detrimental defects have been found. At the cladding surface, a multilayered corrosion oxide film has formed. The U-Al-Si layer resulting from the interaction between the U 3Si 2 fuel and the Al matrix, has been quantified as U(Al,Si) 4.6. It is found that the composition of the fuel particles is not homogenous; zones of USi and U 3Si 2 are observed and measured. The fission gas-related bubbles generated in both phases show a different morphology. In the USi fuel, the bubbles are small and numerous while in U 3Si 2 the bubbles are larger but there are fewer.

  6. Irradiated test fuel shipment plan for the LWR MOX fuel irradiation test project

    Energy Technology Data Exchange (ETDEWEB)

    Shappert, L.B.; Dickerson, L.S.; Ludwig, S.B.

    1998-10-16

    This document outlines the responsibilities of DOE, DOE contractors, the commercial carrier, and other organizations participating in a shipping campaign of irradiated test specimen capsules containing mixed-oxide (MOX) fuel from the Idaho National Engineering and Environmental Laboratory (INEEL) to the Oak Ridge National Laboratory (ORNL). The shipments described here will be conducted according to applicable regulations of the US Department of Transportation (DOT), US Nuclear Regulatory Commission (NRC), and all applicable DOE Orders. This Irradiated Test Fuel Shipment Plan for the LWR MOX Fuel Irradiation Test Project addresses the shipments of a small number of irradiated test specimen capsules and has been reviewed and agreed to by INEEL and ORNL (as participants in the shipment campaign). Minor refinements to data entries in this plan, such as actual shipment dates, exact quantities and characteristics of materials to be shipped, and final approved shipment routing, will be communicated between the shipper, receiver, and carrier, as needed, using faxes, e-mail, official shipping papers, or other backup documents (e.g., shipment safety evaluations). Any major changes in responsibilities or data beyond refinements of dates and quantities of material will be prepared as additional revisions to this document and will undergo a full review and approval cycle.

  7. DISSOLUTION OF IRRADIATED MURR FUEL ASSEMBLIES

    Energy Technology Data Exchange (ETDEWEB)

    Kyser, E.

    2010-06-17

    A literature survey on the dissolution of spent nuclear fuel from the University of Missouri Research Reactor (MURR) has been performed. This survey encompassed both internal and external literature sources for the dissolution of aluminum-clad uranium alloy fuels. The most limiting aspect of dissolution in the current facility configuration involves issues related to the control of the flammability of the off-gas from this process. The primary conclusion of this work is that based on past dissolution of this fuel in H-Canyon, four bundles of this fuel (initial charge) may be safely dissolved in a nitric acid flowsheet catalyzed with 0.002 M mercuric nitrate using a 40 scfm purge to control off-gas flammability. The initial charge may be followed by a second charge of up to five bundles to the same dissolver batch depending on volume and concentration constraints. The safety of this flowsheet relies on composite lower flammability limits (LFL) estimated from prior literature, pilot-scale work on the dissolution of site fuels, and the proposed processing flowsheet. Equipment modifications or improved LFL data offer the potential for improved processing rates. The fuel charging sequence, as well as the acid and catalyst concentrations, will control the dissolution rate during the initial portion of the cycle. These parameters directly impact the hydrogen and off-gas generation and, along with the purge flowrate determine the number of bundles that may be charged. The calculation approach within provides Engineering a means to determine optimal charging patterns. Downstream processing of this material should be similar to that of recent processing of site fuels requiring only minor adjustments of the existing flowsheet parameters.

  8. Power generation from solid fuels

    CERN Document Server

    Spliethoff, Hartmut

    2010-01-01

    Power Generation from Solid Fuels introduces the different technologies to produce heat and power from solid fossil (hard coal, brown coal) and renewable (biomass, waste) fuels, such as combustion and gasification, steam power plants and combined cycles etc. The book discusses technologies with regard to their efficiency, emissions, operational behavior, residues and costs. Besides proven state of the art processes, the focus is on the potential of new technologies currently under development or demonstration. The main motivation of the book is to explain the technical possibilities for reduci

  9. 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.

  10. PIE results on MOX fuel irradiated in MIHAMA Unit 1

    Energy Technology Data Exchange (ETDEWEB)

    Yamate, K. [Kansai Electric Power Co., Inc., Mihama, Fukui (Japan). Mihama Power Station; Abeta, S. [Mitsubishi Heavy Industries Ltd. (Japan); Kosaka, Y. [Nuclear Development Corp., Tokai, Ibaraki (Japan); Abe, Y. [Japan Atomic Power Co. (Japan); Kuwahara, H. [Mitsubishi Atomic Power Industries, Inc., Tokyo (Japan)

    1995-12-31

    This paper describes the results of the post-irradiation examination (PIE) on the MOX fuel rods irradiated in the Japanese commercial PWR, MIHAMA Unit 1. The objective of the PIE is not only to confirm the fuel integrity but also to build up the irradiation data base for MOX fuel design. After three cycles irradiation up to the assembly burnup of 23 GWd/t, they were examined at the site and further PIE was carried out on eight MOX fuel rods at a hot laboratory. The non-destructive test results of PIE proved the integrity of the MOX fuel rods up to the burnup of about 25GWd/t, and revealed the similar irradiation behaviour of dimensional change with the standards UO{sub 2} fuel rods. Results of the following destructive tests also revealed the similar irradiation behaviour of FGR, fuel pellet dimensional change and cladding oxidation with the standard UO{sub 2} fuel, and confirmed no abnormality in microstructure changes of fuel pellets. (author).

  11. Fabrication of DUPIC fuel for the 3rd irradiation test

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Woong Ki; Kim, S. S.; Lee, J. W. and others

    2001-09-01

    In this project, based on the simulated DUPIC fuel fabrication experiment and DUPIC fuel characterization experiment at PIEF, DUPIC fuel manufacturing technologies and processes have been developed at DFDF(DUPIC Fuel Development Facility, IMEF M6). DUPIC fuel has been fabricated for the irradiation test at a research reactor. SIMFUEL and DUPIC fuel fabricated using spent PWR fuel were successfully irradiated at HANARO reactor. In this study, DUPIC fuel pellets and mini-elements were manufactured in March 2001 for the third irradiation test to closely investigate the dynamic characteristics of DUPIC fuel at a reactor core for long period. As a result of the experiment, 15 DUPIC pellets with 10.194 10.312 g/cm{sup 3} of sintered density, 3.53 {delta}9.48 {mu}m of averaged grain size, and less than Ra 0.81 {mu}m of surface roughness satisfying the specifications of DUPIC fuel for the third irradiation test have been remotely fabricated at hot cell. 5 DUPIC pellets were loaded in a mini-element made of Zircaloy-4. The soundness of the weld of the mini-elements has been evaluated by microstructural test, helium leak test, and X-ray inspection. Three DUPIC mini-elements are currently under the third irradiation test at HANARO reactor.

  12. Interim irradiated fuel storage facility for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lolich, Jose [INVAP SE, Bariloche (Argentina)

    2002-07-01

    In most research reactors irradiated fuel discharged from the reactor is initially stored underwater inside the reactor building for along period of time. This allows for heat dissipation and fission product decay. In most cases this initial storage is done in a irradiated fuel storage facility pool located closed to the reactor core. After a certain cooling time, the fuel discharged should be relocated for long-term interim storage in a Irradiated Fuel Storage (IFS) Facility. IFS facilities are required for the safe storage of irradiated nuclear fuel before it is reprocessed or conditioned for disposal as radioactive waste. The IFS Facility described in this report is not an integral part of an operating nuclear reactor. This facility many be either co-located with nuclear facilities (such as a nuclear reactor or reprocessing plant) or sited independently of other nuclear facilities. (author)

  13. KEY RESULTS FROM IRRADIATION AND POST-IRRADIATION EXAMINATION OF AGR-1 UCO TRISO FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Demkowicz, Paul A.; Hunn, John D.; Petti, David A.; Morris, Robert N.

    2016-11-01

    The AGR-1 irradiation experiment was performed as the first test of tristructural isotropic (TRISO) fuel in the US Advanced Gas Reactor Fuel Development and Qualification Program. The experiment consisted of 72 right cylinder fuel compacts containing approximately 3×105 coated fuel particles with uranium oxide/uranium carbide (UCO) fuel kernels. The fuel was irradiated in the Advanced Test Reactor for a total of 620 effective full power days. Fuel burnup ranged from 11.3 to 19.6% fissions per initial metal atom and time average, volume average irradiation temperatures of the individual compacts ranged from 955 to 1136°C. This paper focuses on key results from the irradiation and post-irradiation examination, which revealed a robust fuel with excellent performance characteristics under the conditions tested and have significantly improved the understanding of UCO coated particle fuel irradiation behavior within the US program. The fuel exhibited a very low incidence of TRISO coating failure during irradiation and post-irradiation safety testing at temperatures up to 1800°C. Advanced PIE methods have allowed particles with SiC coating failure to be isolated and meticulously examined, which has elucidated the specific causes of SiC failure in these specimens. The level of fission product release from the fuel during irradiation and post-irradiation safety testing has been studied in detail. Results indicated very low release of krypton and cesium through intact SiC and modest release of europium and strontium, while also confirming the potential for significant silver release through the coatings depending on irradiation conditions. Focused study of fission products within the coating layers of irradiated particles down to nanometer length scales has provided new insights into fission product transport through the coating layers and the role various fission products may have on coating integrity. The broader implications of these results and the application of

  14. Osiris, an irradiation reactor for material and nuclear fuel testing; Osiris, reacteur d'irradiation pour materiaux et combustibles

    Energy Technology Data Exchange (ETDEWEB)

    Loubiere, S.; Durande-Ayme, P. [CEA Saclay, Div. Nucleaire Energie, Dept. Reacteurs et Nucleaire Service, 91 - Gif-sur-Yvette (France)

    2005-07-01

    Since 1966 the Osiris reactor located at Saclay has been participating in French and international irradiation programs for research and development in the field of nuclear fuel and materials. Today the French atomic commission (Cea) pursues irradiation programs in support of existing reactors, mainly PWR, strengthening its own knowledge and the one of its clients on fuel and material behaviour under irradiation, pertaining to plant life-time issues and high burn-up. For instance important programs have been performed on pressure vessel steel aging, pellet-clad interaction, internal component aging and mox fuel qualification. With the arising of the Generation 4 research and development programs, the Osiris reactor has developed capacities to undertake material and fuel irradiation under high temperature conditions. Routine irradiations such as the doping of silicon or the production of radio-nuclides for medical or imaging purposes are made on a daily basis. The specificities of the Osiris reactor are presented in the first part of this paper while the second part focuses on the experimental devices available in Osiris to perform irradiation in light water reactor conditions and in high temperature reactor conditions and on their associated programs.

  15. Production of LEU Fully Ceramic Microencapsulated Fuel for Irradiation Testing

    Energy Technology Data Exchange (ETDEWEB)

    Terrani, Kurt A [ORNL; Kiggans Jr, James O [ORNL; McMurray, Jake W [ORNL; Jolly, Brian C [ORNL; Hunt, Rodney Dale [ORNL; Trammell, Michael P [ORNL; Snead, Lance Lewis [ORNL

    2016-01-01

    Fully Ceramic Microencapsulated (FCM) fuel consists of tristructural isotropic (TRISO) fuel particles embedded inside a SiC matrix. This fuel inherently possesses multiple barriers to fission product release, namely the various coating layers in the TRISO fuel particle as well as the dense SiC matrix that hosts these particles. This coupled with the excellent oxidation resistance of the SiC matrix and the SiC coating layer in the TRISO particle designate this concept as an accident tolerant fuel (ATF). The FCM fuel takes advantage of uranium nitride kernels instead of oxide or oxide-carbide kernels used in high temperature gas reactors to enhance heavy metal loading in the highly moderated LWRs. Production of these kernels with appropriate density, coating layer development to produce UN TRISO particles, and consolidation of these particles inside a SiC matrix have been codified thanks to significant R&D supported by US DOE Fuel Cycle R&D program. Also, surrogate FCM pellets (pellets with zirconia instead of uranium-bearing kernels) have been neutron irradiated and the stability of the matrix and coating layer under LWR irradiation conditions have been established. Currently the focus is on production of LEU (7.3% U-235 enrichment) FCM pellets to be utilized for irradiation testing. The irradiation is planned at INL s Advanced Test Reactor (ATR). This is a critical step in development of this fuel concept to establish the ability of this fuel to retain fission products under prototypical irradiation conditions.

  16. Chemical state of fission products in irradiated uranium carbide fuel

    Science.gov (United States)

    Arai, Yasuo; Iwai, Takashi; Ohmichi, Toshihiko

    1987-12-01

    The chemical state of fission products in irradiated uranium carbide fuel has been estimated by equilibrium calculation using the SOLGASMIX-PV program. Solid state fission products are distributed to the fuel matrix, ternary compounds, carbides of fission products and intermetallic compounds among the condensed phases appearing in the irradiated uranium carbide fuel. The chemical forms are influenced by burnup as well as stoichiometry of the fuel. The results of the present study almost agree with the experimental ones reported for burnup simulated carbides.

  17. Metallographic analysis of irradiated RERTR-3 fuel test specimens.

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, M. K.; Hofman, G. L.; Strain, R. V.; Clark, C. R.; Stuart, J. R.

    2000-11-08

    The RERTR-3 irradiation test was designed to investigate the irradiation behavior of aluminum matrix U-MO alloy dispersion fuels under high-temperature, high-fission-rate conditions. Initial postirradiation examination of RERTR-3 fuel specimens has concentrated on binary U-MO atomized fuels. The rate of matrix aluminum depletion was found to be higher than predictions based on low temperature irradiation data. Wavelength Dispersive X-ray Spectroscopy (WDS) indicates that aluminum is present in the interior of the fuel particles. WDS data is supported by a mass and volume balance calculation performed on the basis of image analysis results. The depletion of matrix aluminum seems to have no detrimental effects on fuel performance under the conditions tested to date.

  18. 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

  19. Metallography and fuel cladding chemical interaction in fast flux test facility irradiated metallic U-10Zr MFF-3 and MFF-5 fuel pins

    Science.gov (United States)

    Carmack, W. J.; Chichester, H. M.; Porter, D. L.; Wootan, D. W.

    2016-05-01

    The Mechanistic Fuel Failure (MFF) series of metal fuel irradiations conducted in the Fast Flux Test Facility (FFTF) provides an important comparison between data generated in the Experimental Breeder Reactor (EBR-II) and that expected in a larger-scale fast reactor. The MFF fuel operated with a peak cladding temperature at the top of the fuel column, but developed peak burnup at the centerline of the core. This places the peak fuel temperature midway between the core center and the top of fuel, lower in the fuel column than in EBR-II experiments. Data from the MFF-3 and MFF-5 assemblies are most comparable to the data obtained from the EBR-II X447 experiment. The two X447 pin breaches were strongly influenced by fuel/cladding chemical interaction (FCCI) at the top of the fuel column. Post irradiation examination data from MFF-3 and MFF-5 are presented and compared to historical EBR-II data.

  20. Irradiation effects on thermal properties of LWR hydride fuel

    Science.gov (United States)

    Terrani, Kurt; Balooch, Mehdi; Carpenter, David; Kohse, Gordon; Keiser, Dennis; Meyer, Mitchell; Olander, Donald

    2017-04-01

    Three hydride mini-fuel rods were fabricated and irradiated at the MIT nuclear reactor with a maximum burnup of 0.31% FIMA or ∼5 MWd/kgU equivalent oxide fuel burnup. Fuel rods consisted of uranium-zirconium hydride (U (30 wt%)ZrH1.6) pellets clad inside a LWR Zircaloy-2 tubing. The gap between the fuel and the cladding was filled with lead-bismuth eutectic alloy to eliminate the gas gap and the large temperature drop across it. Each mini-fuel rod was instrumented with two thermocouples with tips that are axially located halfway through the fuel centerline and cladding surface. In-pile temperature measurements enabled calculation of thermal conductivity in this fuel as a function of temperature and burnup. In-pile thermal conductivity at the beginning of test agreed well with out-of-pile measurements on unirradiated fuel and decreased rapidly with burnup.

  1. Recovery of minor actinides from irradiated superfact fuels

    Energy Technology Data Exchange (ETDEWEB)

    Apoltolidis, C.; Glatz, J.P.; Molinet, R.; Nicholl, A.; Pagliosa, G.; Romer, K.; Bokelund, H.; Koch, L. [European Commission, JRC, Institute fuer Transuranium Elements, Karlsruhe (Germany)

    1995-12-31

    It could be demonstrated that the reprocessing of fast reactor oxide fuels containing up to 45 % MA (Np and Am), irradiated in the PHENIX reactor in the frame of a transmutation study, is possible. The fuels were dissolved under PUREX type conditions in order to determine their behaviour in the head-end step of the reprocessing process. For one of the fuels containing 20 % Am and 20 % Np before irradiation, an almost complete partitioning of actinides from the dissolver solution could be achieved. Chromatographic extraction was used for the separation of the main bulk elements U, Pu and Np, whereas centrifugal extractors were used to separate the minor actinides from the remaining high level liquid wastes (HLLW). For the relevant radio-toxic isotopes a high recovery rate from the irradiation targets was reached. Those elements are thus available for new fuel fabrication. (authors) 12 refs.

  2. Heat Generation by Irradiated Complex Composite Nanostructures

    DEFF Research Database (Denmark)

    Ma, Haiyan; Tian, Pengfei; Pello, Josselin;

    2014-01-01

    Heating of irradiated metallic e-beam generated nanostructures was quantified through direct measurements paralleled by novel model-based numerical calculations. By comparing discs, triangles, and stars we showed how particle shape and composition determines the heating. Importantly, our results ...... revealed that substantial heat is generated in the titanium adhesive layer between gold and glass. Even when the Ti layer is as thin as 2 nm it absorbs as much as a 30 nm Au layer and hence should not be ignored....

  3. 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

  4. Axial gas flow in irradiated PWR fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Dagbjartsson, S.J.; Murdock, B.A.; Owen, D.E.; MacDonald, P.E.

    1977-09-01

    Transient and steady state axial gas flow experiments were performed on six irradiated, commercial pressurized water reactor fuel rods at ambient temperature and 533 K. Laminar flow equations, as used in the FRAP-T2 and SSYST fuel behavior codes, were used with the gas flow results to calculate effective fuel rod radial gaps. The results of these analyses were compared with measured gap sizes obtained from metallographic examination of one fuel rod. Using measured gap sizes as input, the SSYST code was used to calculate pressure drops and mass fluxes and the results were compared with the experimental gas flow data.

  5. Radiochemical Assays of Irradiated VVER-440 Fuel for Use in Spent Fuel Burnup Credit Activities

    Energy Technology Data Exchange (ETDEWEB)

    Jardine, L J

    2005-04-25

    The objective of this spent fuel burnup credit work was to study and describe a VVER-440 reactor spent fuel assembly (FA) initial state before irradiation, its operational irradiation history and the resulting radionuclide distribution in the fuel assembly after irradiation. This work includes the following stages: (1) to pick out and select a specific spent (irradiated) FA for examination; (2) to describe the FA initial state before irradiation; (3) to describe the irradiation history, including thermal calculations; (4) to examine the burnup distribution of select radionuclides along the FA height and cross-section; (5) to examine the radionuclide distributions; (6) to determine the Kr-85 release into the plenum; (7) to select and prepare FA rod specimens for destructive examinations; (8) to determine the radionuclide compositions, isotope masses and burnup in the rod specimens; and (9) to analyze, document and process the results. The specific workscope included the destructive assay (DA) of spent fuel assembly rod segments with an {approx}38.5 MWd/KgU burnup from a single VVER-440 fuel assembly from the Novovorenezh reactor in Russia. Based on irradiation history criteria, four rods from the fuel assembly were selected and removed from the assembly for examination. Next, 8 sections were cut from the four rods and sent for destructive analysis of radionuclides by radiochemical analyses. The results were documented in a series of seven reports over a period of {approx}1 1/2 years.

  6. Radiochemical Assays of Irradiated VVER-440 Fuel for Use in Spent Fuel Burnup Credit Activities

    Energy Technology Data Exchange (ETDEWEB)

    Jardine, L J

    2005-04-25

    The objective of this spent fuel burnup credit work was to study and describe a VVER-440 reactor spent fuel assembly (FA) initial state before irradiation, its operational irradiation history and the resulting radionuclide distribution in the fuel assembly after irradiation. This work includes the following stages: (1) to pick out and select a specific spent (irradiated) FA for examination; (2) to describe the FA initial state before irradiation; (3) to describe the irradiation history, including thermal calculations; (4) to examine the burnup distribution of select radionuclides along the FA height and cross-section; (5) to examine the radionuclide distributions; (6) to determine the Kr-85 release into the plenum; (7) to select and prepare FA rod specimens for destructive examinations; (8) to determine the radionuclide compositions, isotope masses and burnup in the rod specimens; and (9) to analyze, document and process the results. The specific workscope included the destructive assay (DA) of spent fuel assembly rod segments with an {approx}38.5 MWd/KgU burnup from a single VVER-440 fuel assembly from the Novovorenezh reactor in Russia. Based on irradiation history criteria, four rods from the fuel assembly were selected and removed from the assembly for examination. Next, 8 sections were cut from the four rods and sent for destructive analysis of radionuclides by radiochemical analyses. The results were documented in a series of seven reports over a period of {approx}1 1/2 years.

  7. Segmented fuel irradiation program: investigation on advanced materials

    Energy Technology Data Exchange (ETDEWEB)

    Uchida, H. [NUPEC (Japan); Goto, K. [KEPCO, Osaka (Japan); Sabate, R. [A.N. Asco - C.N. Vandellos, Barcelona (Spain); Abeta, S.; Baba, T. [MHI, Nishi-Ku, Yokohama (Japan); Matias, E. de; Alonso, J. [ENUSA, Madrid (Spain)

    1999-07-01

    The Segmented Fuel Irradiation Program, started in 1991, is a collaboration between the Japanese organisations Nuclear Power Engineering Corporation (NUPEC), the Kansai Electric Power Co., Inc. (KEPCO) representing other Japanese utilities, and Mitsubishi Heavy Industries, Ltd. (MHI); and the Spanish Organisations Empresa Nacional de Electricidad, S.A. (ENDESA) representing A.N. Vandellos 2, and Empresa Nacional Uranio, S.A. (ENUSA); with the collaboration of Westinghouse. The objective of the Program is to make substantial contribution to the development of advanced cladding and fuel materials for better performance at high burn-up and under operational power transients. For this Program, segmented fuel rods were selected as the most appropriate vehicle to accomplish the aforementioned objective. Thus, a large number of fuel and cladding combinations are provided while minimising the total amount of new material, at the same time, facilitating an eventual irradiation extension in a test reactor. The Program consists of three major phases: phase I: design, licensing, fabrication and characterisation of the assemblies carrying the segmented rods (1991 - 1994); phase II: base irradiation of the assemblies at Vandellos 2 NPP, and on-site examination at the end of four cycles (1994-1999). Phase III: ramp testing at the Studsvik facilities and hot cell PIE (1996-2001). The main fuel design features whose effects on fuel behaviour are being analysed are: alloy composition (MDA and ZIRLO vs. Zircaloy-4); tubing texture; pellet grain size. The Program is progressing satisfactorily as planned. The base irradiation is completed in the first quarter of 1999, and so far, tests and inspections already carried out are providing useful information on the behaviour of the new materials. Also, the Program is delivering a well characterized fuel material, irradiated in a commercial reactor, which can be further used in other fuel behaviour experiments. The paper presents the main

  8. SLIGHTLY IRRADIATED FUEL (SIF) INTERIM DISPOSITION PROJECT

    Energy Technology Data Exchange (ETDEWEB)

    NORTON SH

    2010-02-23

    CH2M HILL Plateau Remediation Company (CH2M HILL PRC) is proud to submit the Slightly Irradiated Fuel (SIF) Interim Disposition Project for consideration by the Project Management Institute as Project of the Year for 2010. The SIF Project was a set of six interrelated sub-projects that delivered unique stand-alone outcomes, which, when integrated, provided a comprehensive and compliant system for storing high risk special nuclear materials. The scope of the six sub-projects included the design, construction, testing, and turnover of the facilities and equipment, which would provide safe, secure, and compliant Special Nuclear Material (SNM) storage capabilities for the SIF material. The project encompassed a broad range of activities, including the following: Five buildings/structures removed, relocated, or built; Two buildings renovated; Structural barriers, fencing, and heavy gates installed; New roadways and parking lots built; Multiple detection and assessment systems installed; New and expanded communication systems developed; Multimedia recording devices added; and A new control room to monitor all materials and systems built. Project challenges were numerous and included the following: An aggressive 17-month schedule to support the high-profile Plutonium Finishing Plant (PFP) decommissioning; Company/contractor changeovers that affected each and every project team member; Project requirements that continually evolved during design and construction due to the performance- and outcome-based nature ofthe security objectives; and Restrictions imposed on all communications due to the sensitive nature of the projects In spite of the significant challenges, the project was delivered on schedule and $2 million under budget, which became a special source of pride that bonded the team. For years, the SIF had been stored at the central Hanford PFP. Because of the weapons-grade piutonium produced and stored there, the PFP had some of the tightest security on the Hanford

  9. Irradiated MTR fuel assemblies sipping test

    Energy Technology Data Exchange (ETDEWEB)

    Perrotta, J.A.; Terremoto, Luis A.A.; Zeituni, Carlos A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Div. de Engenharia do Nucleo

    1997-10-01

    This paper describes the procedure and methodology used to perform sipping test with the IEA-R1 fuel assemblies at the storage pool, and presents the results obtained for Cs-137 sipping water activity for each fuel assembly analyzed. Discussion is made correlating corrosion pits to the activity values measured. A Cs-137 leaking rate is determined which can be compared to the criteria established for canning spent fuel assemblies inside the pool of for shipment abroad. 3 refs., 13 figs., 1 tab.

  10. 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.

  11. Non destructive testing of irradiated fuel assemblies at the IEA-R1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Jose Eduardo Rosa da; Terremoto, Luis Antonio Albiac; Castanheira, Myrthes; Teodoro, Celso Antonio; Silva, Antonio Teixeira e; Damy, Margaret de Almeida; Lucki, Georgi [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)]. E-mails: jersilva@ipen.br; laaterre@ipen.br; myrthes@ipen.br; cteodoro@ipen.br; teixeira@ipen.br; madamy@ipen.br; glucki@ipen.br

    2007-07-01

    Fuel performance and nuclear fuel qualification require a post-irradiation analysis. Non-destructive methods are utilised both in irradiated fuel storage pools and in hot-cells laboratories. As Brazil does not have hot-cells facilities for post-irradiation analysis, a qualification program for the Material Testing Reactor (MTR) fuel elements made at IPEN/CNEN-SP was adopted, based on non-destructive tests. The IPEN Fuel Engineering Group - CENC developed basic facilities for fuels post-irradiated analysis inside the reactor pool, which gives indications of: general state, by visual inspection; the integrity of the irradiated fuel cladding, by sipping tests; thickness measurements of the fuel miniplates during the irradiation time, for swelling evaluation; and, local burn-up evaluation by gamma spectrometry along the active area of the fuel element. This work describes that facilities, equipment and examples of some irradiated fuels analysis performed. (author)

  12. RECENT DEVELOPMENT IN TEM CHARACTERIZATION OF IRRADIATED RERTR FUELS

    Energy Technology Data Exchange (ETDEWEB)

    J. Gan; B.D. Miller; D.D. Keiser Jr.; A.B. Robinson; J.W. Madden; P.G. Medvedev; D.M. Wachs

    2011-10-01

    The recent development on TEM work of irradiated RERTR fuels includes microstructural characterization of the irradiated U-10Mo/alloy-6061 monolithic fuel plate, the RERTR-7 U-7Mo/Al-2Si and U-7Mo/Al-5Si dispersion fuel plates. It is the first time that a TEM sample of an irradiated nuclear fuel was prepared using the focused-ion-beam (FIB) lift-out technical at the Idaho National Laboratory. Multiple FIB TEM samples were prepared from the areas of interest in a SEM sample. The characterization was carried out using a 200kV TEM with a LaB6 filament. The three dimensional orderings of nanometer-sized fission gas bubbles are observed in the crystalline region of the U-Mo fuel. The co-existence of bubble superlattice and dislocations is evident. Detailed microstructural information along with composition analysis is obtained. The results and their implication on the performance of these fuels are discussed.

  13. Dearomatization of jet fuel on irradiated platinum-supported catalyst

    Science.gov (United States)

    Múčka, V.; Ostrihoňová, A.; Kopernický, I.; Mikula, O.

    The effect of ionizing radiation ( 60Co γ-rays) on Pt-supported catalyst used for the dearomatization of jet fuel with distillation in the range 395-534 K has been studied. Pre-irradiation of the catalyst with doses in the range 10 2-5 × 10 4 Gy leads to the partial catalyst activation. Irradiation of the catalyst enhances its resistance to catalyst poisons, particularly to sulphur-compounds, and this is probably the reason for its catalytic activity being ˜60-100% greater than that of un-irradiated catalyst. Optimum conditions for dearomatization on the irradiated catalyst were found and, by means of a rotary three-factorial experiment, it was shown that these lie at lower temperatures and lower pressures than those for un-irradiated catalyst.

  14. The re-instrumentation irradiation test of nuclear fuel using fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chul Yong; Joung, C. Y.; Hong, J. T.; Ahn, S. H.; Choo, K. N. [KAERI, Daejeon (Korea, Republic of)

    2011-07-15

    This report is the status art report on re-instrumentation. The main techniques described in this report are technology that is developed in Norway HALDEN and domestic research facilities. Although re-instrumentation is not gone vigorously after 1990, HALDEN's re-instrumentation equipment was made until recently. In the meantime, re-instrumentation research was gone in domestic, but irradiation test did not performed actually. But DUPIC fuel irradiation is similar to re-instrumentation, so the irradiation test can be utilized directly to the Fuel Test Loop

  15. DUPIC fuel irradiation test and performance evaluation; the performance analysis of pellet-cladding contact fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ho, K. I.; Kim, H. M.; Yang, K. B.; Choi, S. J. [Suwon University, Whasung (Korea)

    2002-04-01

    Thermal and mechanical models were reviewed, and selected for the analysis of nuclear fuel performance in reactor. 2 dimensional FEM software was developed. Thermal models-gap conductances, thermal conductivity of pellets, fission gas release, temperature distribution-were set and packaged into a software. Both thermal and mechanical models were interrelated to each other, and the final results, fuel performance during irradiation is obtained by iteration calculation. Also, the contact phenomena between pellet and cladding was analysed by mechanical computer software which was developed during this work. dimensional FEM program was developed which estimate the mechanical behavior and the thermal behaviors of nuclear fuel during irradiation. Since there is a importance during the mechanical deformation analysis in describing pellet-cladding contact phenomena, simplified 2 dimensional calculation method is used after the contact. The estimation of thermal fuel behavior during irradiation was compared with the results of other. 8 refs., 17 figs. (Author)

  16. System of leak inspection of irradiated fuel; Sistema de inspeccion de fuga de combustible irradiado

    Energy Technology Data Exchange (ETDEWEB)

    Delfin L, A.; Castaneda J, G.; Mazon R, R.; Aguilar H, F. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: rmr@nuclear.inin.mx

    2007-07-01

    The International Atomic Energy Agency (IAEA) through the project RLA/04/18 Irradiated Fuel Management in Research reactors, recommended among other that the participant countries (Brazil, Argentina, Chile, Peru and Mexico), develop the sipping tool to generate registrations of the state that keep the irradiated fuels in the facilities of each country. The TRIGA Mark lll Reactor (RTMIII) Department, generated a project that it is based on the dimensions of the used fuel by the RTMIII, for design and to build an inspection system of irradiated fuel well known as SIPPING. This technique, provides a high grade of accuracy in the detection of gassy fission products or liquids that escape from the enveloping of fuels that have flaws or flights. The operation process of the SIPPING is carried out generating the migration of fission products through the creation of a pressure differential gas or vacuum to identify fuel assemblies failed by means of the detection of the xenon and/or krypton presence. The SIPPING system, is a device in revolver form with 4 tangential nozzles, which will discharge the fluid between the external surface of the enveloping of the fuel and the interior surface of the encircling one; the device was designed with independent pieces, with threaded joining and with stamps to impede flights of the fluid toward the exterior of the system. The System homogenizes and it distributes the fluid pressure so that the 4 nozzles work to equality of conditions, for what the device was designed in 3 pieces, an internal that is denominated revolver, one external that calls cover, and a joining called mamelon that will unite with the main encircling of the system. The detection of fission products in failed fuels, its require that inside the encircling one where the irradiated fuel element is introduced, be generated a pressure differential of gas or vacuum, and that it allows the samples extraction of water. For what generated a top for the encircling with the

  17. Irradiation characteristics examination technology development of irradiated nuclear material and high burn-up fuels

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Kwon Pyo; Choo, Y. S.; Oh, Y. W. [and others

    2002-12-01

    The research and development for the first year of the project are performed through specialization of researchers, information from aborad and international cooperation, securement of advanced nuclear technology, development and installation of test equipment, application of external man-power, establishment of advanced test techniques, and certified test method. 1. Absolute efficiency measurement examination technology development of gamma scanning system 2. Sample preparation technology development of SEM and EPMA for micro-structural observation and chemical composition analysis 3. Irradiated high burn-up nuclear fuel transportation and test for PWR 4. Development of hot cell examination techniques and equipment 5. Acquirement of KOLAS system. In addition to the project, the following activities are carried out as follows; - PIE of Hanaro fuel(KH99H-001) - PIE of U-Mo advanced nuclear fuel irradiated at Hanaro - PIE of Hi-MET advanced nuclear fuel irradiated at Hanaro - PIE of DUPIC project - Hot cell examination of Hanaro irradiated capsule - Leaching test of PWR fuels - Surveillance test of PWR vessels - Mechanical test of CANDU pressure tubes.

  18. Evaluation of FSV-1 cask for the transport of LWR irradiated fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    1980-05-01

    The Model FSV-1 spent fuel shipping cask was designed by General Atomic Company (GA) to service the Fort St. Vrain (FSV) nuclear generating station, a High Temperature Gas Reactor (HTGR) owned and operated by Public Service Company of Colorado (PSC). This report presents an evaluation of the suitability of the FSV-1 cask for the transport of irradiated Light Water Reactor (LWR) fuel assemblies from both Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR). The FSV-1 cask evaluation parameters covered a wide spectrum of LWR fuel assemblies, based on burnup in Megawatt Days/Metric Ton of Heavy Metal (MWD/MTHM) and years of decay since irradiation. The criteria for suitability included allowable radiation dose rates, cask surface and interior temperatures and the Gross Vehicle Weight (GVW) of the complete shipping system.

  19. Microfluidic fuel cells for energy generation.

    Science.gov (United States)

    Safdar, M; Jänis, J; Sánchez, S

    2016-08-07

    Sustainable energy generation is of recent interest due to a growing energy demand across the globe and increasing environmental issues caused by conventional non-renewable means of power generation. In the context of microsystems, portable electronics and lab-on-a-chip based (bio)chemical sensors would essentially require fully integrated, reliable means of power generation. Microfluidic-based fuel cells can offer unique advantages compared to conventional fuel cells such as high surface area-to-volume ratio, ease of integration, cost effectiveness and portability. Here, we summarize recent developments which utilize the potential of microfluidic devices for energy generation.

  20. NSRR experiment with un-irradiated uranium-zirconium hydride fuel. Design, fabrication process and inspection data of test fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Sasajima, Hideo; Fuketa, Toyoshi; Ishijima, Kiyomi; Kuroha, Hiroshi; Ikeda, Yoshikazu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Aizawa, Keiichi

    1998-08-01

    An experiment plan is progressing in the Nuclear Safety Research Reactor (NSRR) to perform pulse-irradiation with uranium-zirconium hydride (U-ZrH{sub x}) fuel. This fuel is widely used in the training research and isotope production reactor of GA (TRIGA). The objectives of the experiment are to determine the fuel rod failure threshold and to investigate fuel behavior under simulated reactivity initiated accident (RIA) conditions. This report summarizes design, fabrication process and inspection data of the test fuel rods before pulse-irradiation. The experiment with U-ZrH{sub x} fuel will realize precise safety evaluation, and improve the TRIGA reactor performance. The data to be obtained in this program will also contribute development of next-generation TRIGA reactor and its safety evaluation. (author)

  1. 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

  2. Status Report on the Fabrication of Fuel Cladding Chemical Interaction Test Articles for ATR Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-28

    FeCrAl alloys are a promising new class of alloys for light water reactor (LWR) applications due to their superior oxidation and corrosion resistance in high temperature environments. The current R&D efforts have focused on the alloy composition and processing routes to generate nuclear grade FeCrAl alloys with optimized properties for enhanced accident tolerance while maintaining properties needed for normal operation conditions. Therefore, the composition and processing routes must be optimized to maintain the high temperature steam oxidation (typically achieved by increasing the Cr and Al content) while still exhibiting properties conducive to normal operation in a LWR (such as radiation tolerance where reducing Cr content is favorable). Within this balancing act is the addition of understanding the influence on composition and processing routes on the FeCrAl alloys for fuel-cladding chemical interactions (FCCI). Currently, limited knowledge exists on FCCI for the FeCrAl-UO2 clad-fuel system. To overcome the knowledge gaps on the FCCI for the FeCrAl-UO2 clad-fuel system a series of fueled irradiation tests have been developed for irradiation in the Advanced Test Reactor (ATR) housed at the Idaho National Laboratory (INL). The first series of tests has already been reported. These tests used miniaturized 17x17 PWR fuel geometry rodlets of second-generation FeCrAl alloys fueled with industrial Westinghouse UO2 fuel. These rodlets were encapsulated within a stainless steel housing.To provide high fidelity experiments and more robust testing, a new series of rodlets have been developed deemed the Accident Tolerant Fuel Experiment #1 Oak Ridge National Laboratory FCCI test (ATF-1 ORNL FCCI). The main driving factor, which is discussed in detail, was to provide a radiation environment where prototypical fuel-clad interface temperatures are met while still maintaining constant contact between industrial fuel and the candidate cladding alloys

  3. Microbial Biofilm Growth on Irradiated, Spent Nuclear Fuel Cladding

    Energy Technology Data Exchange (ETDEWEB)

    S.M. Frank

    2009-02-01

    A fundamental criticism regarding the potential for microbial influenced corrosion in spent nuclear fuel cladding or storage containers concerns whether the required microorganisms can, in fact, survive radiation fields inherent in these materials. This study was performed to unequivocally answer this critique by addressing the potential for biofilm formation, the precursor to microbial-influenced corrosion, in radiation fields representative of spent nuclear fuel storage environments. This study involved the formation of a microbial biofilm on irradiated spent nuclear fuel cladding within a hot cell environment. This was accomplished by introducing 22 species of bacteria, in nutrient-rich media, to test vessels containing irradiated cladding sections and that was then surrounded by radioactive source material. The overall dose rate exceeded 2 Gy/h gamma/beta radiation with the total dose received by some of the bacteria reaching 5 × 103 Gy. This study provides evidence for the formation of biofilms on spent-fuel materials, and the implication of microbial influenced corrosion in the storage and permanent deposition of spent nuclear fuel in repository environments.

  4. 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.

  5. Radionuclide release from irradiated Th-Pu mox fuel

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez, N.; Quinones, J. [Ciemat., Avda. Complutense 22. E-28040 Madrid (Spain); Cobos, J. [Centro Nacional de Aceleradores, Parque Tecnologico Cartuja 93, Av. Thomas Alva Edison, 7, E-41092 Sevilla (Spain); Rondinella, V.V.; Van Winckel, S.; Somers, J.; Papaioanu, D.; Glatz, J.P. [European Commission, Joint Research Centre, Institute for Transuranium Elements, Postfach 2340, D-76125 Karlsruhe (Germany)

    2010-07-01

    Plutonium and minor actinides produced as by-products of the UO{sub 2} nuclear cycle could be considered as waste or energy source depending on the strategy selected in the nuclear energy programme. Considering Pu and Minor Actinides as a source, they can be burned in existing water reactor for diminishing the radiotoxicity of the spent fuel, it is necessary to use 'inactive' materials as matrix like ThO{sub 2}. ThO{sub 2} matrix has demonstrated its Pu burning efficiency and higher corrosion resistance than UO{sub 2}. Uranium-plutonium mixed oxide (MOX) fuel efficiency is low because the presence of U in MOX results in the creation of some new Pu under irradiation. The dissolution behaviour of irradiated (Th,Pu)O{sub 2} pellets with burn-up of 38.8 MWd/kg Th has been studied in carbonated (20 mM HCO{sub 3}{sup -}), deionised and granite ground water solution in a hot cell. The dissolution behaviour of Th, Pu, U and Np was studied in order to find out whether radionuclides release is depending on the matrix dissolution (solubility control). After irradiating the samples, K-ORIGEN and ORIGEN ARP codes were used to find out the theoretical inventory. Afterwards, fuel samples were dissolved completely and analyzed, in order to determine the experimental radionuclide inventory of the irradiated fuel. Th matrix alteration appears to reach an steady state and radionuclides dissolution shows dependence on the matrix behaviour as can be observed through the FIAP results. (authors)

  6. 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

  7. Report on FY16 Low-dose Metal Fuel Irradiation and PIE

    Energy Technology Data Exchange (ETDEWEB)

    Edmondson, Philip D.

    2016-09-01

    This report gives an overview of the efforts into the low-dose metal fuel irradiation and PIE as part of the Fuel Cycle Research & Development (FCRD) Advanced Fuels Campaign (AFC) milestone M3FT-16OR020303031. The current status of the FCT and FCRP irradiation campaigns are given including a description of the materials that have been irradiated, analysis of the passive temperature monitors, and the initial PIE efforts of the fuel samples.

  8. Deployable Fuel Cell Power Generator - Multi-Fuel Processor

    Science.gov (United States)

    2009-02-01

    apparent difference between the two investigations is the catalyst; however, the larger capacity of the packed-bed over that of microchannel reactor might...Steam Reforming Reactor and the Radiant Burner ................................................................... 7  6: Combustion Fuel Vaporizer...demonstrate the direct steam reforming concept. Packed-bed steam reforming reactor and coiled tube steam generator with radiant burners were used. The

  9. Spent Nuclear Fuel Project (SNFP) gas generation from N-Fuel in multi-canister overpacks

    Energy Technology Data Exchange (ETDEWEB)

    Cooper, T.D.

    1996-08-01

    During the conversion from wet pool storage for spent nuclear fuel at Hanford, gases will be generated from both radiolysis and chemical reactions. The gas generation phenomenon needs to be understood as it applies to safety and design issues,specifically over pressurization of sealed storage containers,and detonation/deflagration of flammable gases. This study provides an initial basis to predict the implications of gas generation on the proposed functional processes for spent nuclear fuel conversion from wet to dry storage. These projections are based upon examination of the history of fuel manufacture at Hanford, irradiation in the reactors, corrosion during wet pool storage, available fuel characterization data and available information from literature. Gas generation via radiolysis and metal corrosion are addressed. The study examines gas generation, the boundary conditions for low medium and high levels of sludge in SNF storage/processing containers. The functional areas examined include: flooded and drained Multi-Canister Overpacks, cold vacuum drying, shipping and staging and long term storage.

  10. 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%.

  11. 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.

  12. Test requirement for PIE of HANARO irradiated fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Lim, I. C.; Cho, Y. G

    2000-06-01

    Since the first criticality of HANARO reached in Feb. of 1995, the rod type U{sub 3}Si-A1 fuel imported from AECL has been used. From the under-water fuel inspection which has been conducted since 1997, a ballooning-rupture type abnormality was observed in several fuel rods. In order to find the root cause of this abnormality and to find the resolution, the post irradiation examination(PIE) was proposed as the best way. In this document, the information from the under-water inspection as well as the PIE requirements are described. Based on the information in this document, a detail test plan will be developed by the project team who shall conduct the PIE.

  13. 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.

  14. Design and manufacturing of 05F-01K instrumented capsule for nuclear fuel irradiation in Hanaro

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, J. M.; Shin, Y. T.; Park, S. J. (and others)

    2007-07-15

    An instrumented capsule was developed to be able to measure fuel characteristics, such as fuel temperature, internal pressure of fuel rod, fuel pellet elongation, and neutron flux, etc., during the irradiation test of nuclear fuel in Hanaro. The instrumented capsule(02F-11K) for measuring and monitoring and monitoring fuel centerline temperature and neutron flux was designed and manufactured. It was successfully irradiated in the test hole OR5 of Hanaro from March 14, 2003 to June 1, 2003 (53.84 full power days at 24 MW). In the year of 2004, 3 test fuel rods and the instrumented capsule(03F-05K) were designed and manufactured to measure fuel centerline temperature, internal pressure of fuel rod, and fuel axial deformation during irradiation test. This capsule was irradiated in the test hole OR5 of Hanaro reactor from April 26, 2004 to October 1, 2004 (59.5 EFPD at 24 {approx} 30 MW). The six typed dual instrumented fuel rods, which allow for two characteristics to be measured simultaneously in one fuel rod, have been designed and manufactured to enhance the efficiency of the irradiation test using the instrumented fuel capsule. The 05F-01K instrumented fuel capsule was designed and manufactured for a design verification test of the three dual instrumented fuel rods. The irradiation test of the 05F-01K instrumented fuel capsule will be carried out at the OR5 vertical experimental hole of Hanaro.

  15. Study on containerisation of irradiated fuel at JRC ISPRA for medium/long-term storage

    Energy Technology Data Exchange (ETDEWEB)

    Bertelli, S.; Bielli, G.; Covini, R.

    2001-07-01

    During the last 40 years big amounts of wastes arising from past experiments have been generated at JRC Ispra. These wastes are now stored on site in unconditioned form and must be characterised and re-conditioned to ensure their acceptance by future repositories. Among the several types of wastes produced, spent fuel has a great importance in JRC waste management. In fact, there are more than the tons of irradiated material, varying widely from commercial to experimental fuel elements or pins, in form of oxides and metal fuel, with very different geometry, dry and wet stored. The biggest part of it can be considered as unirradiated, according to the IAEA regulations (ST-1, 1996), while a relatively small amount (nearly 720 kg U total) are to be considered as irradiated and treated accordingly. Studies on the most suitable solutions for medium/long term storage for such irradiated experimental fuel are being performed, taking into account two main options, containerisation of the nuclear materials, as it is, in suitable casks or reprocessing. The technical aspects of the project for containerisation are here discussed. (Author)

  16. Microscopic analysis of irradiated AGR-1 coated particle fuel compacts

    Energy Technology Data Exchange (ETDEWEB)

    Scott Ploger; Paul Demkowicz; John Hunn; Robert Morris

    2012-10-01

    The AGR-1 experiment involved irradiation of 72 TRISO-coated particle fuel compacts to a peak burnup of 19.5% FIMA with no in-pile failures observed out of 3×105 total particles. Irradiated AGR-1 fuel compacts have been cross-sectioned and analyzed with optical microscopy to characterize kernel, buffer, and coating behavior. Five compacts have been examined so far, spanning a range of irradiation conditions (burnup, fast fluence, and irradiation temperature) and including all four TRISO coating variations irradiated in the AGR-1 experiment. The cylindrical specimens were sectioned both transversely and longitudinally, then polished to expose between approximately 40-80 individual particles on each mount. The analysis focused primarily on kernel swelling and porosity, buffer densification and fracturing, buffer-IPyC debonding, and fractures in the IPyC and SiC layers. Characteristic morphologies have been identified, over 800 particles have been classified, and spatial distributions of particle types have been mapped. No significant spatial patterns were discovered in these cross sections. However, some trends were found between morphological types and certain behavioral aspects. Buffer fractures were found in approximately 23% of the particles, and these fractures often resulted in unconstrained kernel swelling into the open cavities. Fractured buffers and buffers that stayed bonded to IPyC layers appear related to larger pore size in kernels. Buffer-IPyC interface integrity evidently factored into initiation of rare IPyC fractures. Fractures through part of the SiC layer were found in only three particles, all in conjunction with IPyC-SiC debonding. Compiled results suggest that the deliberate coating fabrication variations influenced the frequencies of IPyC fractures, IPyC-SiC debonds, and SiC fractures.

  17. Investigation of Heat Generation from Biomass Fuels

    Directory of Open Access Journals (Sweden)

    Naoharu Murasawa

    2015-06-01

    Full Text Available New biomass fuels are constantly being developed from renewable resources in an effort to counter global warming and to create a sustainable society based on recycling. Among these, biomass fuels manufactured from waste are prone to microbial fermentation, and are likely to cause fires and explosions if safety measures, including sufficient risk assessments and long-term storage, are not considered. In this study, we conducted a series of experiments on several types of newly developed biomass fuels, using combinations of various thermal- and gas-analysers, to identify the risks related to heat- and gas-generation. Since a method for the evaluation of the relative risks of biomass fuels is not yet established in Japan, we also such a method based on our experimental results. The present study found that in cases where safety measures are not thoroughly observed, biomass fuels manufactured from waste materials have a higher possibility of combusting spontaneously at the storage site due to microbial fermentation and heat generation.

  18. Fabrication and Comparison of Fuels for Advanced Gas Reactor Irradiation Tests

    Energy Technology Data Exchange (ETDEWEB)

    Jeffrey Phillips; Charles Barnes; John Hunn

    2010-10-01

    As part of the program to demonstrate TRISO-coated fuel for the Next Generation Nuclear Plant, a series of irradiation tests of Advanced Gas Reactor (AGR) fuel are being performed in the Advanced Test Reactor (ATR) at the Idaho National Laboratory. In the first test, called “AGR-1,” graphite compacts containing approximately 300,000 coated particles were irradiated from December 2006 until November 2009. Development of AGR-1 fuel sought to replicate the properties of German TRISO-coated particles. No particle failures were seen in the nearly 3-year irradiation to a burn up of 19%. The AGR-1 particles were coated in a two-inch diameter coater. Following fabrication of AGR-1 fuel, process improvements and changes were made in each of the fabrication processes. Changes in the kernel fabrication process included replacing the carbon black powder feed with a surface-modified carbon slurry and shortening the sintering schedule. AGR-2 TRISO particles were produced in a six-inch diameter coater using a change size about twenty-one times that of the two-inch diameter coater used to coat AGR-1 particles. Changes were also made in the compacting process, including increasing the temperature and pressure of pressing and using a different type of press. Irradiation of AGR-2 fuel began in late spring 2010. Properties of AGR-2 fuel compare favorably with AGR-1 and historic German fuel. Kernels are more homogeneous in shape, chemistry and density. TRISO-particle sphericity, layer thickness standard deviations, and defect fractions are also comparable. In a sample of 317,000 particles from deconsolidated AGR-2 compacts, 3 exposed kernels were found in a leach test. No SiC defects were found in a sample of 250,000 deconsolidated particles, and no IPyC defects in a sample of 64,000 particles. The primary difference in properties between AGR-1 and AGR-2 compacts is that AGR-2 compacts have a higher matrix density, 1.6 g/cm3 compared to about 1.3 g/cm3 for AGR-1 compacts. Based on

  19. Fuel Cycle Research and Development Accident Tolerant Fuels Series 1 (ATF-1) Irradiation Testing FY 2016 Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Core, Gregory Matthew [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    This report contains a summary of irradiation testing of Fuel Cycle Research and Development (FCRD) Accident Tolerant Fuels Series 1 (ATF 1) experiments performed at Idaho National Laboratory (INL) in FY 2016. ATF 1 irradiation testing work performed in FY 2016 included design, analysis, and fabrication of ATF-1B drop in capsule ATF 1 series experiments and irradiation testing of ATF-1 capsules in the ATR.

  20. Ceramographic Examinations of Irradiated AGR-1 Fuel Compacts

    Energy Technology Data Exchange (ETDEWEB)

    Paul Demkowicz; Scott Ploger; John Hunn; Jay S. Kehn

    2012-09-01

    The AGR 1 experiment involved irradiating 72 cylindrical fuel compacts containing tri-structural isotropic (TRISO)-coated particles to a peak burnup of 19.5% fissions per initial metal atom with no in-pile failures observed out of almost 300,000 particles. Six irradiated AGR 1 fuel compacts were selected for microscopy that span a range of irradiation conditions (temperature, burnup, and fast fluence). These six compacts also included all four TRISO coating variations irradiated in the AGR experiment. The six compacts were cross-sectioned both transversely and longitudinally, mounted, ground, and polished after development of careful techniques for preserving particle structures against preparation damage. From 36 to 79 particles within each cross section were exposed near enough to midplane for optical microscopy of kernel, buffer, and coating behavior. The microstructural analysis focused on kernel swelling and porosity, buffer densification and fracture, debonding between the buffer and inner pyrolytic carbon (IPyC) layers, and fractures in the IPyC and SiC layers. Three basic particle morphologies were established according to the extent of bonding between the buffer and IPyC layers: complete debonding along the interface (Type A), no debonding along the interface (Type B), and partial debonding (Type AB). These basic morphologies were subdivided according to whether the buffer stayed intact or fractured. The resulting six characteristic morphologies were used to classify particles within each cross section, but no spatial patterns were clearly observed in any of the cross-sectional morphology maps. Although positions of particle types appeared random within compacts, examining a total of 931 classified particles allowed other relationships among morphological types to be established.

  1. Ceramographic Examinations of Irradiated AGR-1 Fuel Compacts

    Energy Technology Data Exchange (ETDEWEB)

    Paul Demkowicz; Scott Ploger; John Hunn

    2012-05-01

    The AGR 1 experiment involved irradiating 72 cylindrical fuel compacts containing tri-structural isotropic (TRISO)-coated particles to a peak burnup of 19.5% fissions per initial metal atom with no in-pile failures observed out of almost 300,000 particles. Five irradiated AGR 1 fuel compacts were selected for microscopy that span a range of irradiation conditions (temperature, burnup, and fast fluence). These five compacts also included all four TRISO coating variations irradiated in the AGR experiment. The five compacts were cross-sectioned both transversely and longitudinally, mounted, ground, and polished after development of careful techniques for preserving particle structures against preparation damage. Approximately 40 to 80 particles within each cross section were exposed near enough to mid-plane for optical microscopy of kernel, buffer, and coating behavior. The microstructural analysis focused on kernel swelling and porosity, buffer densification and fracture, debonding between the buffer and inner pyrolytic carbon (IPyC) layers, and fractures in the IPyC and SiC layers. Three basic particle morphologies were established according to the extent of bonding between the buffer and IPyC layers: complete debonding along the interface (Type A), no debonding along the interface (Type B), and partial debonding (Type AB). These basic morphologies were subdivided according to whether the buffer stayed intact or fractured. The resulting six characteristic morphologies were used to classify particles within each cross section, but no spatial patterns were clearly observed in any of the cross-sectional morphology maps. Although positions of particle types appeared random within compacts, examining a total of 830 classified particles allowed other relationships among morphological types to be established.

  2. First Results of Scanning Thermal Diffusivity Microscope (STDM) Measurements on Irradiated Monolithic and Dispersion Fuel

    Energy Technology Data Exchange (ETDEWEB)

    T. K. Huber; M. K. Figg; J. R. Kennedy; A. B. Robinson; D. M. Wachs

    2012-07-01

    The thermal conductivity of the fuel material in a reactor before and during irradiation is a sensitive and fundamental parameter for thermal hydraulic calculations that are useds to correctly determine fuel heat fluxes and meat temperatures and to simulate performance of the fuel elements during operation. Several techniques have been developed to measure the thermal properties of fresh fuel to support these calculations, but it is crucial to also investigate the change of thermal properties during irradiation.

  3. Project Progress of New Domestic Zirconium Alloy Fuel Sub-assembly Irradiation

    Institute of Scientific and Technical Information of China (English)

    ZHANG; Ai-min; ZHANG; Pei-sheng; LIU; Jia-zheng; LIU; Wei

    2015-01-01

    At present,the project of new domestic zirconium alloy fuel sub-assembly irradiation is ongoing according to schedule.This paper presents progress of the project such as fuel sub-assembly detailed design,manufacturing process and fuel transportation method.1 Fuel sub-assembly detailed designing

  4. Fabrication of DUPIC fuel for the 5th irradiation test

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Woong Ki; Kim, S. S.; Lee, J. W. [and others

    2004-03-01

    In this study, 10 DUPIC pellets and two mini-elements were fabricated to investigate the thermal characteristics of DUPIC fuel in December 2003. As a result of the experiment, DUPIC pellets with 10.310{approx}10.415 g/cm{sup 3} (95.3{approx}96.3 % of T.D.) of sintered density and less than Ra 0.76 {mu}m of surface roughness satisfying the specifications of DUPIC fuel for the 5th irradiation test have been remotely fabricated at hot cell. 5 DUPIC pellets including 3 pellets equipped with thermal sensor in the center of the pellet were loaded in a mini-element. Endcap welding of the mini-element was performed by Nd:YAG laser. The soundness of the weld of the mini-element has been confirmed by microstructural test, helium leak test, and X-ray inspection. The DUPIC mini-elements assembled in an instrumented rig are under the irradiation at HANARO reactor.

  5. Separation of Plutonium from Irradiated Fuels and Targets

    Energy Technology Data Exchange (ETDEWEB)

    Gray, Leonard W. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Holliday, Kiel S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Murray, Alice [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Thompson, Major [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Thorp, Donald T. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Yarbro, Stephen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Venetz, Theodore J. [Hanford Site, Benton County, WA (United States)

    2015-09-30

    Spent nuclear fuel from power production reactors contains moderate amounts of transuranium (TRU) actinides and fission products in addition to the still slightly enriched uranium. Originally, nuclear technology was developed to chemically separate and recover fissionable plutonium from irradiated nuclear fuel for military purposes. Military plutonium separations had essentially ceased by the mid-1990s. Reprocessing, however, can serve multiple purposes, and the relative importance has changed over time. In the 1960’s the vision of the introduction of plutonium-fueled fast-neutron breeder reactors drove the civilian separation of plutonium. More recently, reprocessing has been regarded as a means to facilitate the disposal of high-level nuclear waste, and thus requires development of radically different technical approaches. In the last decade or so, the principal reason for reprocessing has shifted to spent power reactor fuel being reprocessed (1) so that unused uranium and plutonium being recycled reduce the volume, gaining some 25% to 30% more energy from the original uranium in the process and thus contributing to energy security and (2) to reduce the volume and radioactivity of the waste by recovering all long-lived actinides and fission products followed by recycling them in fast reactors where they are transmuted to short-lived fission products; this reduces the volume to about 20%, reduces the long-term radioactivity level in the high-level waste, and complicates the possibility of the plutonium being diverted from civil use – thereby increasing the proliferation resistance of the fuel cycle. In general, reprocessing schemes can be divided into two large categories: aqueous/hydrometallurgical systems, and pyrochemical/pyrometallurgical systems. Worldwide processing schemes are dominated by the aqueous (hydrometallurgical) systems. This document provides a historical review of both categories of reprocessing.

  6. Internal hydriding in irradiated defected Zircaloy fuel rods: A review (LWBR Development Program)

    Energy Technology Data Exchange (ETDEWEB)

    Clayton, J C

    1987-10-01

    Although not a problem in recent commercial power reactors, including the Shippingport Light Water Breeder Reactor, internal hydriding of Zircaloy cladding was a persistent cause of gross cladding failures during the 1960s. It occurred in the fuel rods of water-cooled nuclear power reactors that had a small cladding defect. This report summarizes the experimental findings, causes, mechanisms, and methods of minimizing internal hydriding in defected Zircaloy-clad fuel rods. Irradiation test data on the different types of defected fuel rods, intentionally fabricated defected and in-pile operationally defected rods, are compared. Significant factors affecting internal hydriding in defected Zircaloy-clad fuel rods (defect hole size, internal and external sources of hydrogen, Zircaloy cladding surface properties, nickel alloy contamination of Zircaloy, the effect of heat flux and fluence) are discussed. Pertinent in-pile and out-of-pile test results from Bettis and other laboratories are used as a data base in constructing a qualitative model which explains hydrogen generation and distribution in Zircaloy cladding of defected water-cooled reactor fuel rods. Techniques for minimizing internal hydride failures in Zircaloy-clad fuel rods are evaluated.

  7. Fabrication of Non-instrumented capsule for DUPIC simulated fuel irradiation test in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, B.G.; Kang, Y.H.; Park, S.J.; Shin, Y.T. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-10-01

    In order to develope DUPIC nuclear fuel, the irradiation test for simulated DUPIC fuel was planed using a non-instrumented capsule in HANARO. Because DUPIC fuel is highly radioactive material the non-instrumented capsule for an irradiation test of simulated DUPIC fuel in HANARO was designed to remotely assemble and disassemble in hot cell. And then, according to the design requirements the non-instrumented DUPIC capsule was successfully manufactured. Also, the manufacturing technologies of the non-instrumented capsule for irradiating the nuclear fuel in HANARO were established, and the basic technology for the development of the instrumented capsule technology was accumulated. This report describes the manufacturing of the non-instrumented capsule for simulated DUPIC fuel. And, this report will be based to develope the instrumented capsule, which will be utilized to irradiate the nuclear fuel in HANARO. 26 refs., 4 figs. (Author)

  8. Gamma-ray spectroscopy on irradiated MTR fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Terremoto, L.A.A. E-mail: laaterre@net.ipen.br; Zeituni, C.A.; Perrotta, J.A.; Silva, J.E.R. da

    2000-08-11

    The availability of burnup data is an important requirement in any systematic approach to the enhancement of safety, economics and performance of a nuclear research reactor. This work presents the theory and experimental techniques applied to determine, by means of nondestructive gamma-ray spectroscopy, the burnup of Material Testing Reactor (MTR) fuel elements irradiated in the IEA-R1 research reactor. Burnup measurements, based on analysis of spectra that result from collimation and detection of gamma-rays emitted in the decay of radioactive fission products, were performed at the reactor pool area. The measuring system consists of a high-purity germanium (HPGe) detector together with suitable fast electronics and an on-line microcomputer data acquisition module. In order to achieve absolute burnup values, the detection set (collimator tube+HPGe detector) was previously calibrated in efficiency. The obtained burnup values are compared with ones provided by reactor physics calculations, for three kinds of MTR fuel elements with different cooling times, initial enrichment grades and total number of fuel plates. Both values show good agreement within the experimental error limits.

  9. Gel-sphere-pac fuel for thermal reactors: assessment of fabrication technology and irradiation performance

    Energy Technology Data Exchange (ETDEWEB)

    Beatty, R.L. Norman, R.E.; Notz, K.J. (comps.)

    1979-11-01

    Recent interest in proliferation-resistant fuel cycles for light-water reactors has focused attention on spiked plutonium and /sup 233/U-Th fuels, requiring remote refabrication. The gel-sphere-pac process for fabricating metal-clad fuel elements has drawn special attention because it involves fewer steps. Gel-sphere-pac fabrication technology involves two major areas: the preparation of fuel spheres of high density and loading these spheres into rods in an efficiently packed geometry. Gel sphere preparation involves three major steps: preparation of a sol or of a special solution (broth), gelation of droplets of sol or broth to give semirigid spheres of controlled size, and drying and sintering these spheres to a high density. Gelation may be accomplished by water extraction (suitable only for sols) or ammonia gelation (suitable for both sols and broths but used almost exclusively with broths). Ammonia gelation can be accomplished either externally, via ammonia gas and ammonium hydroxide, or internally via an added ammonia generator such as hexamethylenetetramine. Sphere-pac fuel rod fabrication involves controlled blending and metering of three sizes of spheres into the rod and packing by low- to medium-energy vibration to achieve about 88% smear density; these sizes have diametral ratios of about 40:10:1 and are blended in size fraction amounts of about 60% coarse, 18% medium, and 22% fine. Irradiation test results indicate that sphere-pac fuel performs at least as well as pellet fuel, and may in fact offer an advantage in significantly reducing mechanical and chemical interaction between the fuel and cladding. The normal feed for gel sphere preparation, heavy metal nitrate solution, is the usual product of fuel reprocessing, so that fabrication of gel spheres performs all the functions performed by both conversion and pellet fabrication in the case of pellet technology.

  10. Structural and mechanical characterization of ion-irradiated glassy polymeric carbon for TRISO fuel nuclear application

    Science.gov (United States)

    Abunaemeh, Malek; Seif, Mohamed; Elsamadicy, Abdalla; Ila, Daryush

    2012-08-01

    Tristructural isotropic (TRISO) fuel is considered as the fuel design of choice for the next generation of nuclear reactors (Generation IV). Its design consists of a fuel kernel of UO x coated with several layers having different functions. One of these functions is a containment shell/diffusion barrier for the fission fragments. Normally, the material of choice for this shell is pyrolytic carbon (PyC). The material does not offer a perfect barrier, due to its inherent crystalline structure, which is planar (like graphite) and therefore impossible to mold in one continuous sheet around the spherical fuel bead. Plane boundaries allow fragment diffusion at a much higher rate than through the plane. In this study, we investigate the possibility of replacing PyC with a different form of carbon, glassy polymeric carbon (GPC). We prepared samples of GPC and studied the evolution of their physical properties and structure as a function of the radiation environment that they were exposed to. The temperature at which the samples were held during irradiation was very similar to the Generation IV nuclear reactor (∼1000°C). During the fission of U235, the fission fragment mass distribution has two maxima around 98 and 137 amu, which would best correspond to elements Rb and Cs, respectively. However, both ions are hard to produce from our SNICS ion source at the Center for Irradiation of Materials; therefore, we used 107Ag and 197Au as best replacements. The irradiation sessions consisted in various fluences of 5 MeV Ag, and 5 MeV Au. For elemental sample analysis, we used transmission electron microscopy. For mechanical analysis, we used nano-indentation. It is of prime importance to measure the penetration of the implanted 107Ag.and 197Au and the evolution of mechanical properties of GPC irradiated with these ions. A procedure for manufacturing GPC with analysis is presented. This will show how the GPC structure differs as the temperature that it is prepared at increases

  11. Performance evaluation of large U-Mo particle dispersed fuel irradiated in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Ho Jin; Park, Jong Man; Oh, Seok Jin; Jang, Se Jung; Yu, Byung Ok; Lee, Choong Seong; Seo, Chul Gyo; Chae, Hee Taek; Kim, Chang Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-11-15

    U-Mo/Al dispersion fuel is being developed as advanced fuel for research reactors. Irradiation behavior of U-Mo/Al dispersion fuel has been studied to evaluate its fuel performance. One of the performance limiting factors is a chemical interaction between the U-Mo particle and the Al matrix because the thermal conductivity of fuel meat is decreased with the interaction layer growth. In order to overcome the interaction problem, large-sized U-Mo particles were fabricated by controlling the centrifugal atomization conditions. The fuel performance behavior of U-Mo/Al dispersion fuel was estimated by using empirical models formulated based on the microstructural analyses of the post-irradiation examination (PIE) on U-Mo/Al dispersion fuel irradiated in HANARO reactor. Temperature histories of U-Mo/Al dispersion fuel during irradiation tests were estimated by considering the effect of an interaction layer growth on the thermal conductivity of the fuel meat. When the fuel performances of the dispersion fuel rods containing U-Mo particles with various sizes were compared, fuel temperature was decreased as the average U-Mo particles with various sizes were compared, fuel temperature was decreased as the average U-Mo particle size was increases. It was found that the dispersion of a larger U-Mo particle was effective for mitigating the thermal degradation which is associated with an interaction layer growth.

  12. A liquid-fueled electrochemical generator

    Energy Technology Data Exchange (ETDEWEB)

    Yanagikhara, N.; Manadbe, K.

    1983-04-21

    A mixture of fuel and the electrolyte is circulated in the electrochemical generator (EKhG). Electrodes are installed in the circulation system which serve as sensors of the fuel concentration in the electrolyte. The sensors are placed in the TEZ alongside the current outlet anode. The potential of the sensor is identical to the potential of the electrode with which it is connected. The supply of fuel from the tank into the tank with the electrolyte is automatically regulated by a signal from the sensors. A tank is installed between the sensors and the current outlet cathode in the circulation system which is designed for interrupting the electric circuit which is formed as a result of the electrically conducting liquid connection. Substantial current leaks occur in this circuit. In the tank the liquid is fed upward and using different atomization methods, the continuous stream is transformed into individual drops. Falling to the bottom, the drops run together and are discharged from the tank in the form of a continuous jet. Current leaks through the circulation system and the formation of short circuits (KZ) is prevented in the electrochemical generator.

  13. Development of a hybrid photovoltaic-liquid fueled thermoelectric generator for Arctic locations

    Energy Technology Data Exchange (ETDEWEB)

    Kolb, H. (Global Thermoelectric Power Systems Ltd., Bassano, AB (Canada))

    1988-08-01

    The solar irradiation levels in arctic and antarctic regions vary dramatically from summer to winter. It was the objective of this project to develop a photovoltaic-liquid fueled thermoelectric hybrid power system that will take advantage of the available solar irradiation during the period during which the levels are high and switch to a liquid fueled thermoelectric generator during periods when the solar irradiation levels are low. In addition, the system is to provide heating to keep electronics and batteries above a preset minimum temperature. A remote start feature was designed and built into an existing liquid fueled thermoelectric generator. A prototype system was then assembled with a panel factor of about 4.88. Arctic summer conditions of solar irradiation were simulated by adjustment of the panel tilt angle. The performance of the liquid fueled generator was disappointing, numerous failures of the generator were a major impediment to the complete success of the project. It was found that the panel factor should be increased by about 15 to 20% and that the constant voltage battery recharge method is not efficient for this type of system. A cost comparison of the hybrid versus two other alternative remote power systems indicates that it is a cost-effective system. 2 refs., 9 figs., 1 tab.

  14. Metal fuel development and verification for prototype generation- IV Sodium- Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Bock; Cheon, Jin Sik; Kim, Sung Ho; Park, Jeong Yong; Joo, Hyung Kook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR) to be built by 2028. U-Zr fuel is a driver for the initial core of the PGSFR, and U -transuranics (TRU)-Zr fuel will gradually replace U-Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U-Zr fuel, work on U-Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U-TRU-Zr fuel uses TRU recovered through pyroelectrochemical processing of spent PWR (pressurized water reactor) fuels, which contains highly radioactive minor actinides and chemically active lanthanide or rare earth elements as carryover impurities. An advanced fuel slug casting system, which can prevent vaporization of volatile elements through a control of the atmospheric pressure of the casting chamber and also deal with chemically active lanthanide elements using protective coatings in the casting crucible, was developed. Fuel cladding of the ferritic-martensitic steel FC92, which has higher mechanical strength at a high temperature than conventional HT9 cladding, was developed and fabricated, and is being irradiated in the fast reactor.

  15. COMPARATIVE ANALYSIS OF STRUCTURAL CHANGES IN U-MO DISPERSED FUEL OF FULL-SIZE FUEL ELEMENTS AND MINI-RODS IRRADIATED IN THE MIR REACTOR

    Directory of Open Access Journals (Sweden)

    ALEKSEY. L. IZHUTOV

    2013-12-01

    The full-size fuel rods were irradiated up to an average burnup of ∼ 60%235U; the mini-rods were irradiated to an average burnup of ∼ 85%235U. 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%.

  16. Photocatalytic fuel cell (PFC) and dye self-photosensitization photocatalytic fuel cell (DSPFC) with BiOCl/Ti photoanode under UV and visible light irradiation.

    Science.gov (United States)

    Li, Kan; Xu, Yunlan; He, Yi; Yang, Chen; Wang, Yalin; Jia, Jinping

    2013-04-02

    A fuel cell that functioned as a photo fuel cell (PFC) when irradiated with UV light and as a dye self-photosensitization photo fuel cell (DSPFC) when irradiated with visible light was proposed and investigated in this study. The system included a BiOCl/Ti plate photoanode and a Pt cathode, and dye solutions were employed as fuel. Electricity was generated at the same time as the dyes were degraded. 26.2% and 24.4% Coulombic efficiency were obtained when 20 mL of 10 mg · L(-1) Rhodamine B solution was treated with UV for 2 h and visible light for 3 h, respectively. Irradiation with natural and artificial sunlight was also evaluated. UV and visible light could be utilized at the same time and the photogenerated current was observed. The mechanism of electricity generation in BiOCl/Ti PFC and DSPFC was studied through degradation of the colorless salicylic acid solution. Factors that affect the electricity generation and dye degradation performance, such as solution pH and cathode material, were also investigated and optimized.

  17. Segregated exhaust SOFC generator with high fuel utilization capability

    Science.gov (United States)

    Draper, Robert; Veyo, Stephen E.; Kothmann, Richard E.

    2003-08-26

    A fuel cell generator contains a plurality of fuel cells (6) in a generator chamber (1) and also contains a depleted fuel reactor or a fuel depletion chamber (2) where oxidant (24,25) and fuel (81) is fed to the generator chamber (1) and the depleted fuel reactor chamber (2), where both fuel and oxidant react, and where all oxidant and fuel passages are separate and do not communicate with each other, so that fuel and oxidant in whatever form do not mix and where a depleted fuel exit (23) is provided for exiting a product gas (19) which consists essentially of carbon dioxide and water for further treatment so that carbon dioxide can be separated and is not vented to the atmosphere.

  18. Irradiation behavior of modified high-performance nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Jungwirth, Rainer

    2011-11-03

    To overcome the problem of UMo/Al fuel swelling, four different possibilities have been identified: (i) the modification of the Al matrix by adding diffusion limiting elements (ii) the insertion of a diffusion barrier at the interface UMo-Al (iii) further alloying the UMo with a third element to stabilize the γ-UMo phase (iv) a combination of means (i)-(iii). In consequence, 20 different UMoX/AlY (X=Si, Ti, Mg, Bi, with and without oxidation layer; Y=Nb, Ti, Pt) samples have been examined before and after irradiation with Iodine at 80MeV. First it has been shown, that a protective oxidation layer on the UMo grains does not prevent the formation of a interdiffusion layer. In contrast, additions to the Al matrix can be reduced to the self-acting formation of a protective layer at the UMo/Al interface. Additions to the UMo to stabilize the γ-UMo upon heating are of minor importance since irradiation reverses the phase decomposition of UMo.

  19. Hydrogen generation from biogenic and fossil fuels by autothermal reforming

    Science.gov (United States)

    Rampe, Thomas; Heinzel, Angelika; Vogel, Bernhard

    Hydrogen generation for fuel cell systems by reforming technologies from various fuels is one of the main fields of investigation of the Fraunhofer ISE. Suitable fuels are, on the one hand, gaseous hydrocarbons like methane, propane but also, on the other hand, liquid hydrocarbons like gasoline and alcohols, e.g., ethanol as biogenic fuel. The goal is to develop compact systems for generation of hydrogen from fuel being suitable for small-scale membrane fuel cells. The most recent work is related to reforming according to the autothermal principle — fuel, air and steam is supplied to the reactor. Possible applications of such small-scale autothermal reformers are mobile systems and also miniature fuel cell as co-generation plant for decentralised electricity and heat generation. For small stand-alone systems without a connection to the natural gas grid liquid gas, a mixture of propane and butane is an appropriate fuel.

  20. Colloids generation from metallic uranium fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-07-20

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

  1. Post irradiation examination of HANARO nucler mini-element fuel (metallographic and density test)

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Byung Ok; Hong, K. P.; Park, D. G.; Choo, Y. S.; Baik, S. J.; Kim, K. H.; Kim, H. C.; Jung, Y. H

    2001-05-01

    The post irradiation examination of a HANARO mini-element nuclear fuel, KH96C-004, was done in June 6, 2000. The purpose of this project is to evaluate the in-core performance and reliability of mini-element nuclear fuel for HANARO developed by the project ''The Nuclear Fuel Material Development of Research Reactor''. And, in order to examine the performance of mini-element nuclear fuel in normal output condition, the post irradiation examination of a nuclear fuel bundle composed by 6 mini nuclear fuel rods and 12 dummy fuel rods was performed. Based on these examination results, the safety and reliability of HANARO fuel and the basic data on the design of HANARO nuclear fuel can be ensured and obtained,.

  2. Remote fabrication and irradiation test of recycled nuclear fuel prepared by the oxidation and reduction of spent oxide fuel

    Science.gov (United States)

    Jin Ryu, Ho; Chan Song, Kee; Il Park, Geun; Won Lee, Jung; Seung Yang, Myung

    2005-02-01

    A direct dry recycling process was developed in order to reuse spent pressurized light water reactor (LWR) nuclear fuel in CANDU reactors without the separation of sensitive nuclear materials such as plutonium. The benefits of the dry recycling process are the saving of uranium resources and the reduction of spent fuel accumulation as well as a higher proliferation resistance. In the process of direct dry recycling, fuel pellets separated from spent LWR fuel rods are oxidized from UO2 to U3O8 at 500 °C in an air atmosphere and reduced into UO2 at 700 °C in a hydrogen atmosphere, which is called OREOX (oxidation and reduction of oxide fuel). The pellets are pulverized during the oxidation and reduction processes due to the phase transformation between cubic UO2 and orthorhombic U3O8. Using the oxide powder prepared from the OREOX process, the compaction and sintering processes are performed in a remote manner in a shielded hot cell due to the high radioactivity of the spent fuel. Most of the fission gas and volatile fission products are removed during the OREOX and sintering processes. The mini-elements fabricated by the direct dry recycling process are irradiated in the HANARO research reactor for the performance evaluation of the recycled fuel pellets. Post-irradiation examination of the irradiated fuel showed that microstructural evolution and fission gas release behavior of the dry-recycled fuel were similar to high burnup UO2 fuel.

  3. Propagation of Plasma Generated by Intense Pulsed Ion Beam Irradiation

    Institute of Scientific and Technical Information of China (English)

    WU Di; GONG Ye; LIU Jin-Yuan; WANG Xiao-Gang; LIU Yue; MA Teng-Cai

    2006-01-01

    @@ Taking the calculation results based on the established two-dimensional ablation model of the intense-pulsed-ion-beam (IPIB) irradiation process as initial conditions, we build a two-dimensional hydrodynamic ejection model of plasma produced by an IPIB-irradiated metal titanium target into ambient gas. We obtain the conclusions that shock waves generate when the background pressure is around 133 mTorr and also obtain the plume splitting phenomenon that has been observed in the experiments.

  4. Performance evaluation of large U-Mo particle dispersed fuel irradiated in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Ho Jin; Park, Jong Man; Oh, Seok Jin; Jang, Se Jung; Yu, Byung Ok; Lee, Choong Seong; Seo, Chul Gyo; Chae, Hee Taek; Kim, Chang Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-11-15

    U-Mo/Al dispersion fuel is being developed as advanced fuel for research reactors. Irradiation behavior of U-Mo/Al dispersion fuel has been studied to evaluate its fuel performance. One of the performance limiting factors is a chemical interaction between the U-Mo particle and the Al matrix because the thermal conductivity of fuel meat is decreased with the interaction layer growth. In order to overcome the interaction problem, large-sized U-Mo particles were fabricated by controlling the centrifugal atomization conditions. The fuel performance behavior of U-Mo/Al dispersion fuel was estimated by using empirical models formulated based on the microstructural analyses of the post-irradiation examination (PIE) on U-Mo/Al dispersion fuel irradiated in HANARO reactor. Temperature histories of U-Mo/Al dispersion fuel during tests were estimated by considering the effect of an interaction layer growth on the thermal conductivity of the fuel meat. When the fuel performances of the dispersion fuel rods containing U-Mo particles were compared, fuel temperature was decreased as the average U-Mo particle size was increased. It was found that the dispersion of a larger U-Mo particle was effective for mitigating the thermal degradation which is associated with an interaction layer growth.

  5. A new code for predicting the thermo-mechanical and irradiation behavior of metallic fuels in sodium fast reactors

    Science.gov (United States)

    Karahan, Aydın; Buongiorno, Jacopo

    2010-01-01

    An engineering code to predict the irradiation behavior of U-Zr and U-Pu-Zr metallic alloy fuel pins and UO2-PuO2 mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named Fuel Engineering and Structural analysis Tool (FEAST). FEAST has several modules working in coupled form with an explicit numerical algorithm. These modules describe fission gas release and fuel swelling, fuel chemistry and restructuring, temperature distribution, fuel-clad chemical interaction, and fuel and clad mechanical analysis including transient creep-fracture for the clad. Given the fuel pin geometry, composition and irradiation history, FEAST can analyze fuel and clad thermo-mechanical behavior at both steady-state and design-basis (non-disruptive) transient scenarios. FEAST was written in FORTRAN-90 and has a simple input file similar to that of the LWR fuel code FRAPCON. The metal-fuel version is called FEAST-METAL, and is described in this paper. The oxide-fuel version, FEAST-OXIDE is described in a companion paper. With respect to the old Argonne National Laboratory code LIFE-METAL and other same-generation codes, FEAST-METAL emphasizes more mechanistic, less empirical models, whenever available. Specifically, fission gas release and swelling are modeled with the GRSIS algorithm, which is based on detailed tracking of fission gas bubbles within the metal fuel. Migration of the fuel constituents is modeled by means of thermo-transport theory. Fuel-clad chemical interaction models based on precipitation kinetics were developed for steady-state operation and transients. Finally, a transient intergranular creep-fracture model for the clad, which tracks the nucleation and growth of the cavities at the grain boundaries, was developed for and implemented in the code. Reducing the empiricism in the constitutive models should make it more acceptable to extrapolate FEAST-METAL to new fuel compositions and higher burnup, as envisioned in advanced sodium reactors

  6. A new code for predicting the thermo-mechanical and irradiation behavior of metallic fuels in sodium fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Karahan, Aydin, E-mail: karahan@mit.ed [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology (United States); Buongiorno, Jacopo [Center for Advanced Nuclear Energy Systems, Nuclear Science and Engineering Department, Massachusetts Institute of Technology (United States)

    2010-01-31

    An engineering code to predict the irradiation behavior of U-Zr and U-Pu-Zr metallic alloy fuel pins and UO{sub 2}-PuO{sub 2} mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named Fuel Engineering and Structural analysis Tool (FEAST). FEAST has several modules working in coupled form with an explicit numerical algorithm. These modules describe fission gas release and fuel swelling, fuel chemistry and restructuring, temperature distribution, fuel-clad chemical interaction, and fuel and clad mechanical analysis including transient creep-fracture for the clad. Given the fuel pin geometry, composition and irradiation history, FEAST can analyze fuel and clad thermo-mechanical behavior at both steady-state and design-basis (non-disruptive) transient scenarios. FEAST was written in FORTRAN-90 and has a simple input file similar to that of the LWR fuel code FRAPCON. The metal-fuel version is called FEAST-METAL, and is described in this paper. The oxide-fuel version, FEAST-OXIDE is described in a companion paper. With respect to the old Argonne National Laboratory code LIFE-METAL and other same-generation codes, FEAST-METAL emphasizes more mechanistic, less empirical models, whenever available. Specifically, fission gas release and swelling are modeled with the GRSIS algorithm, which is based on detailed tracking of fission gas bubbles within the metal fuel. Migration of the fuel constituents is modeled by means of thermo-transport theory. Fuel-clad chemical interaction models based on precipitation kinetics were developed for steady-state operation and transients. Finally, a transient intergranular creep-fracture model for the clad, which tracks the nucleation and growth of the cavities at the grain boundaries, was developed for and implemented in the code. Reducing the empiricism in the constitutive models should make it more acceptable to extrapolate FEAST-METAL to new fuel compositions and higher burnup, as envisioned in advanced sodium

  7. Changes to Irradiation Conditions of VVER-1000 Surveillance Specimens Resulting from Fuel Assemblies with Greater Fuel Height

    Directory of Open Access Journals (Sweden)

    Panferov Pavel

    2016-01-01

    Full Text Available The goal of the work was to obtain experimental data on the influence of newtype fuel assemblies with higher fuel rods on the irradiation conditions of surveillance specimens installed on the baffe of VVER-1000. For this purpose, two surveillance sets with container assemblies of the same design irradiated in reactors with different fuel assemblies in the core were investigated. Measurements of neutron dosimeters from these sets and retrospective measurements of 54Mn activity accumulated in each irradiated specimen allow a detailed distribution of the fast neutron flux in the containers to be obtained. Neutron calculations have been done using 3D discrete ordinate code KATRIN. On the basis of the obtained results, a change of the lead factor due to newtype fuel assemblies was evaluated for all types of VVER-1000 container assemblies.

  8. Changes to Irradiation Conditions of VVER-1000 Surveillance Specimens Resulting from Fuel Assemblies with Greater Fuel Height

    Science.gov (United States)

    Panferov, Pavel; Kochkin, Viacheslav; Erak, Dmitry; Makhotin, Denis; Reshetnikov, Alexandr; Timofeev, Andrey

    2016-02-01

    The goal of the work was to obtain experimental data on the influence of newtype fuel assemblies with higher fuel rods on the irradiation conditions of surveillance specimens installed on the baffe of VVER-1000. For this purpose, two surveillance sets with container assemblies of the same design irradiated in reactors with different fuel assemblies in the core were investigated. Measurements of neutron dosimeters from these sets and retrospective measurements of 54Mn activity accumulated in each irradiated specimen allow a detailed distribution of the fast neutron flux in the containers to be obtained. Neutron calculations have been done using 3D discrete ordinate code KATRIN. On the basis of the obtained results, a change of the lead factor due to newtype fuel assemblies was evaluated for all types of VVER-1000 container assemblies.

  9. Simulation of the irradiation behaviour of the PBMR fuel in the SAFARI-1 reactor / B.M. Makgopa

    OpenAIRE

    2009-01-01

    Irradiation experiments for the pebble bed modular reactor PBMR fuel (coated fuel particles and pebble fuel) are planned at the South African First Atomic Reactor Installation (SAFARI-1). The experiments are conducted to investigate the behavior of the fuel under normal operating and accelerated/accident simulating conditions because the safe operation of the reactor relies on the integrity of the fuel for retention of radioactivity. For fuel irradiation experiments, the accura...

  10. Design of an experiment to measure the decay heat of an irradiated PWR fuel: MERCI experiment; Conception d'une experience de mesure de la puissance residuelle d'un combustible irradie: l'experience MERCI

    Energy Technology Data Exchange (ETDEWEB)

    Bourganel, St

    2002-11-01

    After a reactor shutdown, a significant quantity of energy known as 'decay heat' continues to be generated from the irradiated fuel. This heat source is due to the disintegration energy of fission products and actinides. Decay heat determination of an irradiated fuel is of the utmost importance for safety analysis as the design cooling systems, spent fuel transport, or handling. Furthermore, the uncertainty on decay heat has a straight economic impact. The unloading fuel spent time is an example. The purpose of MERCI experiment (irradiated fuel decay heat measurement) consists in qualifying computer codes, particularly the DARWIN code system developed by the CEA in relation to industrial organizations, as EDF, FRAMATOME and COGEMA. To achieve this goal, a UOX fuel is irradiated in the vicinity of the OSIRIS research reactor, and then the decay heat is measured by using a calorimeter. The objective is to reduce the decay heat uncertainties from 8% to 3 or 4% at short cooling times. A full simulation on computer of the MERCI experiment has been achieved: fuel irradiation analysis is performed using transport code TRIPOLI4 and evolution code DARWIN/PEPIN2, and heat transfer with CASTEM2000 code. The results obtained are used for the design of this experiment. Moreover, we propose a calibration procedure decreasing the influence of uncertainty measurements and an interpretation method of the experimental results and evaluation of associated uncertainties. (author)

  11. Nuclear power generation and fuel cycle report 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-10-01

    This report presents the current status and projections through 2015 of nuclear capacity, generation, and fuel cycle requirements for all countries using nuclear power to generate electricity for commercial use. It also contains information and forecasts of developments in the worldwide nuclear fuel market. Long term projections of U.S. nuclear capacity, generation, and spent fuel discharges for two different scenarios through 2040 are developed. A discussion on decommissioning of nuclear power plants is included.

  12. Jules Horowitz Reactor Project- Fuel irradiation device, innovative instrumentation proposal for experimental phenomena real time measurement

    Energy Technology Data Exchange (ETDEWEB)

    Gaillot, Stephane; Cheymol, Guy [CEA, Paris (France)

    2013-07-01

    The fuel irradiation devices used for the tests or rods allow reproducing at small scales the conditions of the studied nuclear reactors (as LWR type). During the irradiation phase, the tested fuel rod can be stressed due to thermal, mechanical, nuclear effects which can modify its geometry (dilatation, swelling effects). After the test, the return to normal conditions can have as consequence the disappearance of the phenomenon or give access to partial information (final deformation). Generally, to follow the phenomena related to the irradiation phase, the experimental rod contained in the test device is instrumented with thermocouples and LVDT. As complement of this instrumentation, new sensors using innovating technologies are studied (deformation sensor integrating optical fibres). Through the example of a fuel irradiation device foreseen for the JHR, this paper aims to describe the present development of an innovating instrumentation with the objective to measure, in real time and under flux, the fuel rod deformation phenomena during a ramp test.

  13. Characterization of LWRS Hybrid SiC-CMC-Zircaloy-4 Fuel Cladding after Gamma Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Isabella J van Rooyen

    2012-09-01

    The purpose of the gamma irradiation tests conducted at the Idaho National Laboratory (INL) was to obtain a better understanding of chemical interactions and potential changes in microstructural properties of a mock-up hybrid nuclear fuel cladding rodlet design (unfueled) in a simulated PWR water environment under irradiation conditions. The hybrid fuel rodlet design is being investigated under the Light Water Reactor Sustainability (LWRS) program for further development and testing of one of the possible advanced LWR nuclear fuel cladding designs. The gamma irradiation tests were performed in preparation for neutron irradiation tests planned for a silicon carbide (SiC) ceramic matrix composite (CMC) zircaloy-4 (Zr-4) hybrid fuel rodlet that may be tested in the INL Advanced Test Reactor (ATR) if the design is selected for further development and testing

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

    Science.gov (United States)

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

    2011-12-01

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

  15. 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.

  16. Plan and safety analysis on the high power irradiation test program of full length fuel element for Hanaro

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Y.S.; Kim, C.K.; Park, H.D.; Kim, K.H.; Park, J.M.; Lee, D.B.; Kim, J.D.; Ko, Y.M.; Jang, S.J.; Ahn, H.S.; Woo, Y.M.; Kim, E.S.; Kim, H.R.; Chae, H.T.; Lee, C.S

    1999-06-01

    The advanced research reactor fuel development project has been carried out for a localization of HANARO nuclear fuels. The design and fabrication technologies of the localized fuel are almost developed, and the quality assurance procedure and assessment criteria were established. The characteristics of the fuel fabricated in KAERI were investigated through out-pile test. In order to verify the localized fuel performance, irradiation test plan of the developed fuel has been worked out. It consists of 3 stages. The 1st stage is normal power irradiation test and the final burn-up of the test fuel was supposed to be 85 at%. The fuel has been successfully irradiated until now and will be unloaded in June. The 2nd irradiation test will be done to confirm the fuel performance and to get the in-pile data under the high neutron flux level. This test fuel is identical with the 36-element fuel assembly. After the 1st and 2nd irradiation tests are completed with acceptable results, the 3rd irradiation test of final stage will be carried out as a demonstration. In this report, the results of the 1st irradiation test is introduced. Then the objectives, schedule and test condition, the design documents of fuel elements and bundle, the methods of fabrication, out-pile test results, post-irradiation examination scheme, calculation of linear power distribution, and safety analysis results for the 2nd irradiation test bundle are described. (author). 2 refs., 14 tabs., 12 figs.

  17. Fuel cycle comparison of distributed power generation technologies.

    Energy Technology Data Exchange (ETDEWEB)

    Elgowainy, A.; Wang, M. Q.; Energy Systems

    2008-12-08

    The fuel-cycle energy use and greenhouse gas (GHG) emissions associated with the application of fuel cells to distributed power generation were evaluated and compared with the combustion technologies of microturbines and internal combustion engines, as well as the various technologies associated with grid-electricity generation in the United States and California. The results were primarily impacted by the net electrical efficiency of the power generation technologies and the type of employed fuels. The energy use and GHG emissions associated with the electric power generation represented the majority of the total energy use of the fuel cycle and emissions for all generation pathways. Fuel cell technologies exhibited lower GHG emissions than those associated with the U.S. grid electricity and other combustion technologies. The higher-efficiency fuel cells, such as the solid oxide fuel cell (SOFC) and molten carbonate fuel cell (MCFC), exhibited lower energy requirements than those for combustion generators. The dependence of all natural-gas-based technologies on petroleum oil was lower than that of internal combustion engines using petroleum fuels. Most fuel cell technologies approaching or exceeding the DOE target efficiency of 40% offered significant reduction in energy use and GHG emissions.

  18. Pigmi mechanical fabrication. [Pion Generator for Medical Irradiations (PIGMI)

    Energy Technology Data Exchange (ETDEWEB)

    Hart, V.E.

    1976-01-01

    A prime goal of the mechanical design effort associated with the PIGMI (Pion Generator for Medical Irradiations) program is to investigate new materials and fabrication techniques in an effort to obtain increased machine efficiency and reliability at a reasonable cost. A discussion is given dealing with the modeling program that LASL is pursuing for 450-MHz and 1350-MHz PIGMI development.

  19. Metallography and fuel cladding chemical interaction in fast flux test facility irradiated metallic U-10Zr MFF-3 and MFF-5 fuel pins

    Energy Technology Data Exchange (ETDEWEB)

    Carmack, W. J.; Chichester, H. M.; Porter, D. L.; Wootan, D. W.

    2016-05-01

    Abstract The Mechanistic Fuel Failure (MFF) series of metal fuel irradiations conducted in the Fast Flux Test Facility (FFTF) provides an important potential comparison between data generated in the Experimental Breeder Reactor (EBR-II) and that expected in a larger-scale fast reactor. The irradiations were the beginning tests to qualify U-10wt%Zr as a driver fuel for FFTF. The FFTF core, with a 91.4 cm tall fuel column and a chopped cosine neutron flux profile, operated with a peak cladding temperature at the top of the fuel column, but developed peak burnup at the centerline of the core. This places the peak fuel temperature midway between the core center and the top of fuel, lower in the fuel column than in previous EBR-II experiments that had a 32-cm height core. The MFF-3 and MFF-5 qualification assemblies operated in FFTF to >10 at% burnup, and performed very well with no cladding breaches. The MFF-3 assembly operated to 13.8 at% burnup with a peak inner cladding temperature of 643°C, and the MFF-5 assembly operated to 10.1 at% burnup with a peak inner cladding temperature of 651°C. Because of the very high operating temperatures for both the fuel and the cladding, data from the MFF assemblies are most comparable to the data obtained from the EBR-II X447 experiment, which experienced two pin breaches. The X447 breaches were strongly influenced by a large amount of fuel/cladding chemical interaction (FCCI). The MFF pins benefitted from different axial locations of high burnup and peak cladding temperature, which helped to reduce interdiffusion between rare earth fission products and stainless steel cladding. Post-irradiation examination evidence illustrates this advantage. Comparing other performance data of the long MFF pins to prior EBR-II test data, the MFF fuel inside the cladding grew less axially, and the gas release data did not reveal a definitive difference.

  20. Project of a new circuit for nuclear fuel irradiation; Projeto de um novo circuito para irradiacao de combustiveis nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Zeituni, Carlos A.; Terremoto, Luis A.A.; Perrotta, Jose A.; Silva, Jose E.R. da [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear. Div. de Engenharia do Combustivel. E-mail: czeituni@usp.br

    2000-07-01

    This paper reports information about the operation of the old Irradiated Fuel Assembly for nuclear miniplates irradiation in the reactor IEA-R1, named CICON (Circuit for Nuclear Fuels Irradiation), and presents the project of the new one. This paper also describes the problems of the old capsule and which details we will change in the new project. (author)

  1. The 4th irradiation test of dry process fuel in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Lee, C. Y.; Moon, J. S.; Kang, K. H.; Jung, I. H.; Song, K. C.; Yang, M. S. [KAERI, Taejon (Korea, Republic of)

    2002-10-01

    The 4th irradiation test of dry process pellet using non-instrumented rig is being performed in the HANARO research reactor. Among the three mini-elements for the 4{sup th} irradiation test, the element No.1 is dedicated to the extended irradiation of the DUPIC pellets irradiated in the 3{sup rd} irradiation test, the element No.2 and No.3 are used for the comparative analysis on the in-core behaviors of simulated DUPIC fuel and actual DUPIC fuel. For these purposes, the irradiated rig of the 3{sup rd} irradiation test was disassembled in a hot cell to select the element No.1. Also the SIMFUEL that is fabricated in the DUPIC laboratory is welded by laser in a welding chamber and the DUPIC fuel that is remotely fabricated in DFDF is welded by a laser method as done in the second and third irradiation tests. The rig was remotely assembled using a rig assembler and loaded into the OR5 hole. Since June 2002, the 4{sup th} irradiation test is being performed.

  2. 10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

    Science.gov (United States)

    2010-01-01

    ... fuel and nuclear waste. 71.97 Section 71.97 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PACKAGING... notification of shipment of irradiated reactor fuel and nuclear waste. (a) As specified in paragraphs (b), (c... advance notification of transportation of nuclear waste was published in the Federal Register on June...

  3. Post-irradiation data on fuel elements from KER Loop 4

    Energy Technology Data Exchange (ETDEWEB)

    Bennett, E.C.

    1963-01-10

    Fourteen NAE1 fuel elements were discharged from KER Loop-4, after irradiation to an average exposure of 1250 MWD, at prototype N-Reactor coolant temperature and pressure. The elements were disassembled and measured in the KE fuel examination facility. This report includes all measurements, except the profilometer data.

  4. Status of advanced fuel candidates for Sodium Fast Reactor within the Generation IV International Forum

    Science.gov (United States)

    Delage, F.; Carmack, J.; Lee, C. B.; Mizuno, T.; Pelletier, M.; Somers, J.

    2013-10-01

    The main challenge for fuels for future Sodium Fast Reactor systems is the development and qualification of a nuclear fuel sub-assembly which meets the Generation IV International Forum goals. The Advanced Fuel project investigates high burn-up minor actinide bearing fuels as well as claddings and wrappers to withstand high neutron doses and temperatures. The R&D outcome of national and collaborative programs has been collected and shared between the AF project members in order to review the capability of sub-assembly material and fuel candidates, to identify the issues and select the viable options. Based on historical experience and knowledge, both oxide and metal fuels emerge as primary options to meet the performance and the reliability goals of Generation IV SFR systems. There is a significant positive experience on carbide fuels but major issues remain to be overcome: strong in-pile swelling, atmosphere required for fabrication as well as Pu and Am losses. The irradiation performance database for nitride fuels is limited with longer term R&D activities still required. The promising core material candidates are Ferritic/Martensitic (F/M) and Oxide Dispersed Strengthened (ODS) steels.

  5. AC power generation from microbial fuel cells

    Science.gov (United States)

    Lobo, Fernanda Leite; Wang, Heming; Forrestal, Casey; Ren, Zhiyong Jason

    2015-11-01

    Microbial fuel cells (MFCs) directly convert biodegradable substrates to electricity and carry good potential for energy-positive wastewater treatment. However, the low and direct current (DC) output from MFC is not usable for general electronics except small sensors, yet commercial DC-AC converters or inverters used in solar systems cannot be directly applied to MFCs. This study presents a new DC-AC converter system for MFCs that can generate alternating voltage in any desired frequency. Results show that AC power can be easily achieved in three different frequencies tested (1, 10, 60 Hz), and no energy storage layer such as capacitors was needed. The DC-AC converter efficiency was higher than 95% when powered by either individual MFCs or simple MFC stacks. Total harmonic distortion (THD) was used to investigate the quality of the energy, and it showed that the energy could be directly usable for linear electronic loads. This study shows that through electrical conversion MFCs can be potentially used in household electronics for decentralized off-grid communities.

  6. Generator gas as a fuel to power a diesel engine

    Directory of Open Access Journals (Sweden)

    Tutak Wojciech

    2014-01-01

    Full Text Available The results of gasification process of dried sewage sludge and use of generator gas as a fuel for dual fuel turbocharged compression ignition engine are presented. The results of gasifying showed that during gasification of sewage sludge is possible to obtain generator gas of a calorific value in the range of 2.15  2.59 MJ/m3. It turned out that the generator gas can be effectively used as a fuel to the compression ignition engine. Because of gas composition, it was possible to run engine with partload conditions. In dual fuel operation the high value of indicated efficiency was achieved equal to 35%, so better than the efficiency of 30% attainable when being fed with 100% liquid fuel. The dual fuel engine version developed within the project can be recommended to be used in practice in a dried sewage sludge gasification plant as a dual fuel engine driving the electric generator loaded with the active electric power limited to 40 kW (which accounts for approx. 50% of its rated power, because it is at this power that the optimal conditions of operation of an engine dual fuel powered by liquid fuel and generator gas are achieved. An additional advantage is the utilization of waste generated in the wastewater treatment plant.

  7. Continuous dissolution of irradiated nuclear fuels; Dissolution continue des combustibles nucleaires irradies

    Energy Technology Data Exchange (ETDEWEB)

    Michel, P.; Talmont, X.; Tarnerq, M. [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1967-07-01

    In the case of the continuous dissolution of nuclear fuels, the equations for the calculation of the fuel concentration of the solution flowing out of a pot dissolver have been written. Nitric acid feed flow rates have been calculated in order to obtain an adjusted solution when starting or stopping a dissolution, or when changing the number of rods introduced per hour. Then some transient states brought on by perturbations, have been studied: a) sudden change in nitric acid flow rate; b) continuous drift of the latter; c) sudden change in nitric acid feed concentration; d) transition from a fuel concentration to another by changing the flow rate of nitric acid feed. It has been shown that some transient states cannot be solved with general equations. Computer calculation programs would be probably more useful. (authors) [French] L'etude se rapporte a la dissolution dans l'acide nitrique des combustibles nucleaires irradies, en vue de la recuperation de la matiere fissile qu'ils contiennent. On a etabli, dans le cas de la dissolution continue, les differentes equations permettant le calcul de la concentration en combustible a la sortie d'un dissolveur du type 'marmite'. On a etudie les regimes du debit d'alimentation en acide nitrique a imposer lors du demarrage, de l'arret d'une dissolution, ou lors d'un changement de cadence d'introduction des barreaux, de facon a obtenir une solution ajustee. On a etudie ensuite differents regimes transitoires consecutifs a des perturbations: changement brusque du debit d'acide d'alimentation, derive continue de ce debit, changement brusque de la concentration de l'acide d'alimentation, passage d'une concentration en combustible a une autre par changement du debit d'acide d'alimentation. On a pu montrer que certains regimes transitoires ne peuvent se traiter par des equations generales, et necessiteraient plustot l'etablissement d

  8. Fabrication of DUPIC fuel pellets for irradiation testing in the NRU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Woong Ki; Kim, S. S.; Park, K. I. [and others

    2003-12-01

    In this project, DUPIC fuel manufacturing processes were established, and the irradiation test of DUPIC fuel at NRU in Canada was planned for the evaluation of DUPIC fuel performance. To establish manufacturing processes satisfying the requirements of NRU irradiation test, pre-qualification test and qualification test were performed. As a result of the qualification test, the DUPIC pellet fabrication processes were qualified and accepted by AECL. 8 batches of experiments were performed to fabricate 375 DUPIC pellets satisfying the requirements of NRU irradiation test under control of the quality assurance manual complying with CAN3-Z299.2-85. Sintered densities of the fabricated DUPIC pellets ranged from 10.26 g/cm{sup 3} to 10.43 g/cm{sup 3}. The DUPIC pellets have been stored in a box filled with helium gas. The pellets will be used to fabricate DUPIC elements for the irradiation testing.

  9. Ceramography of Irradiated tristructural isotropic (TRISO) Fuel from the AGR-2 Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Rice, Francine Joyce [Idaho National Lab. (INL), Idaho Falls, ID (United States); Stempien, John Dennis [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    Ceramography was performed on cross sections from four tristructural isotropic (TRISO) coated particle fuel compacts taken from the AGR-2 experiment, which was irradiated between June 2010 and October 2013 in the Advanced Test Reactor (ATR). The fuel compacts examined in this study contained TRISO-coated particles with either uranium oxide (UO2) kernels or uranium oxide/uranium carbide (UCO) kernels that were irradiated to final burnup values between 9.0 and 11.1% FIMA. These examinations are intended to explore kernel and coating morphology evolution during irradiation. This includes kernel porosity, swelling, and migration, and irradiation-induced coating fracture and separation. Variations in behavior within a specific cross section, which could be related to temperature or burnup gradients within the fuel compact, are also explored. The criteria for categorizing post-irradiation particle morphologies developed for AGR-1 ceramographic exams, was applied to the particles in the AGR-2 compacts particles examined. Results are compared with similar investigations performed as part of the earlier AGR-1 irradiation experiment. This paper presents the results of the AGR-2 examinations and discusses the key implications for fuel irradiation performance.

  10. Solar-fuel generation: Towards practical implementation

    Science.gov (United States)

    Dahl, Søren; Chorkendorff, Ib

    2012-02-01

    Limiting reliance on non-renewable fossil fuels inevitably depends on a more efficient utilization of solar energy. Materials scientists discuss the most viable approaches to produce high-energy-density fuels from sunlight that can be implemented in existing infrastructures.

  11. Solar-fuel generation: Towards practical implementation

    DEFF Research Database (Denmark)

    Dahl, Søren; Chorkendorff, Ib

    2012-01-01

    Limiting reliance on non-renewable fossil fuels inevitably depends on a more efficient utilization of solar energy. Materials scientists discuss the most viable approaches to produce high-energy-density fuels from sunlight that can be implemented in existing infrastructures....

  12. Anodic dissolution of irradiated metallic fuels in LiCl-KCl melt

    Science.gov (United States)

    Murakami, T.; Kato, T.; Rodrigues, A.; Ougier, M.; Iizuka, M.; Koyama, T.; Glatz, J.-P.

    2014-09-01

    Electrorefining is the main step in pyro-process of spent nuclear fuels, where actinides are recovered and separated from fission products. In the present study, electrorefining of irradiated metallic fuels called METAPHIX-1 (U-19 wt%Pu-10 wt%Zr alloy irradiated at PHENIX reactor, approximate maximum burn-up 2.5 at%) was performed. A major focus was on minimization of Zr co-dissolution from spent metallic fuels to reduce the burden to the pyro-process. Based on the ICP-MS analysis results and the SEM-EDX observations, the anodic dissolution behavior of the irradiated metallic fuels and the mass balances of actinides and fission products during the electrorefining were evaluated.

  13. Anodic dissolution of irradiated metallic fuels in LiCl–KCl melt

    Energy Technology Data Exchange (ETDEWEB)

    Murakami, T., E-mail: m-tsuyo@criepi.denken.or.jp [Central Research Institute of Electric Power Industry (CRIEPI), Komaeshi, Tokyo 201-8511 (Japan); Kato, T. [Central Research Institute of Electric Power Industry (CRIEPI), Komaeshi, Tokyo 201-8511 (Japan); Rodrigues, A.; Ougier, M. [Joint Research Center–Institute for Transuranium Elements (JRC–ITU), P.O. Box 2340, 76125 Karlsruhe (Germany); Iizuka, M.; Koyama, T. [Central Research Institute of Electric Power Industry (CRIEPI), Komaeshi, Tokyo 201-8511 (Japan); Glatz, J.-P. [Joint Research Center–Institute for Transuranium Elements (JRC–ITU), P.O. Box 2340, 76125 Karlsruhe (Germany)

    2014-09-15

    Electrorefining is the main step in pyro-process of spent nuclear fuels, where actinides are recovered and separated from fission products. In the present study, electrorefining of irradiated metallic fuels called METAPHIX-1 (U–19 wt%Pu–10 wt%Zr alloy irradiated at PHENIX reactor, approximate maximum burn-up 2.5 at%) was performed. A major focus was on minimization of Zr co-dissolution from spent metallic fuels to reduce the burden to the pyro-process. Based on the ICP-MS analysis results and the SEM–EDX observations, the anodic dissolution behavior of the irradiated metallic fuels and the mass balances of actinides and fission products during the electrorefining were evaluated.

  14. Development of Tools for Treating an Irradiated Fuel Rod Assembly in the Pool of HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Hong, J. T.; Ahn, S. H.; Kim, K. H.; Joung, C. Y. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-07-01

    To inspect a fuel rod during irradiation testing at the test loop of a research reactor, the test rig should be disassembled from the IPS (In-pile test section), and the targeted fuel rod assembly should be disassembled from the test rig and encapsulated in a cask to deliver the assembly to the hot cell. In addition, the fuel rod assembly under inspection in the hot cell should be delivered to the reactor pool and reassembled into the test rig to resume the irradiation test. Because the irradiated fuel rod is highly radioactive, all of the assembly and disassembly operations should be carried out in the reactor pool. Therefore, special tools need to be developed to treat the test rig in the pool of a research reactor. In this study, a new mechanically detachable fuel rod assembly has been developed for intermediate inspection during irradiation test at HANARO. A fuel rod assembly can be divided into two parts, such as an instrumented fuel rod assembly and a non-instrumented fuel rod assembly. In particular, an instrumented fuel rod assembly is assembled at the lower part of the test rig, and a non-instrumented fuel rod assembly is assembled at the bottom of the instrumented fuel rod assembly. The non-instrumented fuel rod assembly is locked in the test rig during irradiation test, and is easily disassembled from the instrumented fuel rod assembly by pushing the anchor button and twisting the non-instrumented fuel rod assembly. In addition, because a test rig is 5.4 meters long and the disassembling operation should be carried out at 6 meters deep in the pool of HANARO, tools to help disassemble and assemble the non-instrumented fuel rod assembly have also been developed. All components were designed to operate mechanically and are made of stainless steel and Al 6061 to minimize the effects from the radioactivity. The performance of the developed fuel rod assembly and tools have been verified through an out pile test.

  15. Multiplicative ARMA models to generate hourly series of global irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Mora-Lopez, L. [Universidad de Malaga (Spain). Dpto. Lenguajes y C. Computacion; Sidrach-de-Cardona, M. [Universidad de Malaga (Spain). Dpto. Fisica Aplicada

    1998-11-01

    A methodology to generate hourly series of global irradiation is proposed. The only input parameter which is required is the monthly mean value of daily global irradiation, which is available for most locations. The procedure to obtain new series is based on the use of a multiplicative autoregressive moving-average statistical model for time series with regular and seasonal components. The multiplicative nature of these models enables capture of the two types of relationships observed in recorded hourly series of global irradiation: on the one hand, the relationship between the value at one hour and the value at the previous hour; and on the other hand, the relationship between the value at one hour in one day and the value at the same hour in the previous day. In this paper the main drawback which arises when using these models to generate new series is solved: namely, the need for available recorded series in order to obtain the three parameters contained in the statistical ARMA model which is proposed (autoregressive coefficient, moving-average coefficient and variance of the error term). Specifically, expressions which enable estimation of these parameters using only monthly mean values of daily global irradiation are proposed in this paper. (author)

  16. AFC-1 Transmutation Fuels Post-Irradiation Hot Cell Examination 4-8 at.% - Final Report (Irradiation Experiments AFC-1B, -1F and -1Æ)

    Energy Technology Data Exchange (ETDEWEB)

    Bruce Hilton; Douglas Porter; Steven Hayes

    2006-09-01

    The AFC-1B, AFC-1F and AFC-1Æ irradiation tests are part of a series of test irradiations designed to evaluate the feasibility of the use of actinide bearing fuel forms in advanced fuel cycles for the transmutation of transuranic elements from nuclear waste. The tests were irradiated in the Idaho National Laboratory’s (INL) Advanced Test Reactor (ATR) to an intermediate burnup of 4 to 8 at% (2.7 - 6.8 x 1020 fiss/cm3). The tests contain metallic and nitride fuel forms with non-fertile (i.e., no uranium) and low-fertile (i.e., uranium bearing) compositions. Results of postirradiation hot cell examinations of AFC-1 irradiation tests are reported for eleven metallic alloy transmutation fuel rodlets and five nitride transmutation fuel rodlets. Non-destructive examinations included visual examination, dimensional inspection, gamma scan analysis, and neutron radiography. Detailed examinations, including fission gas puncture and analysis, metallography / ceramography and isotopics and burnup analyses, were performed on five metallic alloy and three nitride transmutation fuels. Fuel performance of both metallic alloy and nitride fuel forms was best correlated with fission density as a burnup metric rather than at.% depletion. The actinide bearing transmutation metallic alloy compositions exhibit irradiation performance very similar to U-xPu-10Zr fuel at equivalent fission densities. The irradiation performance of nitride transmutation fuels was comparable to limited data published on mixed nitride systems.

  17. Fuel cell power generation system. Nenryo denchi hatsuden system

    Energy Technology Data Exchange (ETDEWEB)

    Sato, M.; Shiba, Y.

    1993-06-11

    It is general to fabricate the primary cooling water system including the fuel cell main body using corrosion resistant stainless steel, while the secondary cooling system including absorption type freezer is made of carbon steel. For this structure, returning the cooling water of the secondary cooling system to the primary cooling system can cause the corrosion of the primary cooling system. That is, the water of inferior quality in the secondary system can corrode the primary system including the fuel cell. This invention solves the problem. The fuel cell bypass which is branched from the fuel cell cooling water inlet, detours the fuel cell, and it is connected to the water-vapor separator installed to the fuel cell. And the heat exchanger is installed at any of fuel cooling water outlet line, fuel cell cooling water inlet line, or fuel cell bypass line. With this structure, recovering the heat generated during the power generation by the fuel cell at the secondary side of the heat exchanger can be achieved while separating the primary and secondary cooling water. So that the trouble of fuel cell operation caused by the contamination of the primary cooling water with the secondary cooling water which contains corrosive impurities can be avoided. 6 figs.

  18. Irradiation of Microbes from Spent Nuclear Fuel Storage Pool Environments

    Energy Technology Data Exchange (ETDEWEB)

    Breckenridge, C.R.; Watkins, C.S.; Bruhn, D.F.; Roberto, F.F.; Tsang, M.N.; Pinhero, P.J. [INEEL (US); Brey, R.F. [ISU (US); Wright, R.N.; Windes, W.F.

    1999-09-03

    Microbes have been isolated and identified from spent nuclear fuel storage pools at the Idaho National Engineering and Environmental Laboratory (INEEL). Included among these are Corynebacterium aquaticum, Pseudomonas putida, Comamonas acidovorans, Gluconobacter cerinus, Micrococcus diversus, Rhodococcus rhodochrous, and two strains of sulfate-reducing bacteria (SRB). We examined the sensitivity of these microbes to a variety of total exposures of radiation generated by a 6-MeV linear accelerator (LINAC). The advantage of using a LINAC is that it provides a relatively quick screen of radiation tolerance. In the first set of experiments, we exposed each of the aforementioned microbes along with four additional microbes, pseudomonas aeruginosa, Micrococcus luteus, Escherchia coli, and Deinococcus radiodurans to exposures of 5 x 10{sup 3} and 6 x 10{sup 4} rad. All microbial specimens withstood the lower exposure with little or no reduction in cell population. Upon exposing the microbes to the larger dose of 6 x 10{sup 4} rad, we observed two distinct groupings: microbes that demonstrate resistance to radiation, and microbes that display intolerance through a dramatic reduction from their initial population. Microbes in the radiation tolerant grouping were exposed to 1.1 x 10{sup 5} rad to examine the extent of their resistance. We observe a correlation between radiation resistance and gram stain. The gram-positive species we examined seem to demonstrate a greater radiation resistance.

  19. Results of Recent Microstructural Characterization of Irradiated U-Mo Dispersion Fuels with Al Alloy Matrices that Contain Si

    Energy Technology Data Exchange (ETDEWEB)

    Keiser, Jr, D D; Robinson, A B; Janney, D E; Jue, J F

    2008-03-01

    RERTR U-Mo dispersion fuel plates are being developed for application in research reactors throughout the world. Of particular interest is the irradiation performance of U-Mo dispersion fuels with Si added to the Al matrix. Si is added to improve the performance of U-Mo dispersion fuels. Microstructural examinations have been performed on fuel plates with either Al-0.2Si or 4043 Al (~4.8% Si) alloy matrix in the as-fabricated and/or as-irradiated condition using optical metallography and/or scanning electron microscopy. Fuel plates with either matrix can have Si-rich layers around the U-7Mo particles after fabrication, and during irradiation these layers were observed to grow in thickness and to become Si-deficient in some areas of the fuel plates. For the fuel plates with 4043 Al, this was observed in fuel plate areas that were exposed to very aggressive irradiation conditions.

  20. Summary report on the fuel performance modeling of the AFC-2A, 2B irradiation experiments

    Energy Technology Data Exchange (ETDEWEB)

    Pavel G. Medvedev

    2013-09-01

    The primary objective of this work at the Idaho National Laboratory (INL) is to determine the fuel and cladding temperature history during irradiation of the AFC-2A, 2B transmutation metallic fuel alloy irradiation experiments containing transuranic and rare earth elements. Addition of the rare earth elements intends to simulate potential fission product carry-over from pyro-metallurgical reprocessing. Post irradiation examination of the AFC-2A, 2B rodlets revealed breaches in the rodlets and fuel melting which was attributed to the release of the fission gas into the helium gap between the rodlet cladding and the capsule which houses six individually encapsulated rodlets. This release is not anticipated during nominal operation of the AFC irradiation vehicle that features a double encapsulated design in which sodium bonded metallic fuel is separated from the ATR coolant by the cladding and the capsule walls. The modeling effort is focused on assessing effects of this unanticipated event on the fuel and cladding temperature with an objective to compare calculated results with the temperature limits of the fuel and the cladding.

  1. FLASHPOINT - a tool to routinely calculate the heat load in the irradiated fuel bays

    Energy Technology Data Exchange (ETDEWEB)

    Vyskocil, E.; Morrison, C.; Gifford, E.; Inglot, A.; Kozlowski, K.; Gocmanac, M. [AMEC NSS, Reactor and Radiation Physics, Toronto, Ontario (Canada); Parlatan, Y. [Ontario Power Generation, Safety Analysis Improvement Project Dept., Pickering, Ontario (Canada); Alabasha, H. [Bruce Power, Nuclear Safety Analysis and Support, Toronto, Ontario (Canada)

    2013-07-01

    At the recommendation of the World Association of Nuclear Operators (WANO), a tool was developed as an enhancement of NuFLASH (Nuclear Fuel Location and Storage History) in order to routinely calculate the Irradiated Fuel Bay (IFB) heat load. It uses information stored in NuFLASH regarding the location and details of spent fuel bundle properties to calculate the decay power on a bundle by bundle basis and then sum the decay powers of all bundles in a particular IFB. FLASHPOINT employs a two-step approximation of the bundle irradiation history based on the record of the life cycle for each individual fuel bundle. The primary parameter affecting the decay power of any individual irradiated CANDU fuel bundle following its discharge from core is the period of time elapsed since the bundle last operated at power within the reactor. The remaining factors influencing the decay power of an individual fuel bundle concern the irradiation history of that bundle while in core. The accuracy of the FLASHPOINT methodology has been assessed primarily through comparison of results obtained using the two step history representation implemented in FLASHPOINT against results from a more detailed ORIGEN-S calculation of the decay heat based on the SORO power history for a randomly selected sample of bundles. The results for individual bundles and the aggregate group are presented and the accuracy of the two-step approximation is demonstrated to be acceptable. (author)

  2. EVALUATION OF U10MO FUEL PLATE IRRADIATION BEHAVIOR VIA NUMERICAL AND EXPERIMENTAL BENCHMARKING

    Energy Technology Data Exchange (ETDEWEB)

    Samuel J. Miller; Hakan Ozaltun

    2012-11-01

    This article analyzes dimensional changes due to irradiation of monolithic plate-type nuclear fuel and compares results with finite element analysis of the plates during fabrication and irradiation. Monolithic fuel plates tested in the Advanced Test Reactor (ATR) at Idaho National Lab (INL) are being used to benchmark proposed fuel performance for several high power research reactors. Post-irradiation metallographic images of plates sectioned at the midpoint were analyzed to determine dimensional changes of the fuel and the cladding response. A constitutive model of the fabrication process and irradiation behavior of the tested plates was developed using the general purpose commercial finite element analysis package, Abaqus. Using calculated burn-up profiles of irradiated plates to model the power distribution and including irradiation behaviors such as swelling and irradiation enhanced creep, model simulations allow analysis of plate parameters that are either impossible or infeasible in an experimental setting. The development and progression of fabrication induced stress concentrations at the plate edges was of primary interest, as these locations have a unique stress profile during irradiation. Additionally, comparison between 2D and 3D models was performed to optimize analysis methodology. In particular, the ability of 2D and 3D models account for out of plane stresses which result in 3-dimensional creep behavior that is a product of these components. Results show that assumptions made in 2D models for the out-of-plane stresses and strains cannot capture the 3-dimensional physics accurately and thus 2D approximations are not computationally accurate. Stress-strain fields are dependent on plate geometry and irradiation conditions, thus, if stress based criteria is used to predict plate behavior (as opposed to material impurities, fine micro-structural defects, or sharp power gradients), unique 3D finite element formulation for each plate is required.

  3. UN TRISO Compaction in SiC for FCM Fuel Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Trammell, Michael P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Kiggans, James O. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jolly, Brian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Skitt, Darren J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-11-01

    The U.S. Department of Energy Office of Nuclear Energy (DOE-NE) Advanced Fuels Campaign (AFC) is conducting research and development to elevate the technology readiness level of Fully Ceramic Microencapsulated (FCM) fuels, a candidate nuclear fuel with potentially enhanced accident tolerance due to very high fission product retention. One of the early activities in FY17 was to demonstrate production of FCM pellets with uranium nitride TRISO particles. This was carried out in preparation of the larger pellet production campaign in support of the upcoming irradiation testing of this fuel form at INL’s Advanced Test Reactor.

  4. LIGHT WATER REACTOR ACCIDENT TOLERANT FUELS IRRADIATION TESTING

    Energy Technology Data Exchange (ETDEWEB)

    Carmack, William Jonathan [Idaho National Laboratory; Barrett, Kristine Eloise [Idaho National Laboratory; Chichester, Heather Jean MacLean [Idaho National Laboratory

    2015-09-01

    The purpose of Accident Tolerant Fuels (ATF) experiments is to test novel fuel and cladding concepts designed to replace the current zirconium alloy uranium dioxide (UO2) fuel system. The objective of this Research and Development (R&D) is to develop novel ATF concepts that will be able to withstand loss of active cooling in the reactor core for a considerably longer time period than the current fuel system while maintaining or improving the fuel performance during normal operations, operational transients, design basis, and beyond design basis events. It was necessary to design, analyze, and fabricate drop-in capsules to meet the requirements for testing under prototypic LWR temperatures in Idaho National Laboratory's Advanced Test Reactor (ATR). Three industry led teams and one DOE team from Oak Ridge National Laboratory provided fuel rodlet samples for their new concepts for ATR insertion in 2015. As-built projected temperature calculations were performed on the ATF capsules using the BISON fuel performance code. BISON is an application of INL’s Multi-physics Object Oriented Simulation Environment (MOOSE), which is a massively parallel finite element based framework used to solve systems of fully coupled nonlinear partial differential equations. Both 2D and 3D models were set up to examine cladding and fuel performance.

  5. Modeling of thermo-mechanical and irradiation behavior of mixed oxide fuel for sodium fast reactors

    Science.gov (United States)

    Karahan, Aydın; Buongiorno, Jacopo

    2010-01-01

    An engineering code to model the irradiation behavior of UO2-PuO2 mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named fuel engineering and structural analysis tool (FEAST-OXIDE). FEAST-OXIDE has several modules working in coupled form with an explicit numerical algorithm. These modules describe: (1) fission gas release and swelling, (2) fuel chemistry and restructuring, (3) temperature distribution, (4) fuel-clad chemical interaction and (5) fuel-clad mechanical analysis. Given the fuel pin geometry, composition and irradiation history, FEAST-OXIDE can analyze fuel and cladding thermo-mechanical behavior at both steady-state and design-basis transient scenarios. The code was written in FORTRAN-90 program language. The mechanical analysis module implements the LIFE algorithm. Fission gas release and swelling behavior is described by the OGRES and NEFIG models. However, the original OGRES model has been extended to include the effects of joint oxide gain (JOG) formation on fission gas release and swelling. A detailed fuel chemistry model has been included to describe the cesium radial migration and JOG formation, oxygen and plutonium radial distribution and the axial migration of cesium. The fuel restructuring model includes the effects of as-fabricated porosity migration, irradiation-induced fuel densification, grain growth, hot pressing and fuel cracking and relocation. Finally, a kinetics model is included to predict the clad wastage formation. FEAST-OXIDE predictions have been compared to the available FFTF, EBR-II and JOYO databases, as well as the LIFE-4 code predictions. The agreement was found to be satisfactory for steady-state and slow-ramp over-power accidents.

  6. 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.

  7. POWER GENERATION FROM LIQUID METAL NUCLEAR FUEL

    Science.gov (United States)

    Dwyer, O.E.

    1958-12-23

    A nuclear reactor system is described wherein the reactor is the type using a liquid metal fuel, such as a dispersion of fissile material in bismuth. The reactor is designed ln the form of a closed loop having a core sectlon and heat exchanger sections. The liquid fuel is clrculated through the loop undergoing flssion in the core section to produce heat energy and transferrlng this heat energy to secondary fluids in the heat exchanger sections. The fission in the core may be produced by a separate neutron source or by a selfsustained chain reaction of the liquid fuel present in the core section. Additional auxiliary heat exchangers are used in the system to convert water into steam which drives a turbine.

  8. Simulation of irradiation hardening of Zircaloy within plate-type dispersion nuclear fuel elements

    Science.gov (United States)

    Jiang, Yijie; Wang, Qiming; Cui, Yi; Huo, Yongzhong; Ding, Shurong

    2011-06-01

    Within plate-type dispersion nuclear fuel elements, the metal matrix and cladding attacked continuously by fast neutrons undergo irradiation hardening, which might have remarkable effects upon the mechanical behaviors within fuel elements. In this paper, with the irradiation hardening effect of metal materials mainly considered together with irradiation growth effect of the cladding, the three-dimensional large-deformation constitutive relations for the metal matrix and cladding are developed. The method of virtual temperature increase in the previous studies is further developed to model the irradiation swelling of fuel particles; the method of anisotropic thermal expansion is introduced to model irradiation growth of the cladding; and a method of multi-step-temperature loading is proposed to simulate the coupling features of irradiation-induced swelling of the fuel particles together with irradiation growth of the cladding. Above all, based on the developed relationship between irradiation growth at certain burnup and the loaded virtual temperatures, with considering that certain burnup corresponds to certain fast neutron fluence, the time-dependent constitutive relation due to irradiation hardening effect is replaced by the virtual-temperature-dependent one which is introduced into the commercial software to simulate the irradiation hardening effects of the matrix and cladding. Numerical simulations of the irradiation-induced mechanical behaviors are implemented with the finite element method in consideration of the micro-structure of the fuel meat. The obtained results indicate that when the irradiation hardening effects are introduced into the constitutive relations of the metal matrix and cladding: (1) higher maximum Mises stresses for certain burnup at the matrix exist with the equivalent plastic strains remaining almost the same at lower burnups; (2) the maximum Mises stresses for certain burnup at the cladding are enhanced while the maximum equivalent

  9. Future fossil fuel electricity generation in Europe: options and consequences

    Energy Technology Data Exchange (ETDEWEB)

    Tzimas, E.; Georgakaki, A.; Peteves, S.D.

    2009-07-01

    The study investigates the development of the fossil fuel fired power generation in Europe up to 2030 and identifies the critical factors that influence its evolution. Through the application of the least-cost expansion planning methods, the technology and fuel mix of fossil fuel power plant portfolios emerging from the twenty-four techno-economic scenarios are described. The different scenarios present alternative views for the role of non-fossil fuel power generation, the development of the world fuel and carbon markets and the carbon capture power generating technologies. The study estimates the needs for new fossil fuel capacity and identifies the optimal power plant mix for all possible combinations of the cases mentioned above. The impacts of the resulting portfolios on the objectives of the European energy policy are assessed using as indicators the capital investment fo the construction of the required capacity, the fuel consumption, the composition of the fuel mix, the CO{sub 2} emission levels, and the average production cost of electricity from the fossil fuelled fleet. The report finds that high CO{sub 2} prices need to be maintained and carbon capture technology must be developed and become commercialised. If these conditions re met and medium or high fossil fuel prices prevail, the portfolio of fossil fuel power plants that will be deployed will be compatible wit the European goal for the development of a more sustainable and secure energy system. The key conclusion is that for a sustainable and secure energy system we need to invest, both in the increase of non-fossil fuel power generation and to ensure that carob n capture and storage technologies are ready to be deployed when needed. 46 refs.,

  10. Repurposing an irradiated instrumented TRIGA fuel element for regular use

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Paulo F.; Souza, Luiz C.A., E-mail: pfo@cdtn.br, E-mail: lcas@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    TRIGA IPR-R1 is a research reactor also used for training and radioisotope production, located at the Centro de Desenvolvimento da Tecnologia Nuclear da Comissao Nacional de Energia Nuclear (Nuclear Technology Development Centre, Brazilian National Nuclear Energy Commission - CDTN/CNEN). Its first criticality occurred in November 1960. All original fuel elements were aluminum-clad. In 1971 nine new fuel elements, stainless steel-clad were acquired. One of them was an instrumented fuel element (IFE), equipped with 3 thermocouples. The IFE was introduced into the core only on August 2004, and remained there until July 2007. It was removed from the core after the severing of contacts between the thermocouples and their extension cables. After an unsuccessful attempt to recover electrical access to the thermocouples the IFE was transferred from the reactor pool to an auxiliary spent fuel storage well, with water, in the reactor room. In December 2011 the IFE was transferred to an identical well, dry, where it remains so far. This work is a proposal for recovery of this instrumented fuel element, by removing the cable guide rod and adaptation of a superior terminal plug similar to conventional fuel elements. This will enable its handling through the same tool used for regular fuel elements and its return to the reactor core. This is a delicate intervention in terms of radiological protection, and will require special care to minimize the exposure of operators. (author)

  11. Microstructural Analysis of Irradiated U-Mo Fuel Plates: Recent Results

    Energy Technology Data Exchange (ETDEWEB)

    D. D. Keiser, Jr.; J. Jue; B. D. Miller; J. Gan; A. B. Robinson; P. V. Medvedev

    2012-03-01

    Microstructural characterization of irradiated dispersion and monolithic RERTR fuel plates using scanning electron microscopy (SEM) is being performed in the Electron Microscopy Laboratory at the Idaho National Laboratory. The SEM analysis of samples from U-Mo dispersion fuel plates focuses primarily on the behavior of the Si that has been added to the Al matrix to improve the irradiation performance of the fuel plate and on the overall behavior of fission gases (e.g., Xe and Kr) that develop as bubbles in the fuel microstructure. For monolithic fuel plates, microstructural features of interest, include those found in the U-Mo foil and at the U-Mo/Zr and Zr/6061 Al cladding interfaces. For both dispersion and monolithic fuel plates, samples have been produced using an SEM equipped with a Focused Ion Beam (FIB). These samples are of very high quality and can be used to uncover some very unique microstructural features that are typically not observed when characterizing samples produced using more conventional techniques. Overall, for the dispersion fuel plates with matrices that contained Si, narrower fuel/matrix interaction layers are typically observed compared to the fuel plates with pure Al matrix, and for the monolithic fuel plates microstructural features have been observed in the U-10Mo foil that are similar to what have been observed in the fuel particles found in U-Mo dispersion fuels. Most recently, more prototypic monolithic fuel samples have been characterized and this paper describes the microstructures that have been observed in these samples.

  12. Characterization of fission gas bubbles in irradiated U-10Mo fuel

    Energy Technology Data Exchange (ETDEWEB)

    Casella, Andrew M.; Burkes, Douglas E.; MacFarlan, Paul J.; Buck, Edgar C.

    2017-09-01

    Irradiated U-10Mo fuel samples were prepared with traditional mechanical potting and polishing methods with in a hot cell. They were then removed and imaged with an SEM located outside of a hot cell. The images were then processed with basic imaging techniques from 3 separate software packages. The results were compared and a baseline method for characterization of fission gas bubbles in the samples is proposed. It is hoped that through adoption of or comparison to this baseline method that sample characterization can be somewhat standardized across the field of post irradiated examination of metal fuels.

  13. Key Differences in the Fabrication, Irradiation, and Safety Testing of U.S. and German TRISO-coated Particle Fuel and Their Implications on Fuel Performance

    Energy Technology Data Exchange (ETDEWEB)

    Petti, David Andrew; Maki, John Thomas; Buongiorno, Jacopo; Hobbins, Richard Redfield

    2002-06-01

    High temperature gas reactor technology is achieving a renaissance around the world. This technology relies on high quality production and performance of coated particle fuel. Historically, the irradiation performance of TRISO-coated gas reactor particle fuel in Germany has been superior to that in the United States. German fuel generally displayed in-pile gas release values that were three orders of magnitude lower than U.S. fuel. Thus, we have critically examined the TRISO-coated fuel fabrication processes in the U.S. and Germany and the associated irradiation database with a goal of understanding why the German fuel behaves acceptably, why the U.S. fuel has not faired as well, and what process/ production parameters impart the reliable performance to this fuel form. The postirradiation examination results are also reviewed to identify failure mechanisms that may be the cause of the poorer U.S. irradiation performance. This comparison will help determine the roles that particle fuel process/product attributes and irradiation conditions (burnup, fast neutron fluence, temperature, and degree of acceleration) have on the behavior of the fuel during irradiation and provide a more quantitative linkage between acceptable processing parameters, as-fabricated fuel properties and subsequent in-reactor performance.

  14. Key Differences in the Fabrication, Irradiation, and Safety Testing of U.S. and German TRISO-coated Particle Fuel and Their Implications on Fuel Performance

    Energy Technology Data Exchange (ETDEWEB)

    Petti, David Andrew; Maki, John Thomas; Buongiorno, Jacopo; Hobbins, Richard Redfield

    2002-06-01

    High temperature gas reactor technology is achieving a renaissance around the world. This technology relies on high quality production and performance of coated particle fuel. Historically, the irradiation performance of TRISO-coated gas reactor particle fuel in Germany has been superior to that in the United States. German fuel generally displayed in-pile gas release values that were three orders of magnitude lower than U.S. fuel. Thus, we have critically examined the TRISO-coated fuel fabrication processes in the U.S. and Germany and the associated irradiation database with a goal of understanding why the German fuel behaves acceptably, why the U.S. fuel has not faired as well, and what process/ production parameters impart the reliable performance to this fuel form. The postirradiation examination results are also reviewed to identify failure mechanisms that may be the cause of the poorer U.S. irradiation performance. This comparison will help determine the roles that particle fuel process/product attributes and irradiation conditions (burnup, fast neutron fluence, temperature, and degree of acceleration) have on the behavior of the fuel during irradiation and provide a more quantitative linkage between acceptable processing parameters, as-fabricated fuel properties and subsequent in-reactor performance.

  15. Fabrication of U-10 wt.%Zr Metallic Fuel Rodlets for Irradiation Test in BOR-60 Fast Reactor

    OpenAIRE

    Ki-Hwan Kim; Jong-Hwan Kim; Seok-Jin Oh; Jung-Won Lee; Ho-Jin Lee; Chan-Bock Lee

    2016-01-01

    The fabrication technology for metallic fuel has been developed to produce the driver fuel in a PGSFR in Korea since 2007. In order to evaluate the irradiation integrity and validate the in-reactor of the starting metallic fuel with FMS cladding for the loading of the metallic fuel, U-10 wt.%Zr fuel rodlets were fabricated and evaluated for a verification of the starting driver fuel through an irradiation test in the BOR-60 fast reactor. The injection casting method was applied to U-10 wt.%Zr...

  16. On-site gamma-ray spectroscopic measurements of fission gas release in irradiated nuclear fuel.

    Science.gov (United States)

    Matsson, I; Grapengiesser, B; Andersson, B

    2007-01-01

    An experimental, non-destructive in-pool, method for measuring fission gas release (FGR) in irradiated nuclear fuel has been developed. Using the method, a significant number of experiments have been performed in-pool at several nuclear power plants of the BWR type. The method utilises the 514 keV gamma-radiation from the gaseous fission product (85)Kr captured in the fuel rod plenum volume. A submergible measuring device (LOKET) consisting of an HPGe-detector and a collimator system was utilised allowing for single rod measurements on virtually all types of BWR fuel. A FGR database covering a wide range of burn-ups (up to average rod burn-up well above 60 MWd/kgU), irradiation history, fuel rod position in cross section and fuel designs has been compiled and used for computer code benchmarking, fuel performance analysis and feedback to reactor operators. Measurements clearly indicate the low FGR in more modern fuel designs in comparison to older fuel types.

  17. Final Report on IFA-10, the first Swedish Instrumented Fuel Assembly Irradiated in HBWR, Norway

    Energy Technology Data Exchange (ETDEWEB)

    Gyllander, J.Aa.

    1967-12-15

    A final report is given on IFA-10, the first Swedish instrumented fuel assembly irradiated in HBWR. The post-irradiation data are presented and correlated with the irradiation statistics. No bowing of the bundle was observed, no equi-axed grain growth was discernible, the fission gas release was very small, and the relative dimensional changes in length and diameter were of the order of magnitude 9 x 10{sup -4} The hydride content of the can increased from 35 ppm to 65 ppm and, in the contact point of the spacer, to 180 ppm.

  18. HTGR fuels and core development program. Quarterly progress report for the period ending November 30, 1976. [Graphite and fuel irradiations; fission product release

    Energy Technology Data Exchange (ETDEWEB)

    1976-12-27

    The work reported includes studies of reactions between core materials and coolant impurities, basic fission product transport mechanisms, core graphite development and testing, the development and testing of recyclable fuel systems, and physics and fuel management studies. Materials studies include irradiation capsule tests of both fuel and graphite. Experimental procedures and results are discussed and the data are presented in tables, graphs, and photographs.

  19. HTGR Fuels and Core Development Program. Quarterly progress report for the period ending August 31, 1977. [Graphite and fuel irradiation; fission product release

    Energy Technology Data Exchange (ETDEWEB)

    1977-09-01

    The work reported includes studies of reactions between core materials and coolant impurities, basic fission product transport mechanisms, core graphite development and testing, the development and testing of recyclable fuel systems, and physics and fuel management studies. Materials studies include irradiation capsule tests of both fuel and graphite. Experimental procedures and results are discussed and data are presented.

  20. Four-point Bend Testing of Irradiated Monolithic U-10Mo Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Rabin, B. H. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Lloyd, W. R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schulthess, J. L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wright, J. K. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Lind, R. P. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Scott, L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wachs, K. M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    This paper presents results of recently completed studies aimed at characterizing the mechanical properties of irradiated U-10Mo fuel in support of monolithic base fuel qualification. Mechanical properties were evaluated in four-point bending. Specimens were taken from fuel plates irradiated in the RERTR-12 and AFIP-6 Mk. II irradiation campaigns, and tests were conducted in the Hot Fuel Examination Facility (HFEF) at Idaho National Laboratory (INL). The monolithic fuel plates consist of a U-10Mo fuel meat covered with a Zr diffusion barrier layer fabricated by co-rolling, clad in 6061 Al using a hot isostatic press (HIP) bonding process. Specimens exhibited nominal (fresh) fuel meat thickness ranging from 0.25 mm to 0.64 mm, and fuel plate average burnup ranged from approximately 0.4 x 1021 fissions/cm3 to 6.0 x 1021 fissions/cm3. After sectioning the fuel plates, the 6061 Al cladding was removed by dissolution in concentrated NaOH. Pre- and post-dissolution dimensional inspections were conducted on test specimens to facilitate accurate analysis of bend test results. Four-point bend testing was conducted on the HFEF Remote Load Frame at a crosshead speed of 0.1 mm/min using custom-designed test fixtures and calibrated load cells. All specimens exhibited substantially linear elastic behavior and failed in a brittle manner. The influence of burnup on the observed slope of the stress-strain curve and the calculated fracture strength is discussed.

  1. Study of irradiation induced restructuring of high burnup fuel - Use of computer and accelerator for fuel science and engineering -

    Energy Technology Data Exchange (ETDEWEB)

    Sataka, M.; Ishikawa, N.; Chimn, Y.; Nakamura, J.; Amaya, M. [Japan Atomic Energy Agency, Naka Gun (Japan); Iwasawa, M.; Ohnuma, T.; Sonoda, T. [Central Research Institute of Electric Power Industry, Tokyo (Japan); Kinoshita, M.; Geng, H. Y.; Chen, Y.; Kaneta, Y. [The Univ. of Tokyo, Tokyo (Japan); Yasunaga, K.; Matsumura, S.; Yasuda, K. [Kyushu Univ., Motooka (Japan); Iwase [Osaka Prefecture Univ., Osaka (Japan); Ichinomiya, T.; Nishiuran, Y. [Hokkaido Univ., Kitaku (Japan); Matzke, HJ. [Academy of Ceramics, Karlsruhe (Germany)

    2008-10-15

    In order to develop advanced fuel for future LWR reactors, trials were made to simulate the high burnup restructuring of the ceramics fuel, using accelerator irradiation out of pile and with computer simulation. The target is to reproduce the principal complex process as a whole. The reproduction of the grain subdivision (sub grain formation) was successful at experiments with sequential combined irradiation. It was made by recovery process of the accumulated dislocations, making cells and sub-boundaries at grain boundaries and pore surfaces. Details of the grain sub division mechanism is now in front of us outside of the reactor. Extensive computer science studies, first principle and molecular dynamics gave behavior of fission gas atoms and interstitial oxygen, assisting the high burnup restructuring.

  2. The Design and Manufacturing Report of Non-Instrumented Rig for Dual-cooled Annular Fuel Irradiation Test in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dae Ho; Bang, Je Geon; Lim, Ik Sung; Kim, Sun Ki; Yang, Yong Sik; Song, Kun Woo; Seo, Chul Gyo; Park, Chan Kook

    2008-09-15

    This project is preparing to irradiation test of the developed double cooled annular fuel pellet in HANARO for pursuit advanced performance in High Performance Fuel Technology Development as a part Nuclear Mid and Long-term R and D Program. On the basis test rod is performed the nuclei property and preliminary fuel performance analysis, test rod and non-instrumented rig designed and manufactured for irradiation test in HANARO OR hole. This non- instrumented rig was confirmed the compatibility of HANARO and the integrity of rig structure, and satisfied the quality assurance requirements. This non- instrumented rig is adopt to the irradiation test for double cooled annular fuel pellet in HANARO.

  3. Pre-qualification experiments of DUPIC fuel pellets for irradiation testing in the NRU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Woong Ki; Kim, S. S.; Lee, J. W. and others

    2002-02-01

    DUPIC fuel manufacturing technologies and processes have been developed at DFDF(DUPIC Fuel Development Facility, IMEF M6). Using DUPIC powder prepared by the oxidation and reduction processes, the DUPIC fuel pellets and mini-elements were fabricated for the irradiation test and performance evaluation at HANARO. In this study, the irradiation test was planned for the performance evaluation of DUPIC fuel pellets and elements at NRU. To establish fabrication process satisfying the requirements of NRU irradiation test, sintered DUPIC pellets were fabricated with a variety of process parameters involving compaction pressure and characterized by the inspection system. As a result of the experiment, DUPIC pellets with 12.19 mm of diameter, 10.37{approx}10.45 g/cm{sup 3} of sintered density, and less than Ra 0.8{mu}m of surface roughness have been successfully fabricated at hot cell. The optimum DUPIC pellet fabrication process satisfying the requirements of NRU irradiation has been established based on the result of this experiment.

  4. Gamma spectrometry of irradiated fuel plates; Espectrometria gama em elementos combustiveis tipo placa irradiados

    Energy Technology Data Exchange (ETDEWEB)

    Terremoto, Luis A.A.; Zeituni, Carlos A.; Perrotta, Jose A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil). Div. de Engenharia do Nucleo

    1997-10-01

    This work describes the fundamental aspects of a method which uses gamma-ray spectroscopy in order to perform non-destructive burnup measurements in irradiated MTR fuel elements. Experiments based on such method will be conducted at the storage pool area of the IEA-R1 research reactor. Some preliminary results are presented. (author). 10 refs., 5 figs., 1 tab.

  5. Analysis of sample and fuel pin irradiation experiments in Phenix for basic nuclear data validation

    Energy Technology Data Exchange (ETDEWEB)

    D' Angelo, A.; Cleri, F. (ENEA, Casaccia Nuclear Research Centre, Dipartimento Reattori Veloci, 0060 Casaccia (IT)); Marimbeau, P.; Salvatores, M.; Grouiller, J.P. (Centre d' Etudes Nucleaires de Cadarache, Service de Physique des Reacteurs et du Cycle, Dept. de Recherche Physique, 13108 St. Paul-lez-Durance (FR))

    1990-07-01

    This paper presents comparisons between calculations and experimental data from fuel irradiation experiments performed in Phenix. Both the French CARNAVAL-IV system and the recently developed JEF-1 basic data file are used. The global consistency of the results is excellent and the compared values are, in general, very satisfactory. In some cases, indications for evaluated data modifications are obtained.

  6. Validation of the Physics Analysis used to Characterize the AGR-1 TRISO Fuel Irradiation Test

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James W.; Harp, Jason M.; Demkowicz, Paul A.; Hawkes, Grant L.; Chang, Gray S.

    2015-05-01

    The results of a detailed physics depletion calculation used to characterize the AGR-1 TRISO-coated particle fuel test irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory are compared to measured data for the purpose of validation. The particle fuel was irradiated for 13 ATR power cycles over three calendar years. The physics analysis predicts compact burnups ranging from 11.30-19.56% FIMA and cumulative neutron fast fluence from 2.21?4.39E+25 n/m2 under simulated high-temperature gas-cooled reactor conditions in the ATR. The physics depletion calculation can provide a full characterization of all 72 irradiated TRISO-coated particle compacts during and post-irradiation, so validation of this physics calculation was a top priority. The validation of the physics analysis was done through comparisons with available measured experimental data which included: 1) high-resolution gamma scans for compact activity and burnup, 2) mass spectrometry for compact burnup, 3) flux wires for cumulative fast fluence, and 4) mass spectrometry for individual actinide and fission product concentrations. The measured data are generally in very good agreement with the calculated results, and therefore provide an adequate validation of the physics analysis and the results used to characterize the irradiated AGR-1 TRISO fuel.

  7. Irradiation performance of PFBR MOX fuel after 112 GWd/t burn-up

    Science.gov (United States)

    Venkiteswaran, C. N.; Jayaraj, V. V.; Ojha, B. K.; Anandaraj, V.; Padalakshmi, M.; Vinodkumar, S.; Karthik, V.; Vijaykumar, Ran; Vijayaraghavan, A.; Divakar, R.; Johny, T.; Joseph, Jojo; Thirunavakkarasu, S.; Saravanan, T.; Philip, John; Rao, B. P. C.; Kasiviswanathan, K. V.; Jayakumar, T.

    2014-06-01

    The 500 MWe Prototype Fast Breeder Reactor (PFBR) which is in advanced stage of construction at Kalpakkam, India, will use mixed oxide (MOX) fuel with a target burnup of 100 GWd/t. The fuel pellet is of annular design to enable operation at a peak linear power of 450 W/cm with the requirement of minimum duration of pre-conditioning. The performance of the MOX fuel and the D9 clad and wrapper material was assessed through Post Irradiation Examinations (PIE) after test irradiation of 37 fuel pin subassembly in Fast Breeder Test Reactor (FBTR) to a burn-up of 112 GWd/t. Fission product distribution, swelling and fuel-clad gap evolution, central hole diameter variation, restructuring, fission gas release and clad wastage due to fuel-clad chemical interaction were evaluated through non-destructive and destructive examinations. The examinations have indicated that the MOX fuel can safely attain the desired target burn-up in PFBR.

  8. IRRADIATION TESTING OF THE RERTR FUEL MINIPLATES WITH BURNABLE ABSORBERS IN THE ADVANCED TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    I. Glagolenko; D. Wachs; N. Woolstenhulme; G. Chang; B. Rabin; C. Clark; T. Wiencek

    2010-10-01

    Based on the results of the reactor physics assessment, conversion of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) can be potentially accomplished in two ways, by either using U-10Mo monolithic or U-7Mo dispersion type plates in the ATR fuel element. Both designs, however, would require incorporation of the burnable absorber in several plates of the fuel element to compensate for the excess reactivity and to flatten the radial power profile. Several different types of burnable absorbers were considered initially, but only borated compounds, such as B4C, ZrB2 and Al-B alloys, were selected for testing primarily due to the length of the ATR fuel cycle and fuel manufacturing constraints. To assess and compare irradiation performance of the U-Mo fuels with different burnable absorbers we have designed and manufactured 28 RERTR miniplates (20 fueled and 8 non-fueled) containing fore-mentioned borated compounds. These miniplates will be tested in the ATR as part of the RERTR-13 experiment, which is described in this paper. Detailed plate design, compositions and irradiations conditions are discussed.

  9. Design and manufacturing of non-instrumented capsule for advanced PWR fuel pellet irradiation test in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. H.; Lee, C. B.; Song, K. W. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-04-01

    This project is preparing to irradiation test of the developed large grain UO{sub 2} fuel pellet in HANARO for pursuit fuel safety and high burn-up in 'Advanced LWR Fuel Technology Development Project' as a part Nuclear Mid and Long-term R and D Program. On the basis test rod is performed the nuclei property and preliminary fuel performance analysis, test rod and non-instrumented capsule are designed and manufactured for irradiation test in HANARO. This non-instrumented irradiation capsule of Advanced PWR Fuel pellet was referred the non-instrumented capsule for an irradiation test of simulated DUPIC fuel in HANARO(DUPIC Rig-001) and 18-element HANARO fuel, was designed to ensure the integrity and the endurance of non-instrumented capsule during the long term(2.5 years) irradiation. To irradiate the UO{sub 2} pellets up to the burn-up 70 MWD/kgU, need the time about 60 months and ensure the integrity of non-instrumented capsule for 30 months until replace the new capsule. This non-instrumented irradiation capsule will be based to develope the non-instrumented capsule for the more long term irradiation in HANARO. 22 refs., 13 figs., 5 tabs. (Author)

  10. Build-up of actinides in irradiated fuel rods of the ET-RR-1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Adib, M.; Naguib, K.; Morcos, H.N

    2001-09-01

    The content concentrations of actinides are calculated as a function of operating reactor regime and cooling time at different percentage of fuel burn-up. The build-up transmutation equations of actinides content in an irradiated fuel are solved numerically .A computer code BAC was written to operate on a PC computer to provide the required calculations. The fuel element of 10% {sup 235}U enrichment of ET-RR-1 reactor was taken as an example for calculations using the BAC code. The results are compared with other calculations for the ET-RR-1 fuel rod. An estimation of fissile build-up content of a proposed new fuel of 20% {sup 235}U enrichment for ET-RR-1 reactor is given. The sensitivity coefficients of build-up plutonium concentrations as a function of cross-section data uncertainties are also calculated.

  11. Thermochemical prediction of chemical form distributions of fission products in LWR oxide fuels irradiated to high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Moriyama, Kouki; Furuya, Hirotaka [Kyushu Univ., Fukuoka (Japan). Faculty of Engineering

    1997-09-01

    Based on the result of micro-gamma scanning of a fuel pin irradiated to high burnup in a commercial PWR, the radial distribution of chemical forms of fission products (FPs) in LWR fuel pins was theoretically predicted by a thermochemical computer code SOLGASMIX-PV. The absolute amounts of fission products generated in the fuel was calculated by ORIGEN-2 code, and the radial distributions of temperature and oxygen potential were calculated by taking the neutron depression and oxygen redistribution in the fuel into account. A fuel pellet was radially divided into 51 sections and chemical forms of FPs were calculated in each section. In addition, the effects of linear heat rating (LHR) and average O/U ratio on radial distribution of chemical form were evaluated. It was found that approximately 13 mole% of the total amount of Cs compounds exists as CsI and virtually remaining fraction as Cs{sub 2}MoO{sub 4} under the operation condition of LHR below 400 W/cm. On the other hand, when LHR is beyond 400 W/cm under the transient operation condition, its distribution did not change so much from the one under normal operation condition. (author)

  12. Irradiation Test in HANARO of the Parts of an X-Gen Nuclear Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Choo, K. N.; Kim, B. G.; Kang, Y. H. (and others)

    2008-08-15

    An instrumented capsule of 07M-13N was designed, fabricated and irradiated for an evaluation of the neutron irradiation properties of the parts of an X-Gen nuclear fuel assembly for PWR requested by KNF. Some specimens requested by Westinghouse Co. and Hanyang university were also inserted. 389 KNF specimens such as bucking and spring test specimens of 1x1 cell spacer grid, tensile, microstructure and tensile of welded parts, irradiation growth, spring test specimens made of HANA tube, Zirlo, Zircaloy-4, Inconel-718 were placed in the capsule. The capsule was composed of 5 stages having many kinds of specimens and an independent electric heater at each stage. During the irradiation test, the temperature of the specimens and the thermal/fast neutron fluences were measured by 14 thermocouples and 7 sets of Ni-Ti-Fe (2 sets contain additional Nb-Ag) neutron fluence monitors installed in the capsule. The capsule was irradiated for 59.19days (4 cycles) in the CT test hole of HANARO of a 30MW thermal output at 300 {approx} 420 .deg. C(for KNF specimens) up to a fast neutron fluence of 1.27x10{sup 21}(n/cm{sup 2}) (E>1MeV). After an irradiation test, the main body of the capsule was cut off at the bottom of the protection tube with a cutting system and it was transported to the IMEF (Irradiated Materials Examination Facility). The irradiated specimens were tested to evaluate the irradiation performance of the parts of an X-Gen fuel assembly in the IMEF hot cell.

  13. Transportation impact analysis for shipment of irradiated N-reactor fuel and associated materials

    Energy Technology Data Exchange (ETDEWEB)

    Daling, P.M.; Harris, M.S.

    1994-12-01

    An analysis of the radiological and nonradiological impacts of highway transportation of N-Reactor irradiated fuel (N-fuel) and associated materials is described in this report. N-fuel is proposed to be transported from its present locations in the 105-KE and 105-KW Basins, and possibly the PUREX Facility, to the 327 Building for characterization and testing. Each of these facilities is located on the Hanford Site, which is near Richland, Washington. The projected annual shipping quantity is 500 kgU/yr for 5 years for a total of 2500 kgU. It was assumed the irradiated fuel would be returned to the K- Basins following characterization, so the total amount of fuel shipped was assumed to be 5000 kgU. The shipping campaign may also include the transport and characterization of liquids, gases, and sludges from the storage basins, including fuel assembly and/or canister parts that may also be present in the basins. The impacts of transporting these other materials are bounded by the impacts of transporting 5000 kgU of N-fuel. This report was prepared to support an environmental assessment of the N-fuel characterization program. The RADTRAN 4 and GENII computer codes were used to evaluate the radiological impacts of the proposed shipping campaign. RADTRAN 4 was used to calculate the routine exposures and accident risks to workers and the general public from the N-fuel shipments. The GENII computer code was used to calculate the consequences of the maximum credible accident. The results indicate that the transportation of N-fuel in support of the characterization program should not cause excess radiological-induced latent cancer fatalities or traffic-related nonradiological accident fatalities. The consequences of the maximum credible accident are projected to be small and result in no excess latent cancer fatalities.

  14. Dissolution of Irradiated Commercial UO2 Fuels in Ammonium Carbonate and Hydrogen Peroxide

    Energy Technology Data Exchange (ETDEWEB)

    Soderquist, Chuck Z.; Johnsen, Amanda M.; McNamara, Bruce K.; Hanson, Brady D.; Chenault, Jeffrey W.; Carson, Katharine J.; Peper, Shane M.

    2011-01-18

    We propose and test a disposition path for irradiated nuclear fuel using ammonium carbonate and hydrogen peroxide media. We demonstrate on a 13 g scale that >98% of the irradiated fuel dissolves. Subsequent expulsion of carbonate from the dissolver solution precipitates >95% of the plutonium, americium, curium, and substantial amounts of fission products, effectively partitioning the fuel at the dissolution step. Uranium can be easily recovered from solution by any of several means, such as ion exchange, solvent extraction, or direct precipitation. Ammonium carbonate can be evaporated from solution and recovered for re-use, leaving an extremely compact volume of fission products, transactinides, and uranium. Stack emissions are predicted to be less toxic, less radioactive, chemically simpler, and simpler to treat than those from the conventional PUREX process.

  15. Structural analysis and design optimization of double shell system for fuel irradiation capsule

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Y. S.; Choi, Y. J.; Choi, M. H.; Rhu, C. H.; Go, J. H.; Hong, S. J.; Lee, H. C. [Chungnam National Univ., Taejeon (Korea)

    2001-04-01

    During irradiation tests, the fuel capsule expect that the high temperature will be occur. Thus, to estimate the structural integrity of fuel capsule during irradiation tests, it is needed to perform structural analysis and to obtain the information of mechanical characteristics for the system. In this study, the structure analysis of the circular capsule is performed using the finite element analysis program, ANSYS and analysis calculation. To obtain the mechanical characteristics of the circular capsule structure such as stresses, critical buckling loads and natural frequencies et al. the static nd model analysis are conducted. The effects of various wall thicknesses of capsule outer tube and support tube for circular capsule are obtained. Also, the effects of boundary conditions and principal materials of the fuel capsule on the structural behavior are investigated. The FE results are compared with the analysis results in case of possible. 13 refs., 34 figs., 10 tabs. (Author)

  16. SILICON CARBIDE GRAIN BOUNDARY DISTRIBUTIONS, IRRADIATION CONDITIONS, AND SILVER RETENTION IN IRRADIATED AGR-1 TRISO FUEL PARTICLES

    Energy Technology Data Exchange (ETDEWEB)

    Lillo, T. M.; Rooyen, I. J.; Aguiar, J. A.

    2016-11-01

    Precession electron diffraction in the transmission electron microscope was used to map grain orientation and ultimately determine grain boundary misorientation angle distributions, relative fractions of grain boundary types (random high angle, low angle or coincident site lattice (CSL)-related boundaries) and the distributions of CSL-related grain boundaries in the SiC layer of irradiated TRISO-coated fuel particles. Two particles from the AGR-1 experiment exhibiting high Ag-110m retention (>80%) were compared to a particle exhibiting low Ag-110m retention (<19%). Irradiated particles with high Ag-110m retention exhibited a lower fraction of random, high angle grain boundaries compared to the low Ag-110m retention particle. An inverse relationship between the random, high angle grain boundary fraction and Ag-110m retention is found and is consistent with grain boundary percolation theory. Also, comparison of the grain boundary distributions with previously reported unirradiated grain boundary distributions, based on SEM-based EBSD for similarly fabricated particles, showed only small differences, i.e. a greater low angle grain boundary fraction in unirradiated SiC. It was, thus, concluded that SiC layers with grain boundary distributions susceptible to Ag-110m release were present prior to irradiation. Finally, irradiation parameters were found to have little effect on the association of fission product precipitates with specific grain boundary types.

  17. Behavior of pre-irradiated fuel under a simulated RIA condition. Results of NSRR Test JM-5

    Energy Technology Data Exchange (ETDEWEB)

    Fuketa, Toyoshi; Sasajima, Hideo; Mori, Yukihide; Tanzawa, Sadamitsu; Ishijima, Kiyomi; Kobayashi, Shinsho; Kamata, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Homma, Kozo; Sakai, Haruyuki

    1995-11-01

    This report presents results from the power burst experiment with pre-irradiated fuel rod, Test JM-5, conducted in the Nuclear Safety Research Reactor (NSRR). The data concerning test method, pre-irradiation, pre-pulse fuel examination, pulse irradiation, transient records and post-pulse fuel examination are described, and interpretations and discussions of the results are presented. Preceding to the pulse irradiation in the NSRR, test fuel rod was irradiated in the Japan Materials Testing Reactor (JMTR) up to a fuel burnup of 25.7 MWd/kgU with average linear heat rate of 33.4 kW/m. The fuel rod was subjected to the pulse irradiation resulting in a desposited energy of 223 {+-} 7 cal/g{center_dot}fuel (0.93 {+-} 0.03 kJ/g{center_dot}fuel) and a peak fuel enthalpy of 167 {+-} 5 cal/g{center_dot}fuel (0.70 {+-} 0.02 kJ/g{center_dot}fuel) under stagnant water cooling condition at atmospheric pressure and ambient temperature. Test fuel rod behavior was assessed from pre- and post-pulse fuel examinations and transient records during the pulse. The Test JM-5 resulted in cladding failure. More than twenty small cracks were found in the post-test cladding, and most of the defects located in pre-existing locally hydrided region. The result indicates an occurrence of fuel failure by PCMI (pellet/cladding mechanical interaction) in combination with decreased integrity of hydrided cladding. (author).

  18. Solid Oxide Fuel Cell Hybrid System for Distributed Power Generation

    Energy Technology Data Exchange (ETDEWEB)

    David Deangelis; Rich Depuy; Debashis Dey; Georgia Karvountzi; Nguyen Minh; Max Peter; Faress Rahman; Pavel Sokolov; Deliang Yang

    2004-09-30

    This report summarizes the work performed by Hybrid Power Generation Systems, LLC (HPGS) during the April to October 2004 reporting period in Task 2.3 (SOFC Scaleup for Hybrid and Fuel Cell Systems) under Cooperative Agreement DE-FC26-01NT40779 for the U. S. Department of Energy, National Energy Technology Laboratory (DOE/NETL), entitled ''Solid Oxide Fuel Cell Hybrid System for Distributed Power Generation''. This study analyzes the performance and economics of power generation systems for central power generation application based on Solid Oxide Fuel Cell (SOFC) technology and fueled by natural gas. The main objective of this task is to develop credible scale up strategies for large solid oxide fuel cell-gas turbine systems. System concepts that integrate a SOFC with a gas turbine were developed and analyzed for plant sizes in excess of 20 MW. A 25 MW plant configuration was selected with projected system efficiency of over 65% and a factory cost of under $400/kW. The plant design is modular and can be scaled to both higher and lower plant power ratings. Technology gaps and required engineering development efforts were identified and evaluated.

  19. 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.

  20. 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.

  1. A polymer electrolyte fuel cell stack for stationary power generation from hydrogen fuel

    Energy Technology Data Exchange (ETDEWEB)

    Gottesfeld, S. [Los Alamos National Lab., NM (United States)

    1995-09-01

    The fuel cell is the most efficient device for the conversion of hydrogen fuel to electric power. As such, the fuel cell represents a key element in efforts to demonstrate and implement hydrogen fuel utilization for electric power generation. The low temperature, polymer electrolyte membrane fuel cell (PEMFC) has recently been identified as an attractive option for stationary power generation, based on the relatively simple and benign materials employed, the zero-emission character of the device, and the expected high power density, high reliability and low cost. However, a PEMFC stack fueled by hydrogen with the combined properties of low cost, high performance and high reliability has not yet been demonstrated. Demonstration of such a stack will remove a significant barrier to implementation of this advanced technology for electric power generation from hydrogen. Work done in the past at LANL on the development of components and materials, particularly on advanced membrane/electrode assemblies (MEAs), has contributed significantly to the capability to demonstrate in the foreseeable future a PEMFC stack with the combined characteristics described above. A joint effort between LANL and an industrial stack manufacturer will result in the demonstration of such a fuel cell stack for stationary power generation. The stack could operate on hydrogen fuel derived from either natural gas or from renewable sources. The technical plan includes collaboration with a stack manufacturer (CRADA). It stresses the special requirements from a PEMFC in stationary power generation, particularly maximization of the energy conversion efficiency, extension of useful life to the 10 hours time scale and tolerance to impurities from the reforming of natural gas.

  2. Fabrication of U-10 wt.%Zr Metallic Fuel Rodlets for Irradiation Test in BOR-60 Fast Reactor

    Directory of Open Access Journals (Sweden)

    Ki-Hwan Kim

    2016-01-01

    Full Text Available The fabrication technology for metallic fuel has been developed to produce the driver fuel in a PGSFR in Korea since 2007. In order to evaluate the irradiation integrity and validate the in-reactor of the starting metallic fuel with FMS cladding for the loading of the metallic fuel, U-10 wt.%Zr fuel rodlets were fabricated and evaluated for a verification of the starting driver fuel through an irradiation test in the BOR-60 fast reactor. The injection casting method was applied to U-10 wt.%Zr fuel slugs with a diameter of 5.5 mm. Consequently, fuel slugs per melting batch without casting defects were fabricated through the development of advanced casting technology and evaluation tests. The optimal GTAW welding conditions were also established through a number of experiments. In addition, a qualification test was carried out to prove the weld quality of the end plug welding of the metallic fuel rodlets. The wire wrapping of metallic fuel rodlets was successfully accomplished for the irradiation test. Thus, PGSFR fuel rodlets have been soundly fabricated for the irradiation test in a BOR-60 fast reactor.

  3. 40 CFR 80.531 - How are motor vehicle diesel fuel credits generated?

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 16 2010-07-01 2010-07-01 false How are motor vehicle diesel fuel... (CONTINUED) AIR PROGRAMS (CONTINUED) REGULATION OF FUELS AND FUEL ADDITIVES Motor Vehicle Diesel Fuel... are motor vehicle diesel fuel credits generated? (a) Generation of credits from June 1, 2006...

  4. Mechanical technologies for PIGMI. [Pion Generator for Medical Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Hansborough, L.D.

    1979-01-01

    PIGMI (Pion Generator for Medical Irradiations) is a compact linear proton accelerator designed for a hospital environment. The prototype of the low energy section of PIGMI has been designed and is being fabricated at the Los Alamos Scientific Laboratory. It is an accelerator design which makes use of several advanced or innovative technologies. The PIGMI Prototype consists of a 250 keV injector, a double harmonic buncher, a tape-wound 13 KG solenoid magnet, and four accelerator tanks with a total of 63 drift tubes of which 18 contain strong focusing quadrupoles of permanent magnets. The accelerator tanks are mild steel, copper-plated using a bright acid leveling technique. Drift tubes are stainless steel, fabricated using electron beam welding, shaped in a lathe and then copper plated. Drift tubes loaded with permanent magnets are sealed using laser welding. The samarium cobalt magnets are shaped by cutting and grinding techniques developed at Los Alamos.

  5. A novel procedure for generating solar irradiance TSYs

    Science.gov (United States)

    Fanego, Vicente Lara; Rubio, Jesús Pulgar; Peruchena, Carlos M. Fernández; Romeo, Martín Gastón; Tejera, Sara Moreno; Santigosa, Lourdes Ramírez; Balderrama, Rita X. Valenzuela; Tirado, Luis F. Zarzalejo; Pantaleón, Diego Bermejo; Pérez, Manuel Silva; Contreras, Manuel Pavón; García, Ana Bernardos; Anarte, Sergio Macías

    2017-06-01

    Typical Solar Years (TSYs) are key parameters for the solar energy industry. In particular, TSYs are mainly used for the design and bankability analysis of solar projects. In essence, a TSY intends to describe the expected long-term behavior of the solar resource (direct and/or global irradiance) into a condensed period of one year at the specific location of interest. A TSY differs from a conventional Typical Meteorological Year (TMY) by its absence of meteorological variables other than solar radiation. Concerning the probability of exceedance (Pe) needed for bankability, various scenarios are commonly used, with Pe90, Pe95 or even Pe99 being most usually required as unfavorable scenarios, along with the most widely used median scenario (Pe50). There is no consensus in the scientific community regarding the methodology for generating TSYs for any Pe scenario. Furthermore, the application of two different construction methods to the same original dataset could produce differing TSYs. Within this framework, a group of experts has been established by the Spanish Association for Standardization and Certification (AENOR) in order to propose a method that can be standardized. The method developed by this working group, referred to as the EVA method, is presented in this contribution. Its evaluation shows that it provides reasonable results for the two main irradiance components (direct and global), with low errors in the annual estimations for any given Pe. The EVA method also preserves the long-term statistics when the computed TSYs for a specific Pe are expanded from the monthly basis used in the generation of the TSY to higher time resolutions, such as 1 hour, which are necessary for the precise energy simulation of solar systems.

  6. 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.

  7. 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.

  8. System for operating solid oxide fuel cell generator on diesel fuel

    Science.gov (United States)

    Singh, Prabhu (Inventor); George, Raymond A. (Inventor)

    1997-01-01

    A system is provided for operating a solid oxide fuel cell generator on diesel fuel. The system includes a hydrodesulfurizer which reduces the sulfur content of commercial and military grade diesel fuel to an acceptable level. Hydrogen which has been previously separated from the process stream is mixed with diesel fuel at low pressure. The diesel/hydrogen mixture is then pressurized and introduced into the hydrodesulfurizer. The hydrodesulfurizer comprises a metal oxide such as ZnO which reacts with hydrogen sulfide in the presence of a metal catalyst to form a metal sulfide and water. After desulfurization, the diesel fuel is reformed and delivered to a hydrogen separator which removes most of the hydrogen from the reformed fuel prior to introduction into a solid oxide fuel cell generator. The separated hydrogen is then selectively delivered to the diesel/hydrogen mixer or to a hydrogen storage unit. The hydrogen storage unit preferably comprises a metal hydride which stores hydrogen in solid form at low pressure. Hydrogen may be discharged from the metal hydride to the diesel/hydrogen mixture at low pressure upon demand, particularly during start-up and shut-down of the system.

  9. Transfer of Plutonium-Uranium Extraction Plant and N Reactor irradiated fuel for storage at the 105-KE and 105-KW fuel storage basins, Hanford Site, Richland Washington

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-01

    The U.S. Department of Energy (DOE) needs to remove irradiated fuel from the Plutonium-Uranium Extraction (PUREX) Plant and N Reactor at the Hanford Site, Richland, Washington, to stabilize the facilities in preparation for decontamination and decommissioning (D&D) and to reduce the cost of maintaining the facilities prior to D&D. DOE is proposing to transfer approximately 3.9 metric tons (4.3 short tons) of unprocessed irradiated fuel, by rail, from the PUREX Plant in the 200 East Area and the 105 N Reactor (N Reactor) fuel storage basin in the 100 N Area, to the 105-KE and 105-KW fuel storage basins (K Basins) in the 100 K Area. The fuel would be placed in storage at the K Basins, along with fuel presently stored, and would be dispositioned in the same manner as the other existing irradiated fuel inventory stored in the K Basins. The fuel transfer to the K Basins would consolidate storage of fuels irradiated at N Reactor and the Single Pass Reactors. Approximately 2.9 metric tons (3.2 short tons) of single-pass production reactor, aluminum clad (AC) irradiated fuel in four fuel baskets have been placed into four overpack buckets and stored in the PUREX Plant canyon storage basin to await shipment. In addition, about 0.5 metric tons (0.6 short tons) of zircaloy clad (ZC) and a few AC irradiated fuel elements have been recovered from the PUREX dissolver cell floors, placed in wet fuel canisters, and stored on the canyon deck. A small quantity of ZC fuel, in the form of fuel fragments and chips, is suspected to be in the sludge at the bottom of N Reactor`s fuel storage basin. As part of the required stabilization activities at N Reactor, this sludge would be removed from the basin and any identifiable pieces of fuel elements would be recovered, placed in open canisters, and stored in lead lined casks in the storage basin to await shipment. A maximum of 0.5 metric tons (0.6 short tons) of fuel pieces is expected to be recovered.

  10. Three generation production biotechnology of biomass into bio-fuel

    Science.gov (United States)

    Zheng, Chaocheng

    2017-08-01

    The great change of climate change, depletion of natural resources, and scarcity of fossil fuel in the whole world nowadays have witnessed a sense of urgency home and abroad among scales of researchers, development practitioners, and industrialists to search for completely brand new sustainable solutions in the area of biomass transforming into bio-fuels attributing to our duty-that is, it is our responsibility to take up this challenge to secure our energy in the near future with the help of sustainable approaches and technological advancements to produce greener fuel from nature organic sources or biomass which comes generally from organic natural matters such as trees, woods, manure, sewage sludge, grass cuttings, and timber waste with a source of huge green energy called bio-fuel. Biomass includes most of the biological materials, livings or dead bodies. This energy source is ripely used industrially, or domestically for rather many years, but the recent trend is on the production of green fuel with different advance processing systems in a greener. More sustainable method. Biomass is becoming a booming industry currently on account of its cheaper cost and abundant resources all around, making it fairly more effective for the sustainable use of the bio-energy. In the past few years, the world has witnessed a remarkable development in the bio-fuel production technology, and three generations of bio-fuel have already existed in our society. The combination of membrane technology with the existing process line can play a vital role for the production of green fuel in a sustainable manner. In this paper, the science and technology for sustainable bio-fuel production will be introduced in detail for a cleaner world.

  11. A model for evolution of oxygen potential and stoichiometry deviation in irradiated UO 2 fuel

    Science.gov (United States)

    Ozrin, V. D.

    2011-12-01

    A model for radial redistribution of oxygen in irradiated UO 2 fuel under conditions of temperature and fission rate gradients has been developed. The oxygen transport in irradiated fuel is considered as a two-scale problem. On the local scale defined by the grain size, irradiated fuel is considered as a multi-phase system including solid solution of fission products in UO 2 matrix, solid precipitates (metal phase, grey phase of complex ternary compounds, the phase of condensed CsI) formed at the gas/solid interface and the gas phase in the intergranular bubbles. Intraganular transport of fission products is described by a set of diffusion equations which are supplemented by the condition of partial thermochemical equilibrium in the subsystem "precipitates & gas phase". The boundary conditions are formulated basing on thermochemical equilibrium on the interface of subsystems "solid solution" and "precipitates & gas phase". Calculation of the partial thermochemical equilibrium yields local values of the oxygen chemical potential and the deviation from fuel stoichiometry. On the global scale defined by the fuel pellet size, spatial variations of the oxygen potential caused by the temperature gradients or the presence of sources/sinks at the pellet boundary determine thermal diffusion fluxes resulting in redistribution of oxygen. The whole set of equations describing local equilibration and the transport in the local and global scales is solved in a self-consistent manner. The model results for radial distribution of oxygen potential of UO 2 calculated for typical reactor operating conditions and the fuel burnup up ˜100 MW d/kg HM are in satisfactory agreement with experimental data.

  12. Solid Oxide Fuel Cell Hybrid System for Distributed Power Generation

    Energy Technology Data Exchange (ETDEWEB)

    Faress Rahman; Nguyen Minh

    2004-01-04

    This report summarizes the work performed by Hybrid Power Generation Systems, LLC (HPGS) during the July 2003 to December 2003 reporting period under Cooperative Agreement DE-FC26-01NT40779 for the U. S. Department of Energy, National Energy Technology Laboratory (DOE/NETL) entitled ''Solid Oxide Fuel Cell Hybrid System for Distributed Power Generation''. The main objective of this project is to develop and demonstrate the feasibility of a highly efficient hybrid system integrating a planar Solid Oxide Fuel Cell (SOFC) and a micro-turbine. In addition, an activity included in this program focuses on the development of an integrated coal gasification fuel cell system concept based on planar SOFC technology. Also, another activity included in this program focuses on the development of SOFC scale up strategies.

  13. Loss-of-flow test L5 on FFTF-type irradiated fuel

    Energy Technology Data Exchange (ETDEWEB)

    Simms, R.; Gehl, S.M.; Lo, R.K.; Rothman, A.B.

    1978-03-01

    Test L5 simulated a hypothetical loss-of-flow accident in an LMFBR using three (Pu, U)O/sub 2/ fuel elements of the FTR type. The test elements were irradiated before TREAT Test L5 in the General Electric Test Reactor to 8 at. % burnup at about 40 kW/m. The preirradiation in GETR caused a fuel-restructuring range characteristic of moderate-power structure relative to the FTR. The test transient was devised so that a power burst would be initiated at incipient cladding melting after the loss of flow. The test simulation corresponds to a scenario for FTR in which fuel in high-power-structure subassemblies slump, resulting in a power excursion. The remaining subassemblies are subjected to this power burst. Test L5 addressed the fuel-motion behavior of the subassemblies in this latter category. Data from test-vehicle sensors, hodoscope, and post-mortem examinations were used to construct the sequence of events within the test zone. From these observations, the fuel underwent a predominantly dispersive event just after reaching a peak power six times nominal at, or after, scram. The fuel motion was apparently driven by the release of entrained fission-product gases, since fuel vapor pressure was deliberately kept below significant levels for the transient. The test remains show a wide range of microstructural evolution, depending on the extent of heat deposition along the active fuel column. Extensive fuel swelling was also observed as a result of the lack of the cladding restraint. The results of the thermal-hydraulic calculations with the SAS3A code agreed qualitatively with the postmortem results with respect to the extent of the melting and the dispersal of cladding and fuel. However, the calculated times of certain events did not agree with the observed times.

  14. HIGH EFFICIENCY GENERATION OF HYDROGEN FUELS USING NUCLEAR POWER

    Energy Technology Data Exchange (ETDEWEB)

    BROWN,LC; BESENBRUCH,GE; LENTSCH,RD; SCHULTZ,KR; FUNK,JF; PICKARD,PS; MARSHALL,AC; SHOWALTER,SK

    2003-06-01

    OAK B202 HIGH EFFICIENCY GENERATION OF HYDROGEN FUELS USING NUCLEAR POWER. Combustion of fossil fuels, used to power transportation, generate electricity, heat homes and fuel industry provides 86% of the world's energy. Drawbacks to fossil fuel utilization include limited supply, pollution, and carbon dioxide emissions. Carbon dioxide emissions, thought to be responsible for global warming, are now the subject of international treaties. Together, these drawbacks argue for the replacement of fossil fuels with a less-polluting potentially renewable primary energy such as nuclear energy. Conventional nuclear plants readily generate electric power but fossil fuels are firmly entrenched in the transportation sector. Hydrogen is an environmentally attractive transportation fuel that has the potential to displace fossil fuels. Hydrogen will be particularly advantageous when coupled with fuel cells. Fuel cells have higher efficiency than conventional battery/internal combustion engine combinations and do not produce nitrogen oxides during low-temperature operation. Contemporary hydrogen production is primarily based on fossil fuels and most specifically on natural gas. When hydrogen is produced using energy derived from fossil fuels, there is little or no environmental advantage. There is currently no large scale, cost-effective, environmentally attractive hydrogen production process available for commercialization, nor has such a process been identified. The objective of this work is to find an economically feasible process for the production of hydrogen, by nuclear means, using an advanced high-temperature nuclear reactor as the primary energy source. Hydrogen production by thermochemical water-splitting (Appendix A), a chemical process that accomplishes the decomposition of water into hydrogen and oxygen using only heat or, in the case of a hybrid thermochemical process, by a combination of heat and electrolysis, could meet these goals. Hydrogen produced from

  15. Fabrication of DUPIC Fuel for the 6th Irradiation Test at HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Woong Ki; Lee, D. Y.; Cho, K. H.; Kim, S. S.; Lee, J. W.; Lee, Jae W.; Park, G. I.; Lee, C. Y.; Yang, M. S

    2006-02-15

    In this study, 15 DUPIC pellets and two mini-elements were fabricated to precisely investigate the thermal characteristics of DUPIC fuel. As a result of the experiment, DUPIC pellets with 10.221{approx}10.278 g/cm{sup 3} (94.5{approx}95.0 % of T.D.) of sintered density and less than Ra 0.96 {mu}m of surface roughness satisfying the specifications of DUPIC fuel for the 6th irradiation test have been remotely fabricated at hot cell. 5 DUPIC pellets including 3 pellets equipped with thermal sensor in the center of the pellet were loaded in a mini-element. Endcap welding of the mini-element was performed by Nd:YAG laser. The DUPIC mini-elements assembled in an instrumented rig will be irradiated at HANARO research reactor.

  16. A self-regulating hydrogen generator for micro fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Moghaddam, Saeed; Pengwang, Eakkachai; Shannon, Mark A. [Mechanical Science and Engineering, University of Illinois at Urbana-Champaign, 1206 West Green Street, Urbana, IL 61801 (United States); Masel, Richard I. [Chemical and Biomolecular Engineering, University of Illinois at Urbana-Champaign, 213 Roger Adams Lab, 600 S. Mathews, Urbana, IL 61801 (United States)

    2008-10-15

    The ever-increasing power demands and miniaturization of portable electronics, micro-sensors and actuators, and emerging technologies such as cognitive arthropods have created a significant interest in development of micro fuel cells. One of the major challenges in development of hydrogen micro fuel cells is the fabrication and integration of auxiliary systems for generating, regulating, and delivering hydrogen gas to the membrane electrode assembly (MEA). In this paper, we report the development of a hydrogen gas generator with a micro-scale control system that does not consume any power. The hydrogen generator consists of a hydride reactor and a water reservoir, with a regulating valve separating them. The regulating valve consists of a port from the water reservoir and a movable membrane with via holes that permit water to flow from the reservoir to the hydride reactor. Water flows towards the hydride reactor, but stops within the membrane via holes due to capillary forces. Water vapor then diffuses from the via holes into the hydride reactor resulting in generation of hydrogen gas. When the rate of hydrogen consumed by the MEA is lower than the generation rate, gas pressure builds up inside the hydride reactor, deflecting the membrane, closing the water regulator valve, until the pressure drops, whereby the valve reopens. We have integrated the self-regulating micro hydrogen generator to a MEA and successfully conducted fuel cell tests under varying load conditions. (author)

  17. A self-regulating hydrogen generator for micro fuel cells

    Science.gov (United States)

    Moghaddam, Saeed; Pengwang, Eakkachai; Masel, Richard I.; Shannon, Mark A.

    The ever-increasing power demands and miniaturization of portable electronics, micro-sensors and actuators, and emerging technologies such as cognitive arthropods have created a significant interest in development of micro fuel cells. One of the major challenges in development of hydrogen micro fuel cells is the fabrication and integration of auxiliary systems for generating, regulating, and delivering hydrogen gas to the membrane electrode assembly (MEA). In this paper, we report the development of a hydrogen gas generator with a micro-scale control system that does not consume any power. The hydrogen generator consists of a hydride reactor and a water reservoir, with a regulating valve separating them. The regulating valve consists of a port from the water reservoir and a movable membrane with via holes that permit water to flow from the reservoir to the hydride reactor. Water flows towards the hydride reactor, but stops within the membrane via holes due to capillary forces. Water vapor then diffuses from the via holes into the hydride reactor resulting in generation of hydrogen gas. When the rate of hydrogen consumed by the MEA is lower than the generation rate, gas pressure builds up inside the hydride reactor, deflecting the membrane, closing the water regulator valve, until the pressure drops, whereby the valve reopens. We have integrated the self-regulating micro hydrogen generator to a MEA and successfully conducted fuel cell tests under varying load conditions.

  18. Development of SPND-instrumented rig of HANARO irradiation test of DUPIC fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, C. Y.; Mun, J. S.; Park, H. S.; Song, K. C.; Kang, K. H.; Jeong, I. H.; Yang, M. S. [KAERI, Taejon (Korea, Republic of)

    2001-10-01

    The 3rd irradiation test of DUPIC fuel, which was fabricated in the DFDF has been performed at HANARO. For the objectives of this irradiation test, The SPND-instrumented rig was designed and manufactured on the basis of the design specifications of the non-instrumented rig used in the last irradiation test. The newly designed irradiation rig was equipped with three Rh-type SPND sensors around DUPIC mini-elements for estimating the thermal neutron flux in the OR4 hole. Manufacturing of mini-elements and assembly of the irradiation rig were remotely done in the hot cells using the laser welding system and assembling equipments. The DUPIC rig is installed through the guide tube at the HANARO OR4 hole and the thermal neutron flux measured at that location is transmitted to the monitoring system. This irradiation test launched in June, 2001 at the HANARO OR4, and its anticipated discharge burnup is about 7,000 MWd/tHE.

  19. 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

  20. HIGH-TEMPERATURE SAFETY TESTING OF IRRADIATED AGR-1 TRISO FUEL

    Energy Technology Data Exchange (ETDEWEB)

    Stempien, John D.; Demkowicz, Paul A.; Reber, Edward L.; Chrisensen, Cad L.

    2016-11-01

    High-Temperature Safety Testing of Irradiated AGR-1 TRISO Fuel John D. Stempien, Paul A. Demkowicz, Edward L. Reber, and Cad L. Christensen Idaho National Laboratory, P.O. Box 1625 Idaho Falls, ID 83415, USA Corresponding Author: john.stempien@inl.gov, +1-208-526-8410 Two new safety tests of irradiated tristructural isotropic (TRISO) coated particle fuel have been completed in the Fuel Accident Condition Simulator (FACS) furnace at the Idaho National Laboratory (INL). In the first test, three fuel compacts from the first Advanced Gas Reactor irradiation experiment (AGR-1) were simultaneously heated in the FACS furnace. Prior to safety testing, each compact was irradiated in the Advanced Test Reactor to a burnup of approximately 15 % fissions per initial metal atom (FIMA), a fast fluence of 3×1025 n/m2 (E > 0.18 MeV), and a time-average volume-average (TAVA) irradiation temperature of about 1020 °C. In order to simulate a core-conduction cool-down event, a temperature-versus-time profile having a peak temperature of 1700 °C was programmed into the FACS furnace controllers. Gaseous fission products (i.e., Kr-85) were carried to the Fission Gas Monitoring System (FGMS) by a helium sweep gas and captured in cold traps featuring online gamma counting. By the end of the test, a total of 3.9% of an average particle’s inventory of Kr-85 was detected in the FGMS traps. Such a low Kr-85 activity indicates that no TRISO failures (failure of all three TRISO layers) occurred during the test. If released from the compacts, condensable fission products (e.g., Ag-110m, Cs-134, Cs-137, Eu-154, Eu-155, and Sr-90) were collected on condensation plates fitted to the end of the cold finger in the FACS furnace. These condensation plates were then analyzed for fission products. In the second test, five loose UCO fuel kernels, obtained from deconsolidated particles from an irradiated AGR-1 compact, were heated in the FACS furnace to a peak temperature of 1600 °C. This test had two

  1. 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

  2. HTGR fuels and core development program. Quarterly progress report for the period ending November 30, 1977. [Graphite and fuel irradiation; fission product release

    Energy Technology Data Exchange (ETDEWEB)

    1977-12-01

    The work reported here includes studies of basic fission product transport mechanisms, core graphite development and testing, and the development and testing of recyclable fuel systems. Materials studied include irradiation capsule tests of both fuel and graphite. Experimental procedures and results are discussed and data are presented.

  3. 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.

  4. Results of High-Temperature Heating Test for Irradiated Metallic Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, June-Hyung; Cheon, Jin-Sik; Lee, Byoung-Oon; Kim, Jun-Hwan; Kim, Hee-Moon; Yoo, Boung-Ok; Jung, Yang-Hong; Ahn, Sang-Bok; Lee, Chan-Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The U and Pu constituents in the fuel, however, tend to interact metallurgically with iron-based claddings at elevated temperatures during nominal steady-state operating conditions and off-normal reactor events. In particular, if the temperature is raised above the eutectic temperature of metallic fuel, e.g., in an off-normal reactor event, the fuel can form a mixture of liquid and solid phases that may promote further cladding interaction. Such fuel-cladding chemical interaction, in conjunction with fission gas pressure loading, can potentially shorten fuel pin lifetime and eventually cause cladding breach. In this work, microstructure observation results through microscope, SEM and EPMA are reported for the irradiated U-10Zr and U-10Zr-5Ce fuel slugs with T92 cladding after high-temperature heating test. Also, the measured eutectic penetration rate is compared with the prediction value by the existing eutectic penetration correlation being used for design and modelling purposes. Microstructure of the irradiated U-10Zr and U-10Zr-5Ce fuel slug with T92 cladding after high-temperature heating test were investigated through the microscope, SEM and EPMA. Also, the measured maximum eutectic penetration rate along cladding direction was compared with the prediction value by existing eutectic penetration correlation. In the case of U-10Zr/T92 specimen, migration phenomena of U, Zr, and Fe as well as Nd lanthanide fission product were observed at the eutectic melting region. The measured penetration rate was almost similar to prediction value by existing eutectic penetration rate correlation.

  5. A polymer electrolyte fuel cell stack for stationary power generation from hydrogen fuel

    Energy Technology Data Exchange (ETDEWEB)

    Zawodzinski, C.; Wilson, M.; Gottesfeld, S. [Los Alamos National Lab., NM (United States)

    1996-10-01

    The fuel cell is the most efficient device for the conversion of hydrogen fuel to electric power. As such, the fuel cell represents a key element in efforts to demonstrate and implement hydrogen fuel utilization for electric power generation. A central objective of a LANL/Industry collaborative effort supported by the Hydrogen Program is to integrate PEM fuel cell and novel stack designs at LANL with stack technology of H-Power Corporation (H-Power) in order to develop a manufacturable, low-cost/high-performance hydrogen/air fuel cell stack for stationary generation of electric power. A LANL/H-Power CRADA includes Tasks ranging from exchange, testing and optimization of membrane-electrode assemblies of large areas, development and demonstration of manufacturable flow field, backing and bipolar plate components, and testing of stacks at the 3-5 cell level and, finally, at the 4-5 kW level. The stack should demonstrate the basic features of manufacturability, overall low cost and high energy conversion efficiency. Plans for future work are to continue the CRADA work along the time line defined in a two-year program, to continue the LANL activities of developing and testing stainless steel hardware for longer term stability including testing in a stack, and to further enhance air cathode performance to achieve higher energy conversion efficiencies as required for stationary power application.

  6. Thorium utilization program progress report for January 1, 1974--June 30, 1975. [Reprocessing; refabrication; recycle fuel irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Lotts, A.L.; Kasten, P.R.

    1976-05-01

    Work was carried out on the following: HTGR reprocessing development and pilot plant, refabrication development and pilot plant, recycle fuel irradiations, engineering and economic studies, and conceptual design of a commercial recycle plant. (DLC)

  7. Power generation from furfural using the microbial fuel cell

    Science.gov (United States)

    Luo, Yong; Liu, Guangli; Zhang, Renduo; Zhang, Cuiping

    Furfural is a typical inhibitor in the ethanol fermentation process using lignocellulosic hydrolysates as raw materials. In the literature, no report has shown that furfural can be utilized as the fuel to produce electricity in the microbial fuel cell (MFC), a device that uses microbes to convert organic compounds to generate electricity. In this study, we demonstrated that electricity was successfully generated using furfural as the sole fuel in both the ferricyanide-cathode MFC and the air-cathode MFC. In the ferricyanide-cathode MFC, the maximum power densities reached 45.4, 81.4, and 103 W m -3, respectively, when 1000 mg L -1 glucose, a mixture of 200 mg L -1 glucose and 5 mM furfural, and 6.68 mM furfural were used as the fuels in the anode solution. The corresponding Coulombic efficiencies (CE) were 4.0, 7.1, and 10.2% for the three treatments, respectively. For pure furfural as the fuel, the removal efficiency of furfural reached up to 95% within 12 h. In the air-cathode MFC using 6.68 mM furfural as the fuel, the maximum values of power density and CE were 361 mW m -2 (18 W m -3) and 30.3%, respectively, and the COD removal was about 68% at the end of the experiment (about 30 h). Increase in furfural concentrations from 6.68 to 20 mM resulted in increase in the maximum power densities from 361 to 368 mW m -2, and decrease in CEs from 30.3 to 20.6%. These results indicated that some toxic and biorefractory organics such as furfural might still be suitable resources for electricity generation using the MFC technology.

  8. Uranium silicide pellet fabrication by powder metallurgy for accident tolerant fuel evaluation and irradiation

    Science.gov (United States)

    Harp, Jason M.; Lessing, Paul A.; Hoggan, Rita E.

    2015-11-01

    In collaboration with industry, Idaho National Laboratory is investigating uranium silicide for use in future light water reactor fuels as a more accident resistant alternative to uranium oxide base fuels. Specifically this project was focused on producing uranium silicide (U3Si2) pellets by conventional powder metallurgy with a density greater than 94% of the theoretical density. This work has produced a process to consistently produce pellets with the desired density through careful optimization of the process. Milling of the U3Si2 has been optimized and high phase purity U3Si2 has been successfully produced. Results are presented from sintering studies and microstructural examinations that illustrate the need for a finely ground reproducible particle size distribution in the source powder. The optimized process was used to produce pellets for the Accident Tolerant Fuel-1 irradiation experiment. The average density of these pellets was 11.54 ± 0.06 g/cm3. Additional characterization of the pellets by scanning electron microscopy and X-ray diffraction has also been performed. Pellets produced in this work have been encapsulated for irradiation, and irradiation in the Advanced Test Reactor is expected soon.

  9. A Renewably Powered Hydrogen Generation and Fueling Station Community Project

    Science.gov (United States)

    Lyons, Valerie J.; Sekura, Linda S.; Prokopius, Paul; Theirl, Susan

    2009-01-01

    The proposed project goal is to encourage the use of renewable energy and clean fuel technologies for transportation and other applications while generating economic development. This can be done by creating an incubator for collaborators, and creating a manufacturing hub for the energy economy of the future by training both white- and blue-collar workers for the new energy economy. Hydrogen electrolyzer fueling stations could be mass-produced, shipped and installed in collaboration with renewable energy power stations, or installed connected to the grid with renewable power added later.

  10. An integrated MEMS infrastructure for fuel processing: hydrogen generation and separation for portable power generation

    Science.gov (United States)

    Varady, M. J.; McLeod, L.; Meacham, J. M.; Degertekin, F. L.; Fedorov, A. G.

    2007-09-01

    Portable fuel cells are an enabling technology for high efficiency and ultra-high density distributed power generation, which is essential for many terrestrial and aerospace applications. A key element of fuel cell power sources is the fuel processor, which should have the capability to efficiently reform liquid fuels and produce high purity hydrogen that is consumed by the fuel cells. To this end, we are reporting on the development of two novel MEMS hydrogen generators with improved functionality achieved through an innovative process organization and system integration approach that exploits the advantages of transport and catalysis on the micro/nano scale. One fuel processor design utilizes transient, reverse-flow operation of an autothermal MEMS microreactor with an intimately integrated, micromachined ultrasonic fuel atomizer and a Pd/Ag membrane for in situ hydrogen separation from the product stream. The other design features a simpler, more compact planar structure with the atomized fuel ejected directly onto the catalyst layer, which is coupled to an integrated hydrogen selective membrane.

  11. Test design description Volume 2, Part 1. IFR-1 metal fuel irradiation test (AK-181) element as-built data

    Energy Technology Data Exchange (ETDEWEB)

    Dodds, N. E.

    1986-06-01

    The IFR-1 Test, designated as the AK-181 Test Assembly, will be the first irradiation test of wire wrapped, sodium-bonded metallic fuel elements in the Fast Flux Test Facility (FFTF). The test is part of the Integral Fast Reactor (IFR) fuels program conducted by Argonne National Laboratory (ANL) in support of the Innovative Reactor Concepts Program sponsored by the US Department of Energy (DOE). One subassembly, containing 169 fuel elements, will be irradiated for 600 full power days to achieve 10 at.% burnup. Three metal fuel alloys (U-10Zr, U-8Pu-10Zr) will be irradiated in D9 cladding tubes. The metal fuel elements have a fuel-smeared density of 75% and each contains five slugs. The enriched zone contains three slugs and is 36-in. long. One 6.5-in. long depleted uranium axial blanket slug (DU-10Zr) was loaded at each end of the enriched zone. the fuel elements were fabricated at ANL-W and delivered to Westinghouse-Hanford for wirewrapping and assembly into the test article. This Test Design Description contains relevant data on compositions, densities, dimensions and weights for the cast fuel slugs and completed fuel elements. The elements conform to the requirements in MG-22, "Users` Guide for the Irradiation of Experiments in the FTR."

  12. 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.

  13. Determination of plutonium resent in highly radioactive irradiated fuel solution by spectrophotometric method

    Energy Technology Data Exchange (ETDEWEB)

    Dhamodharam, Krishnan [Reprocessing Group, Indira Gandhi Centre for Atomic Research, Kalpakkam (India); Pius, Anitha [The Gandhigram Rural Institute - Deemed University, Gandhigram (India)

    2016-06-15

    A simple and rapid spectrophotometric method has been developed to enable the determination of plutonium concentration in an irradiated fuel solution in the presence of all fission products. An excess of ceric ammonium nitrate solution was employed to oxidize all the valence states of plutonium to +6 oxidation state. Interference due to the presence of fission products such as ruthenium and zirconium, and corrosion products such as iron in the envisaged concentration range, as in the irradiated fuel solution, was studied in the determination of plutonium concentration by the direct spectrophotometric method. The stability of plutonium in +6 oxidation state was monitored under experimental conditions as a function of time. Results obtained are reproducible, and this method is applicable to radioactive samples resulting before the solvent extraction process during the reprocessing of fast reactor spent fuel. An analysis of the concentration of plutonium shows a relative standard deviation of <1.2% in standard as well as in simulated conditions. This reflects the fast reactor fuel composition with respect to uranium, plutonium, fission products such as ruthenium and zirconium, and corrosion products such as iron.

  14. Partitioning of selected fission products from irradiated oxide fuel induced by thermal treatment

    Science.gov (United States)

    Shcherbina, Natalia; Kivel, Niko; Günther-Leopold, Ines

    2013-06-01

    The release of fission products (FPs) from spent nuclear fuel (SNF) has been studied as a function of the temperature and redox conditions. The present paper concerns essentially the high temperature separation of Cs and Sr from irradiated pressurized (PWR) and boiling water reactor (BWR) fuel of different burn-up levels with use of an in-house designed system for inductive vaporization (InVap). Using thermodynamic calculations with the Module of Fission Product Release (MFPR) code along with annealing experiments on SNF in the InVap it was shown that the speciation of Cs and Sr, hence their release behavior at high temperature, is sensitive to the redox conditions during thermal treatment. It was demonstrated that annealing conditions in the InVap can be adjusted in the way to promote the release of selected FPs without significant loss of the fuel matrix or actinides: complete release of Cs and I was achieved during treatment of irradiated fuel at 1800 °C under reducing atmosphere (0.7% H2/Ar mixture). The developed partitioning procedure can be used for the SNF pretreatment as an advanced head-end step in the hydrometallurgical or pyrochemical reprocessing technology.

  15. Partitioning of selected fission products from irradiated oxide fuel induced by thermal treatment

    Energy Technology Data Exchange (ETDEWEB)

    Shcherbina, Natalia, E-mail: natalia.shcherbina@psi.ch [Department of Nuclear Energy and Safety, Paul Scherrer Institut (PSI), Villigen 5232 (Switzerland); Kivel, Niko; Günther-Leopold, Ines [Department of Nuclear Energy and Safety, Paul Scherrer Institut (PSI), Villigen 5232 (Switzerland)

    2013-06-15

    The release of fission products (FPs) from spent nuclear fuel (SNF) has been studied as a function of the temperature and redox conditions. The present paper concerns essentially the high temperature separation of Cs and Sr from irradiated pressurized (PWR) and boiling water reactor (BWR) fuel of different burn-up levels with use of an in-house designed system for inductive vaporization (InVap). Using thermodynamic calculations with the Module of Fission Product Release (MFPR) code along with annealing experiments on SNF in the InVap it was shown that the speciation of Cs and Sr, hence their release behavior at high temperature, is sensitive to the redox conditions during thermal treatment. It was demonstrated that annealing conditions in the InVap can be adjusted in the way to promote the release of selected FPs without significant loss of the fuel matrix or actinides: complete release of Cs and I was achieved during treatment of irradiated fuel at 1800 °C under reducing atmosphere (0.7% H{sub 2}/Ar mixture). The developed partitioning procedure can be used for the SNF pretreatment as an advanced head-end step in the hydrometallurgical or pyrochemical reprocessing technology.

  16. Comparison of Material Behavior of Matrix Graphite for HTGR Fuel Elements upon Irradiation: A literature Survey

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young-Woo; Yeo, Seunghwan; Cho, Moon Sung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The fuel elements for the HTGRs (i.e., spherical fuel element in pebble-bed type core design and fuel compact in prismatic core design) consists of coated fuel particles dispersed and bonded in a closely packed array within a carbonaceous matrix. This matrix is generally made by mixing fully graphitized natural and needle- or pitchcoke originated powders admixed with a binder material (pitch or phenolic resin), The resulting resinated graphite powder mixture, when compacted, may influence a number of material properties as well as its behavior under neutron irradiation during reactor operation. In the fabrication routes of these two different fuel element forms, different consolidation methods are employed; a quasi-isostatic pressing method is generally adopted to make pebbles while fuel compacts are fabricated by uni-axial pressing mode. The result showed that the hardness values obtained from the two directions showed an anisotropic behavior: The values obtained from the perpendicular section showed much higher micro hardness (176.6±10.5MPa in average) than from the parallel section ((125.6±MPa in average). This anisotropic behavior was concluded to be related to the microstructure of the matrix graphite. This may imply that the uni-axial pressing method to make compacts influence the microstructure of the matrix and hence the material properties of the matrix graphite.

  17. Corrosion of irradiated MOX fuel in presence of dissolved H 2

    Science.gov (United States)

    Carbol, P.; Fors, P.; Van Winckel, S.; Spahiu, K.

    2009-07-01

    The corrosion behaviour of irradiated MOX fuel (47 GWd/tHM) has been studied in an autoclave experiment simulating repository conditions. Fuel fragments were corroded at room temperature in a 10 mM NaCl/2 mM NaHCO 3 solution in presence of dissolved H 2 for 2100 days. The results show that dissolved H 2 in concentration 1 mM and higher inhibits oxidation and dissolution of the fragments. Stable U and Pu concentrations were measured at 7 × 10 -10 and 5 × 10 -11 M, respectively. Caesium was only released during the first two years of the experiment. The results indicate that the UO 2 matrix of a spent MOX fuel is the main contributor to the measured dissolution, while the corrosion of the high burn-up Pu-rich islands appears negligible.

  18. 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

  19. Restructuring and redistribution of actinides in Am-MOX fuel during the first 24 h of irradiation

    Science.gov (United States)

    Tanaka, Kosuke; Miwa, Shuhei; Sekine, Shin-ichi; Yoshimochi, Hiroshi; Obayashi, Hiroshi; Koyama, Shin-ichi

    2013-09-01

    In order to confirm the effect of minor actinide additions on the irradiation behavior of MOX fuel pellets, 3 wt.% and 5 wt.% americium-containing MOX (Am-MOX) fuels were irradiated for 10 min at 43 kW/m and for 24 h at 45 kW/m in the experimental fast reactor Joyo. Two nominal values of the fuel pellet oxygen-to-metal ratio (O/M), 1.95 and 1.98, were used as a test parameter. Emphasis was placed on the behavior of restructuring and redistribution of actinides which directly affect the fuel performance and the fuel design for fast reactors. Microstructural evolutions in the fuels were observed by optical microscopy and the redistribution of constituent elements was determined by EPMA using false color X-ray mapping and quantitative point analyses. The ceramography results showed that structural changes occurred quickly in the initial stage of irradiation. Restructuring of the fuel from middle to upper axial positions developed and was almost completed after the 24-h irradiation. No sign of fuel melting was found in any of the specimens. The EPMA results revealed that Am as well as Pu migrated radially up the temperature gradient to the center of the fuel pellet. The increase in Am concentration on approaching the edge of the central void and its maximum value were higher than those of Pu after the 10-min irradiation and the difference was more pronounced after the 24-h irradiation. The increment of the Am and Pu concentrations due to redistribution increased with increasing central void size. In all of the specimens examined, the extent of redistribution of Am and Pu was higher in the fuel of O/M ratio of 1.98 than in that of 1.95.

  20. Metal Fuel Development and Verification for Prototype Generation IV Sodium-Cooled Fast Reactor

    OpenAIRE

    Chan Bock Lee; Jin Sik Cheon; Sung Ho Kim; Jeong-Yong Park; Hyung-Kook Joo

    2016-01-01

    Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR) to be built by 2028. U–Zr fuel is a driver for the initial core of the PGSFR, and U–transuranics (TRU)–Zr fuel will gradually replace U–Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U–Zr fuel, work on U–Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U–TRU–Zr fuel uses TRU recovered through pyroelectrochem...

  1. Secondary particle tracks generated by ion beam irradiation

    Science.gov (United States)

    García, Gustavo

    2015-05-01

    The Low Energy Particle Track Simulation (LEPTS) procedure is a powerful complementary tool to include the effect of low energy electrons and positrons in medical applications of radiation. In particular, for ion-beam cancer treatments provides a detailed description of the role of the secondary electrons abundantly generated around the Bragg peak as well as the possibility of using transmuted positron emitters (C11, O15) as a complement for ion-beam dosimetry. In this study we present interaction probability data derived from IAM-SCAR corrective factors for liquid environments. Using these data, single electron and positron tracks in liquid water and pyrimidine have been simulated providing information about energy deposition as well as the number and type of interactions taking place in any selected ``nanovolume'' of the irradiated area. In collaboration with Francisco Blanco, Universidad Complutense de Madrid; Antonio Mu noz, Centro de Investigaciones Energéticas Medioambientales y Tecnológicas and Diogo Almeida, Filipe Ferreira da Silva, Paulo Lim ao-Vieira, Universidade Nova de Lisboa. Supported by the Spanish and Portuguese governments.

  2. 40 CFR 80.535 - How are NRLM diesel fuel credits generated?

    Science.gov (United States)

    2010-07-01

    ... PROGRAMS (CONTINUED) REGULATION OF FUELS AND FUEL ADDITIVES Motor Vehicle Diesel Fuel; Nonroad, Locomotive... the standards of § 80.510(a) or (b). V520 = The total volume of motor vehicle diesel fuel produced or... generated by both a foreign refiner and by an importer for the same motor vehicle diesel fuel. (iii)...

  3. AGR-1 Fuel Compact 6-3-2 Post-Irradiation Examination Results

    Energy Technology Data Exchange (ETDEWEB)

    Paul demkowicz; jason Harp; Scott Ploger

    2012-12-01

    Destructive post-irradiation examination was performed on fuel Compact 6-3-2, which was irradiated in the AGR-1 experiment to a final compact average burnup of 11.3% FIMA and a time-average, volume-average temperature of 1070°C. The analysis of this compact was focused on characterizing the extent of fission product release from the particles and examining particles to determine the condition of the kernels and coating layers. The work included deconsolidation of the compact and leach-burn-leach analysis, visual inspection and gamma counting of individual particles, measurement of fuel burnup by several methods, metallurgical preparation of selected particles, and examination of particle cross-sections with optical microscopy. A single particle with a defective SiC layer was identified during deconsolidation-leach-burn-leach analysis, which is in agreement with previous measurements showing elevated cesium in the Capsule 6 graphite fuel holder associated with this fuel compact. The fraction of the compact europium inventory released from the particles and retained in the matrix was relatively high (approximately 6E-3), indicating release from intact particle coatings. The Ag-110m inventory in individual particles exhibited a very broad distribution, with some particles retaining =80% of the predicted inventory and others retaining less than 25%. The average degree of Ag-110m retention in 60 gamma counted particles was approximately 50%. This elevated silver release is in agreement with analysis of silver on the Capsule 6 components, which indicated an average release of 38% of the Capsule 6 inventory from the fuel compacts. In spite of the relatively high degree of silver release from the particles, virtually none of the Ag-110m released was found in the compact matrix, and presumably migrated out of the compact and was deposited on the irradiation capsule components. Release of all other fission products from the particles appears to be less than a single

  4. PIGMI: a design report for Pion Generator for Medical Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Hansborough, L.D. (comp.)

    1981-09-01

    PIGMI (Pion Generator for Medical Irradiations) is an integrated linear accelerator (linac) system developed under the auspices of the National Cancer Institute for specific application to cancer treatment in a hospital environment. In its full configuration, PIGMI is a proton linac that is far smaller, less expensive, and more reliable than previous machines that produce pions. Subsets of PIGMI technology can be used with equal advantage to generate beams of other particles (such as neutrons, protons, or heavy ions) that may be of interest for radiotherapy, radioisotope production, or other applications. The dramatic performance and cost advantages of this new breed of acceleraor result from a number of improvements. In the low-energy portion of the machine, a new type of low-energy linac (the radio-frequency quadrupole(RFQ)) produces an exceptionally good quality beam, and uses a very simple 30-kV injector. In the second part of the machine (the drift-tube linac (DTL)), high accelerating gradients are now achievable with consequent reductions in machine length. Another new structure (the disk and washer (DAW)) will be used in the third and final section of the accelerator; this portion will also be relatively short and require few power amplifiers. The entire machine is designed for ease of operation and high reliability. The pion-production machine, discussed in this report, accelerates a 100-..mu..A average proton-beam current to 650 MeV; use of an efficient pion-collection channel would result in an average pion flux of over 100 rad/min in a volume of about 1 l. Pion-channel design is not treated in this report. Accelerator construction cost is estimated at $10 million (1980 dollars); site preparation and treatment facility costs would bring the cost of a complete facility to an estimated $25 million.

  5. Comparison of US and FRG post-irradiation examination procedures to measure statistically significant failure fractions of irradiated coated-particle fuels. [HTGR

    Energy Technology Data Exchange (ETDEWEB)

    Kania, M.J.; Homan, F.J.; Mehner, A.W.

    1982-08-01

    Two methods for measuring failure fraction on irradiated coated-particle fuels have been developed, one in the United States (the IMGA system - Irradiated-Microsphere Gamma Analyzer) and one in the Federal Republic of Germany (FRG) (the PIAA procedure - Postirradiation Annealing and Beta Autoradiography). A comparison of the two methods on two standardized sets of irradiated particles was undertaken to evaluate the accuracy, operational procedures, and expense of each method in obtaining statistically significant results. From the comparison, the postirradiation examination method employing the IMGA system was found to be superior to the PIAA procedure for measuring statistically significant failure fractions. Both methods require that the irradiated fuel be in the form of loose particles, each requires extensive remote hot-cell facilities, and each is capable of physically separating failed particles from unfailed particles. Important differences noted in the comparison are described.

  6. Irradiated-Microsphere Gamma Analyzer (IMGA): an integrated system for HTGR coated particle fuel performance assessment

    Energy Technology Data Exchange (ETDEWEB)

    Kania, M.J.; Valentine, K.H.

    1980-02-01

    The Irradiated-Microsphere Gamma Analyzer (IMGA) System, designed and built at ORNL, provides the capability of making statistically accurate failure fraction measurements on irradiated HTGR coated particle fuel. The IMGA records the gamma-ray energy spectra from fuel particles and performs quantitative analyses on these spectra; then, using chemical and physical properties of the gamma emitters it makes a failed-nonfailed decision concerning the ability of the coatings to retain fission products. Actual retention characteristics for the coatings are determined by measuring activity ratios for certain gamma emitters such as /sup 137/Cs//sup 95/Zr and /sup 144/Ce//sup 95/Zr for metallic fission product retention and /sup 134/Cs//sup 137/Cs for an indirect measure of gaseous fission product retention. Data from IMGA (which can be put in the form of n failures observed in N examinations) can be accurately described by the binomial probability distribution model. Using this model, a mathematical relationship between IMGA data (n,N), failure fraction, and confidence level was developed. To determine failure fractions of less than or equal to 1% at confidence levels near 95%, this model dictates that from several hundred to several thousand particles must be examined. The automated particle handler of the IMGA system provides this capability. As a demonstration of failure fraction determination, fuel rod C-3-1 from the OF-2 irradiation capsule was analyzed and failure fraction statistics were applied. Results showed that at the 1% failure fraction level, with a 95% confidence level, the fissile particle batch could not meet requirements; however, the fertile particle exceeded these requirements for the given irradiation temperature and burnup.

  7. Evaluation of FFTF fuel pin design procedure vis-a-vis steady state irradiation performance in EBR II

    Energy Technology Data Exchange (ETDEWEB)

    Jackson, R.J.

    1976-11-01

    The FFTF fuel pin design analysis is shown to be conservative through comparison with pin irradiation experience in EBR-II. This comparison shows that the actual lifetimes of EBR-II fuel pins are either greater than 80,000 MWd/MTM or greater than the calculated allowable lifetimes based on thermal creep strain.

  8. Economical analysis of combined fuel cell generators and absorption chillers

    Directory of Open Access Journals (Sweden)

    M. Morsy El-Gohary

    2013-06-01

    Full Text Available This paper presents a co-generation system based on combined heat and power for commercial units. For installation of a co-generation system, certain estimates for this site should be performed through making assessments of electrical loads, domestic water, and thermal demand. This includes domestic hot water, selection of the type of power generator, fuel cell, and the type of air conditioning system, and absorption chillers. As a matter of fact, the co-generation system has demonstrated good results for both major aspects, economic and environmental. From the environmental point of view, this can be considered as an ideal solution for problems concerned with the usage of Chlorofluoro carbons. On the other hand, from the economic point of view, the cost analysis has revealed that the proposed system saves 4% of total cost through using the co-generation system.

  9. A Bio-Based Fuel Cell for Distributed Energy Generation

    Energy Technology Data Exchange (ETDEWEB)

    Anthony Terrinoni; Sean Gifford

    2008-06-30

    The technology we propose consists primarily of an improved design for increasing the energy density of a certain class of bio-fuel cell (BFC). The BFCs we consider are those which harvest electrons produced by microorganisms during their metabolism of organic substrates (e.g. glucose, acetate). We estimate that our technology will significantly enhance power production (per unit volume) of these BFCs, to the point where they could be employed as stand-alone systems for distributed energy generation.

  10. Integration of direct carbon and hydrogen fuel cells for highly efficient power generation from hydrocarbon fuels

    Energy Technology Data Exchange (ETDEWEB)

    Muradov, Nazim; Choi, Pyoungho; Smith, Franklyn; Bokerman, Gary [Florida Solar Energy Center, University of Central Florida, 1679 Clearlake Road, Cocoa, FL 32922-5703 (United States)

    2010-02-15

    In view of impending depletion of hydrocarbon fuel resources and their negative environmental impact, it is imperative to significantly increase the energy conversion efficiency of hydrocarbon-based power generation systems. The combination of a hydrocarbon decomposition reactor with a direct carbon and hydrogen fuel cells (FC) as a means for a significant increase in chemical-to-electrical energy conversion efficiency is discussed in this paper. The data on development and operation of a thermocatalytic hydrocarbon decomposition reactor and its coupling with a proton exchange membrane FC are presented. The analysis of the integrated power generating system including a hydrocarbon decomposition reactor, direct carbon and hydrogen FC using natural gas and propane as fuels is conducted. It was estimated that overall chemical-to-electrical energy conversion efficiency of the integrated system varied in the range of 49.4-82.5%, depending on the type of fuel and FC used, and CO{sub 2} emission per kW{sub el}h produced is less than half of that from conventional power generation sources. (author)

  11. Biogas as a fuel source for SOFC co-generators

    Science.gov (United States)

    Van herle, Jan; Membrez, Yves; Bucheli, Olivier

    This study reports on the combination of solid oxide fuel cell (SOFC) generators fueled with biogas as renewable energy source, recoverable from wastes but at present underexploited. From a mobilisable near-future potential in the European Union (EU-15) of 17 million tonnes oil equivalent (Mtoe), under 15% appears to be converted today into useful heat and power (2 Mtoe). SOFCs could improve and promote the exploitation of biogas on manifold generation sites as small combined heat and power (5-50 kW el), especially for farm and sewage installations, raising the electrical conversion efficiency on such reduced and variable power level. Larger module packs of the high temperature ceramic converter would also be capable of operating on contaminated fuel of low heating value (less than 40% that of natural gas) which can emanate from landfill sites (MW-size). Landfill gas delivers 80% of current world biogas production. This document compiles and estimates biogas data on actual production and future potential and presents the thermodynamics of the biogas reforming and electrochemical conversion processes. A case study is reported of the energy balance of a small SOFC co-generator operated with agricultural biogas, the largest potential source.

  12. Design of high temperature irradiation materials inspection cells. (Spent fuel inspection cells) in the High Temperature Engineering Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ino, Hiroichi; Ueta, Shouhei; Suzuki, Hiroshi; Sawa, Kazuhiro [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Tobita, Tsutomu [Nuclear Engineering Company, Ltd., Tokai, Ibaraki (Japan)

    2002-01-01

    This report summarizes design requirements and design results for shields, ventilation system and fuel handling devices for the high temperature irradiation materials inspection cells (spent fuel inspection cells). These cells are small cells to carry out few post-irradiation examinations of spent fuels, specimen, etc., which are irradiated in the High Temperature Engineering Test Reactor, since the cells should be built in limited space in the HTTR reactor building, the cells are designed considering relationship between the cells and the reactor building to utilize the limited space effectively. The cells consist of three partitioned hot cells with wall for neutron and gamma-ray shields, ventilation system including filtering units and fuel handling devices. The post-irradiation examinations of the fuels and materials are planed by using the cells and the Hot Laboratory of the Japan Materials Testing Reactor to establish the technology basis on high temperature gas-cooled reactors (HTGRs). In future, irradiation tests and post-irradiation examinations will be carried out with the cells to upgrade present HTGR technologies and to make the innovative basic research on high-temperature engineering. (author)

  13. 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.

  14. 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.

  15. Distributed generation system using wind/photovoltaic/fuel cell

    Science.gov (United States)

    Buasri, Panhathai

    This dissertation investigates the performance and the operation of a distributed generation (DG) power system using wind/photovoltaic/fuel cell (W/PV/FC). The power system consists of a 2500 W photovoltaic array subsystem, a 500 W proton exchange membrane fuel cell (PEMFC) stack subsystem, 300 W wind turbine, 500 W wind turbine, and 1500 W wind energy conversion subsystems. To extract maximum power from the PV, a maximum power point tracker was designed and fabricated. A 4 kW single phase inverter was used to convert the DC voltage to AC voltage; also a 44 kWh battery bank was used to store energy and prevent fluctuation of the power output of the DG system. To connect the fuel cell to the batteries, a DC/DC controller was designed and fabricated. To monitor and study the performance of the DG system under variable conditions, a data acquisition system was designed and installed. The fuel cell subsystem performance was evaluated under standalone operation using a variable resistance and under interactive mode, connected to the batteries. The manufacturing data and the experimental data were used to develop an electrical circuit model to the fuel cell. Furthermore, harmonic analysis of the DG system was investigated. For an inverter, the AC voltage delivered to the grid changed depending on the time, load, and electronic equipment that was connected. The quality of the DG system was evaluated by investigating the harmonics generated by the power electronics converters. Finally, each individual subsystem of the DG system was modeled using the neuro-fuzzy approach. The model was used to predict the performance of the DG system under variable conditions, such as passing clouds and wind gust conditions. The steady-state behaviors of the model were validated by the experimental results under different operating conditions.

  16. 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.

  17. Feasibility study of on-line digital X-ray imaging for irradiated fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Parthoens, Y.; Gys, A. [Reactor Material Research Department, SCK-CEN, Mol (Belgium); Smolders, V. [Industrial Engineer Department, Katholieke Hogeschool Kempen, Geel (Belgium)

    2003-07-01

    At the Reactor Material Research Department of the Belgian Nuclear Research Centre SCK-CEN Xray imaging of the internal parts of irradiated fuel rods is done on silver-halide films using a 420 kV X-ray source. The replacement of the films by an on-line digital X-ray imaging system implies several advantages. Images can be evaluated instantly and source parameters can be optimized more easily. Time consuming film development is superfluous. The images can digitally be enhanced, processed, reported and archived. Within this work the feasibility of four commercial on-line digital X-ray imaging systems were studied for post-irradiation examination on fuel rods in a hot cell environment. The criteria to evaluate the systems were image quality, integration in the existing hot cell infrastructure, durability and cost price. For the evaluation and comparison of the image quality a simulation fuel rod was fabricated. Three systems suffered from lack of sensitivity, contrast and/or resolution. Only the CsI-scintillator coupled to a CCD-camera with image intensifier gave a sufficient image quality. On the other hand the image intensifiers' dimensions are difficult to integrate in the existing hot cell infrastructure. Also the durability of intensifier screens is questionable as they are susceptible to image burn. Smaller image intensifiers easier to integrate are commercial available nowadays.

  18. 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.

  19. Pyrometallurgical separation processes of radionuclides contained in the irradiated nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    De Cordoba, Guadalupe; Caravaca, Concha; Quinones, Javier; Gonzalez de la Huebra, Angel

    2005-01-01

    Faced with the new options for the high level waste management, the ''Partitioning and Transmutation (P and T)'' of the radio nuclides contained in the irradiated nuclear fuel appear as a promising option from different points of view, such as environmental risk, radiotoxic inventory reduction, economic, etc.. The present work is part of a research project called ''PYROREP'' of the 5th FWP of the EU that studied the feasibility of the actinide separation from the rest of fission products contained in the irradiated nuclear fuel by pyrometallurgical processes with the aim of their transmutation. In order to design these processes it is necessary to determine basic thermodynamic and kinetic data of the radionuclides contained in the nuclear fuel in molten salt media. The electrochemical study of uranium, samarium and molybdenum in the eutectic melt LiCl - KCl has been performed at a tungsten electrode in the temperature range of 450 - 600 deg C in order to obtain these basic properties. (Author)

  20. Irradiation Performance of U-Mo Alloy Based ‘Monolithic’ Plate-Type Fuel – Design Selection

    Energy Technology Data Exchange (ETDEWEB)

    A. B. Robinson; G. S. Chang; D. D. Keiser, Jr.; D. M. Wachs; D. L. Porter

    2009-08-01

    A down-selection process has been applied to the U-Mo fuel alloy based monolithic plate fuel design, supported by irradiation testing of small fuel plates containing various design parameters. The irradiation testing provided data on fuel performance issues such as swelling, fuel-cladding interaction (interdiffusion), blister formation at elevated temperatures, and fuel/cladding bond quality and effectiveness. U-10Mo (wt%) was selected as the fuel alloy of choice, accepting a somewhat lower uranium density for the benefits of phase stability. U-7Mo could be used, with a barrier, where the trade-off for uranium density is critical to nuclear performance. A zirconium foil barrier between fuel and cladding was chosen to provide a predictable, well-bonded, fuel-cladding interface, allowing little or no fuel-cladding interaction. The fuel plate testing conducted to inform this selection was based on the use of U-10Mo foils fabricated by hot co-rolling with a Zr foil. The foils were subsequently bonded to Al-6061 cladding by hot isostatic pressing or friction stir bonding.

  1. Hydrogen generation at ambient conditions: application in fuel cells.

    Science.gov (United States)

    Boddien, Albert; Loges, Björn; Junge, Henrik; Beller, Matthias

    2008-01-01

    The efficient generation of hydrogen from formic acid/amine adducts at ambient temperature is demonstrated. The highest catalytic activity (TOF up to 3630 h(-1) after 20 min) was observed in the presence of in situ generated ruthenium phosphine catalysts. Compared to the previously known methods to generate hydrogen from liquid feedstocks, the systems presented here can be operated at room temperature without the need for any high-temperature reforming processes, and the hydrogen produced can then be directly used in fuel cells. A variety of Ru precursors and phosphine ligands were investigated for the decomposition of formic acid/amine adducts. These catalytic systems are particularly interesting for the generation of H2 for new applications in portable electric devices.

  2. Results of Recent Microstructural Characterization of Irradiated U-Mo Dispersion Fuels with Al Alloy Matrices that Contain Si

    Energy Technology Data Exchange (ETDEWEB)

    D D. Keiser, Jr.; A. B. Robinson; D. E. Janney; J. F. Jue

    2008-03-01

    RERTR U-Mo dispersion fuel plates are being developed for application in research reactors throughout the world. As part of this development, reactor experiments are being conducted in the Advanced Test Reactor to determine the irradiation performance of different dispersion fuels that contain U-Mo alloys with different Mo contents and Al alloy matrices with different Si contents. Of particular interest is the performance of the dispersion fuels depending on the Si content of the Al alloy matrix, since the addition of Si is being looked to for improving the performance of these dispersion fuels. This paper will describe the results of recent microstructural examinations that have been performed using optical metallography and scanning electron microscopy on as-fabricated and as-irradiated dispersion fuels with different amounts of Si added to the Al matrix. Differences in the microstructural development during irradiation as a function of the Si content in the Al matrix will be discussed, and comments will be made about the development and stability of the fuel/matrix interaction layers that are commonly present in irradiated dispersion fuels.

  3. 75 FR 66008 - Fossil Fuel-Generated Energy Consumption Reduction for New Federal Buildings and Major...

    Science.gov (United States)

    2010-10-27

    ... Parts 433 and 435 RIN 1904-AB96 Fossil Fuel-Generated Energy Consumption Reduction for New Federal... proposed rulemaking (NOPR) regarding the fossil fuel- generated energy consumption ] requirements for new... regarding the fossil fuel-generated energy consumption requirements for new Federal buildings and...

  4. Swelling behavior detection of irradiated U-10Zr alloy fuel using indirect neutron radiography

    Science.gov (United States)

    Sun, Yong; Huo, He-yong; Wu, Yang; Li, Jiangbo; Zhou, Wei; Guo, Hai-bing; Li, Hang; Cao, Chao; Yin, Wei; Wang, Sheng; Liu, Bin; Feng, Qi-jie; Tang, Bin

    2016-11-01

    It is hopeful that fusion-fission hybrid energy system will become an effective approach to achieve long-term sustainable development of fission energy. U-10Zr alloy (which means the mass ratio of Zr is 10%) fuel is the key material of subcritical blanket for fusion-fission hybrid energy system which the irradiation performance need to be considered. Indirect neutron radiography is used to detect the irradiated U-10Zr alloy because of the high residual dose in this paper. Different burnup samples (0.1%, 0.3%, 0.5% and 0.7%) have been tested with a special indirect neutron radiography device at CMRR (China Mianyang Research Reactor). The resolution of the device is better than 50 μm and the quantitative analysis of swelling behaviors was carried out. The results show that the swelling behaviors relate well to burnup character which can be detected accurately by indirect neutron radiography.

  5. Cladding corrosion and hydriding in irradiated defected zircaloy fuel rods (LWBR Development Program)

    Energy Technology Data Exchange (ETDEWEB)

    Clayton, J.C.

    1985-08-01

    Twenty-one LWBR irradiation test rods containing ThO/sub 2/-UO/sub 2/ fuel and Zircaloy cladding with holes or cracks operated successfully. Zircaloy cladding corrosion on the inside and outside diameter surfaces and hydrogen pickup in the cladding were measured. The observed outer surface Zircaloy cladding corrosion oxide thicknesses of the test rods were similar to thicknesses measured for nondefected irradiation test rods. An analysis model, which was developed to calculate outer surface oxide thickness of non-defected rods, gave results which were in reasonable agreement with the outer surface oxide thicknesses of defected rods. When the analysis procedure was modified to account for additional corrosion proportional to fission rate and to time, the calculated values agreed well with measured inner oxide corrosion film values. Hydrogen pickup in the defected rods was not directly proportional to local corrosion oxide weight gain as was the case for non-defected rods. 16 refs., 6 figs., 8 tabs.

  6. High temperature nanoindentation hardness and Young's modulus measurement in a neutron-irradiated fuel cladding material

    Science.gov (United States)

    Kese, K.; Olsson, P. A. T.; Alvarez Holston, A.-M.; Broitman, E.

    2017-04-01

    Nanoindentation, in combination with scanning probe microscopy, has been used to measure the hardness and Young's modulus in the hydride and matrix of a high burn-up neutron-irradiated Zircaloy-2 cladding material in the temperature range 25-300 °C. The matrix hardness was found to decrease only slightly with increasing temperature while the hydride hardness was essentially constant within the temperature range. Young's modulus decreased with increasing temperature for both the hydride and the matrix of the high burn-up fuel cladding material. The hydride Young's modulus and hardness were higher than those of the matrix in the temperature range.

  7. Performance of AGR-1 High-Temperature Reactor Fuel During Post-Irradiation Heating Tests

    Energy Technology Data Exchange (ETDEWEB)

    Morris, Robert Noel [ORNL; Baldwin, Charles A [ORNL; Hunn, John D [ORNL; Demkowicz, Paul [Idaho National Laboratory (INL); Reber, Edward [Idaho National Laboratory (INL)

    2014-01-01

    The fission product retention of irradiated low-enriched uranium oxide/uranium carbide TRISO fuel compacts from the 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% FIMA have been tested using dedicated furnace systems at Idaho National Laboratory and Oak Ridge National Laboratory. The release of fission products 110mAg, 134Cs, 137Cs, 154Eu, 155Eu, 90Sr, and 85Kr 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 10-6 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 85Kr release was very low during the tests (particles with breached SiC, but intact outer pyrocarbon, retained most of their krypton). Krypton release from TRISO failures was only observed after ~210 h at 1800 C in one compact. Post-safety-test examination of fuel compacts and particles has focused on identifying specific particles from each compact with notable fission product release and detailed analysis of the coating layers to understand particle behavior.

  8. Chemical forms of solid fission products in the irradiated uranium—plutonium mixed nitride fuel

    Science.gov (United States)

    Arai, Yasuo; Maeda, Atsushi; Shiozawa, Ken-ichi; Ohmichi, Toshihiko

    1994-06-01

    Chemical forms of solid fission products in the irradiated (U, Pu)N fuel were estimated by both thermodynamic equilibrium calculation and electron microprobe analysis on burnup simulated samples prepared by carbothermic reduction. Besides the MX type matrix phase dissolving zirconium, niobium, yttrium and rare earth elements, the existence of two kinds of inclusion was recognized. One is URu 3 type intermetallic compound constituted by uranium and platinum group elements. The other is an alloy containing molybdenum as a principal constituent. Furthermore, the swelling rate due to solid fission products precipitation was evaluated to be about 0.5% per %FIMA.

  9. Improving the AGR Fuel Testing Power Density Profile Versus Irradiation-Time in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gray S. Chang; David A. Petti; John T. Maki; Misti A. Lillo

    2009-05-01

    The Very High Temperature gas-cooled Reactor (VHTR), which is currently being developed, achieves simplification of safety through reliance on ceramic-coated fuel particles. Each TRISO-coated fuel particle has its own containment which serves as the principal barrier against radionuclide release under normal operating and accident conditions. These fuel particles, in the form of graphite fuel compacts, are currently undergoing a series of irradiation tests in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) to support the Advanced Gas-Cooled Reactor (AGR) fuel qualification program. A representive coated fuel particle with an 235U enrichment of 19.8 wt% was used in this analysis. The fuel burnup analysis tool used to perform the neutronics study reported herein, couples the Monte Carlo transport code MCNP, with the radioactive decay and burnup code ORIGEN2. The fuel burnup methodology known as Monte-Carlo with ORIGEN2 (MCWO) was used to evaluate the AGR experiment assembly and demonstrate compliance with ATR safety requirements. For the AGR graphite fuel compacts, the MCWO-calculated fission power density (FPD) due to neutron fission in 235U is an important design parameter. One of the more important AGR fuel testing requirements is to maintain the peak fuel compact temperature close to 1250°C throughout the proposed irradiation campaign of 550 effective full power days (EFPDs). Based on the MCWO-calculated FPD, a fixed gas gap size was designed to allow regulation of the fuel compact temperatures throughout the entire fuel irradiation campaign by filling the gap with a mixture of helium and neon gases. The chosen fixed gas gap can only regulate the peak fuel compact temperature in the desired range during the irradiation test if the ratio of the peak power density to the time-dependent low power density (P/T) at 550 EFPDs is less than 2.5. However, given the near constant neutron flux within the ATR driver core and the depletion of 235U in

  10. Messiah College Biodiesel Fuel Generation Project Final Technical Report

    Energy Technology Data Exchange (ETDEWEB)

    Zummo, Michael M; Munson, J; Derr, A; Zemple, T; Bray, S; Studer, B; Miller, J; Beckler, J; Hahn, A; Martinez, P; Herndon, B; Lee, T; Newswanger, T; Wassall, M

    2012-03-30

    Many obvious and significant concerns arise when considering the concept of small-scale biodiesel production. Does the fuel produced meet the stringent requirements set by the commercial biodiesel industry? Is the process safe? How are small-scale producers collecting and transporting waste vegetable oil? How is waste from the biodiesel production process handled by small-scale producers? These concerns and many others were the focus of the research preformed in the Messiah College Biodiesel Fuel Generation project over the last three years. This project was a unique research program in which undergraduate engineering students at Messiah College set out to research the feasibility of small-biodiesel production for application on a campus of approximately 3000 students. This Department of Energy (DOE) funded research program developed out of almost a decade of small-scale biodiesel research and development work performed by students at Messiah College. Over the course of the last three years the research team focused on four key areas related to small-scale biodiesel production: Quality Testing and Assurance, Process and Processor Research, Process and Processor Development, and Community Education. The objectives for the Messiah College Biodiesel Fuel Generation Project included the following: 1. Preparing a laboratory facility for the development and optimization of processors and processes, ASTM quality assurance, and performance testing of biodiesel fuels. 2. Developing scalable processor and process designs suitable for ASTM certifiable small-scale biodiesel production, with the goals of cost reduction and increased quality. 3. Conduct research into biodiesel process improvement and cost optimization using various biodiesel feedstocks and production ingredients.

  11. Coupled thermochemical, isotopic evolution and heat transfer simulations in highly irradiated UO2 nuclear fuel

    Science.gov (United States)

    Piro, M. H. A.; Banfield, J.; Clarno, K. T.; Simunovic, S.; Besmann, T. M.; Lewis, B. J.; Thompson, W. T.

    2013-10-01

    Predictive capabilities for simulating irradiated nuclear fuel behavior are enhanced in the current work by coupling thermochemistry, isotopic evolution and heat transfer. Thermodynamic models that are incorporated into this framework not only predict the departure from stoichiometry of UO2, but also consider dissolved fission and activation products in the fluorite oxide phase, noble metal inclusions, secondary oxides including uranates, zirconates, molybdates and the gas phase. Thermochemical computations utilize the spatial and temporal evolution of the fission and activation product inventory in the pellet, which is typically neglected in nuclear fuel performance simulations. Isotopic computations encompass the depletion, decay and transmutation of more than 2000 isotopes that are calculated at every point in space and time. These computations take into consideration neutron flux depression and the increased production of fissile plutonium near the fuel pellet periphery (i.e., the so-called “rim effect”). Thermochemical and isotopic predictions are in very good agreement with reported experimental measurements of highly irradiated UO2 fuel with an average burnup of 102 GW d t(U)-1. Simulation results demonstrate that predictions are considerably enhanced when coupling thermochemical and isotopic computations in comparison to empirical correlations. Notice: This manuscript has been authored by UT-Battelle, LLC, under Contract No. DE-AC05-00OR22725 with the U.S. Department of Energy. The United States Government retains and the publisher, by accepting the article for publication, acknowledges that the United States Government retains a non-exclusive, paid-up, irrevocable, world-wide license to publish or reproduce the published form of this manuscript, or allow others to do so, for United States Government purposes.

  12. UNCERTAINTY QUANTIFICATION OF CALCULATED TEMPERATURES FOR ADVANCED GAS REACTOR FUEL IRRADIATION EXPERIMENTS

    Energy Technology Data Exchange (ETDEWEB)

    Pham, Binh Thi-Cam [Idaho National Laboratory; Hawkes, Grant Lynn [Idaho National Laboratory; Einerson, Jeffrey James [Idaho National Laboratory

    2015-08-01

    This paper presents the quantification of uncertainty of the calculated temperature data for the Advanced Gas Reactor (AGR) fuel irradiation experiments conducted in the Advanced Test Reactor at Idaho National Laboratory in support of the Advanced Reactor Technology Research and Development program. Recognizing uncertainties inherent in physics and thermal simulations of the AGR tests, the results of the numerical simulations are used in combination with statistical analysis methods to improve qualification of measured data. The temperature simulation data for AGR tests are also used for validation of the fission product transport and fuel performance simulation models. These crucial roles of the calculated fuel temperatures in ensuring achievement of the AGR experimental program objectives require accurate determination of the model temperature uncertainties. To quantify the uncertainty of AGR calculated temperatures, this study identifies and analyzes ABAQUS model parameters of potential importance to the AGR predicted fuel temperatures. The selection of input parameters for uncertainty quantification of the AGR calculated temperatures is based on the ranking of their influences on variation of temperature predictions. Thus, selected input parameters include those with high sensitivity and those with large uncertainty. Propagation of model parameter uncertainty and sensitivity is then used to quantify the overall uncertainty of AGR calculated temperatures. Expert judgment is used as the basis to specify the uncertainty range for selected input parameters. The input uncertainties are dynamic accounting for the effect of unplanned events and changes in thermal properties of capsule components over extended exposure to high temperature and fast neutron irradiation. The sensitivity analysis performed in this work went beyond the traditional local sensitivity. Using experimental design, analysis of pairwise interactions of model parameters was performed to establish

  13. Simulation of the heat transfer of a irradiated fuel storage container with code CFD STAR- CCM+; Simulacion de la transferencia de calor de un contenedor de almacenamiento de combustible irradiado con el codigo CFD STAR-CCM+

    Energy Technology Data Exchange (ETDEWEB)

    Barrera matalla, J. E.; Hernandez Gomez, J.; Riverala Gurruchaga, J.

    2012-07-01

    Irradiated fuel has become an object of interest in the industry by the importance of ensuring its safety during long periods of storage time. New containers, stores, methods and codes will be used to ensure a suitable cooling and residual heat removal, and secure the safety of fuel elements in dry storage. The codes CFD (Computational Fluid Dynamics) have great potential to help in design of containers and stores, improving thermal-hydraulic performance and the extraction of heat generated.

  14. Design of a mediated enzymatic fuel cell to generate power from renewable fuel sources.

    Science.gov (United States)

    Korkut, Seyda; Kilic, Muhammet Samet

    2016-01-01

    The present work reported a compartment-less enzymatic fuel cell (EFC) based on newly synthesized Poly(pyrrole-2-carboxylic acid-co-3-thiophene acetic acid) film containing glucose oxidase and laccase effectively wired by p-benzoquinone incorporated into the copolymer structure. The resulting system generated a power density of 18.8 µW/cm(2) with 30 mM of glucose addition at +0.94 V at room temperature. Improvements to maximize the power output were ensured with step-by-step optimization of electrode fabrication design and operational parameters for operating the system with renewable fuel sources. We demonstrated that the improved fuel cell could easily harvest glucose produced during photosynthesis to produce electrical energy in a simple, renewable and sustainable way by generating a power density of 10 nW/cm(2) in the plant leaf within 2 min. An EFC for the first time was successfully operated in municipal wastewater which contained glycolytic substances to generate electrical energy with a power output of 3.3 µW/cm(2).

  15. GEH-4-63, 64: Proposal for irradiation of production brazed Zircaloy-2 clad fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Tverberg, J.C.

    1961-05-18

    A brazed end closure is currently being used on prototypical NPR fuel elements. The production closure will use a braze alloy composed of 5% Be + 95% Zry-2 to braze the Zircaloy-2 cap to the jacket and to the metallic uranium core. A similar MTR test, a GEH-4-57, 58, used a braze alloy of the composition 4% Be + 12% Fe + 84% Zry-2 which melts at a lower temperature. In this previous test, element GEH-4-57 failed through a cladding defect located at the base of the braze heat affected zone. Because of this failure it would be desirable to subject a fuel element, which had been subjected to more severe brazing conditions, to the same conditions as GEH-4-57, 58. For this reason the thermal conditions of this test essentially match those of GEH-4-57, 58. This irradiation test consists of two identical fuel elements. The fuel material is normal metallic uranium, Zircaloy-2 clad of the tubular geometry, NPR inner size. The fuel was coextruded at Hanford by General Electric`s Fuels Preparation Department. Each element is 10.8 inches in length with flat Zircaloy-2 end caps brazed to the jacket and uranium core with the 5 Be + 95 Zry-2 brazing alloy, then TIG welded to further insure closure integrity. The elements ar 1.254 inches OD and 0.439 inches ID. For hydraulic purposes a 0.343 inch diamater flow restrictor has been fitted into the central flow channel of both elements.

  16. Applicability of a set of tomographic reconstruction algorithms for quantitative SPECT on irradiated nuclear fuel assemblies

    Science.gov (United States)

    Jacobsson Svärd, Staffan; Holcombe, Scott; Grape, Sophie

    2015-05-01

    A fuel assembly operated in a nuclear power plant typically contains 100-300 fuel rods, depending on fuel type, which become strongly radioactive during irradiation in the reactor core. For operational and security reasons, it is of interest to experimentally deduce rod-wise information from the fuel, preferably by means of non-destructive measurements. The tomographic SPECT technique offers such possibilities through its two-step application; (1) recording the gamma-ray flux distribution around the fuel assembly, and (2) reconstructing the assembly's internal source distribution, based on the recorded radiation field. In this paper, algorithms for performing the latter step and extracting quantitative relative rod-by-rod data are accounted for. As compared to application of SPECT in nuclear medicine, nuclear fuel assemblies present a much more heterogeneous distribution of internal attenuation to gamma radiation than the human body, typically with rods containing pellets of heavy uranium dioxide surrounded by cladding of a zirconium alloy placed in water or air. This inhomogeneity severely complicates the tomographic quantification of the rod-wise relative source content, and the deduction of conclusive data requires detailed modelling of the attenuation to be introduced in the reconstructions. However, as shown in this paper, simplified models may still produce valuable information about the fuel. Here, a set of reconstruction algorithms for SPECT on nuclear fuel assemblies are described and discussed in terms of their quantitative performance for two applications; verification of fuel assemblies' completeness in nuclear safeguards, and rod-wise fuel characterization. It is argued that a request not to base the former assessment on any a priori information brings constraints to which reconstruction methods that may be used in that case, whereas the use of a priori information on geometry and material content enables highly accurate quantitative assessment, which

  17. 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.

  18. 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.

  19. Influence of ultrasonic irradiation on ozone generation in a dielectric barrier discharge

    DEFF Research Database (Denmark)

    Kusano, Yukihiro; Drews, J.; Leipold, Frank

    2012-01-01

    An atmospheric pressure dielectric barrier discharge (DBD) was generated in an N2/O2 gas mixture at room temperature with and without ultrasonic irradiation to investigate ozone production. Powerful ultrasonic irradiation with the sound pressure level of approximately 150 dB into the DBD can...... enhance ozone production especially when the DBD was driven at a frequency of 15 kHz....

  20. Behavior of irradiated BWR fuel under reactivity-initiated-accident conditions; Results of tests FK-1, -2 and -3

    OpenAIRE

    2004-01-01

    Boiling water reactor (BWR) fuels with burnups of 41 to 45 GWd/tU were pulse-irradiated in the Nuclear Safety Research Reactor (NSRR) to investigate fuel behavior under cold startup reactivity-initiated-accident (RIA) conditions. BWR fuel segment rods of 8times8BJ (STEP I) type from Fukushima-Daiichi Unit 3 nuclear power plant were refabricated into short test rods, and they were subjected to prompt enthalpy insertion from 293 to 607 J/g (70 to 145 cal/g) within about 20 ms. The fuel cladding...

  1. Examination of spent fuel radiation energy conversion for electricity generation

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Haneol; Yim, Man-Sung, E-mail: msyim@kaist.ac.kr

    2016-04-15

    Highlights: • Utilizing conversion of radiation energy of spent fuel to electric energy. • MCNPX modeling and experiment were used to estimate energy conversion. • The converted energy may be useful for nuclear security applications. • The converted energy may be utilized for safety applications through energy storage. - Abstract: Supply of electricity inside nuclear power plant is one of the most important considerations for nuclear safety and security. In this study, generation of electric energy by converting radiation energy of spent nuclear fuel was investigated. Computational modeling work by using MCNPX 2.7.0 code along with experiment was performed to estimate the amount of electric energy generation. The calculation using the developed modeling work was validated through comparison with an integrated experiment. The amount of electric energy generation based on a conceptual design of an energy conversion module was estimated to be low. But the amount may be useful for nuclear security applications. An alternative way of utilizing the produced electric energy could be considered for nuclear safety application through energy storage. Further studies are needed to improve the efficiency of the proposed energy conversion concept and to examine the issue of radiation damage and economic feasibility.

  2. Fuel Cells in the Waste-to-Energy Chain Distributed Generation Through Non-Conventional Fuels and Fuel Cells

    CERN Document Server

    McPhail, Stephen J; Moreno, Angelo

    2012-01-01

    As the availability of fossils fuels becomes more limited, the negative impact of their consumption becomes an increasingly relevant factor in our choices with regards to primary energy sources. The exponentially increasing demand for energy is reflected in the mass generation of by-products and waste flows which characterize current society’s development and use of fossil sources. The potential for recoverable material and energy in these ever-increasing refuse flows is huge, even after the separation of hazardous constituent elements, allowing safe and sustainable further exploitation of an otherwise 'wasted' resource.  Fuel Cells in the Waste-to-Energy Chain explores the concept of waste-to-energy through a 5 step process which reflects the stages during the transformation of  refuse flows to a valuable commodity such as clean energy. By providing selected, integrated alternatives to the current centralized, wasteful, fossil-fuel based infrastructure, Fuel Cells in the Waste-to-Energy Chain explores ho...

  3. Electricity generation from rapeseed straw hydrolysates using microbial fuel cells.

    Science.gov (United States)

    Jablonska, Milena A; Rybarczyk, Maria K; Lieder, Marek

    2016-05-01

    Rapeseed straw is an attractive fuel material for microbial fuel cells (MFCs) due to its high content of carbohydrates (more than 60% carbohydrates). This study has demonstrated that reducing sugars can be efficiently extracted from raw rapeseed straw by combination of hydrothermal pretreatment and enzymatic hydrolysis followed by utilization as a fuel in two-chamber MFCs for electrical power generation. The most efficient method of saccharification of this lignocellulosic biomass (17%) turned out hydrothermal pretreatment followed by enzymatic hydrolysis. Electricity was produced using hydrolysate concentrations up to 150 mg/dm(3). The power density reached 54 mW/m(2), while CEs ranged from 60% to 10%, corresponding to the initial reducing sugar concentrations of 10-150 mg/dm(3). The COD degradation rates based on charge calculation increased from 0.445 g COD/m(2)/d for the hydrolysate obtained with the microwave treatment to 0.602 g COD/m(2)/d for the most efficient combination of hydrothermal treatment followed by enzymatic hydrolysis.

  4. Electricity generation from the mud by using microbial fuel cell

    Directory of Open Access Journals (Sweden)

    Idris Sitinoor Adeib

    2016-01-01

    Full Text Available Microbial fuel cells (MFCs is a bio-electrochemical device that harnesses the power of respiring microbes to convert organic substrates directly into electrical energy. This is achieved when bacteria transfer electrons to an electrode rather than directly to an electron acceptor. Their technical feasibility has recently been proven and there is great enthusiasm in the scientific community that MFCs could provide a source of “green electricity”. Microbial fuel cells work by allowing bacteria to do what they do best, oxidize and reduce organic molecules. Bacterial respiration is basically one big redox reaction in which electrons are being moved around. The objective is to generate electricity throughout the biochemical process using chemical waste basically sludge, via microbial fuel cells. The methodology includes collecting sludge from different locations, set up microbial fuel cells with the aid of salt bridge and observing the results in voltage measurement. The microbial fuel cells consist of two chambers, iron electrodes, copper wire, air pump (to increase the efficiency of electron transfer, water, sludge and salt bridge. After several observations, it is seen that this MFC can achieve up until 202 milivolts (0.202volts with the presence of air pump. It is proven through the experiments that sludge from different locations gives different results in term of the voltage measurement. This is basically because in different locations of sludge contain different type and amount of nutrients to provide the growth of bacteria. Apart from that, salt bridge also play an important role in order to transport the proton from cathode to anode. A longer salt bridge will give a higher voltage compared to a short salt bridge. On the other hand, the limitations that this experiment facing is the voltage that being produced did not last long as the bacteria activity slows down gradually and the voltage produced are not really great in amount. Lastly to

  5. Modeling and control of fuel cell based distributed generation systems

    Science.gov (United States)

    Jung, Jin Woo

    This dissertation presents circuit models and control algorithms of fuel cell based distributed generation systems (DGS) for two DGS topologies. In the first topology, each DGS unit utilizes a battery in parallel to the fuel cell in a standalone AC power plant and a grid-interconnection. In the second topology, a Z-source converter, which employs both the L and C passive components and shoot-through zero vectors instead of the conventional DC/DC boost power converter in order to step up the DC-link voltage, is adopted for a standalone AC power supply. In Topology 1, two applications are studied: a standalone power generation (Single DGS Unit and Two DGS Units) and a grid-interconnection. First, dynamic model of the fuel cell is given based on electrochemical process. Second, two full-bridge DC to DC converters are adopted and their controllers are designed: an unidirectional full-bridge DC to DC boost converter for the fuel cell and a bidirectional full-bridge DC to DC buck/boost converter for the battery. Third, for a three-phase DC to AC inverter without or with a Delta/Y transformer, a discrete-time state space circuit model is given and two discrete-time feedback controllers are designed: voltage controller in the outer loop and current controller in the inner loop. And last, for load sharing of two DGS units and power flow control of two DGS units or the DGS connected to the grid, real and reactive power controllers are proposed. Particularly, for the grid-connected DGS application, a synchronization issue between an islanding mode and a paralleling mode to the grid is investigated, and two case studies are performed. To demonstrate the proposed circuit models and control strategies, simulation test-beds using Matlab/Simulink are constructed for each configuration of the fuel cell based DGS with a three-phase AC 120 V (L-N)/60 Hz/50 kVA and various simulation results are presented. In Topology 2, this dissertation presents system modeling, modified space

  6. Irradiated Xenon Isotopic Ratio Measurement for Failed Fuel Detection and Location in Fast Reactor

    Science.gov (United States)

    Ito, Chikara; Iguchi, Tetsuo; Harano, Hideki

    2009-08-01

    The accuracy of xenon isotopic ratio burn-up calculations used for failed fuel identification was evaluated by an irradiation test of xenon tag gas samples in the Joyo test reactor. The experiment was carried out using pressurized steel capsules containing unique blend ratios of stable xenon tag gases in an on-line creep rupture experiment in Joyo. The tag gas samples were irradiated to total neutron fluences of 1.6 to 4.8 × 1026 n/m2. Laser resonance ionization mass spectrometry was used to analyze the cover gas containing released tag gas diluted to isotopic ratios of 100 to 102 ppb. The isotopic ratios of xenon tag gases after irradiation were calculated using the ORIGEN2 code. The neutron cross sections of xenon nuclides were based on the JENDL-3.3 library. These cross sections were collapsed into one group using the neutron spectra of Joyo. The comparison of measured and calculated xenon isotopic ratios provided C/E values that ranged from 0.92 to 1.10. The differences between calculation and measurement were considered to be mainly due to the measurement errors and the xenon nuclide cross section uncertainties.

  7. Microbial fuel cells: novel biotechnology for energy generation.

    Science.gov (United States)

    Rabaey, Korneel; Verstraete, Willy

    2005-06-01

    Microbial fuel cells (MFCs) provide new opportunities for the sustainable production of energy from biodegradable, reduced compounds. MFCs function on different carbohydrates but also on complex substrates present in wastewaters. As yet there is limited information available about the energy metabolism and nature of the bacteria using the anode as electron acceptor; few electron transfer mechanisms have been established unequivocally. To optimize and develop energy production by MFCs fully this knowledge is essential. Depending on the operational parameters of the MFC, different metabolic pathways are used by the bacteria. This determines the selection and performance of specific organisms. Here we discuss how bacteria use an anode as an electron acceptor and to what extent they generate electrical output. The MFC technology is evaluated relative to current alternatives for energy generation.

  8. Fission product release and microstructure changes of irradiated MOX fuel at high temperatures

    Science.gov (United States)

    Colle, J.-Y.; Hiernaut, J.-P.; Wiss, T.; Beneš, O.; Thiele, H.; Papaioannou, D.; Rondinella, V. V.; Sasahara, A.; Sonoda, T.; Konings, R. J. M.

    2013-11-01

    Samples of irradiated MOX fuel of 44.5 GWd/tHM mean burn-up were prepared by core drilling at three different radial positions of a fuel pellet. They were subsequently heated in a Knudsen effusion mass spectrometer up to complete vaporisation of the sample (˜2600 K) and the release of fission gas (krypton and xenon) as well as helium was measured. Scanning electron microscopy was used in parallel to investigate the evolution of the microstructure of a sample heated under the same condition up to given key temperatures as determined from the gas release profiles. A clear initial difference for fission gas release and microstructure was observed as a function of the radial position of the samples and therefore of irradiation temperature. A good correlation between the microstructure evolution and the gas release peaks could be established as a function of the temperature of irradiation and (laboratory) heating. The region closest to the cladding (0.58 < r/r0 < 0.96), designated as sample type A in Fig. 1. It represents the "cooler" part of the fuel pellet. The irradiation temperatures (Tirrad) in this range are from 854 to 1312 K (ΔT: 458 K). The intermediate radial zone of the pellet (0.42 < r/r0 < 0.81), designated sample type B in Fig. 1, has a Tirrad ranging from 1068 to 1434 K (ΔT: 365 K). The central zone of the pellet (0.003 < r/r0 < 0.41), designated sample type C in Fig. 1, which was close to the hottest part of the pellet, has a Tirrad ranging from 1442 to 1572 K (ΔT: 131 K). The sample irradiation temperatures were determined from the calculated temperature profile (exponential function) knowing the core temperature of the fuel (1573 K) [11], the standard temperature for this type of fuel at the inner side of the cladding (800 K). The average burnup was calculated with TRANSURANUS code [12] and the PA burnup is the average burnup multiplied by the ratio of the fissile Pu concentration in PA over average fissile Pu concentration in fuel [11]. Calculated

  9. Nanoparticle production by UV irradiation of combustion generated soot particles

    Energy Technology Data Exchange (ETDEWEB)

    Stipe, Christopher B.; Choi, Jong Hyun; Lucas, Donald; Koshland, Catherine P.; Sawyer, Robert F.

    2004-07-01

    Laser ablation of surfaces normally produce high temperature plasmas that are difficult to control. By irradiating small particles in the gas phase, we can better control the size and concentration of the resulting particles when different materials are photofragmented. Here, we irradiate soot with 193 nm light from an ArF excimer laser. Irradiating the original agglomerated particles at fluences ranging from 0.07 to 0.26 J/cm{sup 2} with repetition rates of 20 and 100 Hz produces a large number of small, unagglomerated particles, and a smaller number of spherical agglomerated particles. Mean particle diameters from 20 to 50 nm are produced from soot originally having a mean electric mobility diameter of 265nm. We use a non-dimensional parameter, called the photon/atom ratio (PAR), to aid in understanding the photofragmentation process. This parameter is the ratio of the number of photons striking the soot particles to the number of the carbon atoms contained in the soot particles, and is a better metric than the laser fluence for analyzing laser-particle interactions. These results suggest that UV photofragmentation can be effective in controlling particle size and morphology, and can be a useful diagnostic for studying elements of the laser ablation process.

  10. 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)

  11. Iterative ct reconstruction from few projections for the nondestructive post irradiation examination of nuclear fuel assemblies

    Science.gov (United States)

    Abir, Muhammad Imran Khan

    The core components (e.g. fuel assemblies, spacer grids, control rods) of the nuclear reactors encounter harsh environment due to high temperature, physical stress, and a tremendous level of radiation. The integrity of these elements is crucial for safe operation of the nuclear power plants. The Post Irradiation Examination (PIE) can reveal information about the integrity of the elements during normal operations and off?normal events. Computed tomography (CT) is a tool for evaluating the structural integrity of elements non-destructively. CT requires many projections to be acquired from different view angles after which a mathematical algorithm is adopted for reconstruction. Obtaining many projections is laborious and expensive in nuclear industries. Reconstructions from a small number of projections are explored to achieve faster and cost-efficient PIE. Classical reconstruction algorithms (e.g. filtered back projection) cannot offer stable reconstructions from few projections and create severe streaking artifacts. In this thesis, conventional algorithms are reviewed, and new algorithms are developed for reconstructions of the nuclear fuel assemblies using few projections. CT reconstruction from few projections falls into two categories: the sparse-view CT and the limited-angle CT or tomosynthesis. Iterative reconstruction algorithms are developed for both cases in the field of compressed sensing (CS). The performance of the algorithms is assessed using simulated projections and validated through real projections. The thesis also describes the systematic strategy towards establishing the conditions of reconstructions and finds the optimal imaging parameters for reconstructions of the fuel assemblies from few projections.

  12. Evaluation of fretting failures on PWR fuel by post-irradiation examinations and modeling in the DEGRAD-1 code

    Energy Technology Data Exchange (ETDEWEB)

    Castanheira, Myrthes; Silva, Jose Eduardo Rosa da; Lucki, Georgi; Terremoto, Luis A.A.; Silva, Antonio Teixeira e; Teodoro, Celso A.; Damy, Margaret de A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)]. E-mail: myrthes@ipen.br

    2007-07-01

    One of the major recognized causes of fuel rod failures is fretting of the clad due to the entrapment of debris in a fuel rod spacer. Such debris, inadvertently dropped into the primary system during maintenance operations, includes various sizes of particles. Intermediate size particles, such as metal cuttings, electrical connectors, metal fittings, pieces of wire, and small nuts and bolts can become trapped between fuel rods in a spacer where hydraulically induced vibrations can cause fretting failure of the fuel rod. An evaluation of debris fretting failure on PWR fuel is presented. The inquiries on fuel rods failures are based on results of analysis using post-irradiation non-destructive examination. The complementary analysis includes a modeling approach by code DEGRAD-1 to characterize the degradation phenomenon after primary failure integrated in the reactor operational history. (author)

  13. Development of Kinetic Mechanisms for Next-Generation Fuels and CFD Simulation of Advanced Combustion Engines

    Energy Technology Data Exchange (ETDEWEB)

    Pitz, William J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); McNenly, Matt J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Whitesides, Russell [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Mehl, Marco [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Killingsworth, Nick J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Westbrook, Charles K. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-12-17

    Predictive chemical kinetic models are needed to represent next-generation fuel components and their mixtures with conventional gasoline and diesel fuels. These kinetic models will allow the prediction of the effect of alternative fuel blends in CFD simulations of advanced spark-ignition and compression-ignition engines. Enabled by kinetic models, CFD simulations can be used to optimize fuel formulations for advanced combustion engines so that maximum engine efficiency, fossil fuel displacement goals, and low pollutant emission goals can be achieved.

  14. Fuel Savings and Emission Reductions from Next-Generation Mobile Air Conditioning Technology in India: Preprint

    Energy Technology Data Exchange (ETDEWEB)

    Chaney, L.; Thundiyil, K.; Chidambaram, S.; Abbi, Y. P.; Anderson, S.

    2007-05-01

    This paper quantifies the mobile air-conditioning fuel consumption of the typical Indian vehicle, exploring potential fuel savings and emissions reductions these systems for the next generation of vehicles.

  15. Influence of supporting electrolyte in electricity generation and degradation of organic pollutants in photocatalytic fuel cell.

    Science.gov (United States)

    Khalik, Wan Fadhilah; Ong, Soon-An; Ho, Li-Ngee; Wong, Yee-Shian; Voon, Chun-Hong; Yusuf, Sara Yasina; Yusoff, Nik Athirah; Lee, Sin-Li

    2016-08-01

    This study investigated the effect of different supporting electrolyte (Na2SO4, MgSO4, NaCl) in degradation of Reactive Black 5 (RB5) and generation of electricity. Zinc oxide (ZnO) was immobilized onto carbon felt acted as photoanode, while Pt-coated carbon paper as photocathode was placed in a single chamber photocatalytic fuel cell, which then irradiated by UV lamp for 24 h. The degradation and mineralization of RB5 with 0.1 M NaCl rapidly decreased after 24-h irradiation time, followed by MgSO4, Na2SO4 and without electrolyte. The voltage outputs for Na2SO4, MgSO4 and NaCl were 908, 628 and 523 mV, respectively, after 24-h irradiation time; meanwhile, their short-circuit current density, J SC, was 1.3, 1.2 and 1.05 mA cm(-2), respectively. The power densities for Na2SO4, MgSO4 and NaCl were 0.335, 0.256 and 0.245 mW cm(-2), respectively. On the other hand, for without supporting electrolyte, the voltage output and short-circuit current density was 271.6 mV and 0.055 mA cm(-2), respectively. The supporting electrolyte NaCl showed greater performance in degradation of RB5 and generation of electricity due to the formation of superoxide radical anions which enhance the degradation of dye. The mineralization of RB5 with different supporting electrolyte was measured through spectrum analysis and reduction in COD concentration.

  16. Generation of nanometer structures on surfaces of ionic solids generated by laser and electron beam irradiation

    Science.gov (United States)

    Dawes, M. L.; Langford, S. C.; Dickinson, J. Thomas

    2001-03-01

    Radiation effects on hydrated single crystals are poorly understood. We find that dense arrays of nanoscale conical structures, with aspect ratios on the order of 200, are produced when single crystal brushite (CaHPO_4^.2H_2O) is exposed to energetic electrons (2 keV). Other three dimensional nanostructures are generated by exposing brushite to excimer laser irradiation. We show that the mechanism involves: (a) photo/electron stimulated decomposition of the matrix, and (b) thermally stimulated migration of water (in this case, crystalline) and ionic material. We have isolated these factors to some extent and present plausible mechanisms for structure formation. In addition, we have recently exposed non-hydrated ionic crystals to radiation in the presence of background water (pp_water ~ 10-7 Torr), which produces exceedingly fine structures (sub-10 nm). The optical and luminescence properties of these features will be presented. An example of a “stealth surface” will be given with possible applications for the laser generation of x-rays.

  17. SCANNING ELECTRON MICROSCOPY ANALYSIS OF FUEL/MATRIX INTERACTION LAYERS IN HIGHLY-IRRADIATED U-Mo DISPERSION FUEL PLATES WITH Al AND Al–Si ALLOY MATRICES

    Directory of Open Access Journals (Sweden)

    DENNIS D. KEISER, JR.

    2014-04-01

    Full Text Available In order to investigate how the microstructure of fuel/matrix-interaction (FMI layers change during irradiation, different U–7Mo dispersion fuel plates have been irradiated to high fission density and then characterized using scanning electron microscopy (SEM. Specifially, samples from irradiated U–7Mo dispersion fuel elements with pure Al, Al-2Si and AA4043 (∼4.5 wt.%Si matrices were SEM characterized using polished samples and samples that were prepared with a focused ion beam (FIB. Features not observable for the polished samples could be captured in SEM images taken of the FIB samples. For the Al matrix sample, a relatively large FMI layer develops, with enrichment of Xe at the FMI layer/Al matrix interface and evidence of debonding. Overall, a significant penetration of Si from the FMI layer into the U–7Mo fuel was observed for samples with Si in the Al matrix, which resulted in a change of the size (larger and shape (round of the fission gas bubbles. Additionally, solid fission product phases were observed to nucleate and grow within these bubbles. These changes in the localized regions of the microstructure of the U–7Mo may contribute to changes observed in the macroscopic swelling of fuel plates with Al-Si matrices.

  18. HELIOS: the new design of the irradiation of U-free fuels for americium transmutation

    Energy Technology Data Exchange (ETDEWEB)

    D' Agata, E. [European Commission, Joint Research Centre, Institute for Energy, P.O. Box 2, 1755 ZG Petten (Netherlands); Klaassen, F.; Sciolla, C. [Nuclear Research and Consultancy Group, Dept. Life Cycle and Innovations, P.O. Box 25 1755 ZG Petten (Netherlands); Fernandez-Carretero, A. [European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125 Karlsruhe (Germany); Bonnerot, J.M. [Commissariat a l' Energie Atomique, DEC/SESC/LC2I CEA-Cadarache, 13108 St. Paul lez Durance Cedex (France)

    2009-06-15

    Americium is one of the radioactive elements that mostly contribute to the radiotoxicity of the nuclear spent fuel. Transmutation of long-lived nuclides like Americium is an option for the reduction of the mass, the radiotoxicity and the decay heat of nuclear waste. The HELIOS irradiation experiment is the last evolution in a series of experiments on americium transmutation. The previous experiments, EFTTRA-T4 and T4bis, have shown that the release or trapping of helium is the key issue for the design of such kind of target. In fact, the production of helium, which is characteristic of {sup 241}Am transmutation, is quite significant. The experiment is carried out in the framework of the 4-year project EUROTRANS of the EURATOM 6. Framework Programme (FP6). Therefore, the main objective of the HELIOS experiment is to study the in-pile behaviour of U-free fuels such as CerCer (Pu, Am, Zr)O{sub 2} and Am{sub 2}Zr{sub 2}O{sub 7}+MgO or CerMet (Pu, Am)O{sub 2}+Mo in order to gain knowledge on the role of the fuel microstructure and of the temperature on the gas release and on the fuel swelling. The experiment was planned to be conducted in the HFR (High Flux Reactor) in Petten (The Netherlands) starting the first quarter of 2007. Because of the innovative aspects of the fuel, the fabrication has had some delays as well as the final safety analyses of the original design showed some unexpected deviation. Besides, the HFR reactor has been unavailable since August 2008. Due to the reasons described above, the experiment has been postponed. HELIOS should start in the first quarter of 2009 and will last 300 full power days. The paper will cover the description of the new design of the irradiation experiment HELIOS. The experiment has been split in two parts (HELIOS1 and HELIOS2) which will be irradiated together. Moreover, due to the high temperature achieved in cladding and to the high amount of helium produced during transmutation the experiment previously designed for a

  19. 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

  20. Preliminary ecotoxicity assessment of new generation alternative fuels in seawater.

    Science.gov (United States)

    Rosen, Gunther; Dolecal, Renee E; Colvin, Marienne A; George, Robert D

    2014-06-01

    The United States Navy (USN) is currently demonstrating the viability of environmentally sustainable alternative fuels to power its fleet comprised of aircraft and ships. As with any fuel used in a maritime setting, there is potential for introduction into the environment through transport, storage, and spills. However, while alternative fuels are often presumed to be eco-friendly relative to conventional petroleum-based fuels, their environmental fate and effects on marine environments are essentially unknown. Here, standard laboratory-based toxicity experiments were conducted for two alternative fuels, jet fuel derived from Camelina sativa (wild flax) seeds (HRJ5) and diesel fuel derived from algae (HRD76), and two conventional counterparts, jet fuel (JP5) and ship diesel (F76). Initial toxicity tests performed on water-accommodated fractions (WAF) from neat fuels partitioned into seawater, using four standard marine species in acute and chronic/sublethal tests, indicate that the alternative fuels are significantly less toxic to marine organisms.

  1. The Next Generation Nuclear Plant Graphite Creep Experiment Irradiation in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Blaine Grover

    2010-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating six gas reactor graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These graphite irradiations are being accomplished to support development of the next generation reactors in the United States. The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant (NGNP) Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain six stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six stacks will have differing compressive loads applied to the top half of each pair of specimen stacks, while a seventh stack will not have a compressive load. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be the capability of sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during initial start-up of

  2. A new approach to determine {sup 147}Pm in irradiated fuel solutions

    Energy Technology Data Exchange (ETDEWEB)

    Brennetot, R.; Stadelmann, G.; Caussignac, C.; Gombert, C.; Fouque, M.; Lamouroux, Ch. [CEA, Dept Chim Phys, Serv Etud Comportement Radionucleides, Lab Anal Nucl Isotop et Elementaires, Ctr Etud Sacl, F-91191 Gif Sur Yvette, (France)

    2009-07-01

    Developments carried out in the Laboratory of Isotopic Nuclear and Elementary Analyses in order to quantify {sup 147}Pm in spent nuclear fuels analyzed at the CEA within the framework of the Burn Up Credit research program for neutronic code validation are presented here. This determination is essential for safety-criticality Studies. The quantity and the nature of the radionuclides in irradiated fuel solutions force LIS to separate the elements of interest before measuring their isotopic content by mass spectrometry. The main objective of this study is to modify the separation protocol used in our laboratory in order to recover and to measure the {sup 147}Pm at the same time as the other lanthanides and actinides determined by mass spectrometry. A very complete study oil synthetic solution (containing or not {sup 147}Pm) Was undertaken in order to determine the yield of the various stages of separation carried out before obtaining the isolated Pm fraction from the whole of the elements present in the spent fuel Solutions. With the lack of natural tracer to carry out the measurement with the isotope dilution technique, the great number of isotopes in fuel, the originality of this work tests oil the use of another present lanthanide in fuel to define the output of separation. The yields were measured at the conclusion of each stage of separation with two others lanthanides in order to show that one of them could be used as a tracer to correct the measurement of the {sup 147}Pm with the separation yield. The total yield (at the conclusion of the two stages of separation) was measured at the same time by ICP-MS and liquid scintillation. This last determination made it possible to validate the use of the Sm-147 (natural) to measure the {sup 147}Pm in ICP-MS since the outputs determined in liquid scintillation and ICP-MS (starting from the radioactive decrease of the source having been used to make the synthetic solution) were equivalent. It is the first time that such

  3. Energy system analysis of fuel cells and distributed generation

    DEFF Research Database (Denmark)

    Mathiesen, Brian Vad; Lund, Henrik

    2007-01-01

    on the energy system in which they are used. Consequently, coherent energy systems analyses of specific and complete energy systems must be conducted in order to evaluate the benefits of FC technologies and in order to be able to compare alternative solutions. In relation to distributed generation, FC...... can be used for such analyses. Moreover, the chapter presents the results of evaluating the overall system fuel savings achieved by introducing different FC applications into different energy systems. Natural gas-based and hydrogen-based micro FC-CHP, natural gas local FC-CHP plants for district...... technologies have different strengths and weaknesses in different energy systems, but often they do not have the expected effect. Specific analyses of each individual country must be conducted including scenarios of expansion of e.g. wind power in order to evaluate where and when the best use of FC...

  4. High Performance Fuel Desing for Next Generation Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mujid S. Kazimi; Pavel Hejzlar

    2006-01-31

    The use of internally and externally cooled annular fule rods for high power density Pressurized Water Reactors is assessed. The assessment included steady state and transient thermal conditions, neutronic and fuel management requirements, mechanical vibration issues, fuel performance issues, fuel fabrication methods and econmic assessment. The investigation was donducted by a team from MIT, Westinghouse, Gamma Engineering, Framatome ANP, and AECL. The analyses led to the conclusion that raising the power density by 50% may be possible with this advanced fuel. Even at the 150% power level, the fuel temperature would be a few hundred degrees lower than the current fuel temperatre. Significant economic and safety advantages can be obtained by using this fuel in new reactors. Switching to this type of fuel for existing reactors would yield safety advantages, but the economic return is dependent on the duration of plant shutdown to accommodate higher power production. The main feasiblity issue for the high power performance appears to be the potential for uneven splitting of heat flux between the inner and outer fuel surfaces due to premature closure of the outer fuel-cladding gap. This could be overcome by using a very narrow gap for the inner fuel surface and/or the spraying of a crushable zirconium oxide film at the fuel pellet outer surface. An alternative fuel manufacturing approach using vobropacking was also investigated but appears to yield lower than desirable fuel density.

  5. The Hydraulic Test Procedure for Non-instrumented Irradiation Test Rig of Annular Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dae Ho; Lee, Kang Hee; Shin, Chang Hwan; Park, Chan Kook

    2008-08-15

    This report presents the procedure of pressure drop test, vibration test and endurance test for the non-instrumented rig using the irradiation test in HANARO of advanced PWR annular fuel which were designed and fabricated by KAERI. From the out-pile thermal hydraulic tests, confirm the flow rate at the 200 kPa pressure drop and measure the RMS displacement at this time. And the endurance test is confirmed the wear and the integrity of the non-instrumented rig at the 110% design flow rate. This out-pile test perform the Flow-Induced Vibration and Pressure Drop Experimental Tester(FIVPET) facility. The instruments in FIVPET facility was calibrated in KAERI and the pump and the thermocouple were certified by manufacturer.

  6. Copper anode corrosion affects power generation in microbial fuel cells

    KAUST Repository

    Zhu, Xiuping

    2013-07-16

    Non-corrosive, carbon-based materials are usually used as anodes in microbial fuel cells (MFCs). In some cases, however, metals have been used that can corrode (e.g. copper) or that are corrosion resistant (e.g. stainless steel, SS). Corrosion could increase current through galvanic (abiotic) current production or by increasing exposed surface area, or decrease current due to generation of toxic products from corrosion. In order to directly examine the effects of using corrodible metal anodes, MFCs with Cu were compared with reactors using SS and carbon cloth anodes. MFCs with Cu anodes initially showed high current generation similar to abiotic controls, but subsequently they produced little power (2 mW m-2). Higher power was produced with microbes using SS (12 mW m-2) or carbon cloth (880 mW m-2) anodes, with no power generated by abiotic controls. These results demonstrate that copper is an unsuitable anode material, due to corrosion and likely copper toxicity to microorganisms. © 2013 Society of Chemical Industry.

  7. A concise design o the irradiation of U-10Zr metallic fuel at a very low burnup

    Energy Technology Data Exchange (ETDEWEB)

    Guo, Hai Bing; Zhou, Wei; Sun, Yong; Qian, Dazhi; Ma, Jimin; Leng, Jun; Huo, Hyoung; Wang, Shaohua [Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang (China)

    2017-06-15

    In order to investigate the swelling behavior and fuel–cladding interaction mechanism of U–10Zr alloy metallic fuel at very low burnup, an irradiation experiment was concisely designed and conducted on the China Mianyang Research Reactor. Two types of irradiation samples were designed for studying free swelling without restraint and the fuel–cladding interaction mechanism. A new bonding material, namely, pure aluminum powder, was used to fill the gap between the fuel slug and sample shell for reducing thermal resistance and allowing the expansion of the fuel slug. In this paper, the concise irradiation rig design is introduced, and the neutronic and thermal–hydraulic analyses, which were carried out mainly using MCNP (Monte Carlo N-Particle) and FLUENT codes, are presented. Out-of-pile tests were conducted prior to irradiation to verify the manufacturing quality and hydraulic performance of the rig. Nondestructive postirradiation examinations using cold neutron radiography technology were conducted to check fuel cladding integrity and swelling behavior. The results of the preliminary examinations confirmed the safety and effectiveness of the design.

  8. Generation of hourly irradiation synthetic series using the neural network multilayer perceptron

    Energy Technology Data Exchange (ETDEWEB)

    Hontoria, L.; Aguilera, J. [Universidad de Jaen, Linares-Jaen (Spain). Dpto. de Electronica; Zufiria, P. [Ciudad Universitaria, Madrid (Spain). Grupo de Redes Neuronales

    2002-05-01

    In this work, a methodology based on the neural network model called multilayer perceptron (MLP) to solve a typical problem in solar energy is presented. This methodology consists of the generation of synthetic series of hourly solar irradiation. The model presented is based on the capacity of the MLP for finding relations between variables for which interrelation is unknown explicitly. The information available can be included progressively at the series generator at different stages. A comparative study with other solar irradiation synthetic generation methods has been done in order to demonstrate the validity of the one proposed. (author)

  9. Solid Oxide Fuel Cell Hybrid System for Distributed Power Generation

    Energy Technology Data Exchange (ETDEWEB)

    Nguyen Minh

    2005-12-01

    This report summarizes the work performed by Hybrid Power Generation Systems, LLC (HPGS) under Cooperative Agreement DE-FC2601NT40779 for the US Department of Energy, National Energy Technology Laboratory (DoE/NETL) entitled ''Solid Oxide Fuel Cell Hybrid System for Distributed Power Generation''. The main objective of this project is to develop and demonstrate the feasibility of a highly efficient hybrid system integrating a planar Solid Oxide Fuel Cell (SOFC) and a gas turbine. A conceptual hybrid system design was selected for analysis and evaluation. The selected system is estimated to have over 65% system efficiency, a first cost of approximately $650/kW, and a cost of electricity of 8.4 cents/kW-hr. A control strategy and conceptual control design have been developed for the system. A number of SOFC module tests have been completed to evaluate the pressure impact to performance stability. The results show that the operating pressure accelerates the performance degradation. Several experiments were conducted to explore the effects of pressure on carbon formation. Experimental observations on a functioning cell have verified that carbon deposition does not occur in the cell at steam-to-carbon ratios lower than the steady-state design point for hybrid systems. Heat exchanger design, fabrication and performance testing as well as oxidation testing to support heat exchanger life analysis were also conducted. Performance tests of the prototype heat exchanger yielded heat transfer and pressure drop characteristics consistent with the heat exchanger specification. Multicell stacks have been tested and performance maps were obtained under hybrid operating conditions. Successful and repeatable fabrication of large (>12-inch diameter) planar SOFC cells was demonstrated using the tape calendering process. A number of large area cells and stacks were successfully performance tested at ambient and pressurized conditions. A 25 MW plant configuration was

  10. Oxidizing dissolution mechanism of an irradiated MOX fuel in underwater aerated conditions at slightly acidic pH

    Science.gov (United States)

    Magnin, M.; Jégou, C.; Caraballo, R.; Broudic, V.; Tribet, M.; Peuget, S.; Talip, Z.

    2015-07-01

    The (U,Pu)O2 matrix behavior of an irradiated MIMAS-type (MIcronized MASter blend) MOX fuel, under radiolytic oxidation in aerated pure water at pH 5-5.5 was studied by combining chemical and radiochemical analyses of the alteration solution with Raman spectroscopy characterizations of the surface state. Two leaching experiments were performed on segments of irradiated fuel under different conditions: with or without an external γ irradiation field, over long periods (222 and 604 days, respectively). The gamma irradiation field was intended to be representative of the irradiation conditions for a fuel assembly in an underwater interim storage situation. The data acquired enabled an alteration mechanism to be established, characterized by uranium (UO22+) release mainly controlled by solubility of studtite over the long-term. The massive precipitation of this phase was observed for the two experiments based on high uranium oversaturation indexes of the solution and the kinetics involved depended on the irradiation conditions. External gamma irradiation accelerated the precipitation kinetics and the uranium concentrations (2.9 × 10-7 mol/l) were lower than for the non-irradiated reference experiment (1.4 × 10-5 mol/l), as the quantity of hydrogen peroxide was higher. Under slightly acidic pH conditions, the formation of an oxidized UO2+x phase was not observed on the surface and did not occur in the radiolysis dissolution mechanism of the fuel matrix. The Raman spectroscopy performed on the heterogeneous MOX fuel matrix surface, showed that the fluorite structure of the mainly UO2 phase surrounding the Pu-enriched aggregates had not been particularly impacted by any major structural change compared to the data obtained prior to leaching. For the plutonium, its behavior in solution involved a continuous release up to concentrations of approximately 3 × 10-6 mol L-1 with negligible colloid formation. This data appears to support a predominance of the +V oxidation

  11. Carbon 14 distribution in irradiated BWR fuel cladding and released carbon 14 after aqueous immersion of 6.5 years

    Energy Technology Data Exchange (ETDEWEB)

    Sakuragi, T. [Radioactive Waste Management Funding and Research Center, Tsukishima 1-15-7, Chuo City, Tokyo, 104-0052 (Japan); Yamashita, Y.; Akagi, M.; Takahashi, R. [TOSHIBA Corporation, Ukishima Cho 4-1, Kawasaki Ward, Kawasaki, 210-0862 (Japan)

    2016-07-01

    Spent fuel cladding which is highly activated and strongly contaminated is expected to be disposed of in an underground repository. A typical activation product in the activated metal waste is carbon 14 ({sup 14}C), which is mainly generated by the {sup 14}N(n,p){sup 14}C reaction and produces a significant exposure dose due to the large inventory, long half-life (5730 years), rapid release rate, and the speciation and consequent migration parameters. In the preliminary Japanese safety case, the release of radionuclides from the metal matrix is regarded as the corrosion-related congruent release, and the cladding oxide layer is regarded as a source of instant release fraction (IRF). In the present work, specific activity of {sup 14}C was measured using an irradiated BWR fuel cladding (Zircaloy-2, average rod burnup of 41.6 GWd/tU) which has an external oxide film having a thickness of 25.3 μm. The {sup 14}C specific activity of the base metal was 1.49*10{sup 4} Bq/g, which in the corresponding burnup is comparable to values in the existing literature, which were obtained from various irradiated claddings. Although the specific activity in oxide was 2.8 times the base metal activity due to the additive generation by the {sup 17}O(n,α){sup 14}C reaction, the {sup 14}C abundance in oxide was less than 10% of total inventory. A static leaching test using the cladding tube was carried out in an air-tight vessel filled with a deoxygenated dilute NaOH solution (pH of 12.5) at room temperature. After 6.5 years, {sup 14}C was found in each leachate fraction of gas phase and dissolved organics and inorganics, the total of which was less than 0.01% of the {sup 14}C inventory of the immersed cladding tube. A simple calculation based on the congruent release with Zircaloy corrosion has suggested that the 96.7% of released {sup 14}C was from the external oxide layer and 3.3% was from the base Zircaloy metal. However, both the {sup 14}C abundance and the low leaching rate

  12. Irradiation dose and temperature dependence of fracture toughness in high dose HT9 steel from the fuel duct of FFTF

    Energy Technology Data Exchange (ETDEWEB)

    Byun, Thak Sang; Toloczko, Mychailo B.; Saleh, Tarik A.; Maloy, Stuart A.

    2013-01-14

    To expand the knowledge base for fast reactor core materials, fracture toughness has been evaluated for high dose HT9 steel using miniature disk compact tension (DCT) specimens. The HT9 steel DCT specimens were machined from the ACO-3 fuel duct of the Fast Flux Test Facility (FFTF), which achieved high doses in the range of 3–148 dpa at 378–504 C. The static fracture resistance (J-R) tests have been performed in a servohydraulic testing machine in vacuum at selected temperatures including room temperature, 200 C, and each irradiation temperature. Brittle fracture with a low toughness less than 50 MPa pm occurred in room temperature tests when irradiation temperature was below 400 C, while ductile fracture with stable crack growth was observed when irradiation temperature was higher. No fracture toughness less than 100 MPa pm was measured when the irradiation temperature was above 430 C. It was shown that the influence of irradiation temperature was dominant in fracture toughness while the irradiation dose has only limited influence over the wide dose range 3–148 dpa. A slow decrease of fracture toughness with test temperature above room temperature was observed for the nonirradiated and high temperature (>430 *C) irradiation cases, which indicates that the ductile–brittle transition temperatures (DBTTs) in those conditions are lower than room temperature. A comparison with the collection of existing data confirmed the dominance of irradiation temperature in the fracture toughness of HT9 steels.

  13. Comparison of laser welding conditions of Zircaloy-4 and stainless steel for nuclear fuel irradiation rig

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kahye; Hong, Jintae; Joung, Changyoung; Heo, Sungho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Various materials for Zircaloy-4, SUS 316L, such as Inconel, are used as a survey rig that has been produced for fuel irradiation testing. Precision sensors, thermocouples, LVDT, and SPND should also be assembled. Therefore, a welding device for connecting them is necessary. With a high density of energy, laser welding can be properly used in a deep permeation, and in precisely welding narrow and deep joints. In particular, it has been applied to other fields such as metal welding. Since the technology bears no pores or cavities, resulting in a clean surface after the welding process, it does not require an 'after-process' such as grinding or polishing, which is useful where high water-tightness is required. Therefore, we developed and researched a special fiber laser welding system for the production of a nuclear research rig. Through the above test, the different conditions of laser welding were found for Zircaloy-4 and AISI 316L used for producing a nuclear fuel research rig, performing the most optimal welding conditions according to the properties of the materials in the future.

  14. Post-irradiation examination of prototype Al-64 wt% U{sub 3}Si{sub 2} fuel rods from NRU

    Energy Technology Data Exchange (ETDEWEB)

    Sears, D.F.; Primeau, M.F.; Buchanan, C.; Rose, D. [Chalk River Labs., Ontario (Canada)

    1997-08-01

    Three prototype fuel rods containing Al-64 wt% U{sub 3}Si{sub 2} (3.15 gU/cm{sup 3}) have been irradiated to their design burnup in the NRU reactor without incident. The fuel was fabricated using production-scale equipment and processes previously developed for Al-U{sub 3}Si fuel fabrication at Chalk River Laboratories, and special equipment developed for U{sub 3}Si{sub 2} powder production and handling. The rods were irradiated in NRU up to 87 at% U-235 burnup under typical driver fuel conditions; i.e., nominal coolant inlet temperature 37{degrees}C, inlet pressure 654 kPa, mass flow 12.4 L/s, and element linear power ratings up to 73 kW/m. Post-irradiation examinations showed that the fuel elements survived the irradiation without defects. Fuel core diametral increases and volumetric swelling were significantly lower than that of Al-61 wt% U{sub 3}Si fuel irradiated under similar conditions. This irradiation demonstrated that the fabrication techniques are adequate for full-scale fuel manufacture, and qualified the fuel for use in AECL`s research reactors.

  15. A Study on Fuel Options for Power Generation in Thailand

    Directory of Open Access Journals (Sweden)

    Weerin Wangjiraniran

    2010-07-01

    Full Text Available This study focuses on the impact of utilizing gas, coal and nuclear for longterm power generation on generation cost, emission and resource availability. A scenario-based energy accounting model has been applied for creating long-term future scenarios. A baseline scenario has been created on the basis of the existing power development plan (PDP. Three alternative scenarios of coal, nuclear and gas options have been projected for the period beyond the PDP, i.e. 2022-2030. The results indicate that nuclear has high potential for GHG mitigation and cost reduction. For coal option, the benefit of cost reduction would be diminished at carbon price above 40 USD/ton. However, clean technology development as well as the momentum of global trend will be the key factor for coal utilization. The results also show the need of fuel diversification in term of the natural gas reserve depletion. It is clearly seen that natural gas supply in Thailand would inevitably depends very much on the LNG import in long-term. Hence, attraction of natural gas in term of cheap domestic resource utilization will be vanished.

  16. ELECTRICITY GENERATION FROM SWINE WASTEWATER USING MICROBIAL FUEL CELL

    Directory of Open Access Journals (Sweden)

    Chimezie Jason Ogugbue

    2015-11-01

    Full Text Available Electricity generation from swine wastewater using microbial fuel cell (MFC was investigated. Swine wastewater was collected into dual-chambered (aerobic and anaerobic fuel cell. The maximum power output using copper and carbon electrodes were 250.54 and 52.33 µW, while 10.0 and 5.0 cm salt bridge length between the cathode and anode were 279.50 and 355.26 µW, respectively. Potassium permanganate and ordinal water gave a maximum power output of 1287.8 and 13 9.18 µW. MFCs utilize microbial communities to degrade organic materials found within wastewater and converted stored chemical energy to electrical energy in a single step. The initial bacterial and fungal counts were 7.4×106 and 1.1×103 CFU ml-1. Bacterial counts steadily increased with time to 1.40×107 CFU ml-1 while fungal count declined to 4.4×106 CFU ml-1 after day 60. The declined in microbial counts may be attributed to the time necessary for acclimatization of microbes to the anode. The genera identified were Bacillus, Citrobacter, Pseudomonas, Lactobacillus, Escherichia coli, Aspergillus and Rhizopus. These microbes acted as primary and secondary utilizers, utilizing carbon and other organics of the wastewater. Chemical parameters indicated that the biochemical oxygen demand ranged from 91.4–23.2 mg/L, giving 75% while the chemical oxygen demand ranged from 243.1–235.2 mg/L, representing 3.3%. Although, the metabolic activities of microbes were responsible for the observed degradation, leading to electricity, the overall power output depended on the distance between the anode and cathode compartment, types of electrode materials and mediators and oxygen reaction at the cathode.

  17. Next Generation Bipolar Plates for Automotive PEM Fuel Cells

    Energy Technology Data Exchange (ETDEWEB)

    Adrianowycz, Orest; Norley, Julian; Stuart, David J; Flaherty, David; Wayne, Ryan; ; Williams, Warren; Tietze, Roger; Nguyen, Yen-Loan H; Zawodzinski, Tom; Pietrasz, Patrick

    2010-04-15

    The results of a successful U.S. Department of Energy (DoE) funded two-year $2.9 MM program lead by GrafTech International Inc. (GrafTech) are reported and summarized. The program goal was to develop the next generation of high temperature proton exchange membrane (PEM) fuel cell bipolar plates for use in transportation fuel cell applications operating at temperatures up to 120 °C. The bipolar plate composite developed during the program is based on GrafTech’s GRAFCELL resin impregnated flexible graphite technology and makes use of a high temperature Huntsman Advanced Materials resin system which extends the upper use temperature of the composite to the DoE target. High temperature performance of the new composite is achieved with the added benefit of improvements in strength, modulus, and dimensional stability over the incumbent resin systems. Other physical properties, including thermal and electrical conductivity of the new composite are identical to or not adversely affected by the new resin system. Using the new bipolar plate composite system, machined plates were fabricated and tested in high temperature single-cell fuel cells operating at 120 °C for over 1100 hours by Case Western Reserve University. Final verification of performance was done on embossed full-size plates which were fabricated and glued into bipolar plates by GrafTech. Stack testing was done on a 10-cell full-sized stack under a simulated drive cycle protocol by Ballard Power Systems. Freeze-thaw performance was conducted by Ballard on a separate 5-cell stack and shown to be within specification. A third stack was assembled and shipped to Argonne National Laboratory for independent performance verification. Manufacturing cost estimate for the production of the new bipolar plate composite at current and high volume production scenarios was performed by Directed Technologies Inc. (DTI). The production cost estimates were consistent with previous DoE cost estimates performed by DTI for the

  18. Generation of Long-Lived Isomeric States via Bremsstrahlung Irradiation

    CERN Document Server

    Cheng, Y; Tang, C; Liu, Y; Jin, Q; Cheng, Yao; Xia, Bing; Tang, Chuanxiang; Liu, Yinong; Jin, Qingxiu

    2006-01-01

    A method to generate long-lived isomeric states effectively for Mossbauer applications is reported. We demonstrate that this method is better and easier to provide highly sensitive Mossbauer effect of long-lived isomers (>1ms) such as 103Rh. Excitation of (gamma,gamma) process by synchrotron radiation is painful due mainly to their limited linewidth. Instead,(gamma,gamma') process of bremsstrahlung excitation is applied to create these long-lived isomers. Isomers of 45Sc, 107Ag, 109Ag, and 103Rh have been generated from this method. Among them, 103Rh is the only one that we have obtained the gravitational effect at room temperature.

  19. Characterization of Irradiated Metal Waste from the Pyrometallurgical Treatment of Used EBR-II Fuel

    Energy Technology Data Exchange (ETDEWEB)

    B.R. Westphal; K.C. Marsden; W.M. McCartin; S.M. Frank; D.D. Keiser, Jr.; T.S. Yoo; D. Vaden; D.G. Cummings; K.J. Bateman; J. J. Giglio; T. P. O' Holleran; P. A. Hahn; M. N. Patterson

    2013-03-01

    As part of the pyrometallurgical treatment of used Experimental Breeder Reactor-II fuel, a metal waste stream is generated consisting primarily of cladding hulls laden with fission products noble to the electrorefining process. Consolidation by melting at high temperature [1873 K (1600 degrees C)] has been developed to sequester the noble metal fission products (Zr, Mo, Tc, Ru, Rh, Te, and Pd) which remain in the iron-based cladding hulls. Zirconium from the uranium fuel alloy (U-10Zr) is also deposited on the hulls and forms Fe-Zr intermetallics which incorporate the noble metals as well as residual actinides during processing. Hence, Zr has been chosen as the primary indicator for consistency of the metal waste. Recently, the first production-scale metal waste ingot was generated and sampled to monitor Zr content for Fe-Zr intermetallic phase formation and validation of processing conditions. Chemical assay of the metal waste ingot revealed a homogeneous distribution of the noble metal fission products as well as the primary fuel constituents U and Zr. Microstructural characterization of the ingot confirmed the immobilization of the noble metals in the Fe-Zr intermetallic phase.

  20. Performance evaluation and post-irradiation examination of a novel LWR fuel composed of U0.17ZrH1.6 fuel pellets bonded to Zircaloy-2 cladding by lead bismuth eutectic

    Science.gov (United States)

    Balooch, Mehdi; Olander, Donald R.; Terrani, Kurt A.; Hosemann, Peter; Casella, Andrew M.; Senor, David J.; Buck, Edgar C.

    2017-04-01

    A novel light water reactor fuel has been designed and fabricated at the University of California, Berkeley; irradiated at the Massachusetts Institute of Technology Reactor; and examined within the Radiochemical Processing Laboratory at the Pacific Northwest National Laboratory. This fuel consists of U0.17ZrH1.6 fuel pellets core-drilled from TRIGA reactor fuel elements that are clad in Zircaloy-2 and bonded with lead-bismuth eutectic. The performance evaluation and post irradiation examination of this fuel are presented here.

  1. The Case for Natural Gas Fueled Solid Oxide Fuel Cell Power Systems for Distributed Generation

    Energy Technology Data Exchange (ETDEWEB)

    Chick, Lawrence A.; Weimar, Mark R.; Whyatt, Greg A.; Powell, Michael R.

    2015-02-01

    Natural-gas-fueled solid oxide fuel cell (NGSOFC) power systems yield electrical conversion efficiencies exceeding 60% and may become a viable alternative for distributed generation (DG) if stack life and manufacturing economies of scale can be realized. Currently, stacks last approximately 2 years and few systems are produced each year because of the relatively high cost of electricity from the systems. If mass manufacturing (10,000 units per year) and a stack life of 15 years can be reached, the cost of electricity from an NGSOFC system is estimated to be about 7.7 ¢/kWh, well within the price of commercial and residential retail prices at the national level (9.9-10¢/kWh and 11-12 ¢/kWh, respectively). With an additional 5 ¢/kWh in estimated additional benefits from DG, NGSOFC could be well positioned to replace the forecasted 59-77 gigawatts of capacity loss resulting from coal plant closures due to stricter emissions regulations and low natural gas prices.

  2. MODELLING AND FUZZY LOGIC CONTROL OF PEM FUEL CELL SYSTEM POWER GENERATION FOR RESIDENTIAL APPLICATION

    OpenAIRE

    Khaled MAMMAR; CHAKER, Abdelkader

    2010-01-01

    This paper presents a dynamic model of Fuel cell system for residential power generation. The models proposedinclude a fuel cell stack model, reformer model and DC/AC inverter model. More then an analytical details ofhow active and reactive power output of a proton-exchange-membrane (PEM) fuel cell system is controlled.Furthermore a fuzzy logic (FLC) controller is used to control active power of PEM fuel cell system. Thecontroller modifies the hydrogen flow feedback from the terminal load. Si...

  3. Validation of Direct Normal Irradiance from Meteosat Second Generation

    Science.gov (United States)

    Meyer, Angela; Stöckli, Reto; Vuilleumier, Laurent; Wilbert, Stefan; Zarzalejo, Luis

    2016-04-01

    We present a validation study of Direct Normal Irradiance (DNI) derived from MSG/SEVIRI radiance measurements over the site of Plataforma Solar de Almeria (PSA), a solar power plant in Southern Spain. The 1 km x 1 km site of PSA hosts about a dozen pyrheliometers operated by the German Aerospace Centre (DLR) and the Centre for Energy, Environment and Technological Research (CIEMAT). They provide high-quality long-term measurements of surface DNI on a site of the scale of the MSG/SEVIRI pixel resolution. This makes the PSA DNI measurements a dataset particularly well suited for satellite validation purposes. The satellite-based surface DNI was retrieved from MSG/SEVIRI radiances by the HelioMont algorithm (Stöckli 2013) that forms part of the Heliosat algorithm family (e.g. Müller et al., 2004). We have assessed the accuracy of this DNI product for the PSA site by comparing with the in-situ measured DNIs of June 2014 - July 2015. Despite a generally good agreement, the HelioMont DNI exhibits a significant low bias at the PSA site, that is most pronounced during clear-sky periods. We present a bias correction method and discuss (1) the role of circumsolar diffuse radiation and (2) the role of climatological vs. reanalysis-based aerosol optical properties therein. We also characterize and assess the temporal variability of the HelioMont DNI as compared to the in situ measured DNIs, and will discuss and quantify the uncertainties in both DNI datasets.

  4. Investigation of the Feasibility of Utilizing Gamma Emission Computed Tomography in Evaluating Fission Product Migration in Irradiated TRISO Fuel Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Jason M. Harp; Paul A. Demkowicz

    2014-10-01

    In the High Temperature Gas-Cooled Reactor (HTGR) the TRISO particle fuel serves as the primary fission product containment. However the large number of TRISO particles present in proposed HTGRs dictates that there will be a small fraction (~10-4 to 10-5) of as manufactured and in-pile particle failures that will lead to some fission product release. The matrix material surrounding the TRISO particles in fuel compacts and the structural graphite holding the TRISO particles in place can also serve as sinks for containing any released fission products. However data on the migration of solid fission products through these materials is lacking. One of the primary goals of the AGR-3/4 experiment is to study fission product migration from failed TRISO particles in prototypic HTGR components such as structural graphite and compact matrix material. In this work, the potential for a Gamma Emission Computed Tomography (GECT) technique to non-destructively examine the fission product distribution in AGR-3/4 components and other irradiation experiments is explored. Specifically, the feasibility of using the Idaho National Laboratory (INL) Hot Fuels Examination Facility (HFEF) Precision Gamma Scanner (PGS) system for this GECT application is considered. To test the feasibility, the response of the PGS system to idealized fission product distributions has been simulated using Monte Carlo radiation transport simulations. Previous work that applied similar techniques during the AGR-1 experiment will also be discussed as well as planned uses for the GECT technique during the post irradiation examination of the AGR-2 experiment. The GECT technique has also been applied to other irradiated nuclear fuel systems that were currently available in the HFEF hot cell including oxide fuel pins, metallic fuel pins, and monolithic plate fuel.

  5. Capture of mercury in combustion systems by in situ-generated titania particles with UV irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Wu, C.Y.; Lee, T.G.; Tyree, G.; Arar, E.; Biswas, P. [University of Cincinnati, Cincinnati, OH (United States). Dept. of Civil and Environmental Engineering

    1998-10-01

    In situ-generated sorbent titania particles with ultraviolet (UV) irradiation have been shown to be effective in capture of mercury in combustor exhausts. Results of experiments conducted with the (1) sorbent precursor only, (2) mercury only, (3) mercury and UV irradiation, and (4) mercury, titania, and UV irradiation are presented to elucidate the mechanisms of the capture process. Capture efficiencies (percentage of Hg captured on the filter) as high as 96% were measured for mercury by titania with UV irradiation. A very high surface area titania sorbent was first formed, with mercury vapors condensing onto this surface, followed by photocatalytic oxidation and binding with the sorbent particles. The process has significant potential as a low-cost methodology for mercury control in practical combustion systems. Minimal retrofitting may be necessary as conventional particulate control devices such as electrostatic precipitators have coronas with UV radiation present.

  6. Feasibility study of U-235, Pu-239 and Pu-240 content determination in an irradiated fuel by neutron transmission analysis

    Energy Technology Data Exchange (ETDEWEB)

    Naguib, K.; Michaiel, M.L.; Morcos, H.N

    1998-07-01

    A proposed nondestructive method and its feasibility for the determination of U-235, Pu-239 and Pu-240 contents in an irradiated fuel is described. The method is based on the use of shape fit analysis of the Time-Of-Flight (TOF) neutron transmission data of the irradiated fuel for neutron energies below 3 eV. The neutron transmission experiment of the irradiated fuel is planned to carry out using one of the TOF spectrometers installed at ET-RR-1 reactor. The computer code SHAPE is adapted taking into account the known parameters of resonances of certain fissile and fission product nuclei to provide the fit analysis. The content of the gross-fissile and fission product isotopes are determined from the burn-up calculations of the fuel assembly of the ET-RR-1 reactor with defined history. The effect of both uncertainties in resonance parameters on the deduced contents of fissile nuclei and statistical accuracy of the TOF measurements are estimated.

  7. The behaviour under irradiation of molybdenum matrix for inert matrix fuel containing americium oxide (CerMet concept)

    Science.gov (United States)

    D'Agata, E.; Knol, S.; Fedorov, A. V.; Fernandez, A.; Somers, J.; Klaassen, F.

    2015-10-01

    Americium is a strong contributor to the long term radiotoxicity of high activity nuclear waste. Transmutation by irradiation in nuclear reactors or Accelerator Driven System (ADS, subcritical reactors dedicated to transmutation) of long-lived nuclides like 241Am is therefore an option for the reduction of radiotoxicity of waste packages to be stored in a repository. In order to safely burn americium in a fast reactor or ADS, it must be incorporated in a matrix that could be metallic (CerMet target) or ceramic (CerCer target). One of the most promising matrix to incorporate Am is molybdenum. In order to address the issues (swelling, stability under irradiation, gas retention and release) of using Mo as matrix to transmute Am, two irradiation experiments have been conducted recently at the High Flux Reactor (HFR) in Petten (The Netherland) namely HELIOS and BODEX. The BODEX experiment is a separate effect test, where the molybdenum behaviour is studied without the presence of fission products using 10B to ;produce; helium, the HELIOS experiment included a more representative fuel target with the presence of Am and fission product. This paper covers the results of Post Irradiation Examination (PIE) of the two irradiation experiments mentioned above where molybdenum behaviour has been deeply investigated as possible matrix to transmute americium (CerMet fuel target). The behaviour of molybdenum looks satisfying at operating temperature but at high temperature (above 1000 °C) more investigation should be performed.

  8. Digestion of algal biomass for electricity generation in microbial fuel cells.

    Science.gov (United States)

    Nishio, Koichi; Hashimoto, Kazuhito; Watanabe, Kazuya

    2013-01-01

    Algal biomass serves as a fuel for electricity generation in microbial fuel cells. This study constructed a model consortium comprised of an alga-digesting Lactobacillus and an iron-reducing Geobacter for electricity generation from photo-grown Clamydomonas cells. Total power-conversion efficiency (from Light to electricity) was estimated to be 0.47%.

  9. Compost in plant microbial fuel cell for bioelectricity generation.

    Science.gov (United States)

    Moqsud, M A; Yoshitake, J; Bushra, Q S; Hyodo, M; Omine, K; Strik, David

    2015-02-01

    Recycling of organic waste is an important topic in developing countries as well as developed countries. Compost from organic waste has been used for soil conditioner. In this study, an experiment has been carried out to produce green energy (bioelectricity) by using paddy plant microbial fuel cells (PMFCs) in soil mixed with compost. A total of six buckets filled with the same soil were used with carbon fiber as the electrodes for the test. Rice plants were planted in five of the buckets, with the sixth bucket containing only soil and an external resistance of 100 ohm was used for all cases. It was observed that the cells with rice plants and compost showed higher values of voltage and power density with time. The highest value of voltage showed around 700 mV when a rice plant with 1% compost mixed soil was used, however it was more than 95% less in the case of no rice plant and without compost. Comparing cases with and without compost but with the same number of rice plants, cases with compost depicted higher voltage to as much as 2 times. The power density was also 3 times higher when the compost was used in the paddy PMFCs which indicated the influence of compost on bio-electricity generation. Copyright © 2014 Elsevier Ltd. All rights reserved.

  10. Electricity generation from tetrathionate in microbial fuel cells by acidophiles.

    Science.gov (United States)

    Sulonen, Mira L K; Kokko, Marika E; Lakaniemi, Aino-Maija; Puhakka, Jaakko A

    2015-03-02

    Inorganic sulfur compounds, such as tetrathionate, are often present in mining process and waste waters. The biodegradation of tetrathionate was studied under acidic conditions in aerobic batch cultivations and in anaerobic anodes of two-chamber flow-through microbial fuel cells (MFCs). All four cultures originating from biohydrometallurgical process waters from multimetal ore heap bioleaching oxidized tetrathionate aerobically at pH below 3 with sulfate as the main soluble metabolite. In addition, all cultures generated electricity from tetrathionate in MFCs at pH below 2.5 with ferric iron as the terminal cathodic electron acceptor. The maximum current and power densities during MFC operation and in the performance analysis were 79.6 mA m(-2) and 13.9 mW m(-2) and 433 mA m(-2) and 17.6 mW m(-2), respectively. However, the low coulombic efficiency (below 5%) indicates that most of the electrons were directed to other processes, such as aerobic oxidation of tetrathionate and unmeasured intermediates. The microbial community analysis revealed that the dominant species both in the anolyte and on the anode electrode surface of the MFCs were Acidithiobacillus spp. and Ferroplasma spp. This study provides a proof of concept that tetrathionate serves as electron donor for biological electricity production in the pH range of 1.2-2.5. Copyright © 2014 Elsevier B.V. All rights reserved.

  11. Carbon fiber enhanced bioelectricity generation in soil microbial fuel cells.

    Science.gov (United States)

    Li, Xiaojing; Wang, Xin; Zhao, Qian; Wan, Lili; Li, Yongtao; Zhou, Qixing

    2016-11-15

    The soil microbial fuel cell (MFC) is a promising biotechnology for the bioelectricity recovery as well as the remediation of organics contaminated soil. However, the electricity production and the remediation efficiency of soil MFC are seriously limited by the tremendous internal resistance of soil. Conductive carbon fiber was mixed with petroleum hydrocarbons contaminated soil and significantly enhanced the performance of soil MFC. The maximum current density, the maximum power density and the accumulated charge output of MFC mixed carbon fiber (MC) were 10, 22 and 16 times as high as those of closed circuit control due to the carbon fiber productively assisted the anode to collect the electron. The internal resistance of MC reduced by 58%, 83% of which owed to the charge transfer resistance, resulting in a high efficiency of electron transfer from soil to anode. The degradation rates of total petroleum hydrocarbons enhanced by 100% and 329% compared to closed and opened circuit controls without the carbon fiber respectively. The effective range of remediation and the bioelectricity recovery was extended from 6 to 20cm with the same area of air-cathode. The mixed carbon fiber apparently enhanced the bioelectricity generation and the remediation efficiency of soil MFC by means of promoting the electron transfer rate from soil to anode. The use of conductively functional materials (e.g. carbon fiber) is very meaningful for the remediation and bioelectricity recovery in the bioelectrochemical remediation.

  12. Fuel flexibility in power generation onboard offshore floating units

    Energy Technology Data Exchange (ETDEWEB)

    Keep, Jeroen van [Waertsilae Corporation, Helsinki (Finland)

    2012-07-01

    Power Plants for offshore oil and gas installations utilizing dual fuel (DF) reciprocating engines are by many owners seen as an interesting alternative to conventional solutions due to the apparent advantages in fuel flexibility, fuel efficiency and lower emission. The paper summarizes the dual fuel technology, typical solutions for FPSO's and operational. Items that are discussed: DF operation and how it works; fuel flexibility, including transfer between fuel modes; fuel efficiency, also in production an important cost saver; emissions of the different fuel modes; size and weights, constraints; experiences of the P-63 project. With the above it is safe to conclude that the DF-technology is mature with important benefits for the offshore production market in certain specific applications, most notably the FPSO's for fields in low gas to oil ratios, bringing important fuel cost savings and also for new-built F-LNG/FSO/FPSO's where the power plant can be accommodated below decks, freeing up valuable deck space for the process plant. (author)

  13. Multiple Irradiation Capsule Experiment (MICE)-3B Irradiation Test of Space Fuel Specimens in the Advanced Test Reactor (ATR) - Close Out Documentation for Naval Reactors (NR) Information

    Energy Technology Data Exchange (ETDEWEB)

    M. Chen; CM Regan; D. Noe

    2006-01-09

    Few data exist for UO{sub 2} or UN within the notional design space for the Prometheus-1 reactor (low fission rate, high temperature, long duration). As such, basic testing is required to validate predictions (and in some cases determine) performance aspects of these fuels. Therefore, the MICE-3B test of UO{sub 2} pellets was designed to provide data on gas release, unrestrained swelling, and restrained swelling at the upper range of fission rates expected for a space reactor. These data would be compared with model predictions and used to determine adequacy of a space reactor design basis relative to fission gas release and swelling of UO{sub 2} fuel and to assess potential pellet-clad interactions. A primary goal of an irradiation test for UN fuel was to assess performance issues currently associated with this fuel type such as gas release, swelling and transient performance. Information learned from this effort may have enabled use of UN fuel for future applications.

  14. Electricity generation from tetrathionate in microbial fuel cells by acidophiles

    Energy Technology Data Exchange (ETDEWEB)

    Sulonen, Mira L.K., E-mail: mira.sulonen@tut.fi; Kokko, Marika E.; Lakaniemi, Aino-Maija; Puhakka, Jaakko A.

    2015-03-02

    Highlights: • Electricity can be generated from tetrathionate in MFCs at pH below 2.5. • Tetrathionate disproportionated to sulfate and elemental sulfur. • Biohydrometallurgical process waters contained electrochemically active bacteria. • Acidithiobacillus spp. and Ferroplasma spp. were identified from the MFCs. - Abstract: Inorganic sulfur compounds, such as tetrathionate, are often present in mining process and waste waters. The biodegradation of tetrathionate was studied under acidic conditions in aerobic batch cultivations and in anaerobic anodes of two-chamber flow-through microbial fuel cells (MFCs). All four cultures originating from biohydrometallurgical process waters from multimetal ore heap bioleaching oxidized tetrathionate aerobically at pH below 3 with sulfate as the main soluble metabolite. In addition, all cultures generated electricity from tetrathionate in MFCs at pH below 2.5 with ferric iron as the terminal cathodic electron acceptor. The maximum current and power densities during MFC operation and in the performance analysis were 79.6 mA m{sup −2} and 13.9 mW m{sup −2} and 433 mA m{sup −2} and 17.6 mW m{sup −2}, respectively. However, the low coulombic efficiency (below 5%) indicates that most of the electrons were directed to other processes, such as aerobic oxidation of tetrathionate and unmeasured intermediates. The microbial community analysis revealed that the dominant species both in the anolyte and on the anode electrode surface of the MFCs were Acidithiobacillus spp. and Ferroplasma spp. This study provides a proof of concept that tetrathionate serves as electron donor for biological electricity production in the pH range of 1.2–2.5.

  15. Mechanistic approach for nitride fuel evolution and fission product release under irradiation

    Science.gov (United States)

    Dolgodvorov, A. P.; Ozrin, V. D.

    2017-01-01

    A model for describing uranium-plutonium mixed nitride fuel pellet burning was developed. Except fission products generating, the model includes impurities of oxygen and carbon. Nitrogen behaviour in nitride fuel was analysed and the nitrogen chemical potential in solid solution with uranium-plutonium nitride was constructed. The chemical program module was tested with the help of thermodynamic equilibrium phase distribution calculation. Results were compared with analogous data in literature, quite good agreement was achieved, especially for uranium sesquinitride, metallic species and some oxides. Calculation of a process of nitride fuel burning was also conducted. Used mechanistic approaches for fission product evolution give the opportunity to find fission gas release fractions and also volumes of intergranular secondary phases. Calculations present that the most massive secondary phases are the oxide and metallic phases. Oxide phase contain approximately 1 % wt of substance over all time of burning with slightly increasing of content. Metallic phase has considerable rising of mass and by the last stage of burning it contains about 0.6 % wt of substance. Intermetallic phase has less increasing rate than metallic phase and include from 0.1 to 0.2 % wt over all time of burning. The highest element fractions of released gaseous fission products correspond to caesium and iodide.

  16. Construction of a Post-Irradiated Fuel Examination Shielded Enclosure Facility

    Energy Technology Data Exchange (ETDEWEB)

    Michael A. Lehto, Ph.D.; Boyd D. Christensen

    2008-05-01

    The U.S. Department of Energy (DOE) has committed to provide funding to the Idaho National Laboratory (INL) for new post-irradiation examination (PIE) equipment in support of advanced fuels development. This equipment will allow researchers at the INL to accurately characterize the behavior of experimental test fuels after they are removed from an experimental reactor also located at the INL. The accurate and detailed characterization of the fuel from the reactor, when used in conjunction with computer modeling, will allow DOE to more quickly understand the behavior of the fuel and to guide further development activities consistent with the missions of the INL and DOE. Due to the highly radioactive nature of the specimen samples that will be prepared and analyzed by the PIE equipment, shielded enclosures are required. The shielded cells will be located in the existing Analytical Laboratory (AL) basement (Rooms B-50 and B-51) at the INL Material and Fuels Complex (MFC). AL Rooms B-50 and B-51 will be modified to establish an area where sample containment and shielding will be provided for the analysis of radioactive fuels and materials while providing adequate protection for personnel and the environment. The area is comprised of three separate shielded cells for PIE instrumentation. Each cell contains an atmosphere interface enclosure (AIE) for contamination containment. The shielding will provide a work area consistent with the as-low-as-reasonably-achievable (ALARA) concept, assuming a source term of 10 samples in each of the three shielded areas. Source strength is assumed to be a maximum of 3 Ci at 0.75 MeV gamma for each sample. Each instrument listed below will be installed in an individual shielded enclosure: Shielded electron probe micro-analyzer (EPMA) Focused ion beam instrument (FIB) Micro-scale x-ray diffractometer (MXRD). The project is designed and expected to be built incrementally as funds are allocated. The initial phase will be to fund the

  17. Chemical thermodynamics of Cs and Te fission product interactions in irradiated LMFBR mixed-oxide fuel pins

    Science.gov (United States)

    Adamson, M. G.; Aitken, E. A.; Lindemer, T. B.

    1985-02-01

    A combination of fuel chemistry modelling and equilibrium thermodynamic calculations has been used to predict the atom ratios of Cs and Te fission products (Cs:Te) that find their way into the fuel-cladding interface region of irradiated stainless steel-clad mixed-oxide fast breeder reactor fuel pins. It has been concluded that the ratio of condensed, chemically-associated Cs and Te in the interface region,Čs:Te, which in turn determines the Te activity, is controlled by an equilibrium reaction between Cs 2Te and the oxide fuel, and that the value of Čs:Te is, depending on fuel 0:M, either equal to or slightly less than 2:1. Since Cs and Te fission products are both implicated as causative agents in FCCI (fission product-assisted inner surface attack of stainless steel cladding) and in FPLME (fission product-assisted liquid metal embrittlement of AISI-Type 316), the observed out-of-pile Cs:Te thresholds for FCCI (4˜:1) and FPLME (2˜:1) have been rationalized in terms of Cs:Te thermochemistry and phase equilibria. Also described in the paper is an updated chemical evolution model for reactive/volatile fission product behavior in irradiated oxide pins.

  18. Chemical thermodynamics of Cs and Te fission product interactions in irradiated LMFBR mixed-oxide fuel pins

    Energy Technology Data Exchange (ETDEWEB)

    Adamson, M.G.; Aitken, E.A. (General Electric Co., Sunnyvale, CA (USA). Advanced Nuclear Technology Operation)

    1985-02-01

    A combination of fuel chemistry modelling and equilibrium thermodynamic calculations has been used to predict the atom ratios of Cs and Te fission products (Cs:Te) that find their way into the fuel-cladding interface region of irradiated stainless steel-clad mixed-oxide fast breeder reactor fuel pins. It has been concluded that the ratio of condensed, chemically-associated Cs and Te in the interface region, fuel, and that the value of fuel O:M, either equal to or slightly less than 2:1. Since Cs and Te fission products are both implicated as causative agents in FCCI (fission product-assisted inner surface attack of stainless steel cladding) and in FPLME (fission product-assisted liquid metal embrittlement of AISI-Type 316), the observed out-of-pile Cs:Te thresholds for FCCI (proportional4:1) and FPLME (proportional2:1) have been rationalized in terms of Cs:Te thermochemistry and phase equilibria. Also described in the paper is an updated chemical evolution model for reactive/volatile fission product behavior in irradiated oxide pins.

  19. Monolithic fuel cell based power source for burst power generation

    Science.gov (United States)

    Fee, D. C.; Blackburn, P. E.; Busch, D. E.; Dees, D. W.; Dusek, J.; Easler, T. E.; Ellingson, W. A.; Flandermeyer, B. K.; Fousek, R. J.; Heiberger, J. J.

    A unique fuel cell coupled with a low power nuclear reactor presents an attractive approach for SDI burst power requirements. The monolithic fuel cell looks attractive for space applications and represents a quantum jump in fuel cell technology. Such a breakthrough in design is the enabling technology for lightweight, low volume power sources for space based pulse power systems. The monolith is unique among fuel cells in being an all solid state device. The capability for miniaturization, inherent in solid state devices, gives the low volume required for space missions. In addition, the solid oxide fuel cell technology employed in the monolith has high temperature reject heat and can be operated in either closed or open cycles. Both these features are attractive for integration into a burst power system.

  20. Armor: An {alpha}{beta}{gamma} assembly for irradiated fuel analysis; Armor: Chaine {alpha}{beta}{gamma} pour l'analyse des combustibles irradies

    Energy Technology Data Exchange (ETDEWEB)

    Beraud, M. [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1967-04-15

    The assembly ARMOR which was built with a view to carrying out research on irradiated fuels consists of an {alpha}{beta}{gamma} enclosure made up of 11 cells in line. After a general description of the assembly in its present form, the various functions are reviewed: introduction of the samples, chemical de-canning, dissolution of the irradiated uranium pellets, preparation of solutions for mass spectrometric analyses, disposal of the effluents and of the solid waste. The assembly-has been working since 1961. During the 5 to 6 years operation, various improvements have been made and a certain number of observations have been collected concerning the work. (author) [French] Construite en vue de repondre a un programme d'etudes de combustibles Irradies, la chaine Armor est une enceinte {alpha}{beta}{gamma} composee de 11 cellules en ligne. Apres une description generale de la chaine dans son etat actuel, les differentes fonctions sont passees en revue: entree des echantillons, degainage chimique, dissolution des pastilles d'uranium irradie, preparation des solutions pour les analyses par spectrometrie de masse, rejet des effluents et des dechets solides. La chaine est en service depuis 1961. Au cours des cinq a six annees d'exploitation, differentes ameliorations ont ete apportees et un ensemble d'observations sur le travail a ete recueilli. (auteur)

  1. 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.

  2. Analyses of the plasma generated by laser irradiation on sputtered target for determination of the thickness used for plasma generation

    Energy Technology Data Exchange (ETDEWEB)

    Kumaki, Masafumi, E-mail: masafumi.kumaki@riken.jp [Cooperative Major in Nuclear Energy, Waseda University, Shinjuku, Tokyo (Japan); RIKEN, Wako, Saitama (Japan); Ikeda, Shunsuke; Sekine, Megumi; Munemoto, Naoya [RIKEN, Wako, Saitama (Japan); Department of Energy Sciences, Tokyo Institute of Technology, Meguro, Tokyo (Japan); Fuwa, Yasuhiro [RIKEN, Wako, Saitama (Japan); Department of Physics and Astronomy, Kyoto University, Uji, Kyoto (Japan); Cinquegrani, David [American Nuclear Society, University of Michigan, Ann Arbor, Michigan 48109 (United States); Kanesue, Takeshi; Okamura, Masahiro [Collider-Accelerator Department, Brookhaven National Laboratory, Upton, New York 11973 (United States); Washio, Masakazu [Cooperative Major in Nuclear Energy, Waseda University, Shinjuku, Tokyo (Japan)

    2014-02-15

    In Brookhaven National Laboratory, laser ion source has been developed to provide heavy ion beams by using plasma generation with 1064 nm Nd:YAG laser irradiation onto solid targets. The laser energy is transferred to the target material and creates a crater on the surface. However, only the partial material can be turned into plasma state and the other portion is considered to be just vaporized. Since heat propagation in the target material requires more than typical laser irradiation period, which is typically several ns, only the certain depth of the layers may contribute to form the plasma. As a result, the depth is more than 500 nm because the base material Al ions were detected. On the other hand, the result of comparing each carbon thickness case suggests that the surface carbon layer is not contributed to generate plasma.

  3. Preliminary results of post-irradiation examination of the AGR-1 TRISO fuel compacts

    Energy Technology Data Exchange (ETDEWEB)

    Paul Demkowicz; John Hunn; Robert Morris; Jason Harp; Philip Winston; Charles Baldwin; Fred Montgomery; Scott Ploger; Isabella van Rooyen

    2012-10-01

    Five irradiated fuel compacts from the AGR-1 experiment have been examined in detail in order to assess in-pile fission product release behavior. Compacts were electrolytically deconsolidated and analyzed using the leach-burn-leach technique to measure fission product inventory in the compact matrix and identify any particles with a defective SiC layer. Loose particles were then gamma counted to measure the fission product inventory. One particle with a defective SiC layer was found in the five compacts examined. The fractional release of Ag 110m from the particles was significant. The total fraction of silver released from all the particles within a compact ranged from 0-0.63 and individual particles within a single compact often exhibited a very wide range of silver release. The average fractional release of Eu-154 from all particles in a compact was 2.4×10-4—1.3×10-2, which is indicative of release through intact coatings. The fractional Cs-134 inventory in the compact matrix was <2×10-5 when all coatings remained intact, indicating good cesium retention. Approximately 1% of the palladium inventory was found in the compact matrix for two of the compacts, indicating significant release through intact coatings.

  4. Characterization of pitch prepared from pyrolysis fuel oil via electron beam irradiation

    Science.gov (United States)

    Kim, Hong Gun; Park, Mira; Kim, Hak-Yong; Kwac, Lee Ku; Shin, Hye Kyoung

    2017-06-01

    Pitch samples were obtained from pyrolysis fuel oil by thermal treatment for 2 h at 300 °C after electron beam irradiation (EBI) treatment and by thermal treatment alone for different temperature of 250 °C, 300 °C, and 350 °C. EBI treatment was found to be an effective treatment for preparing pitch compare to the pitch obtained without EBI treatment. These results were confirmed by Fourier transform infrared spectroscopy (FT-IR) and Carbon-13 nuclear magnetic resonance (13C NMR) analyses, which showed the increase in the intensities of peaks corresponding to aromatic compounds. In the matrix-assisted laser desorption/ionization time-of-flight (MALDI-TOF) spectra, the amount of components with medium molecular weights in the pitch was found to increase with the temperature; likewise, in the case of the pitch obtained via EBI treatment, we found that the amount of components with higher molecular weight over 1000 (m/v) similarly increased. Further, the thermal stability and carbon yield at 850 °C of the pitch obtained by EBI were greater than those of samples subjected to thermal treatment at 250 and 300 °C.

  5. Direct power generation from waste coffee grounds in a biomass fuel cell

    Science.gov (United States)

    Jang, Hansaem; Ocon, Joey D.; Lee, Seunghwa; Lee, Jae Kwang; Lee, Jaeyoung

    2015-11-01

    We demonstrate the possibility of direct power generation from waste coffee grounds (WCG) via high-temperature carbon fuel cell technology. At 900 °C, the WCG-powered fuel cell exhibits a maximum power density that is twice than carbon black. Our results suggest that the heteroatoms and hydrogen contained in WCG are crucial in providing good cell performance due to its in-situ gasification, without any need for pre-reforming. As a first report on the use of coffee as a carbon-neutral fuel, this study shows the potential of waste biomass (e.g. WCG) in sustainable electricity generation in fuel cells.

  6. STEM-EDS analysis of fission products in neutron-irradiated TRISO fuel particles from AGR-1 experiment

    Science.gov (United States)

    Leng, B.; van Rooyen, I. J.; Wu, Y. Q.; Szlufarska, I.; Sridharan, K.

    2016-07-01

    Historic and recent post-irradiation-examination from the German AVR and Advanced Gas Reactor Fuel Development and Qualification Project have shown that 110 m Ag is released from intact tristructural isotropic (TRISO) fuel. Although TRISO fuel particle research has been performed over the last few decades, little is known about how metallic fission products are transported through the SiC layer, and it was not until March 2013 that Ag was first identified in the SiC layer of a neutron-irradiated TRISO fuel particle. The existence of Pd- and Ag-rich grain boundary precipitates, triple junction precipitates, and Pd nano-sized intragranular precipitates in neutron-irradiated TRISO particle coatings was investigated using Scanning Transmission Electron Microscopy and Energy Dispersive Spectroscopy analysis to obtain more information on the chemical composition of the fission product precipitates. A U-rich fission product honeycomb shape precipitate network was found near a micron-sized precipitate in a SiC grain about ∼5 μm from the SiC-inner pyrolytic carbon interlayer, indicating a possible intragranular transport path for uranium. A single Ag-Pd nano-sized precipitate was found inside a SiC grain, and this is the first research showing such finding in irradiated SiC. This finding may possibly suggest a possible Pd-assisted intragranular transport mechanism for Ag and may be related to void or dislocation networks inside SiC grains. Preliminary semi-quantitative analysis indicated the micron-sized precipitates to be Pd2Si2U with carbon existing inside these precipitates. However, the results of such analysis for nano-sized precipitates may be influenced by the SiC matrix. The results reported in this paper confirm the co-existence of Cd with Ag in triple points reported previously.

  7. Technical Development of Gamma Scanning for Irradiated Fuel Rod after Upgrade of System in Hot-cell

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Woo Seog; Kim, Hee Moon; Baik, Seung Je; Yoo, Byung Ok; Choo, Yong Sun

    2007-06-15

    Non-destructive test system was installed at hot-cell(M1) in IMEF(Irradiated Materials Examination Facility) more than 10 years ago for the diametric measurement and gamma scanning of fuel rod. But this system must be needed to be remodeled for the effective operations. In 2006, the system was upgraded for 3 months. The collimator bench can be movable with horizontal direction(x-direction) by motorized system for sectional gamma scanning and 3-dimensional tomography of fuel rod. So, gamma scanning for fuel rod can be detectable by x, y and rotation directions. It may be possible to obtain the radioactivities with radial and axial directions of pellet. This system is good for the series experiments with several positions. Operation of fuel bench and gamma detection program were linked each other by new program tools. It can control detection and bench moving automatically when gamma inspection of fuel rod is carried out with axial or radial positions. Some of electronic parts were added in PLC panel, and operating panel was re-designed for the remote control. To operate the fuel bench by computer, AD converter and some I/O cards were installed in computer. All of software were developed in Windows-XP system instead of DOS system. Control programs were made by visual-C language. After upgrade of system, DUPIC fuel which was irradiated in HANARO research reactor was detected by gamma scanning. The results were good and operation of gamma scanning showed reduced inspection time and easy control of data on series of detection with axial positions. With consideration of ECT(Eddy Current Test) installation, the computer program and hardware were set up as well. But ECT is not installed yet, so we have to check abnormal situation of program and hardware system. It is planned to install ECT in 2007.

  8. Results of High-Temperature Heating Test for Irradiated U-10Zr(-5Ce with T92 Cladding Fuel

    Directory of Open Access Journals (Sweden)

    June-Hyung Kim

    2016-11-01

    Full Text Available A microstructure observation using an optical microscope, SEM and EPMA was performed for the irradiated U-10Zr and U-10Zr-5Ce fuel slugs with a T92 cladding specimen after a high-temperature heating test. Also, the measured eutectic penetration rate was compared with the value predicted by the existing eutectic penetration correlation being used for design and modeling purposes. The heating temperature and duration time for the U-10Zr/T92 specimen were 750 °C and 1 h, and those for the U-10Zr-5Ce/T92 specimen were 800 °C and 1 h. In the case of the U-10Zr/T92 specimen, the migration phenomena of U, Zr, Fe, and Cr as well as the Nd lanthanide fission product were observed at the eutectic melting region. The measured penetration rate was similar to the value predicted by the existing eutectic penetration rate correlation. In addition, when comparing with measured eutectic penetration rates for the unirradiated U-10Zr fuel slug with FMS (ferritic martensitic steel, HT9 or Gr.91 cladding specimens which had been reported in the literature, the measured eutectic penetration rate for the irradiated fuel specimen was higher than that for the unirradiated U-10Zr specimen. In the case of the U-10Zr-5Ce/T92 specimen in which there had been a gap between the fuel slug and cladding after the irradiation test, the eutectic melting region was not found because contact between the fuel slug and cladding did not take place during the heating test.

  9. Irradiation dose and temperature dependence of fracture toughness in high dose HT9 steel from the fuel duct of FFTF

    Energy Technology Data Exchange (ETDEWEB)

    Byun, Thak Sang, E-mail: byunts@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Toloczko, Mychailo B. [Pacific Northwest National Laboratory, Richland, WA 99352 (United States); Saleh, Tarik A.; Maloy, Stuart A. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

    2013-01-15

    To expand the knowledge base for fast reactor core materials, fracture toughness has been evaluated for high dose HT9 steel using miniature disk compact tension (DCT) specimens. The HT9 steel DCT specimens were machined from the ACO-3 fuel duct of the Fast Flux Test Facility (FFTF), which achieved high doses in the range of 3-148 dpa at 378-504 Degree-Sign C. The static fracture resistance (J-R) tests have been performed in a servohydraulic testing machine in vacuum at selected temperatures including room temperature, 200 Degree-Sign C, and each irradiation temperature. Brittle fracture with a low toughness less than 50 MPa {radical}m occurred in room temperature tests when irradiation temperature was below 400 Degree-Sign C, while ductile fracture with stable crack growth was observed when irradiation temperature was higher. No fracture toughness less than 100 MPa {radical}m was measured when the irradiation temperature was above 430 Degree-Sign C. It was shown that the influence of irradiation temperature was dominant in fracture toughness while the irradiation dose has only limited influence over the wide dose range 3-148 dpa. A slow decrease of fracture toughness with test temperature above room temperature was observed for the nonirradiated and high temperature (>430 Degree-Sign C) irradiation cases, which indicates that the ductile-brittle transition temperatures (DBTTs) in those conditions are lower than room temperature. A comparison with the collection of existing data confirmed the dominance of irradiation temperature in the fracture toughness of HT9 steels.

  10. Oxidizing dissolution mechanism of an irradiated MOX fuel in underwater aerated conditions at slightly acidic pH

    Energy Technology Data Exchange (ETDEWEB)

    Magnin, M., E-mail: magali.magnin@cea.fr; Jégou, C.; Caraballo, R.; Broudic, V.; Tribet, M.; Peuget, S.; Talip, Z.

    2015-07-15

    Highlights: • Oxidizing dissolution mechanism of MOX fuel. • Effect of the influence of the interim storage conditions. • Raman spectroscopy characterizations. • Precipitation of Studtite-type secondary phases. • Heterogeneous microstructure of the (U,Pu)O{sub 2} oxide. - Abstract: The (U,Pu)O{sub 2} matrix behavior of an irradiated MIMAS-type (MIcronized MASter blend) MOX fuel, under radiolytic oxidation in aerated pure water at pH 5–5.5 was studied by combining chemical and radiochemical analyses of the alteration solution with Raman spectroscopy characterizations of the surface state. Two leaching experiments were performed on segments of irradiated fuel under different conditions: with or without an external γ irradiation field, over long periods (222 and 604 days, respectively). The gamma irradiation field was intended to be representative of the irradiation conditions for a fuel assembly in an underwater interim storage situation. The data acquired enabled an alteration mechanism to be established, characterized by uranium (UO{sub 2}{sup 2+}) release mainly controlled by solubility of studtite over the long-term. The massive precipitation of this phase was observed for the two experiments based on high uranium oversaturation indexes of the solution and the kinetics involved depended on the irradiation conditions. External gamma irradiation accelerated the precipitation kinetics and the uranium concentrations (2.9 × 10{sup −7} mol/l) were lower than for the non-irradiated reference experiment (1.4 × 10{sup −5} mol/l), as the quantity of hydrogen peroxide was higher. Under slightly acidic pH conditions, the formation of an oxidized UO{sub 2+x} phase was not observed on the surface and did not occur in the radiolysis dissolution mechanism of the fuel matrix. The Raman spectroscopy performed on the heterogeneous MOX fuel matrix surface, showed that the fluorite structure of the mainly UO{sub 2} phase surrounding the Pu-enriched aggregates had

  11. High fluence laser irradiation induces reactive oxygen species generation in human lung adenocarcinoma cells

    Science.gov (United States)

    Wang, Fang; Xing, Da; Chen, Tong-Sheng

    2006-09-01

    Low-power laser irradiation (LPLI) has been used for therapies such as curing spinal cord injury, healing wound et al. Yet, the mechanism of LPLI remains unclear. Our previous study showed that low fluences laser irradiation induces human lung adenocarcinoma cells (ASTC-a-1) proliferation, but high fluences induced apoptosis and caspase-3 activation. In order to study the mechanism of apoptosis induced by high fluences LPLI further, we have measured the dynamics of generation of reactive oxygen species (ROS) using H IIDCFDA fluorescence probes during this process. ASTC-a-1 cells apoptosis was induced by He-Ne laser irradiation at high fluence of 120J/cm2. A confocal laser scanning microscope was used to perform fluorescence imaging. The results demonstrated that high fluence LPLI induced the increase of mitochondria ROS. Our studies contribute to clarify the biological mechanism of high fluence LPLI-induced cell apoptosis.

  12. CURRENT STATUS OF INTEGRITY ASSESSMENT BY SIPPING SYSTEM OF SPENT FUEL BUNDLES IRRADIATED IN CANDU REACTOR

    Directory of Open Access Journals (Sweden)

    JONG-YOUL PARK

    2014-12-01

    Full Text Available In terms of safety and the efficient management of spent fuel storage, detecting failed fuel is one of the most important tasks in a CANada Deuterium Uranium (CANDU reactor operation. It has been successfully demonstrated that in a CANDU reactor, on-power failed fuel detection and location systems, along with alarm area gamma monitors, can detect and locate defective and suspect fuel bundles before discharging them from the reactor to the spent fuel storage bay. In the reception bay, however, only visual inspection has been used to identify suspect bundles. Gaseous fission product and delayed neutron monitoring systems cannot precisely distinguish failed fuel elements from each fuel bundle. This study reports the use of a sipping system in a CANDU reactor for the integrity assessment of spent fuel bundles. The integrity assessment of spent fuel bundles using this sipping system has shown promise as a nondestructive test for detecting a defective fuel bundle in a CANDU reactor.

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

    Science.gov (United States)

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

    2014-09-01

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

  14. Dynamics of fragmentation and multiple vacancy generation in irradiated single-walled carbon nanotubes

    CERN Document Server

    Javeed, Sumera; Ahmad, Shoaib

    2016-01-01

    The results from mass spectrometry of clusters sputtered from Cs+ irradiated single-walled carbon nano-tubes (SWCNTs) as a function of energy and dose identify the nature of the resulting damage in the form of multiple vacancy generation. For pristine SWCNTs at all Cs+ energies, C2 is the most dominant species, followed by C3, C4 and C1. The experiments were performed in three stages: in the first stage, Cs+ energy E(Cs+) was varied. During the second stage, the nanotubes were irradiated continuously at E(Cs+) = 5 keV for 1,800 s. Afterwards, the entire sequence of irradiation energies was repeated to differentiate between the fragmentation patterns of the pristine and of heavily irradiated SWCNTs. The sputtering and normalized yields identify the quantitative and relative extent of the ion-induced damage by creating double, triple and quadruple vacancies; the single vacancies are least favored. Sputtering from the heavily irradiated SWCNTs occurs not only from the damaged and fragmented nanotubes, but also f...

  15. First elevated-temperature performance testing of coated particle fuel compacts from the AGR-1 irradiation experiment

    Energy Technology Data Exchange (ETDEWEB)

    Charles A. Baldwin; John D. Hunn; Robert N. Morris; Fred C. Montgomery; Chinthaka M. Silva; Paul A. Demkowicz

    2014-05-01

    In the AGR-1 irradiation experiment, 72 coated-particle fuel compacts were taken to a peak burnup of 19.5% fissions per initial metal atom with no in-pile failures. This paper discusses the first post-irradiation test of these mixed uranium oxide/uranium carbide fuel compacts at elevated temperature to examine the fuel performance under a simulated depressurized conduction cooldown event. A compact was heated for 400 h at 1600 degrees C. Release of 85Kr was monitored throughout the furnace test as an indicator of coating failure, while other fission product releases from the compact were periodically measured by capturing them on exchangeable, water-cooled deposition cups. No coating failure was detected during the furnace test, and this result was verified by subsequent electrolytic deconsolidation and acid leaching of the compact, which showed that all SiC layers were still intact. However, the deposition cups recovered significant quantities of silver, europium, and strontium. Based on comparison of calculated compact inventories at the end of irradiation versus analysis of these fission products released to the deposition cups and furnace internals, the minimum estimated fractional losses from the compact during the furnace test were 1.9 x 10-2 for silver, 1.4 x 10-3 for europium, and 1.1 x 10-5 for strontium. Other post-irradiation examination of AGR-1 compacts indicates that similar fractions of europium and silver may have already been released by the intact coated particles during irradiation, and it is therefore likely that the detected fission products released from the compact in this 1600 degrees C furnace test were from residual fission products in the matrix. Gamma analysis of coated particles deconsolidated from the compact after the heating test revealed that silver content within each particle varied considerably; a result that is probably not related to the furnace test, because it has also been observed in other as-irradiated AGR-1 compacts. X

  16. First elevated-temperature performance testing of coated particle fuel compacts from the AGR-1 irradiation experiment

    Energy Technology Data Exchange (ETDEWEB)

    Baldwin, Charles A., E-mail: baldwinca@ornl.gov [Oak Ridge National Laboratory (ORNL), P.O. Box 2008, Oak Ridge, TN 37831-6295 (United States); Hunn, John D.; Morris, Robert N.; Montgomery, Fred C.; Silva, Chinthaka M. [Oak Ridge National Laboratory (ORNL), P.O. Box 2008, Oak Ridge, TN 37831-6295 (United States); Demkowicz, Paul A. [Idaho National Laboratory (INL), P.O. Box 1625, Idaho Falls, ID 83414 (United States)

    2014-05-01

    In the AGR-1 irradiation experiment, 72 coated-particle fuel compacts were taken to a peak burnup of 19.5% fissions per initial metal atom with no in-pile failures. This paper discusses the first post-irradiation test of these mixed uranium oxide/uranium carbide fuel compacts at elevated temperature to examine the fuel performance under a simulated depressurized conduction cooldown event. A compact was heated for 400 h at 1600 °C. Release of {sup 85}Kr was monitored throughout the furnace test as an indicator of coating failure, while other fission product releases from the compact were periodically measured by capturing them on exchangeable, water-cooled deposition cups. No coating failure was detected during the furnace test, and this result was verified by subsequent electrolytic deconsolidation and acid leaching of the compact, which showed that all SiC layers were still intact. However, the deposition cups recovered significant quantities of silver, europium, and strontium. Based on comparison of calculated compact inventories at the end of irradiation versus analysis of these fission products released to the deposition cups and furnace internals, the minimum estimated fractional losses from the compact during the furnace test were 1.9 × 10{sup −2} for silver, 1.4 × 10{sup −3} for europium, and 1.1 × 10{sup −5} for strontium. Other post-irradiation examination of AGR-1 compacts indicates that similar fractions of europium and silver may have already been released by the intact coated particles during irradiation, and it is therefore likely that the detected fission products released from the compact in this 1600 °C furnace test were from residual fission products in the matrix. Gamma analysis of coated particles deconsolidated from the compact after the heating test revealed that silver content within each particle varied considerably; a result that is probably not related to the furnace test, because it has also been observed in other as-irradiated

  17. Symposium on the reprocessing of irradiated fuels. Book 3, Session V

    Energy Technology Data Exchange (ETDEWEB)

    None

    1958-12-31

    Book three of this conference has a single-focused session V entitled Engineering and Economics, with 16 papers. The session is concerned with several phases of chemical reprocessing of fuels which are of a general nature. Hot labs, radiochemical analytical facilities, and high level development cells are described. Dissolution equipment, contactors, flow generation, measurement, and control equipment, samplers, connectors, carriers, valves, filters, and hydroclones are described and discussed. Papers are included on: radiation safety, chemical safety, radiochemical plant operating experience in the U.S., and heavy element isotopic buildup. The general economics of solvent extraction processing is discussed, and capital and operating costs for several U. S. plants given. The Atomic Energy Commission's chemical processing programs and administration are evaluated and the services offered and charges therefore are listed.

  18. Fundamental Studies of Irradiation-Induced Defect Formation and Fission Product Dynamics in Oxide Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Stubbins, James

    2012-12-19

    The objective of this research program is to address major nuclear fuels performance issues for the design and use of oxide-type fuels in the current and advanced nuclear reactor applications. Fuel performance is a major issue for extending fuel burn-up which has the added advantage of reducing the used fuel waste stream. It will also be a significant issue with respect to developing advanced fuel cycle processes where it may be possible to incorporate minor actinides in various fuel forms so that they can be 'burned' rather than join the used fuel waste stream. The potential to fission or transmute minor actinides and certain long-lived fission product isotopes would transform the high level waste storage strategy by removing the need to consider fuel storage on the millennium time scale.

  19. Electricity generation by living plants in a plant microbial fuel cell

    NARCIS (Netherlands)

    Timmers, R.A.

    2012-01-01

    Society is facing local and global challenges to secure needs of people. One of those needs is the increasing demand of energy. Currently most energy is generated by conversion of fossil fuels. The major drawback of using fossil fuels is pollution of the environment by emission of carbon dioxide, ni

  20. Electricity generation by living plants in a plant microbial fuel cell

    NARCIS (Netherlands)

    Timmers, R.A.

    2012-01-01

    Society is facing local and global challenges to secure needs of people. One of those needs is the increasing demand of energy. Currently most energy is generated by conversion of fossil fuels. The major drawback of using fossil fuels is pollution of the environment by emission of carbon dioxide,

  1. Diversity of fuel sources for electricity generation in an evolving U.S. power sector

    Science.gov (United States)

    DiLuccia, Janelle G.

    Policymakers increasingly have shown interest in options to boost the relative share of renewable or clean electricity generating sources in order to reduce negative environmental externalities from fossil fuels, guard against possible resource constraints, and capture economic advantages from developing new technologies and industries. Electric utilities and non-utility generators make decisions regarding their generation mix based on a number of different factors that may or may not align with societal goals. This paper examines the makeup of the electric power sector to determine how the type of generator and the presence (or lack) of competition in electricity markets at the state level may relate to the types of fuel sources used for generation. Using state-level electricity generation data from the U.S. Energy Information Administration from 1990 through 2010, this paper employs state and time fixed-effects regression modeling to attempt to isolate the impacts of state-level restructuring policies and the emergence of non-utility generators on states' generation from coal, from fossil fuel and from renewable sources. While the analysis has significant limitations, I do find that state-level electricity restructuring has a small but significant association with lowering electricity generation from coal specifically and fossil fuels more generally. Further research into the relationship between competition and fuel sources would aid policymakers considering legislative options to influence the generation mix.

  2. Fuels and cycles for the fourth generation systems; Combustibles et cycles pour les systemes de 4. generation

    Energy Technology Data Exchange (ETDEWEB)

    Brossard, Ph. [CEA Saclay, Dir. du Developpement et de l' Innovation Nucleares (DEN/DDIN), 91 - Gif Sur Yvette (France)

    2003-07-01

    This paper presented during the Cea Seminar on the nuclear systems of the future deals with the fuels and cycles of the fourth generation systems. Four goal areas have been defined for these systems: sustainable for the environment, economical for the investment, safety and reliability, a better protection against the proliferation. The different fuels and cycles of the research programs are detailed. (A.L.B.)

  3. GHG PSD Permit: Cheyenne Light, Fuel & Power / Black Hills Power, Inc. – Cheyenne Prairie Generating Station

    Science.gov (United States)

    This page contains the final PSD permit for the Cheyenne Light, Fuel & Power / Black Hills Power, Inc. Cheyenne Prairie Generating Station, located in Laramie, Wyoming, and operated by Black Hills Service Company.

  4. 300 W polymer electrolyte fuel cell generators for educational purposes

    Energy Technology Data Exchange (ETDEWEB)

    Tsukada, A.; Buechi, F.N.; Scherer, G.G.; Haas, O. [Paul Scherrer Inst. (PSI), Villigen (Switzerland); Popelis, I. [Fachhochschule Solothurn Nordwestschweiz (Switzerland)

    1999-08-01

    A 300 W fuel cell power pack has been developed for educational purposes in close collaboration with the Fachhochschule Solothurn Nordwestschweiz. The project was initiated and financed by the Swiss Federal Office of Energy. The outlay and the performance of the power pack are described. (author) 3 figs.

  5. Compost in plant microbial fuel cell for bioelectricity generation

    NARCIS (Netherlands)

    Moqsud, M.A.; Yoshitake, J.; Bushra, Q.S.; Hyodo, M.; Omine, K.; Strik, D.P.B.T.B.

    2015-01-01

    Recycling of organic waste is an important topic in developing countries as well as developed countries. Compost from organic waste has been used for soil conditioner. In this study, an experiment has been carried out to produce green energy (bioelectricity) by using paddy plant microbial fuel cells

  6. Evaluation of solid oxide fuel cell systems for electricity generation

    Science.gov (United States)

    Somers, E. V.; Vidt, E. J.; Grimble, R. E.

    1982-01-01

    Air blown (low BTU) gasification with atmospheric pressure Solid Electrolyte Fuel Cells (SOFC) and Rankine bottoming cycle, oxygen blown (medium BTU) gasification with atmospheric pressure SOFC and Rankine bottoming cycle, air blown gasification with pressurized SOFC and combined Brayton/Rankine bottoming cycle, oxygen blown gasification with pressurized SOFC and combined Brayton/Rankine bottoming cycle were evaluated.

  7. Generation of Solid Recovered Fuel from Sewage Sludge Compost

    Directory of Open Access Journals (Sweden)

    Irina Kliopova

    2013-01-01

    Full Text Available The paper presents results of the research which was carried out in KTU APINI when implementing one stage of the PF7 program project “Polygeneration of energy, fuels, and fertilizers from biomass residues and sewage sludge (ENERCOM” (No TREN/FP7/EN/218916. The research objective was to assess possibilities of producing solid recovered fuel (SRF from compost produced from pre-treated sewage sludge and biomass residuals in “Soil-Concept” plant (Luxemburg. Feasibility of producing pellets and briquettes using the composites of compost, sawdust, and peat was analyzed. Technical and environmental evaluations of SRF production were carried out on the basis of pelleting and briquetting tests. Main chemical and physical parameters of produced SRF were analyzed and compared to the recovered fuel classificatory (CEN/TC 343. All pellets and briquettes, produced during the experiment, were attributed to a certain class of recovered fuel. Results of technical and environmental evaluations of SRF production and their burning as well as conclusions and recommendations made are presented.DOI: http://dx.doi.org/10.5755/j01.erem.62.4.2742

  8. A polymer electrolyte fuel cell stack for stationary power generation from hydrogen fuel

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, M.S.; Moeller-Holst, S.; Webb, D.M.; Zawodzinski, C.; Gottesfeld, S. [Los Alamos National Lab., NM (United States). Materials Science and Technology Div.

    1998-08-01

    The objective is to develop and demonstrate a 4 kW, hydrogen-fueled polymer electrolyte fuel cell (PEFC) stack, based on non-machined stainless steel hardware and on membrane/electrode assemblies (MEAs) of low catalyst loadings. The stack is designed to operate at ambient pressure on the air-side and can accommodate operation at higher fuel pressures, if so required. This is to be accomplished by working jointly with a fuel cell stack manufacturer, based on a CRADA. The performance goals are 57% energy conversion efficiency hydrogen-to-electricity (DC) at a power density of 0.9 kW/liter for a stack operating at ambient inlet pressures. The cost goal is $600/kW, based on present materials costs.

  9. STEM-EDS analysis of fission products in neutron-irradiated TRISO fuel particles from AGR-1 experiment

    Energy Technology Data Exchange (ETDEWEB)

    Leng, B. [University of Wisconsin-Madison, Madison, WI 53706 (United States); Thorium Molten Salts Reactor Center, Shanghai Institute of Applied Physics, Shanghai, 201800 (China); Rooyen, I.J. van, E-mail: Isabella.vanrooyen@inl.gov [Fuel Design and Development Department, Idaho National Laboratory, Idaho Falls, ID 83415-6188 (United States); Wu, Y.Q. [Department of Materials Science and Engineering, Boise State University, Boise, ID 83725-2090 (United States); Center for Advanced Energy Studies, Idaho Falls, ID 83401 (United States); Szlufarska, I.; Sridharan, K. [University of Wisconsin-Madison, Madison, WI 53706 (United States)

    2016-07-15

    Historic and recent post-irradiation-examination from the German AVR and Advanced Gas Reactor Fuel Development and Qualification Project have shown that 110 m Ag is released from intact tristructural isotropic (TRISO) fuel. Although TRISO fuel particle research has been performed over the last few decades, little is known about how metallic fission products are transported through the SiC layer, and it was not until March 2013 that Ag was first identified in the SiC layer of a neutron-irradiated TRISO fuel particle. The existence of Pd- and Ag-rich grain boundary precipitates, triple junction precipitates, and Pd nano-sized intragranular precipitates in neutron-irradiated TRISO particle coatings was investigated using Scanning Transmission Electron Microscopy and Energy Dispersive Spectroscopy analysis to obtain more information on the chemical composition of the fission product precipitates. A U-rich fission product honeycomb shape precipitate network was found near a micron-sized precipitate in a SiC grain about ∼5 μm from the SiC-inner pyrolytic carbon interlayer, indicating a possible intragranular transport path for uranium. A single Ag-Pd nano-sized precipitate was found inside a SiC grain, and this is the first research showing such finding in irradiated SiC. This finding may possibly suggest a possible Pd-assisted intragranular transport mechanism for Ag and may be related to void or dislocation networks inside SiC grains. Preliminary semi-quantitative analysis indicated the micron-sized precipitates to be Pd{sub 2}Si{sub 2}U with carbon existing inside these precipitates. However, the results of such analysis for nano-sized precipitates may be influenced by the SiC matrix. The results reported in this paper confirm the co-existence of Cd with Ag in triple points reported previously. - Highlights: • First research data in neutron irradiated TRISO coated particles showing a Ag-Pd nano-sized precipitate inside a SiC grain. • Intragranular Ag Pd

  10. Irradiation testing of internally pressurized and/or graphite coated Zircaloy-4 clad fuel rods in the NRX Reactor (AWBA Development Program). [LWBR

    Energy Technology Data Exchange (ETDEWEB)

    Hoffman, R.C.; Sherman, J.

    1978-11-01

    Irradiation tests on 0.612 inch O.D. by 117-inch long Zircaloy-4 clad fuel rods were performed to assess the effects on fuel rod performance of (1) internal helium pre-pressurization to 500 psi as fabricated, (2) the presence of a graphite barrier coating on the inside cladding surface, and (3) combined pre-pressurization and graphite coating. Periodic dimensional examinations were performed on the test rods, and the results were compared with data obtained from two previously irradiated test rods--both unpressurized and uncoated and one intentionally defected. These comparisons indicate that both pre-pressurization and graphite coating can substantially improve fuel element performance capability.

  11. Effects of fuel particle size and fission-fragment-enhanced irradiation creep on the in-pile behavior in CERCER composite pellets

    Science.gov (United States)

    Zhao, Yunmei; Ding, Shurong; Zhang, Xunchao; Wang, Canglong; Yang, Lei

    2016-12-01

    The micro-scale finite element models for CERCER pellets with different-sized fuel particles are developed. With consideration of a grain-scale mechanistic irradiation swelling model in the fuel particles and the irradiation creep in the matrix, numerical simulations are performed to explore the effects of the particle size and the fission-fragment-enhanced irradiation creep on the thermo-mechanical behavior of CERCER pellets. The enhanced irradiation creep effect is applied in the 10 μm-thick fission fragment damage matrix layer surrounding the fuel particles. The obtained results indicate that (1) lower maximum temperature occurs in the cases with smaller-sized particles, and the effects of particle size on the mechanical behavior in pellets are intricate; (2) the first principal stress and radial axial stress remain compressive in the fission fragment damage layer at higher burnup, thus the mechanism of radial cracking found in the experiment can be better explained.

  12. Gamma Irradiation of 4th Instar Larva of Angoumois Grain Moth and Effects on Parent and Their Generations

    OpenAIRE

    Boshra, Salwa A. [سلوى عزمي بشرى

    2006-01-01

    Late fourth stage larvae of Angomous grain moth, Sitotroga cerealella (Olivier) were gamma irradiated with doses 0 ( control), 25, 50 75, 100, 125 and 150 Gy. The moths originated from larvae irradiated with 150 Gy became sterile. Irradiation of males as larvae with substerilizing doses of 25 and 50 Gy induced inherited F| sterility which reduced the population. F| progeny exhibited more sterility than their parent generation. Also F| males inherited more sterility than F| females. Adult fert...

  13. Major design issues of molten carbonate fuel cell power generation unit

    Energy Technology Data Exchange (ETDEWEB)

    Chen, T.P.

    1996-04-01

    In addition to the stack, a fuel cell power generation unit requires fuel desulfurization and reforming, fuel and oxidant preheating, process heat removal, waste heat recovery, steam generation, oxidant supply, power conditioning, water supply and treatment, purge gas supply, instrument air supply, and system control. These support facilities add considerable cost and system complexity. Bechtel, as a system integrator of M-C Power`s molten carbonate fuel cell development team, has spent substantial effort to simplify and minimize these supporting facilities to meet cost and reliability goals for commercialization. Similiar to other fuels cells, MCFC faces design challenge of how to comply with codes and standards, achieve high efficiency and part load performance, and meanwhile minimize utility requirements, weight, plot area, and cost. However, MCFC has several unique design issues due to its high operating temperature, use of molten electrolyte, and the requirement of CO2 recycle.

  14. Electricity generation by microbial fuel cells fuelled with wheat straw hydrolysate

    DEFF Research Database (Denmark)

    Thygesen, Anders; Poulsen, Finn Willy; Angelidaki, Irini;

    2011-01-01

    Electricity production from microbial fuel cells fueled with hydrolysate produced by hydrothermal treatment of wheat straw can achieve both energy production and domestic wastewater purification. The hydrolysate contained mainly xylan, carboxylic acids, and phenolic compounds. Power generation...... density with the hydrolysate was higher than the one with only xylan (120 mW m−2) and carboxylic acids as fuel. The higher power density can be caused by the presence of phenolic compounds in the hydrolysates, which could mediate electron transport. Electricity generation with the hydrolysate resulted...

  15. Measurement of Ballooning Gap Size of Irradiated Fuels Using Neutron Radiography Transfer Method and HV Image Filter

    Energy Technology Data Exchange (ETDEWEB)

    Sim, Cheul Muu; Kim, Tae Joo; Oh, Hwa Suk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Joon Cheol [Seonam University, Namwon (Korea, Republic of)

    2013-04-15

    A transfer method of neutron radiography was developed to measure the size of the end plug and a gap of an intact K102L-2, the irradiated fuel of a ballooned K174L-3, a ballooned and ruptured K98L-3. A typical irradiation time of 25 min. was determined to obtain a film density of between 2 and 3 of SR X-ray film with neutrons of 1.5x10{sup 11}n{center_dot}cm{sup -2}. To validate and calibrate the results, a RISO fuel standard sample, Cd plate and ASTM-BPI/SI were used. An activated latent image formed in the 100 {mu}m Dy foil was subsequently transferred in a dark room for more than 8 hours to the SR film which is a maximum of three half-lives. Due to the L/D ratio an unsharpness of 9.82-14{mu}m and a magnification of 1.0003 were given. After digitizing an image of SR film, the ballooning gap of the plug was discernible by an H/V filter of image processing. The gap size of the ballooned element, K174L-3, is equal to or greater than 1.2 mm. The development of a transfer method played a pivotal role in developing high burn-up of Wolsung and PWR nuclear fuel type.

  16. Measured effect of wind generation on the fuel consumption of an isolated diesel power system

    Science.gov (United States)

    Stiller, P. H.; Scott, G. W.; Shaltens, R. K.

    1983-01-01

    The Block Island Power Company (BIPCO), on Block Island, Rhode Island, operates an isolated electric power system consisting of diesel generation and an experimental wind turbine. The 150-kW wind turbine, designated MOD-OA by the U.S. Department of Energy is typically operated in parallel with two diesel generators to serve an average winter load of 350 kW. Wind generation serves up to 60 percent of the system demand depending on wind speed and total system load. Results of diesel fuel consumption measurements are given for the diesel units operated in parallel with the wind turbine and again without the wind turbine. The fuel consumption data are used to calculate the amount of fuel displaced by wind energy. Results indicate that the wind turbine displaced 25,700 lbs. of the diesel fuel during the test period, representing a calculated reduction in fuel consumption of 6.7 percent while generating 11 percent of the total electric energy. The amount of displaced fuel depends on operating conditions and system load. It is also shown that diesel engine throttle activity resulting from wind gusts which rapidly change the wind turbine output do not significantly influence fuel consumption.

  17. Initial performance of the PIGMI prototype. [Pion Generator for Medical Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Stovall, J.E.

    1979-01-01

    The PIGMI (Pion Generator for Medical Irradiations) program at LASL is an accelerator development program aimed at completing the design of an accelerator suitable for use as a pion generator in a hospital-based radiotherapy program. The major thrust of the program has been the design of a 7 MeV prototype accelerator which emphasizes compactness, economy of construction, and operation and reliability. To achieve these goals the design of the prototype has exploited a number of innovations in proton linac technology. An overview of the program discussing the major innovative features of the prototype is presented. The initial operating experience is discussed and initial performance measurements are presented.

  18. Hot deuteron generation and neutron production in deuterated nanowire array irradiated at relativistic intensity

    Science.gov (United States)

    Curtis, Alden; Calvi, Chase; Tinsley, Jim; Hollinger, Reed; Wang, Shoujun; Rockwood, Alex; Wang, Yong; Buss, Conrad; Shlyaptsev, Vyacheslav; Kaymak, V.; Pukhov, Alexander; Rocca, Jorge

    2016-10-01

    Irradiation of arrays of aligned high aspect ratio nanowires with high contrast femtosecond laser pulses of relativistic intensity was recently shown to volumetrically heat near solid density plasmas to multi-KeV energy. Using aligned arrays of deuterated polyethylene nanowires (CD2) irradiated at laser intensities of up to 1 ×1020 W/cm2 we are able to generate near solid density plasmas in which the tail of the deuteron distribution was measured to reach energies of up to 3 MeV, in agreement with particle-in-cell simulations. Comparative measurements conducted using flat CD2 targets irradiated by the same laser pulses show the maximum deuteron energies are sub-MeV. We also observed a 100x increase in the number of neutrons produced as compared to flat CD2 targets irradiated at the same conditions, with the highest yield shots producing above 106 neutrons per Joule of laser energy. Work supported by AFOSR Award FA9560-14-10232 and NSTec SDRD program.

  19. A performance analysis of integrated solid oxide fuel cell and heat recovery steam generator for IGFC system

    DEFF Research Database (Denmark)

    Rudra, Souman; Lee, Jinwook; Rosendahl, Lasse

    2010-01-01

    Solid oxide fuel cell (SOFC) is a promising technology for electricity generation. Sulfur-free syngas from a gas-cleaning unit serves as fuel for SOFC in integrated gasification fuel cell (IGFC) power plants. It converts the chemical energy of fuel gas directly into electric energy, thus high...

  20. Advances in the generation of a new emulsified fuel

    Energy Technology Data Exchange (ETDEWEB)

    Chavez, A. [Technical Consultancy, Energy Plus UC, Huitzilac, Morelos (Mexico); Ramirez, M. [Instituto Mexicano del Petroleo, Programa de Aseguramiento de Hidrocarburos, Mexico, D.F. (Mexico); Medina, E. [Universidad Nacional Autonoma de Mexico, Departamento de Termofluidos, Facultad de Ingenieria, Mexico, D.F. (Mexico); Bolado, R.; Mora, J. [Instituto Mexicano del Petroleo, Laboratorio de Combustion, Veracruz (Mexico)

    2011-08-15

    The development of a new emulsified fuel is described, from the conceptual idea to the semi-industrial tests of the final product. The starting point was the necessity to lower the particulate matter (PM) emissions produced by the combustion of more than 200 MBD of heavy fuel oil (HFO) used for electric power conversion. The major component of HFO is a vacuum residue of the oil refining process mixed with light cycle oils to make it pumpable. An alternative to handle and burn the high viscosity residue (solid at room temperature) is by converting it in an oil-in-water emulsion. The best emulsions resulted of 70% residue in 30% water, Sauter Mean Diameter of 10-20 {mu}m and a stability of more than 90 days. Spray burning tests of the emulsion against HFO in a semi-industrial 500 kW furnace showed a reduction in PM emissions of 24-36%. (orig.)

  1. Substrates and pathway of electricity generation in a nitrification-based microbial fuel cell.

    Science.gov (United States)

    Chen, Hui; Zheng, Ping; Zhang, Jiqiang; Xie, Zuofu; Ji, Junyuan; Ghulam, Abbas

    2014-06-01

    Nitrification-based microbial fuel cell (N-MFC) is a novel inorganic microbial fuel cell based on nitrification in the anode compartment. So far, little information is available on the substrates and pathway of N-MFC. The results of this study indicated that apart from the primary nitrification substrate (ammonium), the intermediates (hydroxylamine and nitrite) could also serve as anodic fuel to generate current, and the end product nitrate showed an inhibitory effect on electricity generation. Based on the research, a pathway of electricity generation was proposed for N-MFC: ammonium was oxidized first to nitrite by ammonia-oxidizing bacteria (AOB), then the nitrite in anolyte and the potassium permanganate in catholyte constituted a chemical cell to generate current. In other words, the electricity generation in N-MFC was not only supported by microbial reaction as we expected, but both biological and electrochemical reactions contributed.

  2. Carbon as a fuel for efficient electricity generation in carbon solid oxide fuel cells

    Directory of Open Access Journals (Sweden)

    Skrzypkiewicz Marek

    2016-01-01

    Full Text Available In this paper, the impact of the physicochemical properties of carbonaceous solid fuels on the performance of a direct carbon solid oxide fuel cell (DC-SOFC was investigated. High-purity synthetic carbon powders such as carbon black N-220 and Carbo Medicinalis FP5 were chosen for analytical and electrochemical investigations in a DC-SOFC. The research focussed on choosing an optimised, cost-effective, high-purity carbon powder which could be applied as a solid reference fuel for all tests performed on a single DC-SOFC cell as well as on DC-SOFC stack constructions. Most of the electrochemical investigations described in this paper were performed using square DCSOFCs with dimensions of 5 × 5 cm. The relationship between structure, physicochemical properties, and electrochemical reactivity in a DC-SOFC was analysed.

  3. Optimization of degradation of Reactive Black 5 (RB5) and electricity generation in solar photocatalytic fuel cell system.

    Science.gov (United States)

    Khalik, Wan Fadhilah; Ho, Li-Ngee; Ong, Soon-An; Voon, Chun-Hong; Wong, Yee-Shian; Yusoff, NikAthirah; Lee, Sin-Li; Yusuf, Sara Yasina

    2017-10-01

    The photocatalytic fuel cell (PFC) system was developed in order to study the effect of several operating parameters in degradation of Reactive Black 5 (RB5) and its electricity generation. Light irradiation, initial dye concentration, aeration, pH and cathode electrode are the operating parameters that might give contribution in the efficiency of PFC system. The degradation of RB5 depends on the presence of light irradiation and solar light gives better performance to degrade the azo dye. The azo dye with low initial concentration decolorizes faster compared to higher initial concentration and presence of aeration in PFC system would enhance its performance. Reactive Black 5 rapidly decreased at higher pH due to the higher amount of OH generated at higher pH and Pt-loaded carbon (Pt/C) was more suitable to be used as cathode in PFC system compared to Cu foil and Fe foil. The rapid decolorization of RB5 would increase their voltage output and in addition, it would also increase their Voc, Jsc and Pmax. The breakage of azo bond and aromatic rings was confirmed through UV-Vis spectrum and COD analysis. Copyright © 2017 Elsevier Ltd. All rights reserved.

  4. Procedure for determining maximum sustainable power generated by microbial fuel cells.

    Science.gov (United States)

    Menicucci, Joseph; Beyenal, Haluk; Marsili, Enrico; Veluchamy, Raajaraajan Angathevar; Demir, Goksel; Lewandowski, Zbigniew

    2006-02-01

    Power generated by microbial fuel cells is computed as a product of current passing through an external resistor and voltage drop across this resistor. If the applied resistance is very low, then high instantaneous power generated by the cell is measured, which is not sustainable; the cell cannot deliver that much power for long periods of time. Since using small electrical resistors leads to erroneous assessment of the capabilities of microbial fuel cells, a question arises: what resistor should be used in such measurements? To address this question, we have defined the sustainable power as the steady state of power delivery by a microbial fuel cell under a given set of conditions and the maximum sustainable power as the highest sustainable power that a microbial fuel cell can deliver under a given set of conditions. Selecting the external resistance that is associated with the maximum sustainable power in a microbial fuel cell (MFC) is difficult because the operator has limited influence on the main factors that control power generation: the rate of charge transfer at the current-limiting electrode and the potential established across the fuel cell. The internal electrical resistance of microbial fuel cells varies, and it depends on the operational conditions of the fuel cell. We have designed an empirical procedure to predict the maximum sustainable power that can be generated by a microbial fuel cell operated under a given set of conditions. Following the procedure, we change the external resistors incrementally, in steps of 500 omega every 10, 60, or 180 s and measure the anode potential, the cathode potential, and the cell current. Power generated in the microbial fuel cell that we were using was limited by the anodic current. The anodic potential was used to determine the condition where the maximum sustainable power is obtained. The procedure is simple, microbial fuel cells can be characterized within an hour, and the results of the measurements can serve

  5. Infrared nanosecond pulsed laser irradiation of stainless steel: micro iron-oxide zones generation.

    Science.gov (United States)

    Ortiz-Morales, M; Frausto-Reyes, C; Soto-Bernal, J J; Acosta-Ortiz, S E; Gonzalez-Mota, R; Rosales-Candelas, I

    2014-07-15

    Nanosecond-pulsed, infrared (1064 nm) laser irradiation was used to create periodic metal oxide coatings on the surface of two samples of commercial stainless steel at ambient conditions. A pattern of four different metal oxide zones was created using a galvanometer scanning head and a focused laser beam over each sample. This pattern is related to traverse direction of the laser beam scanning. Energy-dispersive X-ray spectroscopy (EDS) was used to find the elemental composition and Raman spectroscopy to characterize each oxide zone. Pulsed laser irradiation modified the composition of the stainless steel samples, affecting the concentration of the main components within each heat affected zone. The Raman spectra of the generated oxides have different intensity profiles, which suggest different oxide phases such as magnetite and maghemite. In addition, these oxides are not sensible to the laser power of the Raman system, as are the iron oxide powders reported in the literature. These experiments show that it is possible to generate periodic patterns of various iron oxide zones by laser irradiation, of stainless steel at ambient conditions, and that Raman spectroscopy is a useful punctual technique for the analysis and inspection of small oxide areas.

  6. Scavenging of hydroxyl radicals generated in human plasma following X-ray irradiation.

    Science.gov (United States)

    Hosokawa, Yoichiro; Sano, Tomoaki

    2015-11-01

    There are various antioxidant materials that scavenge free radicals in human plasma. It is possible that the radical-scavenging function causes a radiation protective effect in humans. This study estimated the hydroxyl (OH) radical-scavenging activity induced by X-ray irradiation in human plasma. The test subjects included 111 volunteers (75 males and 36 females) ranging from 22 to 35 years old (average, 24.0). OH radicals generated in irradiated human plasma were measured by electron spin resonance (ESR). The relationships between the amount of the OH radical and chemical and biological parameters [total protein, total cholesterol, triglycerides and hepatitis B surface (HBs) antibodies] were estimated in the plasma of the 111 volunteers by a multivariate analysis. The presence of HBs antibodies had the greatest influence on OH radical-scavenging activity. One volunteer who did not have the HBs antibody was given an inoculation of the hepatitis B vaccine. There was a remarkable decrease in the amount of OH radical generated from plasma after the HBs antibody was produced. The results indicate that the HBs antibody is an important factor for the scavenging of OH radicals initiated by X-ray irradiation in the human body.

  7. Three-component U-Pu-Th fuel for plutonium irradiation in heavy water reactors

    Directory of Open Access Journals (Sweden)

    Peel Ross

    2016-01-01

    Full Text Available This paper discusses concepts for three-component fuel bundles containing plutonium, uranium and thorium for use in pressurised heavy water reactors, and cases for and against implementation of such a nuclear energy system in the United Kingdom. Heavy water reactors are used extensively in Canada, and are deploying within India and China, whilst the UK is considering the use of heavy water reactors to manage its plutonium inventory of 140 tonnes. The UK heavy water reactor proposal uses a mixed oxide (MOX fuel of plutonium in depleted uranium, within the enhanced CANDU-6 (EC-6 reactor. This work proposes an alternative heterogeneous fuel concept based on the same reactor and CANFLEX fuel bundle, with eight large-diameter fuel elements loaded with natural thorium oxide and 35 small-diameter fuel elements loaded with a MOX of plutonium and reprocessed uranium stocks from UK MAGNOX and AGR reactors. Indicative neutronic calculations suggest that such a fuel would be neutronically feasible. A similar MOX may alternatively be fabricated from reprocessed <5% enriched light water reactor fuel, such as the fuel of the AREVA EPR reactor, to consume newly produced plutonium from reprocessing, similar to the DUPIC (direct use of PWR fuel in CANDU process.

  8. Inert matrix fuel performance during the first two irradiation cycles in a test reactor: comparison with modelling results

    Science.gov (United States)

    Hellwig, Ch.; Kasemeyer, U.

    2003-06-01

    In the inert matrix fuel (IMF) type investigated at Paul Scherrer Institut, plutonium is dissolved in the yttrium stabilised zirconium oxide (YSZ), a highly radiation resistant cubic phase with additions of erbium as burnable poison for reactivity control. A first irradiation experiment of YSZ based IMF is ongoing in the OECD Material Test Reactor in Halden together with mixed oxide fuel. The results of the first two cycles for IMF to a burnup of some 105 kW d cm -3 are presented and the modelling results in comparison with the experimental results are shown. A first approximation for a simple swelling model for this YSZ based IMF can be given. Possible fission gas release mechanisms are briefly discussed. The implications of the modelling results are discussed.

  9. Corrigendum to "Coupled thermochemical, isotopic evolution and heat transfer simulations in highly irradiated UO2 nuclear fuel"

    Science.gov (United States)

    Piro, M. H. A.; Banfield, J.; Clarno, K.; Simunovic, S.; Besmann, T. M.; Lewis, B. J.; Thompson, W. T.

    2016-09-01

    Figs. 7-9 in "Coupled thermochemical, isotopic evolution and heat transfer simulations in highly irradiated UO2 nuclear fuel" [1] have a consistent error corresponding to the relative proportions of iodine. Reported concentrations of iodine in the original manuscript are approximately ten times higher than expected, and are comparable in atomic proportions to cesium. One would expect that the amount of cesium would be about one order of magnitude greater than iodine based on the difference in fission yields of 235U and 239Pu. A practical consequence of this error would affect the predicted quantity and chemical composition of iodine on the fuel surface, which is related to iodine-induced stress corrosion cracking [2].

  10. Effect of thermal friction on the generation and transport of interstitial defects in irradiated metals

    CERN Document Server

    Dudarev, S L

    2002-01-01

    Generation of interstitial and vacancy defects under 14.1 MeV neutron irradiation is expected to drive the evolution of microstructure of materials in a future fusion power station. We investigate effects of thermal friction associated with the interaction between mobile clusters of interstitial atoms produced in collision cascades and phonon excitations. Phonons give rise to the random Brownian motion of clusters in the crystal lattice. Phonon excitations are also responsible for the dissipation of energy of rapidly moving clusters formed at the periphery of collision cascades. We investigate how the coefficient of thermal friction depends on the structure of clusters. We also discuss implications of our findings for understanding the origin of higher resistance of bcc metals to irradiation and the connection between this phenomenon and the long-range effect observed in experiments on ion implantation.

  11. 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.

  12. Biomimetic and microbial approaches to solar fuel generation.

    Science.gov (United States)

    Magnuson, Ann; Anderlund, Magnus; Johansson, Olof; Lindblad, Peter; Lomoth, Reiner; Polivka, Tomas; Ott, Sascha; Stensjö, Karin; Styring, Stenbjörn; Sundström, Villy; Hammarström, Leif

    2009-12-21

    Photosynthesis is performed by a multitude of organisms, but in nearly all cases, it is variations on a common theme: absorption of light followed by energy transfer to a reaction center where charge separation takes place. This initial form of chemical energy is stabilized by the biosynthesis of carbohydrates. To produce these energy-rich products, a substrate is needed that feeds in reductive equivalents. When photosynthetic microorganisms learned to use water as a substrate some 2 billion years ago, a fundamental barrier against unlimited use of solar energy was overcome. The possibility of solar energy use has inspired researchers to construct artificial photosynthetic systems that show analogy to parts of the intricate molecular machinery of photosynthesis. Recent years have seen a reorientation of efforts toward creating integrated light-to-fuel systems that can use solar energy for direct synthesis of energy-rich compounds, so-called solar fuels. Sustainable production of solar fuels is a long awaited development that promises extensive solar energy use combined with long-term storage. The stoichiometry of water splitting into molecular oxygen, protons, and electrons is deceptively simple; achieving it by chemical catalysis has proven remarkably difficult. The reaction center Photosystem II couples light-induced charge separation to an efficient molecular water-splitting catalyst, a Mn(4)Ca complex, and is thus an important template for biomimetic chemistry. In our aims to design biomimetic manganese complexes for light-driven water oxidation, we link photosensitizers and charge-separation motifs to potential catalysts in supramolecular assemblies. In photosynthesis, production of carbohydrates demands the delivery of multiple reducing equivalents to CO(2). In contrast, the two-electron reduction of protons to molecular hydrogen is much less demanding. Virtually all microorganisms have enzymes called hydrogenases that convert protons to hydrogen, many of

  13. Pollutants generated by the combustion of solid biomass fuels

    CERN Document Server

    Jones, Jenny M; Ma, Lin; Williams, Alan; Pourkashanian, Mohamed

    2014-01-01

    This book considers the pollutants formed by the combustion of solid biomass fuels. The availability and potential use of solid biofuels is first discussed because this is the key to the development of biomass as a source of energy.This is followed by details of the methods used for characterisation of biomass and their classification.The various steps in the combustion mechanisms are given together with a compilation of the kinetic data. The chemical mechanisms for the formation of the pollutants: NOx, smoke and unburned hydrocarbons, SOx, Cl compounds, and particulate metal aerosols

  14. Electron probe microanalysis of a METAPHIX UPuZr metallic alloy fuel irradiated to 7.0 at.% burn-up

    Science.gov (United States)

    Brémier, S.; Inagaki, K.; Capriotti, L.; Poeml, P.; Ogata, T.; Ohta, H.; Rondinella, V. V.

    2016-11-01

    The METAPHIX project is a collaboration between CRIEPI and JRC-ITU investigating safety and performance of a closed fuel cycle option based on fast reactor metal alloy fuels containing Minor Actinides (MA). The aim of the project is to investigate the behaviour of this type of fuel and demonstrate the transmutation of MA under irradiation. A UPuZr metallic fuel sample irradiated to a burn-up of 7 at.% was examined by electron probe microanalysis. The fuel sample was extensively characterised qualitatively and quantitatively using elemental X-ray imaging and point analysis techniques. The analyses reveal a significant redistribution of the fuel components along the fuel radius highlighting a nearly complete depletion of Zr in the central part of the fuel. Numerous rare earth and fission products secondary phases are present in various compositions. Fuel cladding chemical interaction was observed with creation of a number of intermediary layers affecting a cladding depth of 15-20 μm and migration of cladding elements to the fuel.

  15. DESIGNING AN OPPORTUNITY FUEL WITH BIOMASS AND TIRE-DERIVED FUEL FOR COFIRING AT WILLOW ISLAND GENERATING STATION AND COFIRING SAWDUST WITH COAL AT ALBRIGHT GENERATING STATION

    Energy Technology Data Exchange (ETDEWEB)

    K. Payette; D. Tillman

    2003-07-01

    During the period April 1, 2003--June 30, 2003, Allegheny Energy Supply Co., LLC (Allegheny) proceeded with demonstration operations at the Willow Island Generating Station and improvements to the Albright Generating Station cofiring systems. The demonstration operations at Willow Island were designed to document integration of biomass cofiring into commercial operations. The Albright improvements were designed to increase the resource base for the projects, and to address issues that came up during the first year of operations. This report summarizes the activities associated with the Designer Opportunity Fuel program, and demonstrations at Willow Island and Albright Generating Stations.

  16. Reviews on Fuel Cell Technology for Valuable Chemicals and Energy Co-Generation

    Directory of Open Access Journals (Sweden)

    Wisitsree Wiyaratn

    2010-07-01

    Full Text Available This paper provides a review of co-generation process in fuel cell type reactor to produce valuable chemical compounds along with electricity. The chemicals and energy co-generation processes have been shown to be a promising alternative to conventional reactors and conventional fuel cells with pure water as a byproduct. This paper reviews researches on chemicals and energy co-generation technologies of three types of promising fuel cell i.e. solid oxide fuel cell (SOFC, alkaline fuel cell (AFC, and proton exchange membrane fuel cell (PEMFC. In addition, the research studies on applications of SOFCs, AFCs, and PEMFCs with chemical production (i.e. nitric oxide, formaldehyde, sulfur oxide, C2 hydrocarbons, alcohols, syngas and hydrogen peroxide were also given. Although, it appears that chemicals and energy co-generation processes have potential to succeed in commercial applications, the development of cheaper catalyst materials with longer stability ,and understanding in thermodynamic are still challenging to improve the overall system performance and enable to use in commercial market.

  17. Nuclear power generation and fuel cycle report 1997

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-09-01

    Nuclear power is an important source of electric energy and the amount of nuclear-generated electricity continued to grow as the performance of nuclear power plants improved. In 1996, nuclear power plants supplied 23 percent of the electricity production for countries with nuclear units, and 17 percent of the total electricity generated worldwide. However, the likelihood of nuclear power assuming a much larger role or even retaining its current share of electricity generation production is uncertain. The industry faces a complex set of issues including economic competitiveness, social acceptance, and the handling of nuclear waste, all of which contribute to the uncertain future of nuclear power. Nevertheless, for some countries the installed nuclear generating capacity is projected to continue to grow. Insufficient indigenous energy resources and concerns over energy independence make nuclear electric generation a viable option, especially for the countries of the Far East.

  18. Development of practical stirling engine for co-generation system using woody biomass fuels

    Energy Technology Data Exchange (ETDEWEB)

    Hoshi, Akira; Sasaki, Seizi [Ichinoseki National Coll. of Tech., Iwate (Japan); Tezuka, Nobutoshi [Stirling Engine Co., Ltd., Kawasaki-City (Japan); Fujimoto, Isao [Kansai Electric Power Co., Inc., Hyogo (Japan); Yamada, Noboru [Nagaoka Univ. of Technology (Japan)

    2008-07-01

    In recent years, fossil fuels such as petroleum, coal, and natural gas have become limited resources. In addition, global warming due to carbon dioxide (CO{sub 2}) emission has become a serious environmental issue. Since current living and economical standards depend strongly on fossil energy sources, it is necessary to realize a new society that utilizes biomass as a source of energy. With this background, in 2005, we manufactured a practical Stirling engine using biomass fuels. And we proposed a unique co-generation system using a practical Stirling engine that utilizes woody biomass fuel such as sawdust, firewood, and wood pellets. A burner uses the woody biomass fuel to heat the air in the expansion room to about 650 C and a water cooling system cools the air in the compression room to about 40 C. Under these operating conditions, the new engine generated about 3kW of electricity. (orig.)

  19. Energy generation from biomass with the aid of fuel cells; Energetische Nutzung von Biomasse mit Brennstoffzellenverfahren

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-07-01

    To provide an opportunity for information exchange at the interface between biomass use for energy generation and developers of fuel cells, the workshop 'Energy generation from biomass with the aid of fuel cells' was held by the Fachagentur Nachwachsende Rohstoffe on 9 and 10 December 1998. The lectures and discussions permit to assess better the opportunities and restraints resulting from the use of biogenous fuel gas in fuel cells. (orig.) [German] Um an der Schnittstelle zwischen der energetischen Nutzung von Biomasse und den Entwicklern von Brennstoffzellen einen Informationsaustausch zu ermoeglichen, wurde am 9. und 10. Dezember 1998 der Workshop 'Energetische Nutzung von Biomasse mit Brennstoffzellenverfahren' von der FNR veranstaltet. Die Vortraege und die Diskussion erlauben eine bessere Einschaetzung der Moeglichkeiten und Restriktionen, die sich bei dem Einsatz von biogenen Brenngasen in Brennstoffzellen ergeben. (orig.)

  20. Dynamic Analysis of Nuclear Waste Generation Based on Nuclear Fuel Cycle Transition Scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, S. R. [University of Science and Technology, Daejeon (Korea, Republic of); Ko, W. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    According to the recommendations submitted by the Public Engagement Commission on Spent Nuclear Fuel Management (PECOS), the government was advised to pick the site for an underground laboratory and interim storage facilities before the end of 2020 followed by the related research for permanent and underground disposal of spent fuel after 10 years. In the middle of the main issues, the factors of environmentally friendly and safe way to handle nuclear waste are inextricable from nuclear power generating nation to ensure the sustainability of nuclear power. For this purposes, the closed nuclear fuel cycle has been developed regarding deep geological disposal, pyroprocessing, and burner type sodium-cooled fast reactors (SFRs) in Korea. Among two methods of an equilibrium model and a dynamic model generally used for screening nuclear fuel cycle system, the dynamic model is more appropriate to envisage country-specific environment with the transition phase in the long term and significant to estimate meaningful impacts based on the timedependent behavior of harmful wastes. This study aims at analyzing the spent nuclear fuel generation based on the long-term nuclear fuel cycle transition scenarios considered at up-to-date country specific conditions and comparing long term advantages of the developed nuclear fuel cycle option between once-through cycle and Pyro-SFR cycle. In this study, a dynamic analysis was carried out to estimate the long-term projection of nuclear electricity generation, installed capacity, spent nuclear fuel arising in different fuel cycle scenarios based on the up-to-date national energy plans.

  1. 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.

  2. The reprocessing of irradiated fuels by halides and their compounds; Le traitement des combustibles irradies par les halogenes et leurs composes

    Energy Technology Data Exchange (ETDEWEB)

    Bourgeois, M.; Faugeras, P. [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1964-07-01

    A brief description is given of the experiments leading to the choice of the process volatilization of fluorides by gas phase attack. The chemical process is described for certain current types of clad Fuels: the aluminium or the zirconium cladding is first volatilized as chloride by attack with gaseous hydrogen chloride. The uranium is then transformed into volatile hexafluoride by attack with fluorine. These reactions are carried out consecutively in the same reactor in the presence of a fluidized bed of alumina which facilitates heat exchange. The experiments have been carried out in quantities from 100 gms to several kilograms of fuel, first without activity, and then with tracers. A description is given of the laboratory research which was carried out simultaneously on the separation of uranium and plutonium fluorides. Finally, an apparatus is described which is intended to test the process on irradiated fuel at an activity level of several thousands of curies of fission products. (authors) [French] On rappelle brievement les experimentations qui nous ont permis de decider du procede adopte volatilisation des fluorures par attaque en phase gazeuse. On decrit le processus chimique pour certains types courants de combustibles Gaines: dans un premier stade, l'aluminium ou le zirconium est volatilise sous forme de chlorure par action de l'acide chlorhydrique. Ensuite, l'uranium est transforme en hexafluorure volatil par action du fluor. Ces operations se font successivement dans un meme reacteur, en presence d'un lit fluidise d'alumine qui a pour but de faciliter les echanges thermiques. L'experimentation a ete conduite sur des quantites allant de 100 g a plusieurs kg de combustibles, en inactif, puis avec des traceurs. On decrit les etudes de laboratoire menees parallelement sur la separation des fluorures d'uranium et de plutonium. Enfin, on decrit une installation en construction destinee a experimenter le procede sur

  3. Spatial and Temporal Homogeneity of Solar Surface Irradiance across Satellite Generations

    Directory of Open Access Journals (Sweden)

    Rebekka Posselt

    2011-05-01

    Full Text Available Solar surface irradiance (SIS is an essential variable in the radiation budget of the Earth. Climate data records (CDR’s of SIS are required for climate monitoring, for climate model evaluation and for solar energy applications. A 23 year long (1983–2005 continuous and validated SIS CDR based on the visible channel (0.45–1 μm of the MVIRI instruments onboard the first generation of Meteosat satellites has recently been generated using a climate version of the well established Heliosat method. This version of the Heliosat method includes a newly developed self-calibration algorithm and an improved algorithm to determine the clear sky reflection. The climate Heliosat version is also applied to the visible narrow-band channels of SEVIRI onboard the Meteosat Second Generation Satellites (2004–present. The respective channels are observing the Earth in the wavelength region at about 0.6 μm and 0.8 μm. SIS values of the overlapping time period are used to analyse whether a homogeneous extension of the MVIRI CDR is possible with the SEVIRI narrowband channels. It is demonstrated that the spectral differences between the used visible channels leads to significant differences in the solar surface irradiance in specific regions. Especially, over vegetated areas the reflectance exhibits a high spectral dependency resulting in large differences in the retrieved SIS. The applied self-calibration method alone is not able to compensate the spectral differences of the channels. Furthermore, the extended range of the input values (satellite counts enhances the cloud detection of the SEVIRI instruments resulting in lower values for SIS, on average. Our findings have implications for the application of the Heliosat method to data from other geostationary satellites (e.g., GOES, GMS. They demonstrate the need for a careful analysis of the effect of spectral and technological differences in visible channels on the retrieved solar irradiance.

  4. Using strong sustainability to optimize electricity generation fuel mixes

    Energy Technology Data Exchange (ETDEWEB)

    Bishop, Justin D.K.; Amaratunga, Gehan A.J.; Rodriguez, Cuauhtemoc [University of Cambridge Department of Engineering, 9 JJ Thomson Avenue, Cambridge CB3 0FA, England (United Kingdom)

    2008-03-15

    This work represents a contribution to the field of sustainable electricity system design by using an optimization tool to specify the final mix composition, subject to the constraints of: emissions that are within the biocapacity of the region; a diverse and robust electricity supply system; and supply that at least meets current demand. The 25-country European Union (EU-25) is used as a case study. All the goals, save diversity, can be met by re-structuring the current fuel mix, thus maintaining current consumption levels. The diversity target is only met when consumption is reduced by 10-15% and the constraint on maximum material throughput is relaxed. Re-structuring the mix and reducing consumption is insufficient to achieve a sustainable EU carbon footprint. However, the solution proposed singlehandedly allows the EU to meet its Kyoto emissions target as well as its 2007 policy of a reduction of 20% in greenhouse gas emissions by 2020. (author)

  5. In situ Gas Conditioning in Fuel Reforming for Hydrogen Generation

    Energy Technology Data Exchange (ETDEWEB)

    Bandi, A.; Specht, M.; Sichler, P.; Nicoloso, N.

    2002-09-20

    The production of hydrogen for fuel cell applications requires cost and energy efficient technologies. The Absorption Enhanced Reforming (AER), developed at ZSW with industrial partners, is aimed to simplify the process by using a high temperature in situ CO2 absorption. The in situ CO2 removal results in shifting the steam reforming reaction equilibrium towards increased hydrogen concentration (up to 95 vol%). The key part of the process is the high temperature CO2 absorbent. In this contribution results of Thermal Gravimetric Analysis (TGA) investigations on natural minerals, dolomites, silicates and synthetic absorbent materials in regard of their CO2 absorption capacity and absorption/desorption cyclic stability are presented and discussed. It has been found that the inert parts of the absorbent materials have a structure stabilizing effect, leading to an improved cyclic stability of the materials.

  6. Usage of Local Fuel for Combined Generation of Thermal and Electric Power

    Directory of Open Access Journals (Sweden)

    G. I. Zhihar

    2011-01-01

    Full Text Available The paper reveals that it is necessary to ensure interaction of the various concerned ministries and institutions involved in storage and transportation of local fuel in order to increase its portion in the power balance of Belarus. Nowadays there is a problem to determine the most efficient and reliable designs of electric power plants for usage of local fuel while executing a combined generation of thermal and electric power on a large scale in respect of Belarus

  7. Production Cycle for Large Scale Fission Mo-99 Separation by the Processing of Irradiated LEU Uranium Silicide Fuel Element Targets

    Directory of Open Access Journals (Sweden)

    Abdel-Hadi Ali Sameh

    2013-01-01

    Full Text Available Uranium silicide fuels proved over decades their exceptional qualification for the operation of higher flux material testing reactors with LEU elements. The application of such fuels as target materials, particularly for the large scale fission Mo-99 producers, offers an efficient and economical solution for the related facilities. The realization of such aim demands the introduction of a suitable dissolution process for the applied U3Si2 compound. Excellent results are achieved by the oxidizing dissolution of the fuel meat in hydrofluoric acid at room temperature. The resulting solution is directly behind added to an over stoichiometric amount of potassium hydroxide solution. Uranium and the bulk of fission products are precipitated together with the transuranium compounds. The filtrate contains the molybdenum and the soluble fission product species. It is further treated similar to the in-full scale proven process. The generated off gas stream is handled also as experienced before after passing through KOH washing solution. The generated alkaline fluoride containing waste solution is noncorrosive. Nevertheless fluoride can be selectively bonded as in soluble CaF2 by addition of a mixture of solid calcium hydroxide calcium carbonate to the sand cement mixture used for waste solidification. The generated elevated amounts of LEU remnants can be recycled and retargeted. The related technology permits the minimization of the generated fuel waste, saving environment, and improving processing economy.

  8. KüFA safety testing of HTR fuel pebbles irradiated in the High Flux Reactor in Petten

    Energy Technology Data Exchange (ETDEWEB)

    Seeger, O., E-mail: oliver.seeger@rwth-aachen.de [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Safety of Irradiated Nuclear Materials Unit, Postfach 2340, 76125 Karlsruhe (Germany); Laurie, M., E-mail: mathias.laurie@ec.europa.eu [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Safety of Irradiated Nuclear Materials Unit, Postfach 2340, 76125 Karlsruhe (Germany); Abjani, A. El; Ejton, J.; Boudaud, D.; Freis, D.; Carbol, P.; Rondinella, V.V. [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Safety of Irradiated Nuclear Materials Unit, Postfach 2340, 76125 Karlsruhe (Germany); Fütterer, M. [European Commission, Joint Research Centre (JRC), Institute for Energy and Transport (IET), Nuclear Reactor Integrity Assessment and Knowledge Management Unit, PO Box 2, 1755 ZG Petten (Netherlands); Allelein, H.-J. [Lehrstuhl für Reaktorsicherheit und -technik an der RWTH Aachen, Kackertstraße 9, 52072 Aachen (Germany)

    2016-09-15

    The Cold Finger Apparatus (KühlFinger-Apparatur—KüFA) in operation at JRC-ITU is designed to experimentally scrutinize the effects of Depressurization LOss of Forced Circulation (D-LOFC) accident scenarios on irradiated High Temperature Reactor (HTR) fuel pebbles. Up to 1600 °C, the reference maximum temperature for these accidents, high-quality German HTR fuel pebbles have already demonstrated a small fission product release. This paper discusses and compares the releases obtained from KüFA-testing the pebbles HFR-K5/3 and HFR-EU1/3, which were both irradiated in the High Flux Reactor (HFR) in Petten. We present the time-dependent fractional release of the volatile fission product {sup 137}Cs as well as the fission gas {sup 85}Kr for both pebbles. For HFR-EU1/3 the isotopes {sup 134}Cs and {sup 154}Eu as well as the shorter-lived {sup 110m}Ag have also been measured. A detailed description of the experimental setup and its accuracy is given. The data for the recently tested pebbles is discussed in the context of previous results.

  9. First results of the irradiation program of inert matrices, targets and fuels for minor actinides transmutation in fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bonnerot, Jean-Marc; Ferroud-Plattet, Marie-Pierre; Lamontagne, Jerome [CEA Cadarache, Nuclear Energy Direction, Saint-Paul les Durance Cedex, 13108 (France); Warin, Dominique [CEA Valrho, Nuclear Energy Direction, DRCP, Bagnols-sur-Ceze Cedex, 30207 (France); Gosmain, Lionel [CEA Saclay, Nuclear Energy Direction, DMN, Gif sur Yvette, 91190 (France)

    2008-07-01

    A comprehensive irradiation program was started in France in 1992 to demonstrate the technical feasibility of the transmutation of minor actinides in current and future nuclear reactors, by means of inert support targets or dedicated fuels. The first step of the program (MATINA program) consisted in the irradiation of various inert materials intended as support matrix for transmutation targets, in the fast reactor Phenix, to select the best candidates. These inert materials included as well oxide and nitride ceramics - MgO, MgAl{sub 2}O{sub 4}, Al{sub 2}O{sub 3}, Y{sub 3}Al{sub 5}O{sub 12} and TiN - as refractory metals - W, Nb, Cr and V- and were irradiated under fast neutron flux at temperatures ranged between 650 and 1040 deg. C. The results show that in comparison to MgO, MgAl{sub 2}O{sub 4} and Al{sub 2}O{sub 3} inert matrices irradiated alone, the composite pellets containing UO{sub 2} particles, showed very different behaviors under irradiation. The swelling of MgO pellets is enhanced in the presence of fissile material whereas it is lowered for the Al{sub 2}O{sub 3}-UO{sub 2} pellets. MgAl{sub 2}O{sub 4}-UO{sub 2} pellets remained stable. The second step of the program aimed at testing the behavior of inert support targets containing americium. A new experiment ECRIX H involving composite pellets with an MgO matrix and AmO{sub 2-x} particles was performed in Phenix and completed in 2006. A rather low elongation of the pellet stack was observed and no significant diameter deformation of cladding was detected after irradiation. The analysis of the filling gas of the pin after puncturing, revealed that respectively 28% and 5% of the He and Xe+Kr created under irradiation were released in the expanding volume of the pin. ECRIX H, which is the first experiment on Am base target in Phenix, will undoubtedly represent a very important step in the general design approach about inert matrix support targets once the complete results should be available by the end of

  10. Triolein reduces MMP-1 upregulation in dermal fibroblasts generated by ROS production in UVB-irradiated keratinocytes.

    Science.gov (United States)

    Leirós, Gustavo J; Kusinsky, Ana Gabriela; Balañá, María Eugenia; Hagelin, Karin

    2017-02-01

    Cytokine production and oxidative stress generated by ultraviolet radiation B (UVB) skin exposure are main factors of skin photoaging. Interleukin-6 (IL-6) produced by irradiated keratinocytes is proposed to have a role in metalloproteinases (MMPs) expression activation in dermal fibroblasts. We examined the effect of triolein treatment of UVB-irradiated keratinocytes on MMP1 (interstitial collagenase) expression response of dermal fibroblasts. We assayed UVB-irradiated keratinocytes soluble signals, mainly IL-6 and reactive oxygen species (ROS). IL-6 expression and ROS generation were assayed in UVB-irradiated keratinocytes. MMP1 mRNA expression response was assayed in fibroblasts grown in keratinocytes conditioned medium. We evaluated the effect of treating keratinocytes with triolein on IL-6 expression and ROS generation in keratinocytes, and MMP1 expression in fibroblasts. The irradiation of epidermal cells with sublethal UVB doses increased IL-6 expression and ROS generation. Conditioned culture medium collected from keratinocytes was used to culture dermal fibroblasts. MMP1 mRNA expression increase was observed in fibroblasts cultured in medium collected from UVB-irradiated keratinocytes. Triolein treatment reduced the IL-6 expression and ROS generation in keratinocytes and this effect was reflected in downregulation of MMP1 expression in fibroblasts. Triolein reduces both the expression of IL-6 and ROS generation in irradiated keratinocytes. It seems to exert an anti-inflammatory and anti-oxidative stress effect on irradiated keratinocytes that in turn reduces MMP1 expression in dermal fibroblasts. Collectively, these results indicate that triolein could act as a photoprotective agent. Copyright © 2016 Japanese Society for Investigative Dermatology. Published by Elsevier B.V. All rights reserved.

  11. HIGH-TEMPERATURE TUBULAR SOLID OXIDE FUEL CELL GENERATOR DEVELOPMENT

    Energy Technology Data Exchange (ETDEWEB)

    S.E. Veyo

    1998-09-01

    During the Westinghouse/USDOE Cooperative Agreement period of November 1, 1990 through November 30, 1997, the Westinghouse solid oxide fuel cell has evolved from a 16 mm diameter, 50 cm length cell with a peak power of 1.27 watts/cm to the 22 mm diameter, 150 cm length dimensions of today's commercial prototype cell with a peak power of 1.40 watts/cm. Accompanying the increase in size and power density was the elimination of an expensive EVD step in the manufacturing process. Demonstrated performance of Westinghouse's tubular SOFC includes a lifetime cell test which ran for a period in excess of 69,000 hours, and a fully integrated 25 kWe-class system field test which operated for over 13,000 hours at 90% availability with less than 2% performance degradation over the entire period. Concluding the agreement period, a 100 kW SOFC system successfully passed its factory acceptance test in October 1997 and was delivered in November to its demonstration site in Westervoort, The Netherlands.

  12. Calculation Simulation of Equivalent Irradiation Swelling for Dispersion Nuclear Fuel%弥散核燃料等效辐照肿胀计算模拟

    Institute of Scientific and Technical Information of China (English)

    蔡维; 赵云妹; 龚辛; 丁淑蓉; 霍永忠

    2015-01-01

    The dispersion nuclear fuel was regarded as a kind of special particle compos‐ites .Assuming that the fuel particles are periodically distributed in the dispersion nucle‐ar fuel meat ,the finite element model to calculate its equivalent irradiation swelling was developed with the method of computational micro‐mechanics .Considering irradiation swelling in the fuel particles and the irradiation hardening effect in the metal matrix ,the stress update algorithms were established respectively for the fuel particles and metal matrix .The corresponding user subroutines were programmed ,and the finite element simulation of equivalent irradiation swelling for the fuel meat was performed in Abaqus . The effects of the particle size and volume fraction on the equivalent irradiation swelling were investigated ,and the fitting formula of equivalent irradiation swelling was obtained .The results indicate that the main factors to influence equivalent irradiation swelling of the fuel meat are the irradiation swelling and volume fraction of fuel particles .%本文将弥散核燃料芯体看作一种特殊的颗粒复合材料,利用细观计算力学的方法,假设燃料颗粒在芯体中周期性分布,建立了对芯体等效辐照肿胀进行计算模拟的有限元模型。考虑颗粒的辐照肿胀和基体材料的辐照硬化效应,分别建立了燃料颗粒和基体材料的应力更新算法,编制了用户材料子程序,在A baqus软件中实现了芯体等效辐照肿胀的有限元模拟。计算分析了颗粒大小和体积含量对芯体等效辐照肿胀的影响,并得到了等效辐照肿胀的拟合公式。研究结果表明,影响芯体等效辐照肿胀的主要因素是颗粒的辐照肿胀和体积含量。

  13. New generation nuclear fuel structures: dense particles in selectively soluble matrix

    Energy Technology Data Exchange (ETDEWEB)

    Sickafus, Kurt E [Los Alamos National Laboratory; Devlin, David J [Los Alamos National Laboratory; Jarvinen, Gordon D [Los Alamos National Laboratory; Patterson, Brian M [Los Alamos National Laboratory; Pattillo, Steve G [Los Alamos National Laboratory; Valdez, James [Los Alamos National Laboratory; Phillips, Jonathan [Los Alamos National Laboratory

    2009-01-01

    We have developed a technology for dispersing sub-millimeter sized fuel particles within a bulk matrix that can be selectively dissolved. This may enable the generation of advanced nuclear fuels with easy separation of actinides and fission products. The large kinetic energy of the fission products results in most of them escaping from the sub-millimeter sized fuel particles and depositing in the matrix during burning of the fuel in the reactor. After the fuel is used and allowed to cool for a period of time, the matrix can be dissolved and the fission products removed for disposal while the fuel particles are collected by filtration for recycle. The success of such an approach would meet a major goal of the GNEP program to provide advanced recycle technology for nuclear energy production. The benefits of such an approach include (1) greatly reduced cost of the actinide/fission product separation process, (2) ease of recycle of the fuel particles, and (3) a radiation barrier to prevent theft or diversion of the recycled fuel particles during the time they are re-fabricated into new fuel. In this study we describe a method to make surrogate nuclear fuels of micrometer scale W (shell)/Mo (core) or HfO2 particles embedded in an MgO matrix that allows easy separation of the fission products and their embedded particles. In brief, the method consists of physically mixing W-Mo or hafnia particles with an MgO precursor. Heating the mixture, in air or argon, without agitation, to a temperature is required for complete decomposition of the precursor. The resulting material was examined using chemical analysis, scanning electron microscopy, X-ray diffraction and micro X-ray computed tomography and found to consist of evenly dispersed particles in an MgO + matrix. We believe this methodology can be extended to actinides and other matrix materials.

  14. Preparation and characterization of mono-sheet bipolar membranes by pre-irradiation grafting method for fuel cell applications

    Science.gov (United States)

    Guan, Yingjie; Fang, Jun; Fu, Tao; Zhou, Huili; Wang, Xin; Deng, Zixiang; Zhao, Jinbao

    2016-09-01

    A new method for the preparation of the mono-sheet bipolar membrane applied to fuel cells was developed based on the pre-irradiation grafting technology. A series of bipolar membranes were successfully prepared by simultaneously grafting of styrene onto one side of the poly(ethylene-co-tetrafluoroethylene) base film and 1-vinylimidazole onto the opposite side, followed by the sulfonation and alkylation, respectively. The chemical structures and microstructures of the prepared membranes were investigated by ATR-FTIR and SEM-EDS. The TGA measurements demonstrated the prepared bipolar membranes have reasonable thermal stability. The ion exchange capacity, water uptake and ionic conductivity of the membranes were also characterized. The H2/O2 single fuel cells using these membranes were evaluated and revealed a maximum power density of 107 mW cm-2 at 35 °C with unhumidified hydrogen and oxygen. The preliminary performances suggested the great prospect of these membranes in application of bipolar membrane fuel cells.

  15. A methodology for the stochastic generation of hourly synthetic direct normal irradiation time series

    Science.gov (United States)

    Larrañeta, M.; Moreno-Tejera, S.; Lillo-Bravo, I.; Silva-Pérez, M. A.

    2017-06-01

    Many of the available solar radiation databases only provide global horizontal irradiance (GHI) while there is a growing need of extensive databases of direct normal radiation (DNI) mainly for the development of concentrated solar power and concentrated photovoltaic technologies. In the present work, we propose a methodology for the generation of synthetic DNI hourly data from the hourly average GHI values by dividing the irradiance into a deterministic and stochastic component intending to emulate the dynamics of the solar radiation. The deterministic component is modeled through a simple classical model. The stochastic component is fitted to measured data in order to maintain the consistency of the synthetic data with the state of the sky, generating statistically significant DNI data with a cumulative frequency distribution very similar to the measured data. The adaptation and application of the model to the location of Seville shows significant improvements in terms of frequency distribution over the classical models. The proposed methodology applied to other locations with different climatological characteristics better results than the classical models in terms of frequency distribution reaching a reduction of the 50% in the Finkelstein-Schafer (FS) and Kolmogorov-Smirnov test integral (KSI) statistics.

  16. Generation of soft x-ray radiation by laser irradiation of a gas puff xenon target

    Energy Technology Data Exchange (ETDEWEB)

    Fiedorowicz, H.; Bartnik, A.; Szczurek, M. [Military Univ. of Technology, Warsaw (Poland). Inst. of Optoelectronics] [and others

    1995-12-31

    Plasmas produced from laser-irradiated gas puff xenon targets, created by pulsed injection of xenon with high-pressure solenoid valve, offer the possibility of realizing a debrisless x-ray point source for the x-ray lithography applications. In this paper the authors present results of the experimental investigations on the x-ray generation from a gas puff xenon target irradiated with nanosecond high-power laser pulses produced using two different laser facilities: a Nd:glass laser operating at 1.06 {micro}m, which generated 10--15 J pulses in 1 ns FWHM, and Nd:glass slab laser, producing pulses of 10 ns duration with energy reaching 12 J for a 0.53 {micro}m wavelength or 20 J for 1.05 {micro}m. To study the x-ray emission different x-ray diagnostic methods have been used. X-ray spectra were registered using a flat CsAP crystal spectrograph with an x-ray film or a curved KAP crystal spectrograph with a convex curvature to an x-ray CCD readout detector. X-ray images have been taken using pinhole cameras with an x-ray film or a CCD array. X-ray yield was measured with the use of semiconductor detectors (silicon photodiodes or diamond photoconductors).

  17. DESIGN OF COMBINED CYCLE GENERATION SYSTEM WITH HIGH TEMPERATURE FUEL CELL AND STEAM TURBINE

    Institute of Scientific and Technical Information of China (English)

    Yu Lijun; Yuan Junqi; Cao Guangyi

    2003-01-01

    For environment protection and high efficiency, development of new concept power plant has been required in China. The fuel cell is expected to be used in a power plant as a centralized power station or distributed power plant. It is a chemical power generation device that converts the energy of a chemical reaction directly into electrical energy and not limited by Carnot cycle efficiency. The molten carbonate fuel cell (MCFC) power plant has several attractive features I.e. High efficiency and lower emission of Nox and Sox. A combined cycle generation system with MCFC and steam turbine is designed. Its net electrical efficiency LHV is about 55%.

  18. Expert System analysis of non-fuel assembly hardware and spent fuel disassembly hardware: Its generation and recommended disposal

    Energy Technology Data Exchange (ETDEWEB)

    Williamson, Douglas Alan [Univ. of Florida, Gainesville, FL (United States)

    1991-01-01

    Almost all of the effort being expended on radioactive waste disposal in the United States is being focused on the disposal of spent Nuclear Fuel, with little consideration for other areas that will have to be disposed of in the same facilities. one area of radioactive waste that has not been addressed adequately because it is considered a secondary part of the waste issue is the disposal of the various Non-Fuel Bearing Components of the reactor core. These hardware components fall somewhat arbitrarily into two categories: Non-Fuel Assembly (NFA) hardware and Spent Fuel Disassembly (SFD) hardware. This work provides a detailed examination of the generation and disposal of NFA hardware and SFD hardware by the nuclear utilities of the United States as it relates to the Civilian Radioactive Waste Management Program. All available sources of data on NFA and SFD hardware are analyzed with particular emphasis given to the Characteristics Data Base developed by Oak Ridge National Laboratory and the characterization work performed by Pacific Northwest Laboratories and Rochester Gas & Electric. An Expert System developed as a portion of this work is used to assist in the prediction of quantities of NFA hardware and SFD hardware that will be generated by the United States` utilities. Finally, the hardware waste management practices of the United Kingdom, France, Germany, Sweden, and Japan are studied for possible application to the disposal of domestic hardware wastes. As a result of this work, a general classification scheme for NFA and SFD hardware was developed. Only NFA and SFD hardware constructed of zircaloy and experiencing a burnup of less than 70,000 MWD/MTIHM and PWR control rods constructed of stainless steel are considered Low-Level Waste. All other hardware is classified as Greater-ThanClass-C waste.

  19. Generation of daily solar irradiation by means of artificial neural net works

    Energy Technology Data Exchange (ETDEWEB)

    Siqueira, Adalberto N.; Tiba, Chigueru; Fraidenraich, Naum [Departamento de Energia Nuclear, da Universidade Federal de Pernambuco, Av. Prof. Luiz Freire, 1000 - CDU, CEP 50.740-540 Recife, Pernambuco (Brazil)

    2010-11-15

    The present study proposes the utilization of Artificial Neural Networks (ANN) as an alternative for generating synthetic series of daily solar irradiation. The sequences were generated from the use of daily temporal series of a group of meteorological variables that were measured simultaneously. The data used were measured between the years of 1998 and 2006 in two temperate climate localities of Brazil, Ilha Solteira (Sao Paulo) and Pelotas (Rio Grande do Sul). The estimates were taken for the months of January, April, July and October, through two models which are distinguished regarding the use or nonuse of measured bright sunshine hours as an input variable. An evaluation of the performance of the 56 months of solar irradiation generated by way of ANN showed that by using the measured bright sunshine hours as an input variable (model 1), the RMSE obtained were less or equal to 23.2% being that of those, although 43 of those months presented RMSE less or equal to 12.3%. In the case of the model that did not use the measured bright sunshine hours but used a daylight length (model 2), RMSE were obtained that varied from 8.5% to 37.5%, although 38 of those months presented RMSE less or equal to 20.0%. A comparison of the monthly series for all of the years, achieved by means of the Kolmogorov-Smirnov test (to a confidence level of 99%), demonstrated that of the 16 series generated by ANN model only two, obtained by model 2 for the months of April and July in Pelotas, presented significant difference in relation to the distributions of the measured series and that all mean deviations obtained were inferior to 0.39 MJ/m{sup 2}. It was also verified that the two ANN models were able to reproduce the principal statistical characteristics of the frequency distributions of the measured series such as: mean, mode, asymmetry and Kurtosis. (author)

  20. 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.)

  1. Electricity generation from wastewaters with starch as carbon source using a mediatorless microbial fuel cell.

    Science.gov (United States)

    Herrero-Hernandez, E; Smith, T J; Akid, R

    2013-01-15

    Microbial fuel cells represent a new method for producing electricity from the oxidation of organic matter. A mediatorless microbial fuel cell was developed using Escherichia coli as the active bacterial component with synthetic wastewater of potato extract as the energy source. The two-chamber fuel cell, with a relation of volume between anode and cathode chamber of 8:1, was operated in batch mode. The response was similar to that obtained when glucose was used as the carbon source. The performance characteristics of the fuel cell were evaluated with two different anode and cathode shapes, platinised titanium strip or mesh; the highest maximum power density (502mWm(-2)) was achieved in the microbial fuel cell with mesh electrodes. In addition to electricity generation, the MFC exhibited efficient treatment of wastewater so that significant reduction of initial oxygen demand of wastewater by 61% was observed. These results demonstrate that potato starch can be used for power generation in a mediatorless microbial fuel cell with high removal efficiency of chemical oxygen demand.

  2. Migration from Gasoline to Gaseous Fuel for Small-scale Electricity Generation Systems

    Directory of Open Access Journals (Sweden)

    Sukandar Sukandar

    2013-03-01

    Full Text Available This paper describes a study that gives a consideration to change fuel source for electricity generator from gasoline to combustible gas. A gaseous fuel conversion technology is presented and its performance is compared with gasoline. In the experiment, two types of load were tested, resistive and resistive-inductive. By using both fuels mostly the power factor (Cos ? of resistive-inductive load variations were greater than 0.8, and they had slight difference on operational voltage. The drawback of using gaseous fuel is the frequency of the electricity might be up to 10 Hz deviated from the standard frequency (i.e. 50 Hz. In the lab scale experiment, the gasoline consumption increased proportionally with the load increase, while using gaseous fuel the consumption of gas equal for two different load value in the range of 50% maximum load, which is 100 gram per 15 minutes operation. Therefore, the use of gaseous generation system should have average power twice than the required load. The main advantage using gaseous fuel (liquefied petroleum gas or biogas compared to gasoline is a cleaner emitted gas after combustion.

  3. Compaction Scale Up and Optimization of Cylindrical Fuel Compacts for the Next Generation Nuclear Plant

    Energy Technology Data Exchange (ETDEWEB)

    Jeffrey J. Einerson; Jeffrey A. Phillips; Eric L. Shaber; Scott E. Niedzialek; W. Clay Richardson; Scott G. Nagley

    2012-10-01

    Multiple process approaches have been used historically to manufacture cylindrical nuclear fuel compacts. Scale-up of fuel compacting was required for the Next Generation Nuclear Plant (NGNP) project to achieve an economically viable automated production process capable of providing a minimum of 10 compacts/minute with high production yields. In addition, the scale-up effort was required to achieve matrix density equivalent to baseline historical production processes, and allow compacting at fuel packing fractions up to 46% by volume. The scale-up approach of jet milling, fluid-bed overcoating, and hot-press compacting adopted in the U.S. Advanced Gas Reactor (AGR) Fuel Development Program involves significant paradigm shifts to capitalize on distinct advantages in simplicity, yield, and elimination of mixed waste. A series of designed experiments have been completed to optimize compaction conditions of time, temperature, and forming pressure using natural uranium oxycarbide (NUCO) fuel. Results from these experiments are included. The scale-up effort is nearing completion with the process installed and operational using nuclear fuel materials. The process is being certified for manufacture of qualification test fuel compacts for the AGR-5/6/7 experiment at the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL).

  4. Micro-bubble generated by laser irradiation on an individual carbon nanocoil

    Energy Technology Data Exchange (ETDEWEB)

    Sun, Yanming, E-mail: amandaming@mail.dlut.edu.cn [School of Physics and Optoelectronic Technology, DUT, Linggong Road, Dalian 116024 (China); Pan, Lujun, E-mail: lpan@dlut.edu.cn [School of Physics and Optoelectronic Technology, DUT, Linggong Road, Dalian 116024 (China); Liu, Yuli, E-mail: liuyuli2005@163.com [School of Physics and Optoelectronic Technology, DUT, Linggong Road, Dalian 116024 (China); Sun, Tao, E-mail: 332077309@qq.com [School of Energy and Power Engineering, DUT, Linggong Road, Dalian 116024 (China)

    2015-08-01

    Highlights: • We have investigated laser irradiated microbubbles which can be generated at fixed point on surface of an individual carbon nanocoil (CNC) immerged in deionized water. • The microbubble can be operated easily and flexibly. • Based on classical heat and mass transfer theories, the bubble growth data is in good agreement with the simplified model. - Abstract: We have investigated the micro-bubbles generated by laser induction on an individual carbon nanocoil (CNC) immerged in deionized water. The photon energy of the incident focused laser beam is absorbed by CNC and converted to thermal energy, which efficiently vaporizes the surrounding water, and subsequently a micro-bubble is generated at the laser location. The dynamics behavior of bubble generation, including its nucleation, expansion and steady-state, has been studied experimentally and theoretically. We have derived equations to analyze the expansion process of a bubble based on classical heat and mass transfer theories. The conclusion is in good agreement with the experiment. CNC, which acts as a realistic micro-bubble generator, can be operated easily and flexibly.

  5. Controllable generation of reactive oxygen species by femtosecond-laser irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Yan, Wei; He, Hao, E-mail: haohe@tju.edu.cn; Wang, Yintao; Wang, Yisen; Hu, Minglie; Wang, Chingyue [Ultrafast Laser Laboratory, Key Laboratory of Optoelectronic Information Technology (Ministry of Education), College of Precision Instrument and Optoelectronics Engineering, Tianjin University, Tianjin (China)

    2014-02-24

    Femtosecond lasers have been advancing Biophotonics research in the past two decades with multiphoton microscopy, microsurgery, and photodynamic therapy. Nevertheless, laser irradiation is identified to bring photodamage to cells via reactive oxygen species (ROS) generation with unclear mechanism. Meanwhile, currently in biological researches, there is no effective method to provide controllable ROS production precisely, which originally is leaked from mitochondria during respiration and plays a key role in a lot of important cellular processes and cellular signaling pathways. In this study, we show the process of how the tightly focused femtosecond-laser induces ROS generation solely in mitochondria at the very beginning and then release to cytosol if the stimulus is intense enough. At certain weak power levels, the laser pulses induce merely moderate Ca{sup 2+} release but this is necessary for the laser to generate ROS in mitochondria. Cellular original ROS are also involved with a small contribution. When the power is above a threshold, ROS are then released to cytosol, indicating photodamage overwhelming cellular repair ability. The mechanisms in those two cases are quite different. Those results clarify parts of the mechanism in laser-induced ROS generation. Hence, it is possible to further this optical scheme to provide controllable ROS generation for ROS-related biological researches including mitochondrial diseases and aging.

  6. Controllable generation of reactive oxygen species by femtosecond-laser irradiation

    Science.gov (United States)

    Yan, Wei; He, Hao; Wang, Yintao; Wang, Yisen; Hu, Minglie; Wang, Chingyue

    2014-02-01

    Femtosecond lasers have been advancing Biophotonics research in the past two decades with multiphoton microscopy, microsurgery, and photodynamic therapy. Nevertheless, laser irradiation is identified to bring photodamage to cells via reactive oxygen species (ROS) generation with unclear mechanism. Meanwhile, currently in biological researches, there is no effective method to provide controllable ROS production precisely, which originally is leaked from mitochondria during respiration and plays a key role in a lot of important cellular processes and cellular signaling pathways. In this study, we show the process of how the tightly focused femtosecond-laser induces ROS generation solely in mitochondria at the very beginning and then release to cytosol if the stimulus is intense enough. At certain weak power levels, the laser pulses induce merely moderate Ca2+ release but this is necessary for the laser to generate ROS in mitochondria. Cellular original ROS are also involved with a small contribution. When the power is above a threshold, ROS are then released to cytosol, indicating photodamage overwhelming cellular repair ability. The mechanisms in those two cases are quite different. Those results clarify parts of the mechanism in laser-induced ROS generation. Hence, it is possible to further this optical scheme to provide controllable ROS generation for ROS-related biological researches including mitochondrial diseases and aging.

  7. Generation and validation of a prognostic score to predict outcome after re-irradiation of recurrent glioma

    Energy Technology Data Exchange (ETDEWEB)

    Combs, Stephanie E.; Welzel, Thomas; Debus, Juergen [Univ. Hospital of Heidelberg, Dept. of Radiation Oncology, Heidelberg (Germany)], E-mail: Stephanie.combs@med.uni-heidelberg.de; Edler, Lutz; Rausch, Renate [German Cancer Research Center (dkfz), Dept. of Biostatistics, Heidelberg (Germany); Wick, Wolfgang [Univ. Hospital of Heidelberg, Dept. of Neurooncology, Heidelberg (Germany)

    2013-01-15

    Re-irradiation using high-precision radiation techniques has been established within the clinical routine for patients with recurrent gliomas. In the present work, we developed a practical prognostic score to predict survival outcome after re-irradiation. Patients and methods. Fractionated stereotactic radiotherapy (FSRT) was applied in 233 patients. Primary histology included glioblastoma (n = 89; 38%), WHO Grade III gliomas (n = 52; 22%) and low-grade glioma (n = 92; 40%). FSRT was applied with a median dose of 36 Gy in 2 Gy single fractions. We evaluated survival after re-irradiation as well as progression-free survival after re-irradiation; prognostic factors analyzed included age, tumor volume at re-irradiation, histology, time between initial radiotherapy and re-irradiation, age and Karnofsky Performance Score. Results. Median survival after FSRT was 8 months for glioblastoma, 20 months for anaplastic gliomas, and 24 months for recurrent low-grade patients. The strongest prognostic factors significantly impacting survival after re-irradiation were histology (p <0.0001) and age (<50 vs. ={>=}50, p < 0.0001) at diagnosis and the time between initial radiotherapy and re-irradiation {<=}12 vs. >12 months (p < 0.0001). We generated a four-class prognostic score to distinguish patients with excellent (0 points), good (1 point), moderate (2 points) and poor (3-4 points) survival after re-irradiation. The difference in outcome was highly significant (p < 0.0001). Conclusion. We generated a practical prognostic score index based on three clinically relevant factors to predict the benefit of patients from re-irradiation. This score index can be helpful in patient counseling, and for the design of further clinical trials. However, individual treatment decisions may include other patient-related factors not directly influencing outcome.

  8. Impacts of Wind and Solar on Fossil-Fueled Generators: Preprint

    Energy Technology Data Exchange (ETDEWEB)

    Lew, D.; Brinkman, G.; Kumar, N.; Besuner, P.; Agan, D.; Lefton, S.

    2012-08-01

    High penetrations of wind and solar power will impact the operations of the remaining generators on the power system. Regional integration studies have shown that wind and solar may cause fossil-fueled generators to cycle on and off and ramp down to part load more frequently and potentially more rapidly. Increased cycling, deeper load following, and rapid ramping may result in wear-and-tear impacts on fossil-fueled generators that lead to increased capital and maintenance costs, increased equivalent forced outage rates, and degraded performance over time. Heat rates and emissions from fossil-fueled generators may be higher during cycling and ramping than during steady-state operation. Many wind and solar integration studies have not taken these increased cost and emissions impacts into account because data have not been available. This analysis considers the cost and emissions impacts of cycling and ramping of fossil-fueled generation to refine assessments of wind and solar impacts on the power system.

  9. On the potential of third generation biofuels as a sustainable fuel source

    Energy Technology Data Exchange (ETDEWEB)

    Buckermann, Wilhelm A. [Hochschule Esslingen (Germany). Faculty of Natural Sciences

    2013-06-01

    Compared to other alternative transportation fuels, such as hydrogen or electricity, biofuels have the advantage of high energy density and easy handling. This means that they can be used in a comparable way and with the same logistic systems as classic fossil fuels. Furthermore, it is expected that their utilisation will provide a positive environmental impact by reducing greenhouse gas emissions. Hence, driven by environmental and energy-political ambitions, biofuels will attain increasing importance in the future. Some forecasts see a portion of up to 50% which these fuels might contribute to the global fuel demand by the middle of the century. In contrast to those of the first and second generation, third generation biofuels, which are based on specifically cultivated plants and micro-organisms, are not in competition with the provision of food and have distinctly lower requirements for agricultural land use. This contribution will outline the aspects of availability and sustainability of this kind of biogenic fuels and will analyse in particular which role third generation biofuels might play. (orig.)

  10. 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.

  11. White-light emission from solid carbon in aqueous solution during hydrogen generation induced by nanosecond laser pulse irradiation

    Science.gov (United States)

    Akimoto, Ikuko; Yamamoto, Shota; Maeda, Kosuke

    2016-07-01

    We previously discovered a novel method of hydrogen generation from high-grade charcoal in an aqueous solution using nanosecond laser pulse irradiation. In this paper, white-light emission during this reaction is reported: A broad spectrum over the visible range is observed above a threshold excitation energy density. The white-light emission is a simultaneous product of the hydrogen generation reaction and is attributed to blackbody radiation in accordance with Planck's Law at a temperature above 3800 K. Consequently, we propose that hydrogen generation induced by laser irradiation proceeds similarly to classical coal gasification, which features reactions at high pressure and high temperature.

  12. 78 FR 31821 - Physical Protection of Shipments of Irradiated Reactor Fuel

    Science.gov (United States)

    2013-05-28

    ...: Nuclear Regulatory Commission. ACTION: NUREG; issuance. SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is issuing Revision 2 of NUREG-0561, ``Physical Protection of Shipments of Irradiated Reactor... individuals granted unescorted access to SNF during transportation. DATES: Revision 2 of NUREG-0561...

  13. AGR-1 Post Irradiation Examination Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Demkowicz, Paul Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-08-01

    on the as-fabricated fuel characterization and irradiation data. In addition to the extensive volume of results generated, the work also resulted in a number of novel analysis techniques and lessons learned that are being applied to the examination of fuel from subsequent TRISO fuel irradiations. This report provides a summary of the results obtained as part of the AGR-1 PIE campaign over its approximately 5-year duration.

  14. Fuel and Greenhouse Gas Emission Reduction Potentials by Appropriate Fuel Switching and Technology Improvement in the Canadian Electricity Generation Sector

    Directory of Open Access Journals (Sweden)

    Farshid Zabihian

    2010-01-01

    Full Text Available Problem statement: In recent years, Greenhouse Gas (GHG emissions and their potential effects on global climate change have been a worldwide concern. According to International Energy Agency (IEA, power generation contributes more than half of the global GHG emissions. Approach: Purpose of this study is to examine GHG emission reduction potentials in the Canadian electricity generation sector through fuel switching and adoption of advanced power generation systems. To achieve this objective, eight different scenarios were introduced. In the first scenario, existing power stations’ fuel was switched to natural gas. Existing power plants were replaced by Natural Gas Combined Cycle (NGCC, Integrated Gasification Combined Cycle (IGCC, Solid Oxide Fuel Cell (SOFC, hybrid SOFC and SOFC-IGCC hybrid power stations in scenario numbers 2 to 6, respectively. In last two scenarios, CO2 capture systems were installed in the existing power plants and in the second scenario, respectively. Results: The results showed that Canada’s GHG emissions can be reduced by 33, 59, 20, 64, 69, 29, 86 and 94% based on the first to eighth scenarios, respectively. On the other hand, the second scenario is the most practical and its technology has already matured and is available. In this scenario by replacing existing power plants by NGCC power plants, Canada can fulfill more than 25% of its 238,000 kt year-1 commitment of GHG emission reduction to the Kyoto Protocol. In addition, the GHG emission reduction potentials for each province and Canada as a whole were presented and compared. Based on the results, Alberta, Ontario and Saskatchewan are the biggest producers of GHG in Canada by emitting 49, 21 and 14% of Canada’s GHG emissions, respectively. Therefore, they have higher potential to reduce GHG emissions. The comparison of the results for different provinces revealed that based on efficiency of electricity generation and consumed fuel distribution; specific scenario

  15. Electricity generation from macroalgae Enteromorpha prolifera hydrolysates using an alkaline fuel cell.

    Science.gov (United States)

    Liu, Susu; Liu, Xianhua; Wang, Ying; Zhang, Pingping

    2016-12-01

    The goal of this work was to develop a method for the direct power generation using macroalgae Enteromorpha prolifera. The process conditions for the saccharification of macroalgae were optimized and a type of alkaline fuel cell contained no precious metal catalysts was developed. Under optimum conditions (170°C and 2% hydrochloric acid for 45min), dilute acid hydrolysis of the homogenized plants yielded 272.25g reducing sugar/kg dry algal biomass. The maximum power density reached 3.81W/m(2) under the condition of 3M KOH and 18.15g/L reducing sugar in hydrolysate, higher than any other reported algae-fed fuel cells. This study represents the first report on direct electricity generation from macroalgae using alkaline fuel cells, suggesting that there is great potential for the production of renewable energy using marine biomass. Copyright © 2016 Elsevier Ltd. All rights reserved.

  16. An Experimental Investigation of Performance and Emissions of LPG as Dual Fuel in Diesel Engine Generator

    Directory of Open Access Journals (Sweden)

    K. Mohan Kumar

    2014-11-01

    Full Text Available The usage of diesel engine generating set (Gen set increasing day by day where the places without connection to power grid or emergency power supply when the grid fails. Worldwide dual fuel engines are becoming popular because of high performance and low emissions. LPG with diesel is a proven technology in case of vehicles, but in diesel engine power plants it is far so. The proposed work is concentrated on higher load of Diesel Engine Generator with LPG as dual fuel by keeping environmental concern. A test is conducted on performance of engine along with emissions at different proportions of Diesel and LPG including 100% diesel. An experimental set up is made with simple modifications on existing genset to supply LPG as secondary fuel into Diesel.

  17. Neutron Flux Depression in the UO{sub 2}-PuO{sub 2}(15 to 30%) Fuel Rods from IVO-FR2-Vg7-Irradiation Experiment; Depresion de flujo neutronico en las barras combustibles de UO2-PuO2(15 al 30%) del experimento de irradiacion IVO-FR2-Vg7

    Energy Technology Data Exchange (ETDEWEB)

    Lopez, J.; Fernandez, J. L.

    1983-07-01

    The thermal-neutron flux depression within a fuel rod has a great influence in the radial temperature profile of the rod, especially for high enrichment fuel. For this reason, a study was made about the UO{sub 2}-PUO{sub 2} (15 to 30% PUO{sub 2}) fuel pins for the KfK-JEN joint irradiation program IVO, in the FR2 reactor. Different methods (diffusion, Bonalumi, successive generations) were compared and a new approach (parabolic approximation) was developed. (Author) 22 refs.

  18. Autogenic reaction synthesis of photocatalysts for solar fuel generation

    Energy Technology Data Exchange (ETDEWEB)

    Ingram, Brian J.; Pol, Vilas G.; Cronauer, Donald C.; Ramanathan, Muruganathan

    2016-04-19

    In one preferred embodiment, a photocatalyst for conversion of carbon dioxide and water to a hydrocarbon and oxygen comprises at least one nanoparticulate metal or metal oxide material that is substantially free of a carbon coating, prepared by heating a metal-containing precursor compound in a sealed reactor under a pressure autogenically generated by dissociation of the precursor material in the sealed reactor at a temperature of at least about 600.degree. C. to form a nanoparticulate carbon-coated metal or metal oxide material, and subsequently substantially removing the carbon coating. The precursor material comprises a solid, solvent-free salt comprising a metal ion and at least one thermally decomposable carbon- and oxygen-containing counter-ion, and the metal of the salt is selected from the group consisting of Mn, Ti, Sn, V, Fe, Zn, Zr, Mo, Nb, W, Eu, La, Ce, In, and Si.

  19. Autogenic reaction synthesis of photocatalysts for solar fuel generation

    Energy Technology Data Exchange (ETDEWEB)

    Ingram, Brian J.; Pol, Vilas G.; Cronauer, Donald C.; Ramanathan, Muruganathan

    2016-04-19

    In one preferred embodiment, a photocatalyst for conversion of carbon dioxide and water to a hydrocarbon and oxygen comprises at least one nanoparticulate metal or metal oxide material that is substantially free of a carbon coating, prepared by heating a metal-containing precursor compound in a sealed reactor under a pressure autogenically generated by dissociation of the precursor material in the sealed reactor at a temperature of at least about 600.degree. C. to form a nanoparticulate carbon-coated metal or metal oxide material, and subsequently substantially removing the carbon coating. The precursor material comprises a solid, solvent-free salt comprising a metal ion and at least one thermally decomposable carbon- and oxygen-containing counter-ion, and the metal of the salt is selected from the group consisting of Mn, Ti, Sn, V, Fe, Zn, Zr, Mo, Nb, W, Eu, La, Ce, In, and Si.

  20. Impact of the High Flux Isotope Reactor HEU to LEU Fuel Conversion on Cold Source Nuclear Heat Generation Rates

    Energy Technology Data Exchange (ETDEWEB)

    Chandler, David [ORNL

    2014-03-01

    Under the sponsorship of the US Department of Energy National Nuclear Security Administration, staff members at the Oak Ridge National Laboratory have been conducting studies to determine whether the High Flux Isotope Reactor (HFIR) can be converted from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. As part of these ongoing studies, an assessment of the impact that the HEU to LEU fuel conversion has on the nuclear heat generation rates in regions of the HFIR cold source system and its moderator vessel was performed and is documented in this report. Silicon production rates in the cold source aluminum regions and few-group neutron fluxes in the cold source moderator were also estimated. Neutronics calculations were performed with the Monte Carlo N-Particle code to determine the nuclear heat generation rates in regions of the HFIR cold source and its vessel for the HEU core operating at a full reactor power (FP) of 85 MW(t) and the reference LEU core operating at an FP of 100 MW(t). Calculations were performed with beginning-of-cycle (BOC) and end-of-cycle (EOC) conditions to bound typical irradiation conditions. Average specific BOC heat generation rates of 12.76 and 12.92 W/g, respectively, were calculated for the hemispherical region of the cold source liquid hydrogen (LH2) for the HEU and LEU cores, and EOC heat generation rates of 13.25 and 12.86 W/g, respectively, were calculated for the HEU and LEU cores. Thus, the greatest heat generation rates were calculated for the EOC HEU core, and it is concluded that the conversion from HEU to LEU fuel and the resulting increase of FP from 85 MW to 100 MW will not impact the ability of the heat removal equipment to remove the heat deposited in the cold source system. Silicon production rates in the cold source aluminum regions are estimated to be about 12.0% greater at BOC and 2.7% greater at EOC for the LEU core in comparison to the HEU core. Silicon is aluminum s major transmutation product and

  1. Influence of the pressure holding time on strain generation in fuel injection lines

    Energy Technology Data Exchange (ETDEWEB)

    Basara, Adis, E-mail: adis.basara@evonik.com [Process Technology and Engineering, Evonik Degussa GmbH, Rodenbacher Chaussee 4, 63457 Hanau-Wolfgang (Germany); Alt, Nicolas; Schluecker, Eberhard [Institute for Process Technology and Machinery, Friedrich-Alexander University Erlangen-Nuremberg, Cauerstrasse 4, 91058 Erlangen (Germany)

    2011-04-15

    An influence of the pressure holding time on residual strain generation during the autofrettage process was studied experimentally for the first time in the present work. It is the state of the art that fuel injection lines are held at the autofrettage pressure for only a few seconds in an industrial production. In doing so, it is assumed that a desirable residual stress-strain pattern is generated. However, the results of the experimental investigations outlined in this work indicated that completion of the plastic deformation caused by the autofrettage process and generation of the desirable stress-strain pattern require a much longer period. As shown, a third-order polynomial equation best described the interdependence between the time required for the completion of the process, the corresponding autofrettage pressure and the generated strain state. The method presented can be used as a tool for the determination of the optimal autofrettage process parameters in industrial production of fuel injection lines.

  2. Numerical solution of moving boundary problem for deposition process in solid fuel gas generator

    Science.gov (United States)

    Volokhov, V. M.; Dorofeenko, S. O.; Sharov, M. S.; Toktaliev, P. D.

    2016-11-01

    Moving boundary problem in application to process of depositions formation in gas generator are considered. Gas generator, as a part of fuel preparation system of high-speed vehicle, convert solid fuel into multicomponent multiphase mixture, which further burned down in combustion chamber. Mathematical model of two-phase “gas-solid particles” flow, including Navier-Stokes equations for turbulent flow in gas generator and mass, impulse conservations laws for elementary depositions layer are proposed. Verification of proposed mathematical model for depositions mass in gas generator conditions is done. Further possible improvements of proposed model, based on more detail accounting of particle-wall interaction and wall's surface adhesion properties are analyzed.

  3. Progress on using deuteron-deuteron fusion generated neutrons for 40Ar/39Ar sample irradiation

    Science.gov (United States)

    Rutte, Daniel; Renne, Paul R.; Becker, Tim; Waltz, Cory; Ayllon Unzueta, Mauricio; Zimmerman, Susan; Hidy, Alan; Finkel, Robert; Bauer, Joseph D.; Bernstein, Lee; van Bibber, Karl

    2017-04-01

    We present progress on the development and proof of concept of a deuteron-deuteron fusion based neutron generator for 40Ar/39Ar sample irradiation. Irradiation with deuteron-deuteron fusion neutrons is anticipated to reduce Ar recoil and Ar production from interfering reactions. This will allow dating of smaller grains and increase accuracy and precision of the method. The instrument currently achieves neutron fluxes of ˜9×107 cm-2s-1 as determined by irradiation of indium foils and use of the activation reaction 115In(n,n')115mIn. Multiple foils and simulations were used to determine flux gradients in the sample chamber. A first experiment quantifying the loss of 39Ar is underway and will likely be available at the time of the presentation of this abstract. In ancillary experiments via irradiation of K salts and subsequent mass spectrometric analysis we determined the cross-sections of the 39K(n,p)39Ar reaction at ˜2.8 MeV to be 160 ± 35 mb (1σ). This result is in good agreement with bracketing cross-section data of ˜96 mb at ˜2.45 MeV and ˜270 mb at ˜4 MeV [Johnson et al., 1967; Dixon and Aitken, 1961 and Bass et al. 1964]. Our data disfavor a much lower value of ˜45 mb at 2.59 MeV [Lindström & Neuer, 1958]. In another ancillary experiment the cross section for 39K(n,α)36Cl at ˜2.8 MeV was determined as 11.7 ± 0.5 mb (1σ), which is significant for 40Ar/39Ar geochronology due to subsequent decay to 36Ar as well as for the determination of production rates of cosmogenic 36Cl. Additional experiments resolving the cross section functions on 39K between 1.5 and 3.6 MeV are on their way using the LICORNE neutron source of the IPN Orsay tandem accelerator. Results will likely be available at the time of the presentation of this abstract. While the neutron generator is designed for fluxes of ˜109 cm-2s-1, arcing in the sample chamber currently limits the power—straightforwardly correlated to the neutron flux—the generator can safely be run at. Further

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-10-01

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

  5. Increased power generation from primary sludge by a submersible microbial fuel cell and optimum operational conditions

    DEFF Research Database (Denmark)

    Vologni, Valentina; Kakarla, Ramesh; Angelidaki, Irini;

    2013-01-01

    Microbial fuel cells (MFCs) have received attention as a promising renewable energy technology for waste treatment and energy recovery. We tested a submersible MFC with an innovative design capable of generating a stable voltage of 0.250 ± 0.008 V (with a fixed 470 Ω resistor) directly from primary...

  6. Biomass & Natural Gas Based Hydrogen Fuel For Gas Turbine (Power Generation)

    Science.gov (United States)

    Significant progress has been made by major power generation equipment manufacturers in the development of market applications for hydrogen fuel use in gas turbines in recent years. Development of a new application using gas turbines for significant reduction of power plant CO2 e...

  7. DESIGNING AN OPPORTUNITY FUEL WITH BIOMASS AND TIRE-DERIVED FUEL FOR COFIRING AT WILLOW ISLAND GENERATING STATION AND COFIRING SAWDUST WITH COAL AT ALBRIGHT GENERATING STATION

    Energy Technology Data Exchange (ETDEWEB)

    K. Payette; D. Tillman

    2003-10-01

    During the period July 1, 2003-September 30, 2003, Allegheny Energy Supply Co., LLC (Allegheny) proceeded with demonstration operations at the Willow Island Generating Station and improvements to the Albright Generating Station cofiring systems. The demonstration operations at Willow Island were designed to document integration of bio mass cofiring into commercial operations, including evaluating new sources of biomass supply. The Albright improvements were designed to increase the resource base for the projects, and to address issues that came up during the first year of operations. During this period, a major presentation summarizing the program was presented at the Pittsburgh Coal Conference. This report summarizes the activities associated with the Designer Opportunity Fuel program, and demonstrations at Willow Island and Albright Generating Stations.

  8. Next-generation batteries and fuel cells for commercial, military, and space applications

    CERN Document Server

    Jha, A R

    2012-01-01

    Distilling complex theoretical physical concepts into an understandable technical framework, Next-Generation Batteries and Fuel Cells for Commercial, Military, and Space Applications describes primary and secondary (rechargeable) batteries for various commercial, military, spacecraft, and satellite applications for covert communications, surveillance, and reconnaissance missions. It emphasizes the cost, reliability, longevity, and safety of the next generation of high-capacity batteries for applications where high energy density, minimum weight and size, and reliability in harsh conditions are

  9. Design of the Next Generation Nuclear Plant Graphite Creep Experiments for Irradiation in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover

    2009-05-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating six gas reactor graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These graphite irradiations are being accomplished to support development of the next generation reactors in the United States. The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain seven separate stacks of graphite specimens. Six of the specimen stacks will have half of their graphite specimens under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six stacks will be organized into pairs with a different compressive load being applied to the top half of each pair of specimen stacks. The seventh stack will not have a compressive load on the graphite specimens during irradiation. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be the capability of sampling the sweep gas effluent to determine if any

  10. Fuel irradiation research of Japan at OECD Halden Reactor Project. Achievement of joint researches between JAERI and other organizations in the period from 1994 to 1996

    Energy Technology Data Exchange (ETDEWEB)

    Uetsuka, Hiroshi; Nakamura, Jinichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kinoshita, Motoyasu [and others

    1998-01-01

    JAERI has performed cooperative researches with many Japanese agencies and companies by means of the Halden Boiling Heavy Water Reactor (HBWR) which is located at Halden in Norway. These cooperative researches are carried out based on the contracts of the cooperative researches, which are revised every three years, in accordance with the renewal of the participation of JAERI to the OECD Halden Reactor Project. This report summaries the objectives, contents and the outlines of the achievements of the cooperative researches during the three years from 1994 January to 1996 December. During the period, ten cooperative researches had been carried out, and two of them had finished during the period and other eight researches has been continued to the next three year period. There are many research items, and most of them are irradiation test researches of advanced fuel and cladding concerned with the high burnup utilization of LWR fuel or MOX fuel irradiation researches to prepare for the introduction of Plutonium utilization in LWRs. The researches of fuel irradiation usually take long time because of the characteristics of these kind of research work, and three years are usually not enough to obtain some achievements from the irradiation tests. Therefore, eight tests have been continued after the three year period. In this report, the achievements of the continued researches to the next three year period are not final one but a kind of progress report. (author)

  11. Fuel irradiation research of Japan at halden reactor. Achievement of cooperative researches between JAERI and several organizations in the period from 1997 to 1999

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-11-01

    JAERI has performed cooperative researches with several Japanese agencies and companies by means of the Halden Boiling Heavy Water Reactor (HBWR) which is located at Halden in Norway. These researches are carried out based on the contracts of the cooperative researches, which are revised every three years, in accordance with the renewal of the participation of JAERI to the OECD Halden Reactor Project. This report summaries the objectives, contents and the outlines of the achievements of the cooperative researches during the three years from 1997 January to 1999 December. During the period, nine cooperative researches had been carried out. Two of them had been completed and other seven researches has been continued to the next three years period. Most of them are irradiation test researches of advanced fuel and cladding in order to prepare the higher burnup utilization and introduction of LWR fuel and MOX fuel in LWRs of Japan. The researches of fuel irradiation usually take long time for preparing test and irradiation, then three years are usually not enough to obtain some achievements from the irradiation tests. Therefore, seven tests have been continued to the next three year period. In this report, the achievements of the continued researches to the next period are not final one but a kind of progress report. (author)

  12. Fuel-Cell Power Systems Incorporating Mg-Based H2 Generators

    Science.gov (United States)

    Kindler, Andrew; Narayan, Sri R.

    2009-01-01

    Two hydrogen generators based on reactions involving magnesium and steam have been proposed as means for generating the fuel (hydrogen gas) for such fuel-cell power systems as those to be used in the drive systems of advanced motor vehicles. The hydrogen generators would make it unnecessary to rely on any of the hydrogen storage systems developed thus far that are, variously, too expensive, too heavy, too bulky, and/or too unsafe to be practical. The two proposed hydrogen generators are denoted basic and advanced, respectively. In the basic hydrogen generator (see figure), steam at a temperature greater than or equals 330 C would be fed into a reactor charged with magnesium, wherein hydrogen would be released in the exothermic reaction Mg + H2O yields MgO + H2. The steam would be made in a flash boiler. To initiate the reaction, the boiler could be heated electrically by energy borrowed from a storage battery that would be recharged during normal operation of the associated fuel-cell subsystem. Once the reaction was underway, heat from the reaction would be fed to the boiler. If the boiler were made an integral part of the hydrogen-generator reactor vessel, then the problem of transfer of heat from the reactor to the boiler would be greatly simplified. A pump would be used to feed water from a storage tank to the boiler.

  13. Health assessment of gasoline and fuel oxygenate vapors: generation and characterization of test materials.

    Science.gov (United States)

    Henley, Michael; Letinski, Daniel J; Carr, John; Caro, Mario L; Daughtrey, Wayne; White, Russell

    2014-11-01

    In compliance with the Clean Air Act regulations for fuel and fuel additive registration, the petroleum industry, additive manufacturers, and oxygenate manufacturers have conducted comparative toxicology testing on evaporative emissions of gasoline alone and gasoline containing fuel oxygenates. To mimic real world exposures, a generation method was developed that produced test material similar in composition to the re-fueling vapor from an automotive fuel tank at near maximum in-use temperatures. Gasoline vapor was generated by a single-step distillation from a 1000-gallon glass-lined kettle wherein approximately 15-23% of the starting material was slowly vaporized, separated, condensed and recovered as test article. This fraction was termed vapor condensate (VC) and was prepared for each of the seven test materials, namely: baseline gasoline alone (BGVC), or gasoline plus an ether (G/MTBE, G/ETBE, G/TAME, or G/DIPE), or gasoline plus an alcohol (G/EtOH or G/TBA). The VC test articles were used for the inhalation toxicology studies described in the accompanying series of papers in this journal. These studies included evaluations of subchronic toxicity, neurotoxicity, immunotoxicity, genotoxicity, reproductive and developmental toxicity. Results of these studies will be used for comparative risk assessments of gasoline and gasoline/oxygenate blends by the US Environmental Protection Agency.

  14. Generation and reduction of nitrogen oxides in firing different kinds of fuel in a circulating fluidized bed

    Science.gov (United States)

    Munts, V. A.; Munts, Yu. G.; Baskakov, A. P.; Proshin, A. S.

    2013-11-01

    The processes through which nitrogen oxides are generated and reduced in the course of firing different kinds of fuel in a circulating fluidized bed are addressed. All experimental studies were carried by the authors on their own laboratory installations. To construct a model simulating the generation of nitrogen oxides, the fuel combustion process in a fluidized bed was subdivided into two stages: combustion of volatiles and combustion of coke residue. The processes through which nitrogen oxides are generated and reduced under the conditions of firing fuel with shortage of oxygen (which is one of efficient methods for reducing nitrogen oxide emissions in firing fuel in a fluidized bed) are considered.

  15. Influence of Proton Irradiation on Angular Dependence of Second Generation (2G)HTS

    Energy Technology Data Exchange (ETDEWEB)

    Shiroyanagi, Y.; Greene, G.; Gupta, R.; Sampson, W.

    2011-05-01

    In the Facility for Rare Isotope Beams (FRIB) the quadrupoles in the fragment separator are exposed to very high radiation and heat loads. High Temperature Superconductors (HTS) are a good candidate for these magnets because they can be used at {approx}30-50 K and tolerate higher heat generation than Nb-Ti magnets. Radiation damage studies of HTS wires are crucial to ensure that they will survive in a high radiation environment. HTS wires from two vendors were studied. Samples of 2G HTS wires from SuperPower and American Superconductor (ASC) were irradiated with a 42 {mu}A, 142 MeV proton beam from the Brookhaven Linac Isotope Producer (BLIP). The angular dependence of the critical current was measured in magnetic fields at 77K.

  16. Signal generation in highly irradiated silicon microstrip detectors for the ATLAS experiment

    CERN Document Server

    Ruggiero, G

    2003-01-01

    Silicon detectors are the most diffused tracking devices in High Energy Physics (HEP). The reason of such success can be found in the characteristics of the material together with the existing advanced technology for the fabrication of these devices. Nevertheless in many modem HEP experiments the observation of vary rare events require data taking at high luminosity with a consequent extremely intense hadron radiation field that damages the silicon and degrades the performance of these devices. In this thesis work a detailed study of the signal generation in microstrip detectors has been produced with a special care for the ATLAS semiconductor tracker geometry. This has required a development of an appropriate setup to perform measurements with Transient Current/ Charge Technique. This has allowed studying the evolution of the signal in several microstrips detector samples irradiated at fluences covering the range expected in the ATLAS Semiconductor Tracker. For a better understanding of these measurements a ...

  17. Stochastic model of wind-fuel cell for a semi-dispatchable power generation

    DEFF Research Database (Denmark)

    Alvarez-Mendoza, Fernanda; Bacher, Peder; Madsen, Henrik

    2017-01-01

    Hybrid systems are implemented to improve the efficiency of individual generation technologies by complementing each other. Intermittence is a challenge to overcome especially for renewable energy sources for electric generation, as in the case of wind power. This paper proposes a hybrid system...... for short-term wind power generation and electric generation as the outcome of the hybrid system. A method for a semi-dispatchable electric generation based on time series analysis is presented, and the implementation of wind power and polymer electrolyte membrane fuel cell models controlled by a model...... as an approach for reducing and overcoming the volatility of wind power, by implementing storage technology, forecasts and predictive control. The proposed hybrid system, which is suitable for the distributed generation level, consists of a wind generator, an electrolyzer, hydrogen storage and a polymer...

  18. Modeling the reaction kinetics of a hydrogen generator onboard a fuel cell -- Electric hybrid motorcycle

    Science.gov (United States)

    Ganesh, Karthik

    Owing to the perceived decline of the fossil fuel reserves in the world and environmental issues like pollution, conventional fuels may be replaced by cleaner alternative fuels. The potential of hydrogen as a fuel in vehicular applications is being explored. Hydrogen as an energy carrier potentially finds applications in internal combustion engines and fuel cells because it is considered a clean fuel and has high specific energy. However, at 6 to 8 per kilogram, not only is hydrogen produced from conventional methods like steam reforming expensive, but also there are storage and handling issues, safety concerns and lack of hydrogen refilling stations across the country. The purpose of this research is to suggest a cheap and viable system that generates hydrogen on demand through a chemical reaction between an aluminum-water slurry and an aqueous sodium hydroxide solution to power a 2 kW fuel cell on a fuel cell hybrid motorcycle. This reaction is essentially an aluminum-water reaction where sodium hydroxide acts as a reaction promoter or catalyst. The Horizon 2000 fuel cell used for this purpose has a maximum hydrogen intake rate of 28 lpm. The study focuses on studying the exothermic reaction between the reactants and proposes a rate law that best describes the rate of generation of hydrogen in connection to the surface area of aluminum available for the certain reaction and the concentration of the sodium hydroxide solution. Further, the proposed rate law is used in the simulation model of the chemical reactor onboard the hybrid motorcycle to determine the hydrogen flow rate to the fuel cell with time. Based on the simulated rate of production of hydrogen from the chemical system, its feasibility of use on different drive cycles is analyzed. The rate of production of hydrogen with a higher concentration of sodium hydroxide and smaller aluminum powder size was found to enable the installation of the chemical reactor on urban cycles with frequent stops and starts

  19. The generation of hourly diffuse irradiation: A model from the analysis of the fluctuation of global irradiance series

    Energy Technology Data Exchange (ETDEWEB)

    Posadillo, R.; Lopez Luque, R. [Grupo de Investigacion de Fisica para las Energias y Recursos Renovables, Dpto. de Fisica Aplicada, UCO, Edificio C2 Campus de Rabanales, 14071 Cordoba (Spain)

    2010-04-15

    An analysis of models for the estimation of hourly diffuse irradiation based on the interrelations between the hourly diffuse fraction k{sub d} and the hourly clearness index k{sub t}, has concluded that k{sub t} is not a sufficient variable for parametrizing the effect of clouds on diffuse irradiation. A detailed study of the dispersion recorded by this diffuse component for a specific clearness index under partly cloudy sky conditions has led to analyzing how the variability in the instantaneous clearness index influences this dispersion. The data sets correspond to 10 years of hourly and instantaneous value records of global and diffuse radiation collected in Cordoba, Spain. In addition to the inclusion of the sine of solar elevation as a variable into the k{sub d}-k{sub t} correlations, this model propose the inclusion of others parameters related to the variability in the normalized clearness index within an hour and with the fluctuations presented by the time series of the instantaneous values of that index. Also presented is the implementation of an algorithm permitting both the determination of the hourly diffuse irradiation and the discrimination between the different sky conditions in those situations known by the designation partly cloudy sky. (author)

  20. Oxidation-extraction spectrometry of reactive oxygen species (ROS) generated by chlorophyllin magnesium (Chl-Mg) under ultrasonic irradiation.

    Science.gov (United States)

    Guo, Yuwei; Cheng, Chunping; Wang, Jun; Jin, Xudong; Liu, Bin; Wang, Zhiqiu; Gao, Jingqun; Kang, Pingli

    2011-09-01

    In order to examine the mechanism and process of sonodynamic reaction, the chlorophyllin magnesium (Chl-Mg) acting as a sonosensitizer was irradiated by ultrasound, and the generation of reactive oxygen species (ROS) were detected by the method of oxidation-extraction spectrometry (OES). That is, under ultrasonic irradiation in the presence of Chl-Mg, the 1,5-diphenyl carbazide (DPCI) is oxidized by generated ROS into 1,5-diphenyl carbazone (DPCO), which can be extracted by mixed organic solvent and display a obvious visible absorption at 563 nm wavelength. Besides, the generation conditions of ROS were also reviewed. The results demonstrated that the quantities of generated ROS increased with the increase of ultrasonic irradiation time, Chl-Mg concentration and DPCI concentration. Finally, several radical scavengers (l-Histidine (His), 2,6-Di-tert-butyl-methylphenol (BHT) and Vitamin C (VC)) were used to determine the kind of the generated ROS. It was found that at least the hydroxyl radical (OH) and singlet oxygen (1O2) were generated in the presence of Chl-Mg under ultrasonic irradiation. It is wish that this paper might offer some valuable references for the study on the mechanism of SDT and the application of Chl-Mg in tumor treatment.

  1. Intelligent Power Management of hybrid Wind/ Fuel Cell/ Energy Storage Power Generation System

    Directory of Open Access Journals (Sweden)

    A. Hajizadeh

    2013-12-01

    Full Text Available This paper presents an intelligent power management strategy for hybrid wind/ fuel cell/ energy storage power generation system. The dynamic models of wind turbine, fuel cell and energy storage have been used for simulation of hybrid power system. In order to design power flow control strategy, a fuzzy logic control has been implemented to manage the power between power sources. The optimal operation of the hybrid power system is a main goal of designing power management strategy. The hybrid power system is simulated in MATLAB/ SIMIULINK environment and different operating conditions have been considered to evaluate the response of power management strategy.

  2. Environmental benchmarking of the largest fossil-fueled electricity generating plants in the U.S

    Science.gov (United States)

    Sarkis, Joseph

    2004-02-01

    Environmental management, to be effective, requires performance evaluation and process improvement. This is especially the case in fossil-fueled electricity generating plants. Although eco-efficient management of these types of organizations are critical to local, national and global environmental issues, few studies have focused on performance measurement and eco-efficiency improvements in this industry. This study evaluates the eco-efficiencies of the top 100 major U.S. fossil-fueled electricity generating plants from 1998 data. Using a multi-criteria non-parametric productivity model (data envelopment analysis) efficiency scores are determined. These efficiency scores are treated by a clustering method in identifying benchmarks for improving poorly performing plants. Efficiency measures are based on three resource input measures including boiler generating capacity, total fuel heat used, and total generator capacity, and four output measures including actual energy generated, SO2, NOx, and CO2 emissions. The purpose of this paper is two-fold, to introduce the methodology"s application to eco-efficiency performance measurement and show some characteristics of the benchmarked plants and groups.

  3. 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.

  4. Depletion analysis and sensitivity study of PHENIX fuel-irradiation experiments

    Energy Technology Data Exchange (ETDEWEB)

    Biswas, D.; Kallfelz, J.M.; White, J.R.

    1981-01-01

    The experimental results are in the form of various U and Pu atom density ratios (R/sub E/) and burnup (BU) values. The results were for samples irradiated during the first three cycles in the central zone of PHENIX. The time-dependent sensitivity study was performed with the depletion generalized perturbation code DEPTH-CHARGE, to investigate the sensitivity of R/sub E/ to cross sections and to absolute flux level changes. The depletion analysis was performed using ENDS/B-IV data, an (R-Z) model, and the VENTURE Code system. 2 tables.

  5. Performance of Microbial Fuel Cell for Wastewater Treatment and Electricity Generation

    Directory of Open Access Journals (Sweden)

    Z Yavari

    2013-06-01

    Full Text Available Renewable energy will have an important role as a resource of energy in the future. Microbial fuel cell (MFC is a promising method to obtain electricity from organic matter andwastewater treatment simultaneously. In a pilot study, use of microbial fuel cell for wastewater treatment and electricity generation investigated. The bacteria of ruminant used as inoculums. Synthetic wastewater used at different organic loading rate. Hydraulic retention time was aneffective factor in removal of soluble COD and more than 49% removed. Optimized HRT to achieve the maximum removal efficiency and sustainable operation could be regarded 1.5 and 2.5 hours. Columbic efficiency (CE affected by organic loading rate (OLR and by increasing OLR, CE reduced from 71% to 8%. Maximum voltage was 700mV. Since the microbial fuel cell reactor considered as an anaerobic process, it may be an appropriate alternative for wastewater treatment

  6. Committing to coal and gas: Long-term contracts, regulation, and fuel switching in power generation

    Science.gov (United States)

    Rice, Michael

    Fuel switching in the electricity sector has important economic and environmental consequences. In the United States, the increased supply of gas during the last decade has led to substantial switching in the short term. Fuel switching is constrained, however, by the existing infrastructure. The power generation infrastructure, in turn, represents commitments to specific sources of energy over the long term. This dissertation explores fuel contracts as the link between short-term price response and long-term plant investments. Contracting choices enable power plant investments that are relationship-specific, often regulated, and face uncertainty. Many power plants are subject to both hold-up in investment and cost-of-service regulation. I find that capital bias is robust when considering either irreversibility or hold-up due to the uncertain arrival of an outside option. For sunk capital, the rental rate is inappropriate for determining capital bias. Instead, capital bias depends on the regulated rate of return, discount rate, and depreciation schedule. If policies such as emissions regulations increase fuel-switching flexibility, this can lead to capital bias. Cost-of-service regulation can shorten the duration of a long-term contract. From the firm's perspective, the existing literature provides limited guidance when bargaining and writing contracts for fuel procurement. I develop a stochastic programming framework to optimize long-term contracting decisions under both endogenous and exogenous sources of hold-up risk. These typically include policy changes, price shocks, availability of fuel, and volatility in derived demand. For price risks, the optimal contract duration is the moment when the expected benefits of the contract are just outweighed by the expected opportunity costs of remaining in the contract. I prove that imposing early renegotiation costs decreases contract duration. Finally, I provide an empirical approach to show how coal contracts can limit

  7. Swelling of U(Mo)-Al(Si) dispersion fuel under irradiation - Non-destructive analyses of the LEONIDAS E-FUTURE plates

    Science.gov (United States)

    Van den Berghe, S.; Parthoens, Y.; Charollais, F.; Kim, Y. S.; Leenaers, A.; Koonen, E.; Kuzminov, V.; Lemoine, P.; Jarousse, C.; Guyon, H.; Wachs, D.; Keiser, D., Jr.; Robinson, A.; Stevens, J.; Hofman, G.

    2012-11-01

    In the framework of the elimination of High-Enriched Uranium (HEU) from the civil circuit, the search for an appropriate fuel to replace the high-enriched research reactor fuel in those reactors that currently still require it for their operation has led to the development of a U-7 wt.%Mo alloy based dispersion fuel with an Al-Si matrix. The European LEONIDAS program, joining SCK•CEN, ILL, CEA and AREVA-CERCA, is aimed at the qualification of such a fuel for the use in high power conditions. The first experiment of the program, designated E-FUTURE, was performed to select the appropriate matrix Si concentration and fuel plate post-production heat treatment parameters for further qualification. It consisted of the irradiation of four distinct (4% and 6% Si, 3 different heat treatments) full size, flat fuel plates in the BR2 reactor. The irradiation conditions were relatively severe: 470 W/cm2 peak BOL power, with a ˜70% 235U peak burnup.

  8. Ultra-Deep Adsorptive Desulfurization of Light-Irradiated Diesel Fuel over Supported TiO2-CeO2 Adsorbents

    Energy Technology Data Exchange (ETDEWEB)

    Xiao, Jing; Wang, Xiaoxing; Chen, Yongsheng; Fujii, Mamoru; Song, Chunshan [SCUT-China; (Penn)

    2014-02-13

    This study investigates ultra-deep adsorptive desulfurization (ADS) from light-irradiated diesel fuel over supported TiO2–CeO2 adsorbents. A 30-fold higher desulfurization capacity of 95 mL of fuel per gram of adsorbent (mL-F/g-sorb) or 1.143 mg of sulfur per gram of adsorbent (mg-S/g-sorb) was achieved from light-irradiated fuel over the original low-sulfur fuel containing about 15 ppm by weight (ppmw) of sulfur. The sulfur species on spent TiO2–CeO2/MCM-48 adsorbent was identified by sulfur K-edge XANES as sulfones and the adsorption selectivity to different compounds tested in a model fuel decreases in the order of indole > dibenzothiophenesulfone → dibenzothiophene > 4-methyldibenzothiophene > benzothiophene > 4,6-dimethyldibenzothiophene > phenanthrene > 2-methylnaphthalene ~ fluorene > naphthalene. The results suggest that during ADS of light-irradiated fuel, the original sulfur species were chemically transformed to sulfones, resulting in the significant increase in desulfurization capacity. For different supports for TiO2–CeO2 oxides, the ADS capacity increases with a decrease in the point of zero charge (PZC) value; for silica-supported TiO2–CeO2 oxides (the lowest PZC value of 2–4) with different surface areas, the ADS capacity increases monotonically with increasing surface area. The supported TiO2–CeO2/MCM-48 adsorbent can be regenerated using oxidative air treatment. The present study provides an attractive new path to achieve ultraclean fuel more effectively.

  9. Electricity Generation from Organic Matters in Biocatalyst-Based Microbial Fuel Cells (MFCs)

    DEFF Research Database (Denmark)

    Min, Booki; Zhang, Yifeng; Angelidaki, Irini

    Microbial fuel cells (MFCs) are a novel technology for converting organic matter directly to electricity via biocatalytic reactions by microorganisms. MFCs can also be used for wastewater treatment by the oxidations of organic pollutants during the electricity generation. Several factors for opti......Microbial fuel cells (MFCs) are a novel technology for converting organic matter directly to electricity via biocatalytic reactions by microorganisms. MFCs can also be used for wastewater treatment by the oxidations of organic pollutants during the electricity generation. Several factors...... for optimum power generation in MFC have been investigated at previous studies. A submersible microbial fuel cell (SMFC), which is a novel configuration, was developed by immersing an anode electrode and a cathode chamber in an anaerobic reactor. Domestic wastewater without any amendments was used...... as the medium and the inoculum in the experiments. The SMFC could successfully generate a stable voltage of 0.428±0.003V with a fixed 470Ω resistor from acetate. From the polarization test, the maximum power density of 204mWm−2 was obtained at current density of 595mAm−2 (external resistance = 180Ω). The power...

  10. Experimental Analysis of a Small Generator set Operating on Dual Fuel Diesel-Ethanol

    Directory of Open Access Journals (Sweden)

    Marcel Alex Vailatti

    2017-08-01

    Full Text Available This work aims to analyze the operation of a generator set on single fuel mode with diesel oil, and on dual fuel mode using diesel–ethanol blends. The engine used to realize the experimental analysis was a diesel cycle model, single cylinder, direct injection, air refrigerated and coupled to a three-phase electric generator, whose set capacity was 8.0 kVA. The generated electric energy was dissipated in electrical resistances inside a reservoir with running water. Fuels were blended in different volumetric ratios, using a small portion of vegetable castor oil to promote the homogenization. The percentages of substitutions of diesel oil were by 10% to 50%, increasing by 10% the replacement for each sample. Also, the engine was operated with 100% substitution of diesel oil, i.e., for this condition, the samples were composed of ethanol/castor oil 90/10 (volume/volume, 80/20 and 75/25. The blends of diesel and ethanol did not obtain good performance, mainly in taxes of substitution above 40%, causing combustion failures, operational instability, and increase of fuel consumption, although it has achieved a greatly reduction on opacity percentages. The blends with 100% of substitution of diesel oil obtained good performance except to blend with 90% ethanol, where occurred combustion failures, which caused operational instability. To these conditions, the results achieved are increase of consumption by 17%, decrease of opacity by 79%, decrease of exhaust gas temperature by 3.5% and increase of engine thermal efficiency by 1.3%. At the ethanol – castor oil blends there was a decrease in the percentage of opacity by 96%, decrease of exhaust gas temperature by 17.6%, with a minimum of operational irregularities, although fuel consumption has increased by 52.4% and the engine thermal efficiency has decreased almost 1.7%.

  11. [Power generation from glucose and nitrobenzene degradation using the microbial fuel cell].

    Science.gov (United States)

    Li, Jie; Liu, Guang-Li; Zhang, Ren-Duo; Luo, Yong; Zhang, Cui-Ping; Li, Ming-Chen; Quan, Xiang-Chun

    2010-11-01

    By constructing a dual-chamber microbial fuel cell (MFC), experiments were carried out using an initial glucose concentration of 1 000 mg/L with different nitrobenzene (NB) concentrations (0, 50, 150 and 250 mg/L) as the MFC's fuel. Results showed that with an external resistance of 1 000 omega, the initial glucose concentration of 1 000 mg/L and the initial NB concentrations of 0, 50, 150, 250 mg/L, the operation periods were 55.7, 51.6, 45.9 and 32.2 h, respectively, the maximum voltage outputs were 670, 597, 507, and 489 mV, the maximum volumetric power densities were 28.57, 20.42, 9.29, and 8.47 W/m3, and the electric charges were 65.10, 43.50, 35.48, and 30.32 C. The MFC could use the NB and glucose mixtures as fuel and generated stable electricity outputs. The degradation rates of NB in the MFC in all cases reached up to 100% and COD removals in the MFC were 87% - 98%. However, the electricity generation was negligible when using 250 mg/L NB as the sole fuel. Denaturing gradient gel electrophoresis (DGGE) profiles demonstrated that the presence of NB resulted in changes of the dominant bacterial species on the electrodes.

  12. MODELING, SIMULATON AND SIZING OF PHOTOVOLTAIC/WIND/FUEL CELL HYBRID GENERATION SYSTEM

    Directory of Open Access Journals (Sweden)

    Dr.S.LATHA

    2012-05-01

    Full Text Available The depleting fossil fuel reserves and increasing concern towards global warming have created the need to surge for the alternative power generation options. Renewable energy sources like Wind, Solar-PV, Biomass and fuel cells are gaining prominence nowadays, as they are more energy efficient, reduce pollution and also they serveas a promising solution to the toughest energy crisis faced during the recent years. This paper focuses on the modeling and simulation of solar – photovoltaic, wind and fuel cell hybrid energy systems using MATLAB/Simulink software. The intermittent nature of solar and wind energy sources make them unreliable. Hence Maximum Power Point Tracking (MPPT is used to extract maximum power from the wind and sunwhen it is available. The standard perturb and observe method of MPPT is used for the PV system and for the wind generation system. The simulation results of the PV/Wind /Fuel cell hybrid system are presented in graph showing the effectiveness of the proposed system model. Also, hardware implementation of microcontroller based MPPT for solar-PV alone and unit sizing of the hybrid system for the PG simulation lab in EEE Dept. of Thiagarajar College of Engineering is depicted in the paper.

  13. Effects of water-emulsified fuel on a diesel engine generator's thermal efficiency and exhaust.

    Science.gov (United States)

    Syu, Jin-Yuan; Chang, Yuan-Yi; Tseng, Chao-Heng; Yan, Yeou-Lih; Chang, Yu-Min; Chen, Chih-Chieh; Lin, Wen-Yinn

    2014-08-01

    Water-emulsified diesel has proven itself as a technically sufficient improvement fuel to improve diesel engine fuel combustion emissions and engine performance. However, it has seldom been used in light-duty diesel engines. Therefore, this paper focuses on an investigation into the thermal efficiency and pollution emission analysis of a light-duty diesel engine generator fueled with different water content emulsified diesel fuels (WD, including WD-0, WD-5, WD-10, and WD-15). In this study, nitric oxide, carbon monoxide, hydrocarbons, and carbon dioxide were analyzed by a vehicle emission gas analyzer and the particle size and number concentration were measured by an electrical low-pressure impactor. In addition, engine loading and fuel consumption were also measured to calculate the thermal efficiency. Measurement results suggested that water-emulsified diesel was useful to improve the thermal efficiency and the exhaust emission of a diesel engine. Obviously, the thermal efficiency was increased about 1.2 to 19.9%. In addition, water-emulsified diesel leads to a significant reduction of nitric oxide emission (less by about 18.3 to 45.4%). However the particle number concentration emission might be increased if the loading of the generator becomes lower than or equal to 1800 W. In addition, exhaust particle size distributions were shifted toward larger particles at high loading. The consequence of this research proposed that the water-emulsified diesel was useful to improve the engine performance and some of exhaust emissions, especially the NO emission reduction. Implications: The accumulated test results provide a good basis to resolve the corresponding pollutants emitted from a light-duty diesel engine generator. By measuring and analyzing transforms of exhaust pollutant from this engine generator, the effects of water-emulsified diesel fuel and loading on emission characteristics might be more clear. Understanding reduction of pollutant emissions during the use

  14. Separation of actinides from irradiated An–Zr based fuel by electrorefining on solid aluminium cathodes in molten LiCl–KCl

    Energy Technology Data Exchange (ETDEWEB)

    Souček, P., E-mail: Pavel.Soucek@ec.europa.eu [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, 76125 Karlsruhe (Germany); Murakami, T. [Central Research Institute of Electric Power Industry (CRIEPI), Komae-shi, Tokyo 201-8511 (Japan); Claux, B.; Meier, R.; Malmbeck, R. [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, 76125 Karlsruhe (Germany); Tsukada, T. [Central Research Institute of Electric Power Industry (CRIEPI), Komae-shi, Tokyo 201-8511 (Japan); Glatz, J.-P. [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Postfach 2340, 76125 Karlsruhe (Germany)

    2015-04-15

    Highlights: • Electrorefining process in molten LiCl-KCl using solid Al electrodes was demonstrated. • High separation factors of actinides over lanthanides were achieved. • Efficient recovery of actinides from irradiated nuclear fuel was achieved. • Uniform, dense and well adhered deposits were obtained and characterised. • Kinetic parameters of actinide–aluminium alloy formation were evaluated. - Abstract: An electrorefining process for metallic spent nuclear fuel treatment is being investigated in ITU. Solid aluminium cathodes are used for homogeneous recovery of all actinides within the process carried out in molten LiCl–KCl eutectic salt at a temperature of 500 °C. As the selectivity, efficiency and performance of solid Al has been already shown using un-irradiated An–Zr alloy based test fuels, the present work was focused on laboratory-scale demonstration of the process using irradiated METAPHIX-1 fuel composed of U{sub 67}–Pu{sub 19}–Zr{sub 10}–MA{sub 2}–RE{sub 2} (wt.%, MA = Np, Am, Cm, RE = Nd, Ce, Gd, Y). Different electrorefining techniques, conditions and cathode geometries were used during the experiment yielding evaluation of separation factors, kinetic parameters of actinide–aluminium alloy formation, process efficiency and macro-structure characterisation of the deposits. The results confirmed an excellent separation and very high efficiency of the electrorefining process using solid Al cathodes.

  15. Simulation of transient heat transfer in MACSTOR/KN-400 module storing irradiated CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sabourin, G. [Atomic Energy of Canada Limited, Montreal, Quebec (Canada); Lee, K.-H.; Yoon, J.-H.; Choi, B.-I.; Lee, H.-Y.; Song, M.-J. [KHNP, Nuclear Environment Technology Inst., Taejon (Korea, Republic of)

    2004-07-01

    Korea Hydro and Nuclear Power (KHNP), in collaboration with Atomic Energy of Canada Limited (AECL), are developing a new module for the dry storage of spent fuel from the four CANDU 6 nuclear reactors at the Wolsong site in South Korea, the MACSTOR/KN-400. The simulation of transient conditions for AECL's spent fuel dry storage systems, presented in this paper, has not been performed before and is considered a major achievement of the present work. In a fist step, CATHENA was compared to MACSTOR-200 temperature measurements and the accuracy of the results were very good. In a second step, CATHENA was applied to the MACSTOR/KN-400. Four cases were performed for the MACSTOR/KN-400: Off-normal cases in summer and winter and reduced air flow cases in summer and winter. The maximum local concrete temperatures were predicted to be 63{sup o}C for the off-normal case and 65{sup o}C in the reduced air flow case. The maximum temperature gradients in the concrete are predicted to be 28{sup o}C for the off-normal case and 30{sup o}C in the reduced air flow case, incorporating a 3{sup o}C uncertainty. This paper shows that the maximum temperature for the module is expected to meet the temperature limitations of appropriate standards. (author)

  16. Blue-Violet Light Irradiation Dose Dependently Decreases Carotenoids in Human Skin, Which Indicates the Generation of Free Radicals

    Directory of Open Access Journals (Sweden)

    Staffan Vandersee

    2015-01-01

    Full Text Available In contrast to ultraviolet and infrared irradiation, which are known to facilitate cutaneous photoaging, immunosuppression, or tumour emergence due to formation of free radicals and reactive oxygen species, potentially similar effects of visible light on the human skin are still poorly characterized. Using a blue-violet light irradiation source and aiming to characterize its potential influence on the antioxidant status of the human skin, the cutaneous carotenoid concentration was measured noninvasively in nine healthy volunteers using resonance Raman spectroscopy following irradiation. The dose-dependent significant degradation of carotenoids was measured to be 13.5% and 21.2% directly after irradiation at 50 J/cm² and 100 J/cm² (P<0.05. The irradiation intensity was 100 mW/cm². This is above natural conditions; the achieved doses, though, are acquirable under natural conditions. The corresponding restoration lasted 2 and 24 hours, respectively. The degradation of cutaneous carotenoids indirectly shows the amount of generated free radicals and especially reactive oxygen species in human skin. In all volunteers the cutaneous carotenoid concentration dropped down in a manner similar to that caused by the infrared or ultraviolet irradiations, leading to the conclusion that also blue-violet light at high doses could represent a comparably adverse factor for human skin.

  17. Solid oxide fuel cells powered by biomass gasification for high efficiency power generation

    DEFF Research Database (Denmark)

    Gadsbøll, Rasmus Østergaard; Thomsen, Jesper; Bang-Møller, Christian

    2017-01-01

    efficiencies, flexibility and possibly costs of current biomass power generating systems, a power plant concept combining solid oxide fuel cells (SOFC) and gasification is investigated experimentally. The aim of the study is to examine the commercial operation system potential of these two technologies......Increased use of bioenergy is a very cost-effective and flexible measure to limit changes in the climate and the infrastructure. One of the key technologies toward a higher implementation of biomass is thermal gasification, which enables a wide span of downstream applications. In order to improve....... Investigations are done by combining the commercial TwoStage Viking gasifier developed at the Technical University of Denmark and a state-of-the-art SOFC stack from Topsoe Fuel Cell for high efficiency power generation. A total of 5 tests were performed including polarization tests at various gas flows to study...

  18. Utilization of hydrolysate from lignocellulosic biomass pretreatment to generate electricity by enzymatic fuel cell system.

    Science.gov (United States)

    Kim, Sung Bong; Kim, Dong Sup; Yang, Ji Hyun; Lee, Junyoung; Kim, Seung Wook

    2016-04-01

    The waste hydrolysate after dilute acid pretreatment (DAP) of lignocellulosic biomass was utilized to generate electricity using an enzymatic fuel cell (EFC) system. During DAP, the components of biomass containing hemicellulose and other compounds are hydrolyzed, and glucose is solubilized into the dilute acid solution, called as the hydrolysate liquid. Glucose oxidase (GOD) and laccase (Lac) were assembled on the electrode of the anode and cathode, respectively. Cyclic voltammetry (CV) and electrochemical impedance spectroscopy (EIS) were measured, and the maximum power density was found to be 1.254×10(3) μW/cm(2). The results indicate that the hydrolysate from DAP is a reliable electrolyte containing the fuel of EFC. Moreover, the impurities in the hydrolysate such as phenols and furans slightly affected the charge transfer on the surface of the electrode, but did not affect the power generation of the EFC system in principal.

  19. Effects of furan derivatives and phenolic compounds on electricity generation in microbial fuel cells

    Science.gov (United States)

    Catal, Tunc; Fan, Yanzhen; Li, Kaichang; Bermek, Hakan; Liu, Hong

    Lignocellulosic biomass is an attractive fuel source for MFCs due to its renewable nature and ready availability. Furan derivatives and phenolic compounds could be potentially formed during the pre-treatment process of lignocellulosic biomass. In this study, voltage generation from these compounds and the effects of these compounds on voltage generation from glucose in air-cathode microbial fuel cells (MFCs) were examined. Except for 5-hydroxymethyl furfural (5-HMF), all the other compounds tested were unable to be utilized directly for electricity production in MFCs in the absence of other electron donors. One furan derivate, 5-HMF and two phenolic compounds, trans-cinnamic acid and 3,5-dimethoxy-4-hydroxy-cinnamic acid did not affect electricity generation from glucose at a concentration up to 10 mM. Four phenolic compounds, including syringaldeyhde, vanillin, trans-4-hydroxy-3-methoxy, and 4-hydroxy cinnamic acids inhibited electricity generation at concentrations above 5 mM. Other compounds, including 2-furaldehyde, benzyl alcohol and acetophenone, inhibited the electricity generation even at concentrations less than 0.2 mM. This study suggests that effective electricity generation from the hydrolysates of lignocellulosic biomass in MFCs may require the employment of the hydrolysis methods with low furan derivatives and phenolic compounds production, or the removal of some strong inhibitors prior to the MFC operation, or the improvement of bacterial tolerance against these compounds through the enrichment of new bacterial cultures or genetic modification of the bacterial strains.

  20. Effects of furan derivatives and phenolic compounds on electricity generation in microbial fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Catal, Tunc [Department of Biological and Ecological Engineering, Oregon State University, 116 Gilmore Hall, Corvallis, OR 97331 (United States); Department of Wood Science and Engineering, Oregon State University, 102 97331, Corvallis, OR (United States); Department of Molecular Biology and Genetics, Istanbul Technical University, 34469-Maslak, Istanbul (Turkey); Fan, Yanzhen; Liu, Hong [Department of Biological and Ecological Engineering, Oregon State University, 116 Gilmore Hall, Corvallis, OR 97331 (United States); Li, Kaichang [Department of Wood Science and Engineering, Oregon State University, 102 97331, Corvallis, OR (United States); Bermek, Hakan [Department of Molecular Biology and Genetics, Istanbul Technical University, 34469-Maslak, Istanbul (Turkey)

    2008-05-15

    Lignocellulosic biomass is an attractive fuel source for MFCs due to its renewable nature and ready availability. Furan derivatives and phenolic compounds could be potentially formed during the pre-treatment process of lignocellulosic biomass. In this study, voltage generation from these compounds and the effects of these compounds on voltage generation from glucose in air-cathode microbial fuel cells (MFCs) were examined. Except for 5-hydroxymethyl furfural (5-HMF), all the other compounds tested were unable to be utilized directly for electricity production in MFCs in the absence of other electron donors. One furan derivate, 5-HMF and two phenolic compounds, trans-cinnamic acid and 3,5-dimethoxy-4-hydroxy-cinnamic acid did not affect electricity generation from glucose at a concentration up to 10 mM. Four phenolic compounds, including syringaldeyhde, vanillin, trans-4-hydroxy-3-methoxy, and 4-hydroxy cinnamic acids inhibited electricity generation at concentrations above 5 mM. Other compounds, including 2-furaldehyde, benzyl alcohol and acetophenone, inhibited the electricity generation even at concentrations less than 0.2 mM. This study suggests that effective electricity generation from the hydrolysates of lignocellulosic biomass in MFCs may require the employment of the hydrolysis methods with low furan derivatives and phenolic compounds production, or the removal of some strong inhibitors prior to the MFC operation, or the improvement of bacterial tolerance against these compounds through the enrichment of new bacterial cultures or genetic modification of the bacterial strains. (author)

  1. Chemical states of fission products in irradiated (U 0.3Pu 0.7)C 1+ x fuel at high burn-ups

    Science.gov (United States)

    Agarwal, Renu; Venugopal, V.

    2006-12-01

    The chemical states of fission products have been theoretically determined for the irradiated carbide fuel of Fast Breeder Test Reactor (FBTR) at Kalpakkam, India, at different burn-ups. The SOLGASMIX-PV computer code was used to determine the equilibrium chemical composition of the fuel. The system was assumed to be composed of a gaseous phase at one atmosphere pressure, and various solid phases. The distribution of elements in these phases and their chemical states at different temperatures were calculated as a function of burn-up. The FBTR fuel, (U 0.3Pu 0.7)C 1+ x, was loaded with C/M values in the range, 1.03-1.06. The present calculations indicated that even for the lowest starting C/M of 1.03 in the FBTR fuel, the liquid metal phase of (U, Pu), should not appear at a burn-up as high as 150 GWd/t.

  2. Effects of low doses of short-term gamma irradiation on growth and development through two generations of Pisum sativum

    Energy Technology Data Exchange (ETDEWEB)

    Zaka, R.; Misset, M.T. [UMR-CNRS 6553 Ecobio, Equipe Evolution des Populations et des Especes, Universite de Rennes 1, Campus de Beaulieu Bat. 14, Rennes Cedex F 35042 (France); Chenal, C. [Laboratoire de Radiobiologie, Universite de Rennes 1, Centre Regional de Lutte contre le Cancer, Rennes Cedex F 35062 (France)

    2004-03-29

    The effects of short-term gamma radiation on pea plants were investigated by exposing 5-day-old seedlings with doses ranging from 0 to 60 Gy, and studying plant growth and development over two generations after irradiation. Doses higher than 6 Gy significantly inhibited the G1 plant growth and productivity, and no seedling survived irradiation with 40 Gy and above. These effects were transmitted and were even more severe in the next generation, G2. Irradiated G1 ({>=}10 Gy) and G2 ({>=}0.4 Gy) plants were significantly smaller than controls. The mean number of pods produced per plant was reduced by at least 20% at all doses in both G1 and G2. In parallel, the mean numbers of ovules and normally developed seeds per pod were significantly reduced after 10 Gy in G1 and after 0.4 Gy in G2, leading to a significant drop in seed production. This effect was correlated with a linear decrease in male fertility linked to abnormal meiosis (tetrads with micronuclei) as a function of doses from 0 to 10 Gy, in G1 and G2 plants. These long-term changes in plant development demonstrate a genomic instability induced by irradiation. However, there were neither quantitative nor qualitative changes in storage proteins in G1 seeds at any of the irradiation doses tested from 0 to 10 Gy.

  3. Systems of three generation using fuel cells; Sistemas de trigeracao com o uso de celulas a combustivel

    Energy Technology Data Exchange (ETDEWEB)

    San Martin Diaz, Jose Ignacio; Aperribay Maiztegui, Victor; San Martin Diaz, Jose Javier [Escuela Universitaria de Ingenieria Tecnica Industrial de Eibar, Guipuzcoa, Pais Vasco (Spain); Zamora Belver, Inmaculada; Eguia Lopez, Pablo [Escuela Tecnica Superior de Ingenieria de Bilbao (Spain)

    2010-10-15

    The three generation includes production processes and simultaneous use oe electric power, heat and cold from a fuel source providing global high energy efficiency, reduction of emission of gases in the atmosphere, losses of investments, and more reliability in the supply. This paper approaches the use of fuel cells for three generation, and presents different subsystems of generation with the main characteristics and applications. (author)

  4. Status and Prognosis of Future-Generation Photoconversion to Photovoltaics and Solar Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Beard, Matthew C.; Blackburn, Jeffrey L.; Johnson, Justin C.; Rumbles, Garry

    2016-08-12

    Professor Arthur J. Nozik has fought for, inspired, cajoled, and led a generation of scientists in the pursuit of the science of solar photoconversion, photovoltaics, and solar fuels. On March 25th, 2016, a group of former colleagues, co-workers, and friends met to recognize Prof. Nozik's contribution to their work, excellence in science, and life. While the event was a celebration of his many scientific contributions, it served mostly to honor his leadership and vision.

  5. Bioinspired Nanosucker Array for Enhancing Bioelectricity Generation in Microbial Fuel Cells.

    Science.gov (United States)

    Wang, Wei; You, Shijie; Gong, Xiaobo; Qi, Dianpeng; Chandran, Bevita K; Bi, Lanpo; Cui, Fuyi; Chen, Xiaodong

    2016-01-13

    A bioinspired active anode with a suction effect is demonstrated for microbial fuel cells by constructing polypyrrole (PPy) nanotubular arrays on carbon textiles. The oxygen in the inner space of the nanosucker can be depleted by micro-organisms with the capability of facul-tative respiration, forming a vacuum, which then activates the electrode to draw the microorganism by suction and thus improve the bioelectricity generation. © 2015 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  6. Performance of Klebsiella oxytoca to generate electricity from POME in microbial fuel cell

    OpenAIRE

    Islam Md. Amirul; Rahman Maksudur; Yousuf Abu; Cheng Chin Kui; Wai Woon Chee

    2016-01-01

    This study is aimed to evaluate the electricity generation from microbial fuel cell (MFC) and to analyze the microbial community structure of city wastewater and anaerobic sludge to enhance the MFC performance. MFCs, enriched with palm oil mill effluent (POME) were employed to harvest electricity by innoculating of Klebsiella oxytoca, collected from city wastewater and other microbes from anaerobic sludge (AS). The MFC showed maximum power density of 207.28 mW/m3 with continuous feeding of PO...