WorldWideScience

Sample records for fuel development progress

  1. AECL's progress in DUPIC fuel development

    International Nuclear Information System (INIS)

    Sullivan, J.D.; Ryz, M.A.; Lee, J.W.

    1997-01-01

    Previous papers described progress in choosing a fabrication route for the DUPIC (Direct Use of Spent PWR Fuel in CANDU) fuel cycle [1], details of the OREOX (Oxidation Reduction of Oxide fuel) process, and preliminary results of out-cell and small-scale in-cell experiments [2]. AECL's project to develop the DUPIC fuel cycle has now progressed to the stage of fabricating DUPIC fuel elements for irradiation testing in a research reactor. Because of the high radiation fields around the spent PWR fuel, all work is being done in hot cells. The equipment used for fabrication of the DUPIC fuel elements is described in this paper. The commissioning, in-cell installation and current status of the fabrication process are also described and plans for the completion of this phase of the DUPIC project are outlined. The goal of this phase of the project is demonstration of the technical feasibility of the DUPIC fuel cycle. (author)

  2. AECL's progress in developing the DUPIC fuel fabrication process

    International Nuclear Information System (INIS)

    Sullivan, J.D.; Cox, D.S.

    1995-01-01

    Spent Pressurized Water Reactor (PWR) fuel can be used directly in CANDU reactors without the need for wet chemical reprocessing or reenrichment. Considerable experimental progress has been made in verifying the practicality of this fuel cycle, including hot-cell experiments using spent PWR fuels and out-cell trials using surrogate fuels. This paper describes the current status of these experiments. (author)

  3. Quarterly Progress Report Fuels Development Operation: October - December 1959

    Energy Technology Data Exchange (ETDEWEB)

    Cadwell, J. J. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation; Tobin, J. C. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Physical Metallurgy; Minor, J. E. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Fuel Element Design; Evans, E. A. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Ceramic Fuels Development; Bush, S. H. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Fuels Fabrication Development

    1960-01-15

    The present Quarterly Report is the continuation of a series issued by the new Fuels Development operation. Reports in this series combine portions of the quarterly reports by the former Metallurgy Research and Fuel Technology Sub-Sections. Work reported includes research conducted by the Physical Metallurgy Operation, and research and development conducted by Fuel Design, Fuels Fabrication Development and Ceramic Fuels Development Operations. Studies formerly reported by the Radiometallurgy, Metallography, and Welding and Corrosion Units, in addition to portions of the Fuels Technology work, are reported elsewhere.

  4. Quarterly Progress Report Fuels Development Operation: January - March 1958

    Energy Technology Data Exchange (ETDEWEB)

    Cadwell, J. J. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation; Tobin, J. C. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Physical Metallurgy; Minor, J. E. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Fuel Element Design; Evans, E. A. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Ceramic Fuels Development; Bush, S. H. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Fuels Fabrication Development

    1958-04-15

    The present Quarterly Report is the continuation of a series issued by the new Fuels Development operation. Reports in this series combine portions of the quarterly reports by the former Metallurgy Research and Fuel Technology Sub-Sections. Work reported includes research conducted by the Physical Metallurgy Operation, and research and development conducted by Fuel Design, Fuels Fabrication Development and Ceramic Fuels Development Operations. Studies formerly reported by the Radiometallurgy, Metallography, and Welding and Corrosion Units, in addition to portions of the Fuels Technology work, are reported elsewhere.

  5. Quarterly Progress Report Fuels Development Operation: July - September 1957

    Energy Technology Data Exchange (ETDEWEB)

    Bush, S. H. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Physical Metallurgy; Minor, J. E. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Fuel Element Design; Evans, E. A. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Ceramic Fuels Development; Wallace, W. P. [Hanford Site (HNF), Richland, WA (United States). Fuels Development Operation. Fuels Fabrication Development

    1957-10-15

    The present Quarterly Report is the continuation of a series issued by the new Fuels Development operation. Reports in this series combine portions of the quarterly reports by the former Metallurgy Research and Fuel Technology Sub-Sections. Work reported includes research conducted by the Physical Metallurgy Operation, and research and development conducted by Fuel Design, Fuels Fabrication Development and Ceramic Fuels Development Operations. Studies formerly reported by the Radiometallurgy, Metallography, and Welding and Corrosion Units, in addition to portions of the Fuels Technology work, are reported elsewhere.

  6. Recent Progress in the Development of Diesel Surrogate Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Pitz, W J; Mueller, C J

    2009-12-09

    There has been much recent progress in the area of surrogate fuels for diesel. In the last few years, experiments and modeling have been performed on higher molecular weight components of relevance to diesel fuel such as n-hexadecane (n-cetane) and 2,2,4,4,6,8,8-heptamethylnonane (iso-cetane). Chemical kinetic models have been developed for all the n-alkanes up to 16 carbon atoms. Also, there has been much experimental and modeling work on lower molecular weight surrogate components such as n-decane and n-dodecane that are most relevant to jet fuel surrogates, but are also relevant to diesel surrogates where simulation of the full boiling point range is desired. For two-ring compounds, experimental work on decalin and tetralin recently has been published. For multi-component surrogate fuel mixtures, recent work on modeling of these mixtures and comparisons to real diesel fuel is reviewed. Detailed chemical kinetic models for surrogate fuels are very large in size. Significant progress also has been made in improving the mechanism reduction tools that are needed to make these large models practicable in multi-dimensional reacting flow simulations of diesel combustion. Nevertheless, major research gaps remain. In the case of iso-alkanes, there are experiments and modeling work on only one of relevance to diesel: iso-cetane. Also, the iso-alkanes in diesel are lightly branched and no detailed chemical kinetic models or experimental investigations are available for such compounds. More components are needed to fill out the iso-alkane boiling point range. For the aromatic class of compounds, there has been no new work for compounds in the boiling point range of diesel. Most of the new work has been on alkyl aromatics that are of the range C7 to C8, below the C10 to C20 range that is needed. For the chemical class of cycloalkanes, experiments and modeling on higher molecular weight components are warranted. Finally for multi-component surrogates needed to treat real

  7. US Progress on Property Characterization to Support LEU U-10 Mo Monolithic Fuel Development

    Energy Technology Data Exchange (ETDEWEB)

    Cole, James Irvin [Idaho National Laboratory; Rabin, Barry H [Idaho National Laboratory; Smith, James Arthur [Idaho National Laboratory; Scott, Clark Landon [Idaho National Laboratory; Benefiel, Bradley Curtis [Idaho National Laboratory; Larsen, Eric David [Idaho National Laboratory; Lind, Robert Paul [Idaho National Laboratory; Sell, David Alan [Idaho National Laboratory

    2016-03-01

    The US High Performance Research Reactor program is pursuing development and qualification of a new high density monolithic LEU fuel to facilitate conversion of five higher power research reactors located in the US (ATR, HFIR, NBSR, MIT and MURR). In order to support fabrication development and fuel performance evaluations, new testing capabilities are being developed to evaluate the properties of fuel specimens. Residual stress and fuel-cladding bond strength are two characteristics related to fuel performance that are being investigated. In this overview, new measurement capabilities being developed to assess these characteristics in both fresh and irradiated fuel are described. Progress on fresh fuel testing is summarized and on-going hot-cell implementation efforts to support future PIE campaigns are detailed. It is anticipated that benchmarking of as-fabricated fuel characteristics will be critical to establishing technical bases for specifications that optimize fuel fabrication and ensure acceptable in-reactor fuel performance.

  8. Development of the international trade in wood fuels. Progress report

    International Nuclear Information System (INIS)

    Loenner, Goeran

    2000-04-01

    The imports of different biofuels have increased drastically over the last years on the expanding Swedish market. Apart from peat and refuse, imports cover wood fibre from clearcuts and forest industry and recycled wood from the rapidly expanding recycling industry in Central European countries. The objectives of this research project has been to analyse how the Swedish market for wood fuels has developed and interacted with the international market. The integration aspects have been difficult to cover due to the limited means of the project; a general data gathering and contact making have been prioritised. Knowledge has also been gathered to be used for an application for support for a new project within the EU fifth framework program. In 1997 some 6-8 TWh (21-29 PJ) of biofuels were imported, out of which 50-60% were wood fuels. There is a strong need for a continued analysis of how national policy instruments influence and interact with different markets. This becomes all the more important when facing an emerging European energy policy and the EU white paper on renewable energy

  9. Recent progress in the development of metallic fuel

    International Nuclear Information System (INIS)

    Seidel, B.R.; Batte, G.L.; Dodds, N.E.; Lahm, C.E.; Pahl, R.G.; Tsai, H.C.

    1990-01-01

    Tests to date demonstrate that metallic fuel for advanced liquid metal reactors performs well, is easily reprocessed and refabricated and provides inherent reactor safety within an economic design. The behavior and performance of metallic fuel is key to the demonstration of the Integral Fast Reactor (IFR) concept at Argonne National Laboratory. Since 1985, more than 40 assemblies of experimental fuel in addition to the standard metallic driver fuel for Experimental Breeder Reactor 2 (EBR-2)have been irradiated; several more continue to be designed and fabricated. Results have characterized the influence of a wide range of fabrication, design and material variables upon irradiation behavior throughout the fuel lifetime under normal and upset conditions including operation with breached cladding. Results of test, both in- and out-of-reactor, indicate that metallic fuel is readily and economically fabricated, capable of achieving high exposure and long reactor residence times, and possesses unique and promising safety features. 9 refs., 6 figs

  10. Progress and status of the Integral Fast Reactor (IFR) fuel cycle development

    International Nuclear Information System (INIS)

    Till, C.E.; Chang, Y.I.

    1993-01-01

    The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and the future plans and directions

  11. Progress and status of the Integral Fast Reactor (IFR) fuel cycle development

    International Nuclear Information System (INIS)

    Till, C.E.; Chang, Y.I.

    1991-01-01

    The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and the future plans and directions. 10 refs

  12. Progress and status of the Integral Fast Reactor (IFR) fuel cycle development

    International Nuclear Information System (INIS)

    Till, C.E.; Chang, Y.I.

    1991-01-01

    The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and the future plans and directions. (author)

  13. Progress and status of the Integral Fast Reactor (IFR) fuel cycle development

    Energy Technology Data Exchange (ETDEWEB)

    Till, C.E.; Chang, Y.I.

    1991-01-01

    The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and the future plans and directions. 10 refs.

  14. Progress and status of the Integral Fast Reactor (IFR) fuel cycle development

    Energy Technology Data Exchange (ETDEWEB)

    Till, C.E.; Chang, Y.I.

    1993-03-01

    The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and the future plans and directions.

  15. Progress and status of the Integral Fast Reactor (IFR) fuel cycle development

    Energy Technology Data Exchange (ETDEWEB)

    Till, C.E.; Chang, Y.I.

    1993-01-01

    The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and the future plans and directions.

  16. QUARTERLY PROGRESS REPORT JANUARY, FEBRUARY, MARCH, 1968 REACTOR FUELS AND MATERIALS DEVELOPMENT PROGRAMS FOR FUELS AND MATERIALS BRANCH OF USAEC DIVISION OF REACTOR DEVELOPMENT AND TECHNOLOGY

    Energy Technology Data Exchange (ETDEWEB)

    Cadwell, J. J.; de Halas, D. R.; Nightingale, R. E.; Worlton, D. C.

    1968-06-01

    Progress is reported in these areas: nuclear graphite; fuel development for gas-cooled reactors; HTGR graphite studies; nuclear ceramics; fast-reactor nitrides research; non-destructive testing; metallic fuels; basic swelling studies; ATR gas and water loop operation and maintenance; reactor fuels and materials; fast reactor dosimetry and damage analysis; and irradiation damage to reactor metals.

  17. Progress and status of the integral fast reactor (IFR) fuel cycle development

    International Nuclear Information System (INIS)

    Till, C.E.; Chang, Y.I.

    1993-01-01

    The Integral Fast Reactor (IFR) fuel cycle holds promise for substantial improvements in economics, diversion-resistance, and waste management. This paper discusses technical features of the IFR fuel cycle, its technical progress, the development status, and the future plans and directions. The Integral Fast Reactor (IFR) fuel cycle, is based on the use of a metallic fuel alloy (U-Pu-Zr) that permits use of an innovative method for processing of spent fuel. This method, a combination of pyrometallurgical and electrochemical processes, has been termed pyroprocessing. It offers the advantages of a simple, compact processing system and limited volumes of stabilized high-level wastes. This translates to an economically viable system that is likely to receive favorable public response, particularly when combined with the other attributes of the Integral Fast Reactor. Substantial progress has been made in the development of the IFR pyroprocessing method. A comprehensive demonstration of the process will soon begin at the Argonne National Laboratory Idaho site, using spent fuel from the EBR-II reactor. An important advantage of the IFR is its ability to recycle fuel in the process of power generation, extending fuel resources by a considerable amount and assuring the continued viability of nuclear power stations by reducing dependence on external fuel supplies. Pyroprocessing is the means whereby the recycle process is accomplished. It can also be applied to the recovery of fuel constituents from spent fuel generated in the process of operation of conventional light water reactor power plants, offering the means to recover the valuable fuel resources remaining in that material

  18. Progress in the development of uranium silicide (U3Si2) fuel at BATAN

    International Nuclear Information System (INIS)

    Suripto, A.; Soentono, S.

    1995-01-01

    After successful fabrication of two full-size prototype fuel elements containing ∼3.0 gU/cm 3 in the form of U 3 Si 2 -Al dispersion now undergoing irradiation in the Reaktor Serba Guna G.A. Siwabessy (RSG-GAS) core since 1990, further development in U 3 Si 2 -A2 dispersion fuel element manufacturing has been pursued, whose progress in discussed in this paper, with a special attention on the use of much higher-loading aimed at obtaining a better understanding on the influence of higher-loading on fuel core and plate manufacturing and quality. At present, high-loading U 3 Si 2 -AI dispersion miniplates are being manufactured for preparing some mini-fuel elements to be test-irradiated in the new MTR in-pile loop of the RSG-GAS. (author)

  19. Spent Fuel and Waste Management Technology Development Program. Annual progress report

    Energy Technology Data Exchange (ETDEWEB)

    Bryant, J.W.

    1994-01-01

    This report provides information on the progress of activities during fiscal year 1993 in the Spent Fuel and Waste Management Technology Development Program (SF&WMTDP) at the Idaho Chemical Processing Plant (ICPP). As a new program, efforts are just getting underway toward addressing major issues related to the fuel and waste stored at the ICPP. The SF&WMTDP has the following principal objectives: Investigate direct dispositioning of spent fuel, striving for one acceptable waste form; determine the best treatment process(es) for liquid and calcine wastes to minimize the volume of high level radioactive waste (HLW) and low level waste (LLW); demonstrate the integrated operability and maintainability of selected treatment and immobilization processes; and assure that implementation of the selected waste treatment process is environmentally acceptable, ensures public and worker safety, and is economically feasible.

  20. Recent progress in the development of anode materials for solid oxide fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Cowin, Peter I.; Petit, Christophe T.G.; Lan, Rong; Tao, Shanwen [Department of Chemical and Process Engineering, University of Strathclyde, Glasgow G1 1XJ (United Kingdom); Irvine, John T.S. [School of Chemistry, University of St Andrews, St Andrews, Fife KY16 9ST (United Kingdom)

    2011-05-15

    The field of research into solid oxide fuel cell (SOFC) anode materials has been rapidly moving forward. In the four years since the last in-depth review significant advancements have been made in the reduction of the operating temperature and improvement of the performance of SOFCs. This progress report examines the developments in the field and looks to draw conclusions and inspiration from this research. A brief introduction is given to the field, followed by an overview of the principal previous materials. A detailed analysis of the developments of the last 4 years is given using a selection of the available literature, concentrating on metal-fluorite cermets and perovskite-based materials. This is followed by a consideration of alternate fuels for use in SOFCs and their associated problems and a short discussion on the effect of synthesis method on anode performance. The concluding remarks compile the significant developments in the field along with a consideration of the promise of future research. The recent progress in the development of anode materials for SOFCs based on oxygen ion conducting electrolytes is reviewed. (Copyright copyright 2011 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  1. Progress in the development of very high density research and test reactor fuels

    Energy Technology Data Exchange (ETDEWEB)

    Wachs, D.M. [Idaho National Laboratory, P.O. Box 2528, Idaho Falls, Idaho 83415 (United States)

    2009-06-15

    New nuclear fuels are being developed to enable many of the most important research and test reactors worldwide to convert from high enriched uranium (HEU) fuels to low enriched uranium (LEU) fuels without significant loss in performance. The last decade of work has focused on the development of uranium-molybdenum alloy (U-Mo) based fuels and is an international effort that includes the active participation of more than ten national programs. The US RERTR program, under the NNSA's Global Threat Reduction Initiative (GTRI), is in the process of developing both dispersion and monolithic U-Mo fuel designs. While the U-Mo fuel alloy has behaved extremely well under irradiation, initial testing (circa 2003) revealed that the U-Mo fuels dispersed in aluminum had an unexpected tendency toward unstable swelling (pillowing) under high-power conditions. Technical investigations were initiated worldwide at this time by the partner programs to understand this behavior as well as to develop and test remedies. The behavior was corrected by modifying the chemistry of the U-Mo/Al interfaces in both fuel designs. In the dispersion fuel design, this was accomplished by the addition of small amounts of silicon to the aluminum matrix material. Two methods are under development for the monolithic fuel design, which include the application of a thin layer of silicon or a thin zirconium based diffusion barrier at the fuel/clad interface. This paper gives an overview of the current status of U-Mo fuel development, including basic research results, manufacturing aspects, results of the latest irradiations and post irradiation examinations, the approach to fuel performance qualification, and the scale-up and commercialization of fabrication technology. (authors)

  2. U.S. progress in the development of very high density low enrichment research reactor fuels

    International Nuclear Information System (INIS)

    Meyer, M. K.; Wachs, D. M.; Jue, J.-F.; Keiser, D. D.; Gan, J.; Rice, F.; Robinson, A.; Woolstenhulme, N. E.; Medvedev, P.; Hofman, G. L.; Kim, Y.-S.

    2012-01-01

    The effort to develop low-enriched fuels for high power research reactors began world-wide in 1996. Since that time, hundreds of fuel specimens have been tested to investigate the operational limits of many variations of U-Mo alloy dispersion and monolithic fuels. In the U.S., the fuel development program has focused on the development of monolithic fuel, and is currently transitioning from conducting research experiments to the demonstration of large scale, prototypic element assemblies. These larger scale, integral fuel performance demonstrations include the AFIP-7 test of full-sized, curved plates configured as an element, the RERTR-FE irradiation of hybrid fuel elements in the Advanced Test Reactor, reactor specific Design Demonstration Experiments, and a multi-element Base Fuel Demonstration. These tests are conducted alongside mini-plate tests designed to prove fuel stability over a wide range of operating conditions. Along with irradiation testing, work on collecting data on fuel plate mechanical integrity, thermal conductivity, fission product release, and microstructural stability is underway. (authors)

  3. Sphere-pac fuel development program. First semi-annual progress report, October 1979-March 1980

    International Nuclear Information System (INIS)

    Felt, R.E.

    1980-05-01

    Development of processes by Exxon Nuclear Company, Inc. for fabricating spherical particle nuclear fuel (sphere-pac) during October 1979 through March 1980 is reported. The program surveyed available technology to develop an initial flowsheet as a design basis for process development. An 0.1 ton/day pilot plant was built to develop and demonstrate the fabrication of sphere-pac fuel. Process and equipment efforts have been directed towards the demonstration of processes and equipment necessary to fabricate sphere-pac fuel on a commercial scale

  4. Progress on LEU very high density fuel and target development in Argentina

    International Nuclear Information System (INIS)

    Balart, S.; Cabot, P.; Calzetta, O.; Duran, A.; Garces, J.; Hermida, J.D.; Manzini, A.; Pasqualini, E.; Taboada, H.

    2006-01-01

    Since last RRFM meeting, CNEA has continued on new LEU fuel and target development activities. Main goals are the plan to convert our RA-6 reactor from HEU to a new LEU core, to get a comprehensive understanding of U-Mo/Al compounds phase formation in dispersed and monolithic fuels, to develop possible solutions to VHD dispersed and monolithic fuels technical problems, to optimize techniques to recover U from silicide scrap samples as cold test for radiowaste separation for final conditioning of silicide spent fuels. and to improve the diffusion of LEU target and radiochemical technology for radioisotope production. Future plans include: - Completion of the RA-6 reactor conversion to LEU; - Improvement on fuel development and production facilities to implement new technologies, including NDT techniques to assess bonding quality; - Irradiation of miniplates and full scale fuel assembly at RA-3 and plans to perform irradiation on higher power and temperature regime reactors; - Optimization of LEU target and radiochemical techniques for radioisotope production. (author)

  5. Fuel Chemistry Division: progress report for 1987

    International Nuclear Information System (INIS)

    1990-01-01

    The progress of research and development activities of the Fuel Chemistry Division of the Bhabha Atomic Research Centre, Bombay, during 1987 is reported in the form of summaries which are arranged under the headings: Fuel Development Chemistry, Chemistry of Actinides, Chemical Quality Control of Fuel, and Studies related to Nuclear Material Accounting. A list of publications by the members of the Division during the report period is given at the end of the report. (M.G.B.). refs., 15 figs., 85 tabs

  6. Progress of fusion fuel processing system development at the Japan Atomic Energy Research Institute

    International Nuclear Information System (INIS)

    Nishi, Masataka; Yamanishi, Toshihiko; Kawamura, Yoshinori; Iwai, Yasunori; Isobe, Kanetsugu; O'Hira, Shigeru; Hayashi, Takumi; Nakamura, Hirofumi; Kobayashi, Kazuhiro; Suzuki, Takumi; Yamada, Masayuki; Konishi, Satoshi

    2000-01-01

    The Tritium Process Laboratory (TPL) at the Japan Atomic Energy Research Institute has been working on the development of fuel processing technology for fusion reactors as a major activity. A fusion fuel processing loop was installed and is being tested with tritium under reactor relevant conditions. The loop at the TPL consists of ZrCo based tritium storage beds, a plasma exhaust processing system using a palladium diffuser and an electrolytic reactor, cryogenic distillation columns for isotope separation, and analytical systems based on newly developed micro gas chromatographs and Raman Spectroscopy. Several extended demonstration campaigns were performed under realistic reactor conditions to test tritiated impurity processing. A sophisticated control technique of distillation column was performed at the same time, and integrated fuel circulation was successfully demonstrated. Major recent design work on the International Thermonuclear Experimental Reactor (ITER) tritium plant at the TPL is devoted to water detritiation based on liquid phase catalytic exchange for improved tritium removal from waste water

  7. Progress in hydrogen fueled busses

    International Nuclear Information System (INIS)

    Scott, P.B.; Mazaika, D.M.; Tyler, T.

    2004-01-01

    'Full text:' The Thor/ISE fuel cell bus has been in demonstration and revenue service during 2002-2003 at sites including SunLine Transit, Chula Vista Transit, Los Angeles County Metropolitan Transit Authority, and AC Transit in Oakland. By taking advantage of ISE's advanced hybrid-electric drive technology, this 30-foot bus operates with a much smaller fuel cell than those used in other buses of this class. Further, stress on the fuel cell is diminished. Based on the exceptional performance of this prototype bus, the transit agencies listed above have concluded that hybrid electric hydrogen fueled buses are attractive. Two types of hydrogen fueled hybrid electric buses will be described: - fuel cell powered, and - HICE (Hydrogen Internal Combustion Engine) This progress report will include: 1. Experience with the Thor/ISE fuel cell bus, including results from revenue service at two transit locations, 2. Design and fabrication status of the advanced fuel cell buses being built for AC Transit and SunLine Transit, 3. Design and fabrication status of the prototype HHICE (Hybrid electric Hydrogen fueled Internal Combustion Engine) bus that uses a Ford hydrogen burning engine, mated to a generator, rather than a fuel cell. Other than the engine, the drive train in the HHICE bus is nearly identical to that of a fuel cell hybrid-electric bus. Canadian participation in the HHICE bus is extensive, it is a New Flyer platform and will be winter tested in Winnipeg. (author)

  8. Recent progress in gasoline surrogate fuels

    KAUST Repository

    Sarathy, Mani

    2017-12-06

    Petroleum-derived gasoline is currently the most widely used fuel for transportation propulsion. The design and operation of gasoline fuels is governed by specific physical and chemical kinetic fuel properties. These must be thoroughly understood in order to improve sustainable gasoline fuel technologies in the face of economical, technological, and societal challenges. For this reason, surrogate mixtures are formulated to emulate the thermophysical, thermochemical, and chemical kinetic properties of the real fuel, so that fundamental experiments and predictive simulations can be conducted. Early studies on gasoline combustion typically adopted single component or binary mixtures (n-heptane/isooctane) as surrogates. However, the last decade has seen rapid progress in the formulation and utilization of ternary mixtures (n-heptane/isooctane/toluene), as well as multicomponent mixtures that span the entire carbon number range of gasoline fuels (C4–C10). The increased use of oxygenated fuels (ethanol, butanol, MTBE, etc.) as blending components/additives has also motivated studies on their addition to gasoline fuels. This comprehensive review presents the available experimental and chemical kinetic studies which have been performed to better understand the combustion properties of gasoline fuels and their surrogates. Focus is on the development and use of surrogate fuels that emulate real fuel properties governing the design and operation of engines. A detailed analysis is presented for the various classes of compounds used in formulating gasoline surrogate fuels, including n-paraffins, isoparaffins, olefins, naphthenes, and aromatics. Chemical kinetic models for individual molecules and mixtures of molecules to emulate gasoline surrogate fuels are presented. Despite the recent progress in gasoline surrogate fuel combustion research, there are still major gaps remaining; these are critically discussed, as well as their implications on fuel formulation and engine

  9. Recent progress in gasoline surrogate fuels

    KAUST Repository

    Sarathy, Mani; Farooq, Aamir; Kalghatgi, Gautam T.

    2017-01-01

    Petroleum-derived gasoline is currently the most widely used fuel for transportation propulsion. The design and operation of gasoline fuels is governed by specific physical and chemical kinetic fuel properties. These must be thoroughly understood in order to improve sustainable gasoline fuel technologies in the face of economical, technological, and societal challenges. For this reason, surrogate mixtures are formulated to emulate the thermophysical, thermochemical, and chemical kinetic properties of the real fuel, so that fundamental experiments and predictive simulations can be conducted. Early studies on gasoline combustion typically adopted single component or binary mixtures (n-heptane/isooctane) as surrogates. However, the last decade has seen rapid progress in the formulation and utilization of ternary mixtures (n-heptane/isooctane/toluene), as well as multicomponent mixtures that span the entire carbon number range of gasoline fuels (C4–C10). The increased use of oxygenated fuels (ethanol, butanol, MTBE, etc.) as blending components/additives has also motivated studies on their addition to gasoline fuels. This comprehensive review presents the available experimental and chemical kinetic studies which have been performed to better understand the combustion properties of gasoline fuels and their surrogates. Focus is on the development and use of surrogate fuels that emulate real fuel properties governing the design and operation of engines. A detailed analysis is presented for the various classes of compounds used in formulating gasoline surrogate fuels, including n-paraffins, isoparaffins, olefins, naphthenes, and aromatics. Chemical kinetic models for individual molecules and mixtures of molecules to emulate gasoline surrogate fuels are presented. Despite the recent progress in gasoline surrogate fuel combustion research, there are still major gaps remaining; these are critically discussed, as well as their implications on fuel formulation and engine

  10. Progress of the Russian RERTR program: Development of new-type fuel elements for Russian-built research reactors

    International Nuclear Information System (INIS)

    Vatulin, A. V.; Stetskiy, Y.A.; Mishunin, V.A.; Suprun, V.B.; Dobrikova, I.V.

    2002-01-01

    The new design of pin-type fuel elements and fuel assembly on their basis for Russian research reactors has been developed. The number of following activities has been performed: computational and experimental substantiation of fuel element design; development of fabrication process of fuel elements; manufacturing of experimental assembly for lifetime in-pile tests. The relevant fuel assemblies are considered to be perspective for usage as low-enriched fuel for Russian research reactors. (author)

  11. Progress in development of low-enriched U-Mo dispersion fuels

    International Nuclear Information System (INIS)

    Hofman, G.L.; Snelgrove, J.L.; Hayes, S.L.; Meyer, M.K.

    2002-01-01

    Results from post irradiation examinations and analyses of U-Mo/Al dispersion mini plates are presented. Irradiation test RERTR-5 contained mini- fuel plates with fuel loadings of 6 and 8 g U cm -3 . The fuel material consisted of 6, 7 and 10 wt. % Mo-uranium-alloy powders in atomized and machined form. The swelling behavior of the various fuel types is analyzed, indicating athermal swelling of the U-Mo alloy and temperature-dependent swelling owing to U-Mo/Al interdiffusion. (author)

  12. Oxide fuel element and blanket element development programs. Quarterly progress report, January-February-March, 1979

    International Nuclear Information System (INIS)

    1979-01-01

    Fuel pin profilometry of some 9% burnup F20-F5 pins showed small diameter increases at the fuel-insulator interface at the top of the core. Neither these secondary peaks nor the larger diameter increases near the core midplane exhibited any relationship to the local presence of once-molten fuel in any F20 fuel pin. Augmented safety analysis computations for experiment AB-1 (additional transients suggested by HEDL) showed that cumulative damage fractions from the additional transients were in every case less than 10 -4 . Mechanical tests have been performed that confirm previous computations for the removal end plugs to be used in a characterizer subassembly for AB-1. The resulting pin removal forces are well within the design envelope

  13. Fuel Chemistry Division: progress report for 1985

    International Nuclear Information System (INIS)

    1988-01-01

    Fuel Chemistry Division was formed in May 1985 to give a larger emphasis on the research and development in chemistry of the nuclear fuel cycle. The areas of research in Fuel Chemistry Division are fuel development and its chemical quality control, understanding of the fuel behaviour and post irradiation examinations, chemistry of reprocessing and waste management processes as also the basic aspects of actinide and relevant fission product elements. This report summarises the work by the staff of the Division during 1985 and also some work from the previous periods which was not reported in the progress reports of the Radiochemistry Division. The work related to the FBTR fuel was one of the highlights during this period. In the area of process chemistry useful work has been carried out for processing of plutonium bearing solutions. In the area of mass spectrometry, the determination of trace constituents by spark source mass spectrometry has been a major area of research. Significant progress has also been made in the use of alpha spectromet ry techniques for the determination of plutonium in dissolver solution and other samples. The technology of plutonium utilisation is quite complex and the Division would continue to look into the chemical aspects of this technology and provide the necessary base for future developments in this area. (author)

  14. Progress in developing very-high-density low-enriched-uranium fuels

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  15. FY2014 Fuel & Lubricant Technologies Annual Progress Report

    Energy Technology Data Exchange (ETDEWEB)

    Stork, Kevin [Office of Energy Efficiency and Renewable Energy (EERE), Washington, DC (United States)

    2016-02-01

    Annual progress report for Fuel & Lubricant Technologies. The Fuel & Lubricant Technologies Program supports fuels and lubricants research and development (R&D) to provide vehicle manufacturers and users with cost-competitive options that enable high fuel economy with low emissions, and contribute to petroleum displacement.

  16. Thorium fuel cycle development activities in India (a decade of progress: 1981-1990)

    International Nuclear Information System (INIS)

    Basu, T.K.; Srinivasan, M.

    1990-01-01

    The report comprises of 72 papers on various aspects of thorium fuel cycle. These papers report the research work carried out at the Bhabha Atomic Research Centre, Bombay and other units of the Department of Atomic Energy during the decade 1981-1990. figs., tabs

  17. 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.

  18. Fuel development studies

    International Nuclear Information System (INIS)

    Michel, F.

    1986-12-01

    This paper describes the main lines of the studies carried out to develop the Fast Neutron Fuel Element, from the ''SPX1-first load'' version, to progress to high performance which will be required for the project 1500 and for the fast neutron series [fr

  19. U.S. DOE Progress Towards Developing Low-Cost, High Performance, Durable Polymer Electrolyte Membranes for Fuel Cell Applications.

    Science.gov (United States)

    Houchins, Cassidy; Kleen, Greg J; Spendelow, Jacob S; Kopasz, John; Peterson, David; Garland, Nancy L; Ho, Donna Lee; Marcinkoski, Jason; Martin, Kathi Epping; Tyler, Reginald; Papageorgopoulos, Dimitrios C

    2012-12-18

    Low cost, durable, and selective membranes with high ionic conductivity are a priority need for wide-spread adoption of polymer electrolyte membrane fuel cells (PEMFCs) and direct methanol fuel cells (DMFCs). Electrolyte membranes are a major cost component of PEMFC stacks at low production volumes. PEMFC membranes also impose limitations on fuel cell system operating conditions that add system complexity and cost. Reactant gas and fuel permeation through the membrane leads to decreased fuel cell performance, loss of efficiency, and reduced durability in both PEMFCs and DMFCs. To address these challenges, the U.S. Department of Energy (DOE) Fuel Cell Technologies Program, in the Office of Energy Efficiency and Renewable Energy, supports research and development aimed at improving ion exchange membranes for fuel cells. For PEMFCs, efforts are primarily focused on developing materials for higher temperature operation (up to 120 °C) in automotive applications. For DMFCs, efforts are focused on developing membranes with reduced methanol permeability. In this paper, the recently revised DOE membrane targets, strategies, and highlights of DOE-funded projects to develop new, inexpensive membranes that have good performance in hot and dry conditions (PEMFC) and that reduce methanol crossover (DMFC) will be discussed.

  20. U.S. DOE Progress Towards Developing Low-Cost, High Performance, Durable Polymer Electrolyte Membranes for Fuel Cell Applications

    Directory of Open Access Journals (Sweden)

    Dimitrios C. Papageorgopoulos

    2012-12-01

    Full Text Available Low cost, durable, and selective membranes with high ionic conductivity are a priority need for wide-spread adoption of polymer electrolyte membrane fuel cells (PEMFCs and direct methanol fuel cells (DMFCs. Electrolyte membranes are a major cost component of PEMFC stacks at low production volumes. PEMFC membranes also impose limitations on fuel cell system operating conditions that add system complexity and cost. Reactant gas and fuel permeation through the membrane leads to decreased fuel cell performance, loss of efficiency, and reduced durability in both PEMFCs and DMFCs. To address these challenges, the U.S. Department of Energy (DOE Fuel Cell Technologies Program, in the Office of Energy Efficiency and Renewable Energy, supports research and development aimed at improving ion exchange membranes for fuel cells. For PEMFCs, efforts are primarily focused on developing materials for higher temperature operation (up to 120 °C in automotive applications. For DMFCs, efforts are focused on developing membranes with reduced methanol permeability. In this paper, the recently revised DOE membrane targets, strategies, and highlights of DOE-funded projects to develop new, inexpensive membranes that have good performance in hot and dry conditions (PEMFC and that reduce methanol crossover (DMFC will be discussed.

  1. Technical Division quarterly progress report, October 1--December 31, 1977. [Fuel cycle research and development; special materials production

    Energy Technology Data Exchange (ETDEWEB)

    Slansky, C.M. (ed.)

    1978-02-01

    Results are presented on the fluidized-bed calcination of simulated radioactive waste from the reprocessing of spent commercial nuclear fuel and on the removal of actinide elements from the waste prior to calcination. Other programs include the development of storage technology for /sup 85/Kr waste; and the behavior of volatile radionuclides during the combustion of HTGR graphite-based fuel. The long-term management of defense waste from the ICPP covers post-calcination treatment of ICPP calcined waste; the removal of actinide elements from first-cycle raffinate; the retrieval and handling of calcined waste from ICPP storage vaults; and the preparation of the Programmatic Environmental Impact Statement on ICPP waste. Process improvements are reported on the Fluorinel headend process for Zircaloy-clad fuels and on uranium accountability measurements. Other development results cover the process for recovering spent Rover fuel, buried pipeline transfer systems, support to the Waste Management Program, Waste Calcining Facility support, New Waste Calcining Facility support, and effluent monitoring methods evaluation and development. In this category are studies on nuclear materials security, application of a liquid-solid fluidized-bed heat exchanger to the recovery of geothermal heat, inplant reactor source term measurements, burnup methods for fast breeder reactor fuels, absolute thermal fission yield measurements, analytical support to light water breeder reactor development, research on analytical methods, and the behavior of environmental species of iodine.

  2. LMFBR fuel cycle studies progress report, August 1972, No. 42

    International Nuclear Information System (INIS)

    Unger, W.E.; Blanco, R.E.; Crouse, D.J.; Irvine, A.R.; Watson, C.D.

    1972-10-01

    This report continues a series outlining progress in the development of methods for reprocessing of LMFBR fuels. Development work is reported on problems of irradiated fuel transport to the processing facility, the dissolution of the fuel and the chemical recovery of PuO 2 --UO 2 values, the containment of volatile fission products, product purification, conversion of fuel processing plant product nitrate solutions to solids suitable for shipping and for subsequent fuel fabrication. Pertinent experimental results are presented for the information of those immediately concerned with the field. Detailed description of experimental work and data are included in the topical reports and in the Chemical Technology Division Annual Reports

  3. High-temperature gas-cooled reactor fuel recycle development. Annual progress report for period ending September 30, 1977

    International Nuclear Information System (INIS)

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

    1978-09-01

    The status of the following tasks is reported: program management, studies and analysis, fuel processing, refabrication development, in-plant waste treatment, research general support, and major facilities including HTGR recycle reference facility, hot engineering test facility and cold prototype test facility-refabrication

  4. Logistic Fuel Processor Development

    National Research Council Canada - National Science Library

    Salavani, Reza

    2004-01-01

    The Air Base Technologies Division of the Air Force Research Laboratory has developed a logistic fuel processor that removes the sulfur content of the fuel and in the process converts logistic fuel...

  5. Progress of fossil fuel science

    Energy Technology Data Exchange (ETDEWEB)

    Demirbas, M.F.

    2007-07-01

    Coal is the most abundant and widely distributed fossil fuel. More than 45% of the world's electricity is generated from coal, and it is the major fuel for generating electricity worldwide. The known coal reserves in the world are enough for more than 215 years of consumption, while the known oil reserves are only about 39 times of the world's consumption and the known natural gas reserves are about 63 times of the world's consumption level in 1998. In recent years, there have been effective scientific investigations on Turkish fossil fuels, which are considerable focused on coal resources. Coal is a major fossil fuel source for Turkey. Turkish coal consumption has been stable over the past decade and currently accounts for about 24% of the country's total energy consumption. Lignite coal has had the biggest share in total fossil fuel production, at 43%, in Turkey. Turkish researchers may investigate ten broad pathways of coal species upgrading, such as desulfurization and oxydesulfurization, pyrolysis and hydropyrolysis, liquefaction and hydroliquefaction, extraction and supercritical fluid extraction, gasification, oxidation, briquetting, flotation, and structure identification.

  6. Fuel Chemistry Division: annual progress report for 1988

    International Nuclear Information System (INIS)

    Vaidyanathan, S.

    1991-01-01

    The progress report gives the brief descriptions of various activites of the Fuel Chemistry Division of Bhabha Atomic Research Centre, Bombay for the year 1988. The descriptions of activities are arranged under the headings: Fuel Development Chemistry of Actinides, Quality Control of Fuel, and Studies related to Nuclear Material Accounting. At the end of report, a list of publications published in journals and papers presented at various conferences/symposia during 1988 is given. (author). 13 figs., 61 tabs

  7. Fuel Chemistry Division annual progress report for 1990

    International Nuclear Information System (INIS)

    Vaidyanathan, R.

    1993-01-01

    The progress report gives brief descriptions of the various activities of the Fuel Chemistry Division of Bhabha Atomic Research Centre, Bombay for the year 1990. The descriptions of activities are arranged under the headings: Fuel Development Chemistry, Chemistry of Actinides, Quality Control of Nuclear Fuels, and studies related to Nuclear Materials Accounting. At the end of the report, a list of papers published in journals and presented at various conferences/symposia is also given. (author). 7 figs., 52 tabs

  8. Integrated fuel processor development

    International Nuclear Information System (INIS)

    Ahmed, S.; Pereira, C.; Lee, S. H. D.; Krumpelt, M.

    2001-01-01

    The Department of Energy's Office of Advanced Automotive Technologies has been supporting the development of fuel-flexible fuel processors at Argonne National Laboratory. These fuel processors will enable fuel cell vehicles to operate on fuels available through the existing infrastructure. The constraints of on-board space and weight require that these fuel processors be designed to be compact and lightweight, while meeting the performance targets for efficiency and gas quality needed for the fuel cell. This paper discusses the performance of a prototype fuel processor that has been designed and fabricated to operate with liquid fuels, such as gasoline, ethanol, methanol, etc. Rated for a capacity of 10 kWe (one-fifth of that needed for a car), the prototype fuel processor integrates the unit operations (vaporization, heat exchange, etc.) and processes (reforming, water-gas shift, preferential oxidation reactions, etc.) necessary to produce the hydrogen-rich gas (reformate) that will fuel the polymer electrolyte fuel cell stacks. The fuel processor work is being complemented by analytical and fundamental research. With the ultimate objective of meeting on-board fuel processor goals, these studies include: modeling fuel cell systems to identify design and operating features; evaluating alternative fuel processing options; and developing appropriate catalysts and materials. Issues and outstanding challenges that need to be overcome in order to develop practical, on-board devices are discussed

  9. Progress on KMRR fuel fabrication

    International Nuclear Information System (INIS)

    Kuk, I.H.; Lee, J.B.; Rhee, C.K.; Kim, K.W.

    1991-01-01

    In order to increase the practical applicability of powder heat-treatment in KMRR fuel fabrication, efforts were made to reduce the critical size. Primary U 3 Si 2 particle size was reviewed in terms of cooling rate. Temperature dependence of peritectoid reaction was reviewed as well. (1) Cooling rate of U 3 Si molten alloy was calculated by ADINA program. In practice, particle size of the primary U 3 Si 2 varies radially. U 3 Si 2 size increases as it goes deeper from the surface. As cooling rate increases, primary U 3 Si 2 size decreases. (2) Peritectoid reaction occurs in two unique groups of temperature; one is below 790 C where β-U and U 3 Si 2 reaction occurs, and the other above 790 C where γ-U and U 3 Si 2 reaction occurs. 780 C is most completely reacting temperature in β-U region, and 810 C is so in γ-U region. Reaction is completed more perfectly in γ-U region than in β-U region. 810 C is found to be the optimized heat-treatment temperature, but it is desirable not to approach to 790 C in heat-treatment. (3) The critical powder size in powder heat-treatment is dependent on the primary U 3 Si 2 particle size. The smaller the primary U 3 Si 2 particle size, the smaller the critical particle size of the powder. At present, the primary U 3 Si 2 particle size can be reduced to 3∝5 μm at 4∝5 mm deep from surface in Cu mold. This may be reduced further by rapid solidification process. (orig.)

  10. 2000 Annual Progress Report for Fuels for Advanced CIDI Engines and Fuel Cells

    Energy Technology Data Exchange (ETDEWEB)

    Chalk, S.

    2000-12-11

    The Department of Energy's Office of Transportation Technologies Fiscal Year (FY) 2000 Annual Progress Report for the Fuels for Advanced CIDI Engines and Fuel Cells Program highlights progress achieved during FY 2000 and comprises 22 summaries of industry and National Laboratory projects that were conducted. The report provides an overview of the exciting work being conducted to tackle the tough technical challenges associated with developing clean burning fuels that will enable meeting the performance goals of the Emission Control R and D for Advanced CIDI Engines and the Transportation Fuel Cell Power Systems Programs. The summaries cover the effects of CIDI engine emissions and fuel cell power system performance, the effects of lubricants on engine emissions, the effects of fuel and consumed lubricants on exhaust emission control devices and the health and safety, materials compatibility, and economics of advanced petroleum-based fuels.

  11. Logistic Fuel Processor Development

    National Research Council Canada - National Science Library

    Salavani, Reza

    2004-01-01

    ... to light gases then steam reform the light gases into hydrogen rich stream. This report documents the efforts in developing a fuel processor capable of providing hydrogen to a 3kW fuel cell stack...

  12. 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.

  13. Progress on the development of a new fuel management code to simulate the movement of pebble and block type fuel elements in a very high temperature reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Xhonneux, Andre, E-mail: a.xhonneux@fz-juelich.de [Forschungszentrum Jülich, 52425 Jülich (Germany); Institute for Reactor Safety and Reactor Technology, RWTH-Aachen, 52064 Aachen (Germany); Kasselmann, Stefan; Rütten, Hans-Jochem [Forschungszentrum Jülich, 52425 Jülich (Germany); Becker, Kai [Institute for Reactor Safety and Reactor Technology, RWTH-Aachen, 52064 Aachen (Germany); Allelein, Hans-Josef [Forschungszentrum Jülich, 52425 Jülich (Germany); Institute for Reactor Safety and Reactor Technology, RWTH-Aachen, 52064 Aachen (Germany)

    2014-05-01

    The history of gas-cooled high-temperature reactor prototypes in Germany is closely related to Forschungszentrum Jülich and its “Institute of Nuclear Waste Disposal and Reactor Safety (IEK-6)”. A variety of computer codes have been developed, validated and optimized to simulate the different safety and operational aspects of V/HTR. In order to overcome the present limitations of these codes and to exploit the advantages of modern computer clusters, a project has been initiated to integrate these individual programs into a consistent V/HTR code package (VHCP) applying state-of-the-art programming techniques and standards. One important aspect in the simulation of a V/HTR is the modeling of a continuous moving pebble bed or the periodic rearrangement of prismatic block type fuel. Present models are either too coarse to take special issues (e.g. pebble piles) into account or are too detailed and therefore too time consuming to be applicable in the HCP. The new Software for Handling Universal Fuel Elements (SHUFLE) recently being developed is well suited to close this gap. Although at first the code has been designed for pebble bed reactors, it can in principal be applied to all other types of nuclear fuel. The granularity of the mesh grid meets the requirements to consider these special issues while keeping the used computing power within reasonable limits. New features are for example the possibility to consider azimuthally differing flow velocities in the case of a pebble bed reactor or individual void factors to simulate effects to seismic events. The general idea behind this new approach to the simulation of pebble bed reactors is the following: In the preprocessing step, experimental flow lines or flow lines simulated by more detailed codes serve as an input. For each radial mesh column a representative flow line is then determined by interpolation. These representative flow lines are finally mapped to a user defined rectangular grid forming chains of meshes

  14. Progress on the development of a new fuel management code to simulate the movement of pebble and block type fuel elements in a very high temperature reactor core

    International Nuclear Information System (INIS)

    Xhonneux, Andre; Kasselmann, Stefan; Rütten, Hans-Jochem; Becker, Kai; Allelein, Hans-Josef

    2014-01-01

    The history of gas-cooled high-temperature reactor prototypes in Germany is closely related to Forschungszentrum Jülich and its “Institute of Nuclear Waste Disposal and Reactor Safety (IEK-6)”. A variety of computer codes have been developed, validated and optimized to simulate the different safety and operational aspects of V/HTR. In order to overcome the present limitations of these codes and to exploit the advantages of modern computer clusters, a project has been initiated to integrate these individual programs into a consistent V/HTR code package (VHCP) applying state-of-the-art programming techniques and standards. One important aspect in the simulation of a V/HTR is the modeling of a continuous moving pebble bed or the periodic rearrangement of prismatic block type fuel. Present models are either too coarse to take special issues (e.g. pebble piles) into account or are too detailed and therefore too time consuming to be applicable in the HCP. The new Software for Handling Universal Fuel Elements (SHUFLE) recently being developed is well suited to close this gap. Although at first the code has been designed for pebble bed reactors, it can in principal be applied to all other types of nuclear fuel. The granularity of the mesh grid meets the requirements to consider these special issues while keeping the used computing power within reasonable limits. New features are for example the possibility to consider azimuthally differing flow velocities in the case of a pebble bed reactor or individual void factors to simulate effects to seismic events. The general idea behind this new approach to the simulation of pebble bed reactors is the following: In the preprocessing step, experimental flow lines or flow lines simulated by more detailed codes serve as an input. For each radial mesh column a representative flow line is then determined by interpolation. These representative flow lines are finally mapped to a user defined rectangular grid forming chains of meshes

  15. Metallic fuel development

    International Nuclear Information System (INIS)

    Walters, L.C.

    1987-01-01

    Metallic fuels are capable of achieving high burnup as a result of design modifications instituted in the late 1960's. The gap between the fuel slug and the cladding is fixed such that by the time the fuel swells to the cladding the fission gas bubbles interconnect and release the fission gas to an appropriately sized plenum volume. Interconnected porosity thus provides room for the fuel to deform from further swelling rather than stress the cladding. In addition, the interconnected porosity allows the fuel pin to be tolerant to transient events because as stresses are generated during a transient event the fuel flows rather than applying significant stress to the cladding. Until 1969 a number of metallic fuel alloys were under development in the US. At that time the metallic fuel development program in the US was discontinued in favor of ceramic fuels. However, development had proceeded to the point where it was clear that the zirconium addition to uranium-plutonium fuel would yield a ternary fuel with an adequately high solidus temperature and good compatibility with austenitic stainless steel cladding. Furthermore, several U-Pu-Zr fuel pins had achieved about 6 at.% bu by the late 1960's, without failure, and thus the prospect for high burnup was promising

  16. Progress in Electrolyte-Free Fuel Cells

    Energy Technology Data Exchange (ETDEWEB)

    Lu, Yuzheng [Jiangsu Provincial Key Laboratory of Solar Energy Science and Technology, School of Energy and Environment, Southeast University, Nanjing (China); Zhu, Bin, E-mail: binzhu@kth.se [Faculty of Physics and Electronic Technology, Hubei Collaborative Innovation Center for Advanced Organic Materials, Hubei University, Wuhan (China); Department of Energy Technology, Royal Institute of Technology KTH, Stockholm (Sweden); Cai, Yixiao [Ångström Laboratory, Department of Engineering Sciences, Uppsala University, Uppsala (Sweden); Kim, Jung-Sik [Department of Aeronautical and Automotive Engineering, Loughborough University, Loughborough (United Kingdom); Wang, Baoyuan [Faculty of Physics and Electronic Technology, Hubei Collaborative Innovation Center for Advanced Organic Materials, Hubei University, Wuhan (China); Department of Energy Technology, Royal Institute of Technology KTH, Stockholm (Sweden); Wang, Jun, E-mail: binzhu@kth.se; Zhang, Yaoming [Jiangsu Provincial Key Laboratory of Solar Energy Science and Technology, School of Energy and Environment, Southeast University, Nanjing (China); Li, Junjiao [Nanjing Yunna Nano Technology Co., Ltd., Nanjing (China)

    2016-05-02

    Solid oxide fuel cell (SOFC) represents a clean electrochemical energy conversion technology with characteristics of high conversion efficiency and low emissions. It is one of the most important new energy technologies in the future. However, the manufacture of SOFCs based on the structure of anode/electrolyte/cathode is complicated and time-consuming. Thus, the cost for the entire fabrication and technology is too high to be affordable, and challenges still hinder commercialization. Recently, a novel type of electrolyte-free fuel cell (EFFC) with single component was invented, which could be the potential candidate for the next generation of advanced fuel cells. This paper briefly introduces the EFFC, working principle, performance, and advantages with updated research progress. A number of key R&D issues about EFFCs have been addressed, and future opportunities and challenges are discussed.

  17. Progress in Electrolyte-Free Fuel Cells

    Directory of Open Access Journals (Sweden)

    Yuzheng eLu

    2016-05-01

    Full Text Available Solid Oxide Fuel Cell (SOFC represents a clean electrochemical energy conversion technology with characteristics of high conversion efficiency and low emissions. It is one of the most important new energy technologies in the future. However, the manufacture of SOFCs based on the structure of anode/electrolyte/cathode is complicated and time-consuming. Thus, the cost for the entire fabrication and technology is too high to be affordable and challenges still hinder commercialization. Recently, a novel type of Electrolyte -free fuel cell (EFFC with single component was invented which could be the potential candidate for the next generation of advanced fuel cells. This paper briefly introduces the EFFC, working principle, performance and advantages with updated research progress. A number of key R&D issues about EFFCs have been addressed and future opportunities and challenges are discussed.

  18. Progress in Electrolyte-Free Fuel Cells

    International Nuclear Information System (INIS)

    Lu, Yuzheng; Zhu, Bin; Cai, Yixiao; Kim, Jung-Sik; Wang, Baoyuan; Wang, Jun; Zhang, Yaoming; Li, Junjiao

    2016-01-01

    Solid oxide fuel cell (SOFC) represents a clean electrochemical energy conversion technology with characteristics of high conversion efficiency and low emissions. It is one of the most important new energy technologies in the future. However, the manufacture of SOFCs based on the structure of anode/electrolyte/cathode is complicated and time-consuming. Thus, the cost for the entire fabrication and technology is too high to be affordable, and challenges still hinder commercialization. Recently, a novel type of electrolyte-free fuel cell (EFFC) with single component was invented, which could be the potential candidate for the next generation of advanced fuel cells. This paper briefly introduces the EFFC, working principle, performance, and advantages with updated research progress. A number of key R&D issues about EFFCs have been addressed, and future opportunities and challenges are discussed.

  19. Consolidated fuel reprocessing program. Progress report, January 1-March 31, 1981

    International Nuclear Information System (INIS)

    1981-06-01

    Progress and activities are reported on process development, laboratory R and D, engineering research, engineering systems, Integrated Equipment Test (IET) facility operations, and HTGR fuel reprocessing

  20. Progress on the Application of Metallic Fuels for Actinide Transmutation

    International Nuclear Information System (INIS)

    Kennedy, J. Rory; Fielding, Randall; Janney, Dawn; Mariani, Robert; Teague, Melissa; Egeland, Gerald

    2015-01-01

    Full text of publication follows: Idaho National Laboratory (INL) is developing actinide bearing alloy metallic fuels intended for effecting the transmutation of long-lived isotopes in fast reactor application as part of a partitioning and transmutation strategy. This presentation will report on progress in three areas of this effort: demonstration of the fabrication of fuels under remote (hot cell) conditions directly coupled to the product from the Pyro-processing of spent fuel as part of the Joint Fuel Cycle Studies (JFCS) collaboration with the Korean Atomic Energy Research Institute (KAERI); the chemical sequestration of lanthanide fission products to mitigate fuel-cladding-chemical-interaction (FCCI); and transmission electron microscopy (TEM) and atom probe tomography (APT) studies on the as-cast microstructure of the metallic fuel alloy. For the JFCS efforts, we report on the implementation of the Glove-box Advanced Casting System (GACS) as a prototype casting furnace for eventual installation into the INL Hot Fuel Examination Facility (HFEF) where the recycled fuel will be cast. Results from optimising process parameters with respect to fuel characteristics, americium volatility, materials interaction, and lanthanide fission product carry over distribution will be discussed. With respect to the lanthanide carry over from the Pyro-processing product, encouraging studies on concepts to chemically sequester the FCCI promoting lanthanides within the fuel matrix thus inhibiting migration and interaction with the cladding will be presented. Finally, in relation to advanced modelling and simulation efforts, detailed investigations and interpretation on the nano-scale as cast microstructure of possible recycle fuel composition containing U, Pu, Am, Np as well as carry-over lanthanide species will be discussed. These studies are important for establishing the initial conditions from which advanced physics based fuel performance codes will run. (authors)

  1. Progress on the Hanford K basins spent nuclear fuel project

    International Nuclear Information System (INIS)

    Culley, G.E.; Fulton, J.C.; Gerber, E.W.

    1996-01-01

    This paper highlights progress made during the last year toward removing the Department of Energy's (DOE) approximately, 2,100 metric tons of metallic spent nuclear fuel from the two outdated K Basins at the Hanford Site and placing it in safe, economical interim dry storage. In the past year, the Spent Nuclear Fuel (SNF) Project has engaged in an evolutionary process involving the customer, regulatory bodies, and the public that has resulted in a quicker, cheaper, and safer strategy for accomplishing that goal. Development and implementation of the Integrated Process Strategy for K Basins Fuel is as much a case study of modern project and business management within the regulatory system as it is a technical achievement. A year ago, the SNF Project developed the K Basins Path Forward that, beginning in December 1998, would move the spent nuclear fuel currently stored in the K Basins to a new Staging and Storage Facility by December 2000. The second stage of this $960 million two-stage plan would complete the project by conditioning the metallic fuel and placing it in interim dry storage by 2006. In accepting this plan, the DOE established goals that the fuel removal schedule be accelerated by a year, that fuel conditioning be closely coupled with fuel removal, and that the cost be reduced by at least $300 million. The SNF Project conducted coordinated engineering and technology studies over a three-month period that established the technical framework needed to design and construct facilities, and implement processes compatible with these goals. The result was the Integrated Process Strategy for K Basins Fuel. This strategy accomplishes the goals set forth by the DOE by beginning fuel removal a year earlier in December 1997, completing it by December 1999, beginning conditioning within six months of starting fuel removal, and accomplishes it for $340 million less than the previous Path Forward plan

  2. Recent progresses in materials for the direct methanol fuel cell

    Energy Technology Data Exchange (ETDEWEB)

    Lamy, C; Leger, J M [Centre National de la Recherche Scientifique (CNRS), 86 - Poitiers (France)

    1998-12-31

    Research programs are being conducted worldwide to develop a clean, zero emissions electric vehicle. However, even with the most advanced batteries, such as nickel/metal hydride, or lithium ion batteries, the driving range is limited and the recharging time is long. Only fuel cells which can convert chemical energy directly into electrical energy can compete with internal combustion engines. This paper reviewed the recent progress made in the development of a direct methanol fuel cell using the concept developed for the proton exchange membrane fuel cell (PEMFC). It was noted that the electrode materials, at the methanol anode and oxygen cathode need to be improved by using multifunctional electrocatalysts. The development of new temperature resistant proton exchange membranes with good ionic conductivity and low methanol cross-over, which resulted from the need to increase operating temperatures above 100 degrees C was also reviewed. 35 refs., 1 tab., 2 figs.

  3. Fuel Chemistry Division annual progress report for 1989

    International Nuclear Information System (INIS)

    Singh Mudher, K.D.

    1993-01-01

    The progress report gives a brief description of the various activities of the Fuel Chemistry Division of Bhabha Atomic Research Centre, Bombay for the year 1989. The descriptions of activities are arranged under the headings: Fuel Development Chemistry, Chemical Quality Control, Chemistry of Actinides, Sol-Gel process for the non Nuclear Ceramics and Studies related to Nuclear Material Accounting.At the end of the report, a list of papers published in journals and presented at various conferences/symposia is also given. (author). 69 tabs., 6 figs

  4. Developments in fuel manufacturing

    International Nuclear Information System (INIS)

    Williams, T.

    1997-01-01

    BNFL has a long tradition of willingness to embrace technological challenge and a dedication to quality. This paper describes advances in the overall manufacturing philosophy at BNFL's Fuel Business Group and then covers how some new technologies are currently being employed in BNFL Fuel Business Group's flagship oxide complex (OFC), which is currently in its final stages of commissioning. This plant represents a total investment of some Pound 200 million. This paper also describes how these technologies are also being deployed in BNFL's MOX plant now being built at Sellafield and, finally, covers some new processes being developed for advanced fuel manufacture. (author)

  5. Canadian fuel development program

    International Nuclear Information System (INIS)

    Gacesa, M.; Young, E.G.

    1992-11-01

    CANDU power reactor fuel has demonstrated an enviable operational record. More than 99.9% of the bundles irradiated have provided defect-free service. Defect excursions are responsible for the majority of reported defects. In some cases research and development effort is necessary to resolve these problems. In addition, development initiatives are also directed at improvements of the current design or reduction of fueling cost. The majority of the funding for this effort has been provided by COG (CANDU Owners' Group) over the past 10 to 15 years. This paper contains an overview of some key fuel technology programs within COG. The CANDU reactor is unique among the world's power reactors in its flexibility and its ability to use a number of different fuel cycles. An active program of analysis and development, to demonstrate the viability of different fuel cycles in CANDU, has been funded by AECL in parallel with the work on the natural uranium cycle. Market forces and advances in technology have obliged us to reassess and refocus some parts of our effort in this area, and significant success has been achieved in integrating all the Canadian efforts in this area. This paper contains a brief summary of some key components of the advanced fuel cycle program. (Author) 4 figs., tab., 18 refs

  6. Recent Progress in Development and Manufacturing of SOFC at Topsoe Fuel Cell A/S and Risø DTU

    DEFF Research Database (Denmark)

    Christiansen, Niels; Holm-Larsen, Helge; Primdahl, Søren

    2011-01-01

    effort is directed towards improvement of current generations as well as development of the next generation SOFC technology. The innovative concept of the next generation, aiming at improved reliability and robustness, is based on metal-supported cells and nano-structured electrodes with perspectives...

  7. Developments in fuel manufacturing

    International Nuclear Information System (INIS)

    Ion, S.E.; Harrop, G.; Maricalva Gonzalez, J.

    1995-01-01

    The status of the investment and R and D programmes in the UK and Spanish fuel fabrication facilities is outlined. Due to a number of circumstances, BNFL and ENUSA have been in the forefront of capital investment, with associated commitment to engineering and scientific research and development. Carrying through this investment has allowed the embodiment of proven state of the art technologies in the design of fuel fabrication plants, with particular emphasis on meeting the future challenge of health and safety, and product quality, at an acceptable cost. ENUSA and BNFL currently supply fuel, not only to their respective 'home' markets but also to France, Belgium, Sweden, and Germany. Both organisations employ an International Business outlook and partake in focused and speculative R and D projects for the design and manufacture of nuclear fuel. (orig./HP)

  8. Breeder reactor fuel fabrication system development

    International Nuclear Information System (INIS)

    Bennett, D.W.; Fritz, R.L.; McLemore, D.R.; Yatabe, J.M.

    1981-01-01

    Significant progress has been made in the design and development of remotely operated breeder reactor fuel fabrication and support systems (e.g., analytical chemistry). These activities are focused by the Secure Automated Fabrication (SAF) Program sponsored by the Department of Energy to provide: a reliable supply of fuel pins to support US liquid metal cooled breeder reactors and at the same time demonstrate the fabrication of mixed uranium/plutonium fuel by remotely operated and automated methods

  9. Project fuel development

    International Nuclear Information System (INIS)

    Stratton, R.W.

    1981-05-01

    The activities continued on lab-scale production of uranium-plutonium carbide fuel for the fast reactor using gelation methods, irradiation testing and performance evaluation. Whereas in earlier years a balance was maintained between research and development or with emphasis on research, 1980 was marked by a concentrated equipment development effort for an increased throughput with improved process control and product reproducability and installation of new equipment for large pin fabrication. (Auth.)

  10. Progress and recent trends in biodiesel fuels

    International Nuclear Information System (INIS)

    Demirbas, Ayhan

    2009-01-01

    Fossil fuel resources are decreasing daily. Biodiesel fuels are attracting increasing attention worldwide as blending components or direct replacements for diesel fuel in vehicle engines. Biodiesel fuel typically comprises lower alkyl fatty acid (chain length C 14 -C 22 ), esters of short-chain alcohols, primarily, methanol or ethanol. Various methods have been reported for the production of biodiesel from vegetable oil, such as direct use and blending, microemulsification, pyrolysis, and transesterification. Among these, transesterification is an attractive and widely accepted technique. The purpose of the transesterification process is to lower the viscosity of the oil. The most important variables affecting methyl ester yield during the transesterification reaction are the molar ratio of alcohol to vegetable oil and the reaction temperature. Methanol is the commonly used alcohol in this process, due in part to its low cost. Methyl esters of vegetable oils have several outstanding advantages over other new-renewable and clean engine fuel alternatives. Biodiesel fuel is a renewable substitute fuel for petroleum diesel or petrodiesel fuel made from vegetable or animal fats; it can be used in any mixture with petrodiesel fuel, as it has very similar characteristics, but it has lower exhaust emissions. Biodiesel fuel has better properties than petrodiesel fuel; it is renewable, biodegradable, non-toxic, and essentially free of sulfur and aromatics. Biodiesel seems to be a realistic fuel for future; it has become more attractive recently because of its environmental benefits. Biodiesel is an environmentally friendly fuel that can be used in any diesel engine without modification

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

    International Nuclear Information System (INIS)

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

    2005-10-01

    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

  12. 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.

  13. Progress of air-breathing cathode in microbial fuel cells

    Science.gov (United States)

    Wang, Zejie; Mahadevan, Gurumurthy Dummi; Wu, Yicheng; Zhao, Feng

    2017-07-01

    Microbial fuel cell (MFC) is an emerging technology to produce green energy and vanquish the effects of environmental contaminants. Cathodic reactions are vital for high electrical power density generated from MFCs. Recently tremendous attentions were paid towards developing high performance air-breathing cathodes. A typical air-breathing cathode comprises of electrode substrate, catalyst layer, and air-diffusion layer. Prior researches demonstrated that each component influenced the performance of air-breathing cathode MFCs. This review summarized the progress in development of the individual component and elaborated main factors to the performance of air-breathing cathode.

  14. Chilean fuel elements fabrication progress report

    International Nuclear Information System (INIS)

    Baeza, J.; Contreras, H.; Chavez, J.; Klein, J.; Mansilla, R.; Marin, J.; Medina, R.

    1993-01-01

    Due to HEU-LEU core conversion necessity for the Chilean MTR reactors, the Fuel Elements Plant is being implemented to LEU nuclear fuel elements fabrication. A glove box line for powder-compact processing designed at CCHEN, which supposed to operate under an automatic control system, is at present under initial tests. Results of first natural uranium fuel plates manufacturing runs are shown

  15. Metallic fuel design development

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-04-01

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

  16. Metallic fuel design development

    International Nuclear Information System (INIS)

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

    1999-04-01

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

  17. Status of LEU fuel development and conversion of NRU

    International Nuclear Information System (INIS)

    Sears, D.F.; Herbert, L.N.; Vaillancourt, K.D.

    1991-01-01

    This paper reviews the status of the LEU conversion program and the progress made in the fuel development program over the last year. The results from post-irradiation examinations of prototype NRU fuel rods containing Al-U 3 Si dispersion fuel, and of mini-elements containing Al-U 3 Si 2 dispersion fuel, are presented. (orig.)

  18. High performance fuel technology development

    Energy Technology Data Exchange (ETDEWEB)

    Koon, Yang Hyun; Kim, Keon Sik; Park, Jeong Yong; Yang, Yong Sik; In, Wang Kee; Kim, Hyung Kyu [KAERI, Daejeon (Korea, Republic of)

    2012-01-15

    {omicron} Development of High Plasticity and Annular Pellet - Development of strong candidates of ultra high burn-up fuel pellets for a PCI remedy - Development of fabrication technology of annular fuel pellet {omicron} Development of High Performance Cladding Materials - Irradiation test of HANA claddings in Halden research reactor and the evaluation of the in-pile performance - Development of the final candidates for the next generation cladding materials. - Development of the manufacturing technology for the dual-cooled fuel cladding tubes. {omicron} Irradiated Fuel Performance Evaluation Technology Development - Development of performance analysis code system for the dual-cooled fuel - Development of fuel performance-proving technology {omicron} Feasibility Studies on Dual-Cooled Annular Fuel Core - Analysis on the property of a reactor core with dual-cooled fuel - Feasibility evaluation on the dual-cooled fuel core {omicron} Development of Design Technology for Dual-Cooled Fuel Structure - Definition of technical issues and invention of concept for dual-cooled fuel structure - Basic design and development of main structure components for dual- cooled fuel - Basic design of a dual-cooled fuel rod.

  19. 2016 Annual Progress Report: DOE Hydrogen and Fuel Cells Program

    Energy Technology Data Exchange (ETDEWEB)

    Satyapal, Sunita [National Renewable Energy Lab. (NREL), Golden, CO (United States)

    2017-02-01

    In the past year, the DOE Hydrogen Program (the Program) made substantial progress toward its goals and objectives. The Program has conducted comprehensive and focused efforts to enable the widespread commercialization of hydrogen and fuel cell technologies in diverse sectors of the economy. With emphasis on applications that will effectively strengthen our nation's energy security and improve our stewardship of the environment, the Program engages in research, development, and demonstration of critical improvements in the technologies. Highlights of the Program's accomplishments can be found in the sub-program chapters of this report.

  20. 2015 Annual Progress Report: DOE Hydrogen and Fuel Cells Program

    Energy Technology Data Exchange (ETDEWEB)

    Popovich, Neil

    2015-12-01

    In the past year, the DOE Hydrogen Program (the Program) made substantial progress toward its goals and objectives. The Program has conducted comprehensive and focused efforts to enable the widespread commercialization of hydrogen and fuel cell technologies in diverse sectors of the economy. With emphasis on applications that will effectively strengthen our nation's energy security and improve our stewardship of the environment, the Program engages in research, development, and demonstration of critical improvements in the technologies. Highlights of the Program's accomplishments can be found in the sub-program chapters of this report.

  1. 2012 Annual Progress Report: DOE Hydrogen and Fuel Cells Program

    Energy Technology Data Exchange (ETDEWEB)

    2012-12-01

    In the past year, the DOE Hydrogen Program (the Program) made substantial progress toward its goals and objectives. The Program has conducted comprehensive and focused efforts to enable the widespread commercialization of hydrogen and fuel cell technologies in diverse sectors of the economy. With emphasis on applications that will effectively strengthen our nation's energy security and improve our stewardship of the environment, the Program engages in research, development, and demonstration of critical improvements in the technologies. Highlights of the Program's accomplishments can be found in the sub-program chapters of this report.

  2. Examination of fast reactor fuels, FBR analytical quality assurance standards and methods, and analytical methods development: irradiation tests. Progress report, April 1--June 30, 1976, and FY 1976

    International Nuclear Information System (INIS)

    Baker, R.D.

    1976-08-01

    Characterization of unirradiated and irradiated LMFBR fuels by analytical chemistry methods will continue, and additional methods will be modified and mechanized for hot cell application. Macro- and microexaminations will be made on fuel and cladding using the shielded electron microprobe, emission spectrograph, radiochemistry, gamma scanner, mass spectrometers, and other analytical facilities. New capabilities will be developed in gamma scanning, analyses to assess spatial distributions of fuel and fission products, mass spectrometric measurements of burnup and fission gas constituents and other chemical analyses. Microstructural analyses of unirradiated and irradiated materials will continue using optical and electron microscopy and autoradiographic and x-ray techniques. Analytical quality assurance standards tasks are designed to assure the quality of the chemical characterizations necessary to evaluate reactor components relative to specifications. Tasks include: (1) the preparation and distribution of calibration materials and quality control samples for use in quality assurance surveillance programs, (2) the development of and the guidance in the use of quality assurance programs for sampling and analysis, (3) the development of improved methods of analysis, and (4) the preparation of continuously updated analytical method manuals. Reliable analytical methods development for the measurement of burnup, oxygen-to-metal (O/M) ratio, and various gases in irradiated fuels is described

  3. Consolidated fuel reprocessing program. Progress report, July 1-September 30, 1981

    International Nuclear Information System (INIS)

    1981-12-01

    Technical progress is reported in overview fashion in the following areas: process development, laboratory R and D, engineering research, engineering systems, integrated equipment test facility (IET) operations, and HTGR fuel reprocessing

  4. Alcohol fuels for developing countries

    International Nuclear Information System (INIS)

    Bhattacharya, Partha

    1993-01-01

    The importance of alcohol as an alternative fuel has been slowly established. In countries such as Brazil, they are already used in transport and other sectors of economy. Other developing countries are also trying out experiments with alcohol fuels. Chances of improving the economy of many developing nations depends to a large extent on the application of this fuel. The potential for alcohol fuels in developing countries should be considered as part of a general biomass-use strategy. The final strategies for the development of alcohol fuel will necessarily reflect the needs, values, and conditions of the individual nations, regions, and societies that develop them. (author). 5 refs

  5. 2015 Annual Progress Report: DOE Hydrogen and Fuel Cells Program

    Energy Technology Data Exchange (ETDEWEB)

    None

    2015-12-23

    The 2015 Annual Progress Report summarizes fiscal year 2015 activities and accomplishments by projects funded by the DOE Hydrogen and Fuel Cells Program. It covers the program areas of hydrogen production; hydrogen delivery; hydrogen storage; fuel cells; manufacturing R&D; technology validation; safety, codes and standards; systems analysis; and market transformation.

  6. 2016 Annual Progress Report: DOE Hydrogen and Fuel Cells Program

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2017-03-09

    The 2016 Annual Progress Report summarizes fiscal year 2016 activities and accomplishments by projects funded by the DOE Hydrogen and Fuel Cells Program. It covers the program areas of hydrogen production; hydrogen delivery; hydrogen storage; fuel cells; manufacturing R&D; technology validation; safety, codes and standards; systems analysis; market transformation; and Small Business Innovation Research projects.

  7. RDBE Development and Progress

    Science.gov (United States)

    Niell, A.; Bark, M.; Beaudoin, C.; Brisken, W.; Frej, H. Ben; Doeleman, S.; Durand, S.; Guerra, M.; Hinton, A.; Luce, M.; hide

    2010-01-01

    A digital backend based on the ROACH board has been developed jointly by the National Radio Astronomy Observatory and MIT Haystack Observatory. The RDBE will have both Polyphase Filterbank and Digital Downconverter personalities. The initial configuration outputs sixteen 32-MHz channels, comprised of half the channels from the PFB processing of the two IF inputs, for use in the VLBI2010 geodetic system and in the VLBA sensitivity upgrade project. The output rate is 2x109 bits/second (1x10(exp 9) bits/sec = 1 Gbps) over a 10 GigE connection to the Mark 5C with the data written in Mark 5B format on disk.

  8. RDBE Development and Progress

    Science.gov (United States)

    Neill, A.; Bark, M.; Beaudoin, C.; Brisken, W.; Ben Frej, H.; Doeleman, S.; Durand, S.; Guerra, Ml; Hinton, A.; Luce, M.; McWhirter, R.; Morris, K.; Peck, G.; Revnell, M.; Rogers, A.; Romney, J.; Ruszczyk, C; Taveniku, M.; Walker, R.; Whitney, A.

    2010-12-01

    A digital backend based on the ROACH board has been developed jointly by the National Radio Astronomy Observatory and MIT Haystack Observatory. The RDBE will have both Polyphase Filterbank and Digital Downconverter personalities. The initial configuration outputs sixteen 32-MHz channels, comprised of half the channels from the PFB processing of the two IF inputs, for use in the VLBI2010 geodetic system and in the VLBA sensitivity upgrade project. The output rate is 2x10^9 bits/second (1x10^9 bits/sec = 1 Gbps) over a 10 GigE connection to the Mark 5C with the data written in Mark 5B format on disk.

  9. Development of CANFLEX fuel bundle

    International Nuclear Information System (INIS)

    Suk, Ho Chun; Hwang, Woan; Jeong, Young Hwan

    1991-12-01

    This research project is underway in cooperation with AECL to develop the CANDU advanced fuel bundle(so-called CANFLEX) which can enhance reactor safety and fuel economy in comparison with the current CANDU fuel and which can be used with natural uranium, slightly enriched uranium and other advanced fuel cycle. As the final schedule, the advanced fuel will be verified by carrying out a large scale demonstration of the bundle irradiation in a commercial CANDU reactors for 1996 and 1997, and consequently will be used in the existing and future reactors in Korea. The research activities during this year include the basic design of CANFLEX fuel with slightly enriched uranium(CANFLEX-SEU), with emphasis on the extension of fuel operation limit. Based on this basic design, CANFLEX fuel was mocked up. Out-of-pile hydraulic scoping tests were conducted with the fuel. (Author)

  10. Fuel-oil boilers are improving; Les chaudieres a fioul progressent

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    2000-08-01

    The share of domestic fuel oil consumption in France is progressively growing up in space heating applications. In order to perennialize this growth, the 'Chauffage Fioul' association has developed an advertising strategy for the promotion of fuel-oil boilers in accommodations. Short paper. (J.S.)

  11. HTR fuel development for advanced application

    International Nuclear Information System (INIS)

    Nickel, H.; Balthesen, E.; Graham, L.W.; Hick, H.

    1975-01-01

    The advantages of the HTR for nuclear steam supply systems are briefly outlined. Due to its great design flexibility a number of different designs have evolved and the main characteristics of existing experimental prototype and power reactor HTR designs are summarized. The present state of coated particle fuel, particularly with regard to performance, is considered. Some implications of producing higher temperatures are discussed. Finally some of the developments in progress such as minimising the temperature drop between fuel and coolant, and of improving fuel performance by better fission product retention, better chemical stability, and the use of alternative coated materials, are discussed. (U.K.)

  12. Recent Progress on the DUPIC Fuel Fabrication Technology at KAERI

    International Nuclear Information System (INIS)

    Jung-Won Lee; Ho-Jin Ryu; Geun-Il Park; Kee-Chan Song

    2008-01-01

    Since 1991, KAERI has been developing the DUPIC fuel cycle technology. The concept of a direct use of spent PWR fuel in Candu reactors (DUPIC) is based on a dry processing method to re-fabricate Candu fuel from spent PWR fuel without any intentional separation of the fissile materials and fission products. A DUPIC fuel pellet was successfully fabricated and the DUPIC fuel element fabrication processes were qualified on the basis of a Quality Assurance program. Consequently, the DUPIC fuel fabrication technology was verified and demonstrated on a laboratory-scale. Recently, the fuel discharge burn-up of PWRs has been extended to reduce the amount of spent fuel and the fuel cycle costs. Considering this trend of extending the fuel burn-up in PWRs, the DUPIC fuel fabrication technology should be improved to process high burn-up spent fuels. Particularly the release behavior of cesium from the pellet prepared with a high burn-up spent fuel was assessed. an improved DUPIC fuel fabrication technology was experimentally established with a fuel burn-up of 65,000 MWd/tU. (authors)

  13. IFR fuel cycle--pyroprocess development

    International Nuclear Information System (INIS)

    Laidler, J.J.; Miller, W.E.; Johnson, T.R.; Ackerman, J.P.; Battles, J.E.

    1992-01-01

    The Integral Fast Reactor (IFR) fuel cycle is based on the use of a metallic fuel alloy, with nominal composition U-2OPu-lOZr. In its present state of development, this fuel system offers excellent high-burnup capabilities. Test fuel has been carried to burnups in excess of 20 atom % in EBR-II irradiations, and to peak burnups over 15 atom % in FFTF. The metallic fuel possesses physical characteristics, in particular very high thermal conductivity, that facilitate a high degree of passive inherent safety in the IFR design. The fuel has been shown to provide very large margins to failure in overpower transient events. Rapid overpower transient tests carried out in the TREAT reactor have shown the capability to withstand up to 400% overpower conditions before failing. An operational transient test conducted in EBR-II at a power ramp rate of 0.1% per second reached its termination point of 130% of normal power without any fuel failures. The IFR metallic fuel also exhibits superior compatibility with the liquid sodium coolant. Equally as important as the performance advantages offered by the use of metallic fuel is the fact that this fuel system permits the use of an innovative reprocessing method, known as ''pyroprocessing,'' featuring fused-salt electrorefining of the spent fuel. Development of the IFR pyroprocess has been underway at the Argonne National Laboratory for over five years, and great progress has been made toward establishing a commercially-viable process. Pyroprocessing offers a simple, compact means for closure of the fuel cycle, with anticipated significant savings in fuel cycle costs

  14. Advanced Research Reactor Fuel Development

    Energy Technology Data Exchange (ETDEWEB)

    Kim, C. K.; Park, H. D.; Kim, K. H. (and others)

    2006-04-15

    RERTR program for non-proliferation has propelled to develop high-density U-Mo dispersion fuels, reprocessable and available as nuclear fuel for high performance research reactors in the world. As the centrifugal atomization technology, invented in KAERI, is optimum to fabricate high-density U-Mo fuel powders, it has a great possibility to be applied in commercialization if the atomized fuel shows an acceptable in-reactor performance in irradiation test for qualification. In addition, if rod-type U-Mo dispersion fuel is developed for qualification, it is a great possibility to export the HANARO technology and the U-Mo dispersion fuel to the research reactors supplied in foreign countries in future. In this project, reprocessable rod-type U-Mo test fuel was fabricated, and irradiated in HANARO. New U-Mo fuel to suppress the interaction between U-Mo and Al matrix was designed and evaluated for in-reactor irradiation test. The fabrication process of new U-Mo fuel developed, and the irradiation test fuel was fabricated. In-reactor irradiation data for practical use of U-Mo fuel was collected and evaluated. Application plan of atomized U-Mo powder to the commercialization of U-Mo fuel was investigated.

  15. Progress in Chemical Kinetic Modeling for Surrogate Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Pitz, W J; Westbrook, C K; Herbinet, O; Silke, E J

    2008-06-06

    Gasoline, diesel, and other alternative transportation fuels contain hundreds to thousands of compounds. It is currently not possible to represent all these compounds in detailed chemical kinetic models. Instead, these fuels are represented by surrogate fuel models which contain a limited number of representative compounds. We have been extending the list of compounds for detailed chemical models that are available for use in fuel surrogate models. Detailed models for components with larger and more complicated fuel molecular structures are now available. These advancements are allowing a more accurate representation of practical and alternative fuels. We have developed detailed chemical kinetic models for fuels with higher molecular weight fuel molecules such as n-hexadecane (C16). Also, we can consider more complicated fuel molecular structures like cyclic alkanes and aromatics that are found in practical fuels. For alternative fuels, the capability to model large biodiesel fuels that have ester structures is becoming available. These newly addressed cyclic and ester structures in fuels profoundly affect the reaction rate of the fuel predicted by the model. Finally, these surrogate fuel models contain large numbers of species and reactions and must be reduced for use in multi-dimensional models for spark-ignition, HCCI and diesel engines.

  16. LEU fuel development at CERCA

    International Nuclear Information System (INIS)

    Durand, Jean Pierre; Ottone, J.C.; Mahe, M.; Ferraz, G.

    1998-01-01

    The aim of this paper is to detail the recent progress on both U 3 Si 2 high loaded fuels and new γ phase fuels. Concerning high density density silicide plates up to 6 g Ut/cm 3 , the CEA irradiation programme is completed. Data are still under analysis but one can state that the behaviour was globally similar to conventional fuels known in SILOE and OSIRIS reactors. From the new γ fuel point of view, after demonstration feasibility in 1997 of U Mo thermally stable plates loaded up to 8.3 g Ut/cm3, CERCA has analysed the technical ability of quality inspection means assuming that is of an utmost interest for the insurance of a proper use of high performances fuel in reactors. There are mainly two differences between U Mo fuels (and more generally γ fuels) and conventional ones. Firstly, X-ray diffraction analysis on the fuel powder are needed because the chemical analysis is not sufficient to characterise the γ structure requested. Secondly, the physical limits of the Ultrasonic inspection have been reached due to transitory effect between the meat and the edges. Therefore this technic can not applied in the transitory areas. From that knowledge, the manufacture specifications for a plate dedicated to an irradiation plan can be discussed with a clearer view of the main differences with the U 3 Si 2 fuel reference. (author)

  17. RERTR program progress in qualifying reduced-enrichment fuels

    International Nuclear Information System (INIS)

    Snelgrove, James L.

    1983-01-01

    In order to provide the technical means for reducing the enrichment of uranium used to fuel research and test reactors, the U.S. Reduced Enrichment Research and Test Reactor (RERTR) Program has been engaged in the development and testing of higher-uranium-density fuels than had been used previously. This fuel development effort included work to increase the density of fuels which were being used at the time the Program began and work on a fuel with the potential for much higher density. The ultimate goal of the fuel development and testing phase of the Program is to 'qualify' the fuel for use. A fuel is considered qualified when a sufficient data base for the fuel exists that it can be approved by regulating bodies for use in reactors. To convert a core to the use of reduced-enrichment fuel it is necessary to show that the core will behave properly during normal and off-normal operating conditions and to show that the fuel will behave properly to a reasonable margin beyond the conditions expected during normal operation. It is this latter area that this paper will address. The main characteristics to be considered in evaluating the performance of a fuel are its swelling, its blister-threshold temperature, and its metallurgical appearance. Data for the qualification of the reduced-enrichment fuels being developed by the RERTR Program are obtained from examination of miniature fuel plates (miniplates) which successfully pass the irradiation screening tests and from examinations of full-sized fuel elements. This paper will summarize the miniplate data reported in other papers presented during this meeting and will give the status of full-sized element irradiations. Finally, the current status of qualification of the various fuel types will be discussed and some projections of the future will be given

  18. Advanced Hydrocarbon Fuel Development

    Science.gov (United States)

    Bai, S. Don; Rodgers, Stephen L. (Technical Monitor)

    2000-01-01

    As a part of a high energy density materials (HEDM) development, the hot fire tests for Quadricyclane, 1,7 Octadiyne, AFRL-1, Biclopropylidene, and CINCH (Dimethyl amino ethyl azide) have been conducted at NASA/MSFC. The first 4 materials for this task are provided from Air Force Research Laboratory at Edward Air Force Base and US Army provided CINCH. The performance of these fuels is compared with RP-1. The preliminary results of these tests are presented. The preliminary results of Quadricyclane tests indicate that the specific impulse and c-star efficiency for quadricyclane at the mixture ratio 1.94 are approximately 5 sec and 105 ft/sec better than the RP-1 at mixture ratio 1.9. The 1,7 Octadiyne test indicate that the specific impulse and c-star efficiency at the mixture ratio 2.1 are approximately -1 sec and 89 ft/sec differ than the RP-1 at mixture ratio 2.04. The Quadricyclane soot buildup at the combustor is a little more than RP-1, but detail study of soot formation is not considered at this time. There was no visual soot buildup for the 1,7 Octadiyne and AFRL-1.

  19. Gammasphere software development. Progress report

    Energy Technology Data Exchange (ETDEWEB)

    Piercey, R.B.

    1994-01-01

    This report describes the activities of the nuclear physics group at Mississippi State University which were performed during 1993. Significant progress has been made in the focus areas: chairing the Gammasphere Software Working Group (SWG); assisting with the porting and enhancement of the ORNL UPAK histogramming software package; and developing standard formats for Gammasphere data products. In addition, they have established a new public ftp archive to distribute software and software development tools and information.

  20. Advanced fuel development at AECL: What does the future hold for CANDU fuels/fuel cycles?

    Energy Technology Data Exchange (ETDEWEB)

    Kupferschmidt, W.C.H. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2013-07-01

    This paper outlines advanced fuel development at AECL. It discusses expanding the limits of fuel utilization, deploy alternate fuel cycles, increase fuel flexibility, employ recycled fuels; increase safety and reliability, decrease environmental impact and develop proliferation resistant fuel and fuel cycle.

  1. 2013 Annual Progress Report: DOE Hydrogen and Fuel Cells Program

    Energy Technology Data Exchange (ETDEWEB)

    none,

    2013-12-01

    The 2013 Annual Progress Report summarizes fiscal year 2013 activities and accomplishments by projects funded by the DOE Hydrogen Program. It covers the program areas of hydrogen production and delivery; hydrogen storage; fuel cells; manufacturing; technology validation; safety, codes and standards; market transformation; and systems analysis.

  2. 2014 Annual Progress Report: DOE Hydrogen and Fuel Cells Program

    Energy Technology Data Exchange (ETDEWEB)

    none,

    2014-11-01

    The 2014 Annual Progress Report summarizes fiscal year 2014 activities and accomplishments by projects funded by the DOE Hydrogen Program. It covers the program areas of hydrogen production and delivery; hydrogen storage; fuel cells; manufacturing; technology validation; safety, codes and standards; market transformation; and systems analysis.

  3. 2011 Annual Progress Report: DOE Hydrogen and Fuel Cells Program

    Energy Technology Data Exchange (ETDEWEB)

    Satyapal, Sunita [Office of Energy Efficiency and Renewable Energy (EERE), Washington, DC (United States)

    2011-11-01

    The 2011 Annual Progress Report summarizes fiscal year 2011 activities and accomplishments by projects funded by the DOE Hydrogen Program. It covers the program areas of hydrogen production and delivery; hydrogen storage; fuel cells; manufacturing; technology validation; safety, codes and standards; education; market transformation; and systems analysis.

  4. Fuel cycle developments

    International Nuclear Information System (INIS)

    Anon.

    1994-01-01

    This article is a review of the end-of-1994 status of world uranium production and fuels processing. The major producing areas/countries of the world are discussed and the production figures for each area/country are provided. The conversion services market is also discussed, as is the enrichment services market. Each of the major enrichment services provider organizations is noted

  5. Development of nuclear fuel. Development of CANDU advanced fuel bundle

    International Nuclear Information System (INIS)

    Suk, Ho Chun; Hwang, Woan; Jeong, Young Hwan; Jung, Sung Hoon

    1991-07-01

    In order to develop CANDU advanced fuel, the agreement of the joint research between KAERI and AECL was made on February 19, 1991. AECL conceptual design of CANFLEX bundle for Bruce reactors was analyzed and then the reference design and design drawing of the advanced fuel bundle with natural uranium fuel for CANDU-6 reactor were completed. The CANFLEX fuel cladding was preliminarily investigated. The fabricability of the advanced fuel bundle was investigated. The design and purchase of the machinery tools for the bundle fabrication for hydraulic scoping tests were performed. As a result of CANFLEX tube examination, the tubes were found to be meet the criteria proposed in the technical specification. The dummy bundles for hydraulic scoping tests have been fabricated by using the process and tools, where the process parameters and tools have been newly established. (Author)

  6. Development of MOX fuel database

    International Nuclear Information System (INIS)

    Ikusawa, Yoshihisa; Ozawa, Takayuki

    2007-03-01

    We developed MOX Fuel Database, which included valuable data from several irradiation tests in FUGEN and Halden reactor, for help of LWR MOX use. This database includes the data of fabrication and irradiation, and the results of post-irradiation examinations for seven fuel assemblies, i.e. P06, P2R, E03, E06, E07, E08 and E09, irradiated in FUGEN. The highest pellet peak burn-up reached ∼48GWd/t in MOX fuels, of which the maximum plutonium content was ∼6 wt%, irradiated in E09 fuel assembly without any failure. Also the data from the instrumented MOX fuels irradiated in HBWR to study the irradiation behavior of BWR MOX fuels under the steady state condition (IFA-514/565 and IFA-529), under the load-follow operation condition (IFA-554/555) and under the transit condition (IFA-591) are included in this database. The highest assembly burn-up reached ∼56 GWd/t in IFA-565 steady state irradiation test, and the maximum linear power of MOX fuel rods was 58.3-68.4 kW/m without any failure in IFA-591 ramp test. In addition, valuable instrument data, i.e. cladding elongation, fuel stack elongation, fuel center temperature and rod inner pressure were obtained from IFA-554/555 load-follow test. (author)

  7. Fuel Cell Manufacturing Research and Development | Hydrogen and Fuel Cells

    Science.gov (United States)

    | NREL Fuel Cell Manufacturing Research and Development Fuel Cell Manufacturing Research and Development NREL's fuel cell manufacturing R&D focuses on improving quality-inspection practices for high costs. A researcher monitoring web-line equipment in the Manufacturing Laboratory Many fuel cell

  8. 2010 Hydrogen and Fuel Cell Global Commercialization & Development Update

    Energy Technology Data Exchange (ETDEWEB)

    none,

    2010-11-01

    This report offers examples of real-world applications and technical progress of hydrogen and fuel cell technologies, including policies adopted by countries to increase technology development and commercialization.

  9. Progress in safety evaluation for the JMTR core conversion to LEU fuel

    International Nuclear Information System (INIS)

    Sakurai, F.; Komori, Y.; Saito, J.; Komukai, B.; Ando, H.; Nakata, H.; Sakakura, A.; Niiho, S.; Saito, M.; Futamura, Y.

    1991-01-01

    The JMTR (50 MWt) has been in steady operation with MEU fuel since July 1986. The effort is still continued to convert the core from MEU to LEU fuel. The LEU silicide fuel element at 4.8 gU/cm 3 with Cd wires as burnable absorbers has been selected in order to achieve upgraded fuel cycle performance of extended cycle length and reduced control rod movement operation. The neutronic calculation methods (diffusion theory model) developed for the LEU core with Cd wires was benchmarked with a detailed Monte Carlo model and verified experimentally using the critical facility, JMTRC. Hydraulic tests of the LEU silicide fuel element with Cd wires were completed with satisfactory results, and measurements of release/born (R/B) ratios of FPs of silicide fuel at high temperature are in progress. (orig.)

  10. Fast breeder fuel element development

    International Nuclear Information System (INIS)

    Marth, W.; Muehling, G.

    1983-08-01

    This report is a compilation of the papers which have been presented during a seminar ''Fast Breeder Fuel Element Development'' held on November 15/16, 1982 at KfK. The papers give a survey of the status, of the obtained results and of the necessary work, which still has to be done in the frame of various development programmes for fast breeder fuel elements. In detail the following items were covered by the presentations: - the requirements and boundary conditions for the design of fuel pins and elements both for the reference concept of the SNR 300 core and for the large, commercial breeder type of the future (presentation 1,2 and 6); - the fabrication, properties and characterization of various mixed oxide fuel types (presentations 3,4 and 5); - the operational fuel pin behaviour, limits of different design concepts and possible mechanism for fuel pin failures (presentations (7 and 8); - the situation of cladding- and wrapper materials development especially with respect to the high burn-up values of commercial reactors (presentations 9 and 10); - the results of the irradiation experiments performed under steady-state and non-stationary operational conditions and with failed fuel pins (presentations 11, 12, 13 and 14). (orig./RW) [de

  11. Romanian nuclear fuel cycle development

    International Nuclear Information System (INIS)

    Rapeanu, S.N.; Comsa, Olivia

    1998-01-01

    Romanian decision to introduce nuclear power was based on the evaluation of electricity demand and supply as well as a domestic resources assessment. The option was the introduction of CANDU-PHWR through a license agreement with AECL Canada. The major factors in this choice have been the need of diversifying the energy resources, the improvement the national industry and the independence of foreign suppliers. Romanian Nuclear Power Program envisaged a large national participation in Cernavoda NPP completion, in the development of nuclear fuel cycle facilities and horizontal industry, in R and D and human resources. As consequence, important support was being given to development of industries involved in Nuclear Fuel Cycle and manufacturing of equipment and nuclear materials based on technology transfer, implementation of advanced design execution standards, QA procedures and current nuclear safety requirements at international level. Unit 1 of the first Romanian nuclear power plant, Cernavoda NPP with a final profile 5x700 Mw e, is now in operation and its production represents 10% of all national electricity production. There were also developed all stages of FRONT END of Nuclear Fuel Cycle as well as programs for spent fuel and waste management. Industrial facilities for uranian production, U 3 O 8 concentrate, UO 2 powder and CANDU fuel bundles, as well as heavy water plant, supply the required fuel and heavy water for Cernavoda NPP. The paper presents the Romanian activities in Nuclear Fuel Cycle and waste management fields. (authors)

  12. The DUPIC fuel development program in KAERI

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

  13. Review of progress on enhanced accident tolerant fuel

    International Nuclear Information System (INIS)

    McCoy, K.; Dunn, B.; Kochendarfer, R.

    2015-01-01

    The accident at Fukushima has resulted in renewed interest in understanding the performance of nuclear power plants under accident conditions. Part of that interest is directed toward determining how to improve the performance of fuel during an accident that involves long exposures of the fuel to high temperatures. This paper describes the method being used by AREVA to select and evaluate approaches for improving the accident tolerance of nuclear fuel. The method involves starting with a large number of approaches that might enhance accident tolerance, and reviewing how well each approach satisfies a set of engineering requirements and goals. Among the approaches investigated we have the development of fuel pellets that contain a second phase to improve thermal conductivity, the use of molybdenum alloy tubing as fuel cladding, the use of oxidation-resistant coatings to zirconium cladding, and the use of nanoparticles in the coolant to improve heat transfer

  14. Tubular solid oxide fuel cell development program

    Energy Technology Data Exchange (ETDEWEB)

    Ray, E.R.; Cracraft, C.

    1995-12-31

    This paper presents an overview of the Westinghouse Solid Oxide Fuel Cell (SOFC) development activities and current program status. The Westinghouse goal is to develop a cost effective cell that can operate for 50,000 to 100,000 hours. Progress toward this goal will be discussed and test results presented for multiple single cell tests which have now successfully exceeded 56,000 hours of continuous power operation at temperature. Results of development efforts to reduce cost and increase power output of tubular SOFCs are described.

  15. Progress in the developing countries

    International Nuclear Information System (INIS)

    Simnad, M.

    1981-01-01

    Nuclear programmes in selective developing countries are briefly discussed. The oil rich countries of Iraq, Libya and Iran all have reactors on order. Turkey has decided to purchase a PWR from the USSR and Egypt's programme anticipates a capacity of 6600 MWe by 2000. The current projections for India are 6000 MWe by 1990 and 20,000 MWe by 2000. The progress of Pakistan, South Korea and other Asian countries are discussed. The predicted growth in reactors and population in Latin America is considered - 17 reactors presently planned for a population of 340 million and 18-57 possible additions in 2000 for an estimated population of 600 million. The role of the IAEA and experience of some Western countries in technology transfer is discussed with the ambitious Spanish nuclear power programme and the experience of Argentina in purchasing Candu reactors. (author)

  16. FY2015 ceramic fuels development annual highlights

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-22

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

  17. FY2016 Ceramic Fuels Development Annual Highlights

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-01-24

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

  18. Development of laser-based techniques for identification of breached nuclear fuel elements in storage. Progress report, 1 October 1978-1 September 1979

    International Nuclear Information System (INIS)

    Anderson, R.J.; Esherick, P.

    1980-06-01

    From many possible laser-based techniques of atomic and molecular spectroscopy we have selected multiphoton ionization spectroscopy (MIS) as that one technique which best conforms to the constraints imposed by the problem of identifying breached nuclear fuel assemblies in storage. We describe experiments utilizing MIS to specifically and sensitively detect xenon (Xe) and nitric oxide (NO). We therefore experimentally demonstrate the applicability of this technique to the detection of volatile fission products leaking from breached nuclear fuel assemblies stored in water cooled basins. Even for the non-ideal circumstances of these preliminary experiments, we estimate a detection limit of 10 10 atoms/cm 3 for xenon, roughly 1% of the atmospheric content of xenon at sea level

  19. The UK fuel poverty strategy: 5th annual progress report 2007

    International Nuclear Information System (INIS)

    2007-12-01

    This fifth annual progress report details government progress in 2007 in tackling fuel poverty and movement towards targets. The United Kingdom were the first country in the world to recognise the issue of fuel poverty and to put in place measures to tackle the issue, including spending 20 billion pounds sterling on benefits and programmes since the year 2000. The report covers progress to date, schemes and initiatives to tackle fuel poverty, the energy market and looks ahead to the future. Progress and development of the schemes across the devolved nations are also considered. This report is the first to publish the Government's proposals for the Carbon Emissions Reduction Target (CERT) priority group, which were laid before Parliament on 5th December 2007. This report is the first to present the fuel poverty figures for 2005, and shows the effects of rising energy prices. The Government continues to take action to ensure that the energy market is working properly, and to encourage reform in the EU on energy market liberalisation - this should reduce pressure on prices. Those in fuel poverty have much to gain by switching supplier and this report outlines the action taken by Ofgem and Energywatch to encourage this

  20. Development of PEM fuel cell technology at international fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Wheeler, D.J.

    1996-04-01

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

  1. Used Fuel Disposition in Crystalline Rocks: FY16 Progress Report

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Y. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hadgu, Teklu [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Kalinina, Elena Arkadievna [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Jerden, James L. [Argonne National Lab. (ANL), Argonne, IL (United States); Copple, Jacqueline M. [Argonne National Lab. (ANL), Argonne, IL (United States); Cruse, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Ebert, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Buck, E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Eittman, R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Tinnacher, R. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Tournassat, Christophe [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Davis, J. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Viswanathan, H. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Chu, S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dittrich, T. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hyman, F. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Karra, S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Makedonska, N. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Reimus, P. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Zavarin, Mavrik [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Joseph, C. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-09-21

    The objective of the Crystalline Disposal R&D Work Package is to advance our understanding of long-term disposal of used fuel in crystalline rocks and to develop necessary experimental and computational capabilities to evaluate various disposal concepts in such media. FY16 continued to be a successful year in both experimental and modeling arenas in evaluation of used fuel disposal in crystalline rocks. The work covers a wide range of research topics identified in the R&D plan.

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

    Energy Technology Data Exchange (ETDEWEB)

    Nuvera Fuel Cells

    2005-04-15

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

  3. Co-Optimization of Fuels & Engines (Co-Optima) Initiative: Recent Progress on Light-Duty Boosted Spark-Ignition Fuels/Engines

    Energy Technology Data Exchange (ETDEWEB)

    Farrell, John

    2017-07-03

    This presentation reports recent progress on light-duty boosted spark-ignition fuels/engines being developed under the Co-Optimization of Fuels and Engines initiative (Co-Optima). Co-Optima is focused on identifying fuel properties that optimize engine performance, independent of composition, allowing the market to define the best means to blend and provide these fuels. However, in support of this, we are pursuing a systematic study of blendstocks to identify a broad range of feasible options, with the objective of identifying blendstocks that can provide target ranges of key fuel properties, identifying trade-offs on consistent and comprehensive basis, and sharing information with stakeholders.

  4. Progress in fuel pin modelling in the USA

    Energy Technology Data Exchange (ETDEWEB)

    Stephen, J D; Biancheria, A; Leibnitz, D; O' Reilly, B D; Liu, Y Y; Labar, M P; Gneiting, B C [General Electric Company, Sunnyvale, CA (United States)

    1979-12-01

    In the USA, the focus for theoretical fuel pin modeling is the LIFE system. This system of codes, algorithms, criteria and analysis guidelines is intended to provide a common basis for communication amongst the development groups, a reference set of analysis guidelines for design, and eventually a consensus on the state-of-the-art for licensing. The technical objective is to predict the effect of design options on fuel pin performance limits, which include fuel temperature, pin deformation and cladding breach during normal operation and design basis transients. The mechanistic approach to modeling is taken in LIFE to the extent possible. That is, the approach is to describe the key phenomena in sufficient detail to provide a fundamental understanding of their synergistic effect on the fuel pin performance limits.

  5. Fuel element database: developer handbook

    International Nuclear Information System (INIS)

    Dragicevic, M.

    2004-09-01

    The fuel elements database which was developed for Atomic Institute of the Austrian Universities is described. The software uses standards like HTML, PHP and SQL. For the standard installation freely available software packages such as MySQL database or the PHP interpreter from Apache Software Foundation and Java Script were used. (nevyjel)

  6. Alternative Fuels and Sustainable Development

    DEFF Research Database (Denmark)

    Jørgensen, Kaj; Nielsen, Lars Henrik

    1996-01-01

    The main report of the project on Transportation Fuels based on Renewable Energy. The report contains a review of potential technologies for electric, hybrid and hydrogen propulsion in the Danish transport sector, including an assessment of their development status. In addition, the energy...

  7. Developing fossil fuel based technologies

    International Nuclear Information System (INIS)

    Manzoori, A.R.; Lindner, E.R.

    1991-01-01

    Some of the undesirable effects of burning fossil fuels in the conventional power generating systems have resulted in increasing demand for alternative technologies for power generation. This paper describes a number of new technologies and their potential to reduce the level of atmospheric emissions associated with coal based power generation, such as atmospheric and pressurized fluid bed combustion systems and fuel cells. The status of their development is given and their efficiency is compared with that of conventional pc fired power plants. 1 tab., 7 figs

  8. Progress in Brucella vaccine development

    Science.gov (United States)

    YANG, Xinghong; SKYBERG, Jerod A.; CAO, Ling; CLAPP, Beata; THORNBURG, Theresa; PASCUAL, David W.

    2012-01-01

    Brucella spp. are zoonotic, facultative intracellular pathogens, which cause animal and human disease. Animal disease results in abortion of fetuses; in humans, it manifests flu-like symptoms with an undulant fever, with osteoarthritis as a common complication of infection. Antibiotic regimens for human brucellosis patients may last several months and are not always completely effective. While there are no vaccines for humans, several licensed live Brucella vaccines are available for use in livestock. The performance of these animal vaccines is dependent upon the host species, dose, and route of immunization. Newly engineered live vaccines, lacking well-defined virulence factors, retain low residual virulence, are highly protective, and may someday replace currently used animal vaccines. These also have possible human applications. Moreover, due to their enhanced safety and efficacy in animal models, subunit vaccines for brucellosis show great promise for their application in livestock and humans. This review summarizes the progress of brucellosis vaccine development and presents an overview of candidate vaccines. PMID:23730309

  9. Nuclear fuel cycle under progressing preparation of its systemisation

    International Nuclear Information System (INIS)

    Anon.

    2001-01-01

    Trends of nuclear development in Japan show more remarkable advancements in 2000, such as new addition of nuclear power plant, nuclear fuel cycling business, and so on. Based on an instruction of the criticality accident in JCO formed on September, 1999, government made efforts on revision of the law on regulation of nuclear reactor and so forth and establishment of a law on protection of nuclear accident as sooner, to enforce nuclear safety management and nuclear accident protective countermeasure. On the other hand, the nuclear industry field develops some new actions such as establishment of Nuclear Safety Network (NSnet)', mutual evaluation of nuclear-relative works (pier review), and so forth. And, on the high level radioactive wastes disposal of the most important subject remained in nuclear development, the Nuclear Waste Management Organization of Japan' of its main business body was established on October, 1999 together with establishment of the new law, to begin a business for embodiment of the last disposal aiming at 2030s to 2040s. On the same October, the Japan Nuclear Fuel Limited. concluded a safety agreement on premise of full-dress transportation of the used fuels to the Rokkasho Reprocessing Plant in Aomori prefecture with local government, to begin their transportation from every electric company since its year end. Here were described on development of the nuclear fuel cycling business in Japan, establishment of nuclear fuel cycling, disposal on the high level radioactive wastes, R and D on geological disposal of the high level radioactive wastes, establishment on cycle back-end of nuclear fuels, and full-dressing of nuclear fuel cycling. (G.K.)

  10. Development of alkaline fuel cells.

    Energy Technology Data Exchange (ETDEWEB)

    Hibbs, Michael R.; Jenkins, Janelle E.; Alam, Todd Michael; Janarthanan, Rajeswari; Horan, James L.; Caire, Benjamin R.; Ziegler, Zachary C.; Herring, Andrew M.; Yang, Yuan; Zuo, Xiaobing; Robson, Michael H.; Artyushkova, Kateryna; Patterson, Wendy; Atanassov, Plamen Borissov

    2013-09-01

    This project focuses on the development and demonstration of anion exchange membrane (AEM) fuel cells for portable power applications. Novel polymeric anion exchange membranes and ionomers with high chemical stabilities were prepared characterized by researchers at Sandia National Laboratories. Durable, non-precious metal catalysts were prepared by Dr. Plamen Atanassovs research group at the University of New Mexico by utilizing an aerosol-based process to prepare templated nano-structures. Dr. Andy Herrings group at the Colorado School of Mines combined all of these materials to fabricate and test membrane electrode assemblies for single cell testing in a methanol-fueled alkaline system. The highest power density achieved in this study was 54 mW/cm2 which was 90% of the project target and the highest reported power density for a direct methanol alkaline fuel cell.

  11. Advanced thermally stable jet fuels: Technical progress report, October 1994--December 1994

    Energy Technology Data Exchange (ETDEWEB)

    Schobert, H.H.; Eser, S.; Song, C.; Hatcher, P.G.; Boehman, A.; Coleman, M.M.

    1995-02-01

    There are five tasks within this project on thermally stable coal-based jet fuels. Progress on each of the tasks is described. Task 1, Investigation of the quantitative degradation chemistry of fuels, has 5 subtasks which are described: Literature review on thermal stability of jet fuels; Pyrolytic and catalytic reactions of potential endothermic fuels: cis- and trans-decalin; Use of site specific {sup 13}C-labeling to examine the thermal stressing of 1-phenylhexane: A case study for the determination of reaction kinetics in complex fuel mixtures versus model compound studies; Estimation of critical temperatures of jet fuels; and Surface effects on deposit formation in a flow reactor system. Under Task 2, Investigation of incipient deposition, the subtask reported is Uncertainty analysis on growth and deposition of particles during heating of coal-derived aviation gas turbine fuels; under Task 3, Characterization of solid gums, sediments, and carbonaceous deposits, is subtask, Studies of surface chemistry of PX-21 activated carbon during thermal degradation of jet A-1 fuel and n-dodecane; under Task 4, Coal-based fuel stabilization studies, is subtask, Exploratory screening and development potential of jet fuel thermal stabilizers over 400 C; and under Task 5, Exploratory studies on the direct conversion of coal to high quality jet fuels, are 4 subtasks: Novel approaches to low-severity coal liquefaction and coal/resid co-processing using water and dispersed catalysts; Shape-selective naphthalene hydrogenation for production of thermally stable jet fuels; Design of a batch mode and a continuous mode three-phase reactor system for the liquefaction of coal and upgrading of coal liquids; and Exploratory studies on coal liquids upgrading using mesopores molecular sieve catalysts. 136 refs., 69 figs., 24 tabs.

  12. Individual moral development and moral progress

    OpenAIRE

    Schinkel, Anders; de Ruyter, Doret J.

    2017-01-01

    At first glance, one of the most obvious places to look for moral progress is in individuals, in particular in moral development from childhood to adulthood. In fact, that moral progress is possible is a foundational assumption of moral education. Beyond the general agreement that moral progress is not only possible but even a common feature of human development things become blurry, however. For what do we mean by ‘progress’? And what constitutes moral progress? Does the idea of individual m...

  13. Advanced research reactor fuel development

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-05-01

    The fabrication technology of the U{sub 3}Si fuel dispersed in aluminum for the localization of HANARO driver fuel has been launches. The increase of production yield of LEU metal, the establishment of measurement method of homogeneity, and electron beam welding process were performed. Irradiation test under normal operation condition, had been carried out and any clues of the fuel assembly breakdown was not detected. The 2nd test fuel assembly has been irradiated at HANARO reactor since 17th June 1999. The quality assurance system has been re-established and the eddy current test technique has been developed. The irradiation test for U{sub 3}Si{sub 2} dispersed fuels at HANARO reactor has been carried out in order to compare the in-pile performance of between the two types of U{sub 3}Si{sub 2} fuels, prepared by both the atomization and comminution processes. KAERI has also conducted all safety-related works such as the design and the fabrication of irradiation rig, the analysis of irradiation behavior, thermal hydraulic characteristics, stress analysis for irradiation rig, and thermal analysis fuel plate, for the mini-plate prepared by international research cooperation being irradiated safely at HANARO. Pressure drop test, vibration test and endurance test were performed. The characterization on powders of U-(5.4 {approx} 10 wt%) Mo alloy depending on Mo content prepared by rotating disk centrifugal atomization process was carried out in order to investigate the phase stability of the atomized U-Mo alloy system. The {gamma}-U phase stability and the thermal compatibility of atomized U-16at.%Mo and U-14at.%Mo-2at.%X(: Ru, Os) dispersion fuel meats at an elevated temperature have been investigated. The volume increases of U-Mo compatibility specimens were almost the same as or smaller than those of U{sub 3}Si{sub 2}. However the atomized alloy fuel exhibited a better irradiation performance than the comminuted alloy. The RERTR-3 irradiation test of nano

  14. Recent developments in biodesulfurization of fossil fuels.

    Science.gov (United States)

    Xu, Ping; Feng, Jinhui; Yu, Bo; Li, Fuli; Ma, Cuiqing

    2009-01-01

    The emission of sulfur oxides can have adverse effects on the environment. Biodesulfurization of fossil fuels is attracting more and more attention because such a bioprocess is environmentally friendly. Some techniques of desulfurization have been used or studied to meet the stricter limitation on sulfur content in China. Recent advances have demonstrated the mechanism and developments for biodesulfurization of gasoline, diesel and crude oils by free cells or immobilized cells. Genetic technology was also used to improve sulfur removal efficiencies. In this review, we summarize recent progress mainly in China on petroleum biodesulfurization.

  15. Progress and prospects for phosphoric acid fuel cell power plants

    Energy Technology Data Exchange (ETDEWEB)

    Bonville, L.J.; Scheffler, G.W.; Smith, M.J. [International Fuel Cells Corp., South Windsor, CT (United States)

    1996-12-31

    International Fuel Cells (IFC) has developed the fuel cell power plant as a new, on-site power generation source. IFC`s commercial fuel cell product is the 200-kW PC25{trademark} power plant. To date over 100 PC25 units have been manufactured. Fleet operating time is in excess of one million hours. Individual units of the initial power plant model, the PC25 A, have operated for more than 30,000 hours. The first model {open_quotes}C{close_quotes} power plant has over 10,000 hours of operation. The manufacturing, application and operation of this power plant fleet has established a firm base for design and technology development in terms of a clear understanding of the requirements for power plant reliability and durability. This fleet provides the benchmark against which power plant improvements must be measured.

  16. Unified fuel elements development for research reactors

    International Nuclear Information System (INIS)

    Vatulin, A.; Stetsky, Y.; Dobrikova, I.

    1998-01-01

    Square cross-section rod type fuel elements have been developed for russian pool-type research reactors. new fuel elements can replace the large nomenclature of tubular fuel elements with around, square and hexahedral cross-sections and to solve a problem of enrichment reduction. the fuel assembly designs with rod type fuel elements have been developed. The overall dimensions of existing the assemblies are preserved in this one. the experimental-industrial fabricating process of fuel elements, based on a joint extrusion method has been developed. The fabricating process has been tested in laboratory conditions, 150 experimental fuel element samples of the various sizes were produced. (author)

  17. Progress in control equipment for fuel-handling machinery

    International Nuclear Information System (INIS)

    Nutting, B.A.

    1986-01-01

    The paper outlines the development of the equipment used to control the fuel-handling machinery associated with nuclear reactors, from the early electromechanical equipment, through solid-state switching logic to programmable controllers and microprocessors. The control techniques have developed along with the technology, and modern systems offer versatility, reliability and ease of design, operation and maintenance. Future trends and developments are discussed together with possible limiting factors. (author)

  18. Demonstrating the benefits of fuel cells: further significant progress towards commercialisation

    Energy Technology Data Exchange (ETDEWEB)

    Anon,

    1995-01-01

    The fourteenth Fuel Cell Seminar held in San Diego, California in 1994 is reported. The phosphoric acid fuel cell (PAFC) is the closest to widespread commercialization. PAFC cogeneration plants have to be shown to compare favourable in reliability with current mature natural gas-fuelled engine and turbine technologies. Although highly efficient, further development is necessary to produce cost effective generators. Progress is being made on proton exchange membrane fuel cell (PEMFC) stationary power plants, too, which may prove to be cost effective. In view of its lower operating temperature, at below 100[sup o]C compared with about 200[sup o]C for the PAFC, the principal use of the PEMFC has been identified as powering vehicles. Fuel cells have significant environmental advantages but further capital cost reductions are necessary if they are to compete with established technologies. (UK)

  19. Progress on matrix SiC processing and properties for fully ceramic microencapsulated fuel form

    International Nuclear Information System (INIS)

    Terrani, K.A.; Kiggans, J.O.; Silva, C.M.; Shih, C.; Katoh, Y.; Snead, L.L.

    2015-01-01

    The consolidation mechanism and resulting properties of the silicon carbide (SiC) matrix of fully ceramic microencapsulated (FCM) fuel form are discussed. The matrix is produced via the nano-infiltration transient eutectic-forming (NITE) process. Coefficient of thermal expansion, thermal conductivity, and strength characteristics of this SiC matrix have been characterized in the unirradiated state. An ad hoc methodology for estimation of thermal conductivity of the neutron-irradiated NITE–SiC matrix is also provided to aid fuel performance modeling efforts specific to this concept. Finally, specific processing methods developed for production of an optimal and reliable fuel form using this process are summarized. These various sections collectively report the progress made to date on production of optimal FCM fuel form to enable its application in light water and advanced reactors

  20. Individual moral development and moral progress

    NARCIS (Netherlands)

    Schinkel, Anders; de Ruyter, Doret J.

    At first glance, one of the most obvious places to look for moral progress is in individuals, in particular in moral development from childhood to adulthood. In fact, that moral progress is possible is a foundational assumption of moral education. Beyond the general agreement that moral progress is

  1. Individual Moral Development and Moral Progress

    NARCIS (Netherlands)

    Schinkel, Anders; de Ruyter, Doret J.

    2017-01-01

    At first glance, one of the most obvious places to look for moral progress is in individuals, in particular in moral development from childhood to adulthood. In fact, that moral progress is possible is a foundational assumption of moral education. Beyond the general agreement that moral progress is

  2. Fuel performance, design and development

    International Nuclear Information System (INIS)

    Prasad, P.N.; Tripathi, Rahul Mani; Soni, Rakesh; Ravi, M.; Vijay Kumar, S.; Dwivedi, K.P.; Pandarinathan, P.R.; Neema, L.K.

    2006-01-01

    The normal fuel configurations for operating 220 MWe and 540 MWe PHWRs are natural uranium dioxide 19-element and 37- element fuel bundle types respectively. The fuel configuration for BWRs is 6 x 6 fuel. So far, about 330 thousand PHWR fuel bundles and 3500 number of BWR bundles have been irradiated in the 14 PHWRs and 2 BWRs. Improvements in fuel design, fabrication, quality control and operating practices are continuously carried out towards improving fuel utilization as well as reducing fuel failure rate. Efforts have been put to improve the fuel bundle utilization by increasing the fuel discharge burnup of the natural uranium bundles The overall fuel failure rate currently is less than 0.1 % . Presently the core discharge burnups in different reactors are around 7500 MWD/TeU. The paper gives the fuel performance experience over the years in the different power reactors and actions taken to improve fuel performance over the years. (author)

  3. 2011 Annual Progress Report: DOE Hydrogen and Fuel Cells Program (Book)

    Energy Technology Data Exchange (ETDEWEB)

    2011-11-01

    In the past year, the DOE Hydrogen and Fuel Cells Program (the Program) made substantial progress toward its goals and objectives. The Program has conducted comprehensive and focused efforts to enable the widespread commercialization of hydrogen and fuel cell technologies in diverse sectors of the economy. With emphasis on applications that will effectively strengthen our nation's energy security and improve our stewardship of the environment, the Program engages in research, development, and demonstration of critical improvements in the technologies. Highlights of the Program's accomplishments can be found in the sub-program chapters of this report.

  4. Flowsheet development for HTGR fuel reprocessing

    International Nuclear Information System (INIS)

    Baxter, B.; Benedict, G.E.; Zimmerman, R.D.

    1976-01-01

    Development studies to date indicate that the HTGR fuel blocks can be effectively crushed with two stages of eccentric jaw crushing, followed by a double-roll crusher, a screener and an eccentrically mounted single-roll crusher for oversize particles. Burner development results indicate successful long-term operation of both the primary and secondary fluidized-bed combustion systems can be performed with the equipment developed in this program. Aqueous separation development activities have centered on adapting known Acid-Thorex processing technology to the HTGR reprocessing task. Significant progress has been made on dissolution of burner ash, solvent extraction feed preparation, slurry transfer, solids drying and solvent extraction equipment and flowsheet requirements

  5. Research and development of thorium fuel cycle

    International Nuclear Information System (INIS)

    Oishi, Jun.

    1994-01-01

    Nuclear properties of thorium are summarized and present status of research and development of the use of thorium as nuclear fuel is reviewed. Thorium may be used for nuclear fuel in forms of metal, oxide, carbide and nitride independently, alloy with uranium or plutonium or mixture of the compound. Their use in reactors is described. The reprocessing of the spent oxide fuel in thorium fuel cycle is called the thorex process and similar to the purex process. A concept of a molten salt fuel reactor and chemical processing of the molten salt fuel are explained. The required future research on thorium fuel cycle is commented briefly. (T.H.)

  6. Development of portable fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Nakatou, K.; Sumi, S.; Nishizawa, N. [Sanyo Electric Co., Ltd., Osaka (Japan)

    1996-12-31

    Sanyo Electric has been concentrating on developing a marketable portable fuel cell using phosphoric acid fuel cells (PAFC). Due to the fact that this power source uses PAFC that operate at low temperature around 100{degrees} C, they are easier to handle compared to conventional fuel cells that operate at around 200{degrees} C , they can also be expected to provide extended reliable operation because corrosion of the electrode material and deterioration of the electrode catalyst are almost completely nonexistent. This power source is meant to be used independently and stored at room temperature. When it is started up, it generates electricity itself using its internal load to raise the temperature. As a result, the phosphoric acid (the electolyte) absorbs the reaction water when the temperature starts to be raised (around room temperature). At the same time the concentration and volume of the phosphoric acid changes, which may adversely affect the life time of the cell. We have studied means for starting, operating PAFC stack using methods that can simply evaluate changes in the concentration of the electrolyte in the stack with the aim of improving and extending cell life and report on them in this paper.

  7. LIFE: Recent Developments and Progress

    International Nuclear Information System (INIS)

    Anklam, T.M.

    2011-01-01

    Test results from the NIF show excellent progress toward achieving ignition. Experiments designed to verify coupling of the laser energy to the fusion target have shown that the efficiency meets that needed for ignition. Several tests with the cryogenic targets needed for ignition have been performed, and world-record neutron output produced. The National Ignition Campaign is on schedule to meet its 2012 ignition milestone, with the next phase in the campaign due to start later this month. It has been a busy and very productive year. The NIF is in full 24/7 operations and has progressed markedly in the path toward ignition. The long-standing goal of the National Ignition Campaign to demonstrate ignition by the end of FY 2012 is on track. The LIFE plant design has matured considerably, and a delivery plan established based on close interactions with vendors. National-level reviews of fusion are underway, and are due to present initial findings later this year. A value proposition has been drafted for review. The LIFE project is ready to move into the delivery phase.

  8. Canadian fuel cell commercialization roadmap update : progress of Canada's hydrogen and fuel cell industry

    International Nuclear Information System (INIS)

    Filbee, S.; Karlsson, T.

    2009-01-01

    Hydrogen and fuel cells are considered an essential part of future low-carbon energy systems for transportation and stationary power. In recognition of this, Industry Canada has worked in partnership with public and private stakeholders to provide an update to the 2003 Canadian Fuel Cell Commercialization Roadmap to determine infrastructure requirements for near-term markets. The update includes technology and market developments in terms of cost and performance. This presentation included an overview of global hydrogen and fuel cell markets as background and context for the activities of the Canadian industry. Approaches toward commercial viability and mass market success were also discussed along with possible scenarios and processes by which these mass markets could develop. Hydrogen and fuel cell industry priorities were outlined along with recommendations for building a hydrogen infrastructure

  9. Used Fuel Disposal in Crystalline Rocks. FY15 Progress Report

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Yifeng [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-08-20

    The objective of the Crystalline Disposal R&D Work Package is to advance our understanding of long-term disposal of used fuel in crystalline rocks and to develop necessary experimental and computational capabilities to evaluate various disposal concepts in such media. Chapter headings are as follows: Fuel matrix degradation model and its integration with performance assessments, Investigation of thermal effects on the chemical behavior of clays, Investigation of uranium diffusion and retardation in bentonite, Long-term diffusion of U(VI) in bentonite: dependence on density, Sorption and desorption of plutonium by bentonite, Dissolution of plutonium intrinsic colloids in the presence of clay and as a function of temperature, Laboratory investigation of colloid-facilitated transport of cesium by bentonite colloids in a crystalline rock system, Development and demonstration of discrete fracture network model, Fracture continuum model and its comparison with discrete fracture network model.

  10. Dry process fuel performance technology development

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kweon Ho; Kim, K. W.; Kim, B. K. (and others)

    2006-06-15

    The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase III R and D. In order to fulfil this objectives, property model development of DUPIC fuel and irradiation test was carried out in Hanaro using the instrumented rig. Also, the analysis on the in-reactor behavior analysis of DUPIC fuel, out-pile test using simulated DUPIC fuel as well as performance and integrity assessment in a commercial reactor were performed during this Phase. The R and D results of the Phase III are summarized as follows: Fabrication process establishment of simulated DUPIC fuel for property measurement, Property model development for the DUPIC fuel, Performance evaluation of DUPIC fuel via irradiation test in Hanaro, Post irradiation examination of irradiated fuel and performance analysis, Development of DUPIC fuel performance code (KAOS)

  11. Dry process fuel performance technology development

    International Nuclear Information System (INIS)

    Kang, Kweon Ho; Kim, K. W.; Kim, B. K.

    2006-06-01

    The objective of the project is to establish the performance evaluation system of DUPIC fuel during the Phase III R and D. In order to fulfil this objectives, property model development of DUPIC fuel and irradiation test was carried out in Hanaro using the instrumented rig. Also, the analysis on the in-reactor behavior analysis of DUPIC fuel, out-pile test using simulated DUPIC fuel as well as performance and integrity assessment in a commercial reactor were performed during this Phase. The R and D results of the Phase III are summarized as follows: Fabrication process establishment of simulated DUPIC fuel for property measurement, Property model development for the DUPIC fuel, Performance evaluation of DUPIC fuel via irradiation test in Hanaro, Post irradiation examination of irradiated fuel and performance analysis, Development of DUPIC fuel performance code (KAOS)

  12. Development of high burnup nuclear fuel technology

    International Nuclear Information System (INIS)

    Suk, Ho Chun; Kang, Young Hwan; Jung, Jin Gone; Hwang, Won; Park, Zoo Hwan; Ryu, Woo Seog; Kim, Bong Goo; Kim, Il Gone

    1987-04-01

    The objectives of the project are mainly to develope both design and manufacturing technologies for 600 MWe-CANDU-PHWR-type high burnup nuclear fuel, and secondly to build up the foundation of PWR high burnup nuclear fuel technology on the basis of KAERI technology localized upon the standard 600 MWe-CANDU- PHWR nuclear fuel. So, as in the first stage, the goal of the program in the last one year was set up mainly to establish the concept of the nuclear fuel pellet design and manufacturing. The economic incentives for high burnup nuclear fuel technology development are improvement of fuel utilization, backend costs plant operation, etc. Forming the most important incentives of fuel cycle costs reduction and improvement of power operation, etc., the development of high burnup nuclear fuel technology and also the research on the incore fuel management and safety and technologies are necessary in this country

  13. State-of-the-Art Report on the Progress of Nuclear Fuel Cycle Chemistry

    International Nuclear Information System (INIS)

    Collins, E.D.; DelCul, G.D.; Spencer, B.B.; Jubin, R.T.; Maher, C.; Kim, I.-T.; Lee, H.; Federov, Yu. S.; Saprykin, V.F.; Beznosyuk, V.I.; Kolyadin, A.B.; Baron, P.; Miguirditchian, M.; Sorel, C.; Morita, Y.; Taylor, R.; Khaperskaya, A.; Hill, C.; Malmbeck, R.; Law, J.; Angelis, G. de; Boucher, L.; Xeres, X.; Collins, E.; Mendes, E.; Lee, H.-S.; Inoue, T.; Glatz, J.P.; Kormilitsyn, M.; Uhlir, J.; Ignatiev, V.; Serp, J.; Delpech, S.

    2018-01-01

    The implementation of advanced nuclear systems requires that new technologies associated with the back end of the fuel cycle are developed. The separation of minor actinides from other fuel components is one of the advanced concepts being studied to help close the nuclear fuel cycle and to improve the long-term effects on the performance of geological repositories. Separating spent fuel elements and subsequently converting them through transmutation into short-lived nuclides should considerably reduce the long-term risks associated with nuclear power generation. R and D programs worldwide are attempting to address such challenges, and many processes for advanced reprocessing and partitioning minor actinides are being developed. This report provides a comprehensive overview of progress on separation chemistry processes, and in particular on the technologies associated with the separation and recovery of minor actinides for recycling so as to help move towards the implementation of advanced fuel cycles. The report examines both aqueous and pyro processes, as well as the status of current and proposed technologies described according to the hierarchy of separations targeting different fuel components. The process criteria that will affect technology down-selection are also reviewed, as are non-proliferation requirements. The maturity of different reprocessing techniques are assessed using a scale based on the technology readiness level, and perspectives for future R and D are reviewed

  14. Fuel utilization in a progressive conversion reactor (PCR)

    International Nuclear Information System (INIS)

    Leyse, C.F.; Judd, J.L.

    1981-05-01

    Preliminary studies indicate that for once-through fuel cycles, the PCR offers potential improvements over current LWRs in the following major areas: improved uranium utilization (reduced uranium demand), degraded plutonium product in spent fuel, reduced plutonium content of spent fuel, reduced amount of spent fuel, reduced fissile content of spent fuel, and reduced separative work

  15. Canadian CANDU fuel development program and recent fuel operating experience

    International Nuclear Information System (INIS)

    Lau, J.H.K.; Inch, W.W.R.; Cox, D.S.; Steed, R.G.; Kohn, E.; Macici, N.N.

    1999-01-01

    This paper reviews the performance of the CANDU fuel in the Canadian CANDU reactors in 1997 and 1998. The operating experience demonstrates that the CANDU fuel has performed very well. Over the 2-year period, the fuel-bundle defect rate for all bundles irradiated in the Canadian CANDU reactors has remained very low, at between 0.006% to 0.016%. On a fuel element basis, this represents an element defect rate of less than about 0.0005%. One of the reasons for the good fuel performance is the support provided by the Canadian fuel research and development programs. These programs address operational issues and provide evolutionary improvements to the fuel products. The programs consist of the Fuel Technology Program, funded by the CANDU Owners Group, and the Advanced Fuel and Fuel Cycles Technology Program, funded by Atomic Energy of Canada Ltd. These 2 programs, which have been in place for many years, complement each other by sharing expert resources and experimental facilities. This paper describes the programs in 1999/2000, to provide an overview of the scope of the programs and the issues that these programs address. (author)

  16. A summary of the program and progress to 1984 December of the Canadian nuclear fuel waste management program

    International Nuclear Information System (INIS)

    Dixon, R.S.

    1986-08-01

    The Canadian Nuclear Fuel Waste Management Program involves research into the storage and transportation of used nuclear fuel, immobilization of fuel waste, and deep geological disposal of the immobilized waste. The program is now in the fifth year of a ten-year generic research and development phase. The objective of this phase of the program is to assess the safety and environmental aspects of the deep underground disposal of immobilized fuel waste in plutonic rock. The objectives of the research for each component of the program and the progress made to the end of 1984 are described in this report. 74 refs

  17. Development of nuclear fuel cycle technologies

    International Nuclear Information System (INIS)

    Suzuoki, Akira; Matsumoto, Takashi; Suzuki, Kazumichi; Kawamura, Fumio

    1995-01-01

    In the long term plan for atomic energy that the Atomic Energy Commission decided the other day, the necessity of the technical development for establishing full scale fuel cycle for future was emphasized. Hitachi Ltd. has engaged in technical development and facility construction in the fields of uranium enrichment, MOX fuel fabrication, spent fuel reprocessing and so on. In uranium enrichment, it took part in the development of centrifuge process centering around Power Reactor and Nuclear Fuel Development Corporation (PNC), and took its share in the construction of the Rokkasho uranium enrichment plant of Japan Nuclear Fuel Service Co., Ltd. Also it cooperates with Laser Enrichment Technology Research Association. In Mox fuel fabrication, it took part in the construction of the facilities for Monju plutonium fuel production of PNC, for pellet production, fabrication and assembling processes. In spent fuel reprocessing, it cooperated with the technical development of maintenance and repair of Tokai reprocessing plant of PNC, and the construction of spent fuel stores in Rokkasho reprocessing plant is advanced. The centrifuge process and the atomic laser process of uranium enrichment are explained. The high reliability of spent fuel reprocessing plants and the advancement of spent fuel reprocessing process are reported. Hitachi Ltd. Intends to exert efforts for the technical development to establish nuclear fuel cycle which increases the importance hereafter. (K.I.)

  18. Progress in fuel processing for PEMFC systems for transport applications

    Energy Technology Data Exchange (ETDEWEB)

    Dams, A J; Hayter, P R; Moore, S C

    1998-07-01

    Wellman CJB Limited has been developing fuel processors for PEMFC systems for transport applications using a range of feedstocks. Feedstocks that can be processed to produce a hydrogen-rich gas stream suitable for use with a PEMFC include methanol, gasoline, diesel, LPG, dimethylether, marine diesel and aviation fuel. The basic fuel processor is a steam reformer combined with a selective carbon monoxide oxidation stage. Depending on the nature of the liquid feedstock, other process steps will be required such as vaporizer, desulfurizer, pre-reformer and high and low temperature shift reactors. Work on methanol reforming has been specifically targeted at a PEMFC driven passenger car as part of a multinational European Community JOULE programme called MERCATOX. The objective is to develop and test a compact and fast response methanol reformer and gas clean-up unit to meet a car manufacturer's specification. The method of construction is to coat a methanol reforming catalyst onto one side of a metal corrugated plate. On the other side is a coated combustion catalyst which burns fuel cell off-gas to provide the endothermic heat for the methanol reforming reaction. A number of these plates are assembled in a compact unit ensuring good heat transfer. The gas clean-up unit is similarly constructed with a selective oxidation catalyst on one side and a thermal fluid on the other. Initial tests have indicated a superior performance to conventional packed bed reformers in terms of response and start-up time. Steam reforming of gasoline, diesel, LPG, dimethylether, marine diesel and aviation fuel has been demonstrated on a bench scale (0.5kW). The process steps commence with vaporization (except for LPG), desulfurization (except for dimethylether), prereforming, reforming, low and high temperature shift and selective oxidation. A simple technology demonstrator has shown that a hydrogen-rich mixture (75% hydrogen, 25% carbon dioxide) with less than 2ppm carbon monoxide can be

  19. Spent fuel: prediction model development

    International Nuclear Information System (INIS)

    Almassy, M.Y.; Bosi, D.M.; Cantley, D.A.

    1979-07-01

    The need for spent fuel disposal performance modeling stems from a requirement to assess the risks involved with deep geologic disposal of spent fuel, and to support licensing and public acceptance of spent fuel repositories. Through the balanced program of analysis, diagnostic testing, and disposal demonstration tests, highlighted in this presentation, the goal of defining risks and of quantifying fuel performance during long-term disposal can be attained

  20. Numerical simulation of progressive BWR fuel inlet orifices

    International Nuclear Information System (INIS)

    Sara Lundgren; Hernan Tinoco; Aleksander Pohl; Wiktor Frid

    2005-01-01

    Full text of publication follows: A 'progressive' orifice is characterized by an edge-shaped hole that gives a Reynolds number dependent resistance coefficient. For Reynolds numbers smaller than a critical one, the resistance coefficient has a high constant value that drops to a much lower value for Reynolds numbers greater than this critical value. A similar effect is widely known for external flows around bodies of different shapes, i. e. spheres, cylinders, etc., and the sudden drop in drag coefficient is due to the shift from laminar to turbulent boundary-layer flow. Experimentally, progressive orifices have been investigated under high-pressure and high-temperature conditions by Akiba et al. (2001) for a reduced set of geometrical parameters. Using the sparse experimental data, a core stability study was carried out by Forsmaks Kraftgrupp AB that showed an improvement in core stability but without the expected reduction in pump power at normal operation. The reason for this partial success was the impossibility of optimizing the fuel inlet pressure drop owing to the limited amount of available data. Due to the high costs associated with the experimental generation of high-pressure, high-temperature data, it was considered that, if possible, the lacking data could be generated numerically at much lower cost. Therefore, the present work deals with the possibility of numerically simulate the flow through progressive orifices, and with the conditions under which to reproduce and generate resistance coefficient data by means of a commercial CFD-code. The results obtained with a two-dimensional, axisymmetric approximation show that Reynolds Averaged Navier-Stokes (RANS) turbulence models are able to qualitatively capture the physics of the phenomenon but with an earlier transition to turbulent boundary-layer flow and with an underestimation of the resistance coefficient by approximately 20 %. This underestimation of the resistance coefficient is related to the two

  1. Numerical simulation of progressive BWR fuel inlet orifices

    Energy Technology Data Exchange (ETDEWEB)

    Sara Lundgren; Hernan Tinoco [Forsmarks Kraftgrupp AB, 742 03 Oesthammar (Sweden); Aleksander Pohl; Wiktor Frid [The Royal Institute of Technology, Dept. Energy Technology, SE-100 44 Stockholm (Sweden)

    2005-07-01

    Full text of publication follows: A 'progressive' orifice is characterized by an edge-shaped hole that gives a Reynolds number dependent resistance coefficient. For Reynolds numbers smaller than a critical one, the resistance coefficient has a high constant value that drops to a much lower value for Reynolds numbers greater than this critical value. A similar effect is widely known for external flows around bodies of different shapes, i. e. spheres, cylinders, etc., and the sudden drop in drag coefficient is due to the shift from laminar to turbulent boundary-layer flow. Experimentally, progressive orifices have been investigated under high-pressure and high-temperature conditions by Akiba et al. (2001) for a reduced set of geometrical parameters. Using the sparse experimental data, a core stability study was carried out by Forsmaks Kraftgrupp AB that showed an improvement in core stability but without the expected reduction in pump power at normal operation. The reason for this partial success was the impossibility of optimizing the fuel inlet pressure drop owing to the limited amount of available data. Due to the high costs associated with the experimental generation of high-pressure, high-temperature data, it was considered that, if possible, the lacking data could be generated numerically at much lower cost. Therefore, the present work deals with the possibility of numerically simulate the flow through progressive orifices, and with the conditions under which to reproduce and generate resistance coefficient data by means of a commercial CFD-code. The results obtained with a two-dimensional, axisymmetric approximation show that Reynolds Averaged Navier-Stokes (RANS) turbulence models are able to qualitatively capture the physics of the phenomenon but with an earlier transition to turbulent boundary-layer flow and with an underestimation of the resistance coefficient by approximately 20 %. This underestimation of the resistance coefficient is related to

  2. Business progress towards sustainable development

    International Nuclear Information System (INIS)

    Stigson, Bjorn

    1998-01-01

    The executive director of the World Business Council for Sustainable Development described the organisation, its membership and its objectives. The organisation believes nuclear energy is needed in support of the goal of eradicating poverty, but it must also make all-round financial sense. If the risks are perceived to be high then investors expect a high financial return. The argument is supported by discussions on: (i) industry and sustainable development; (ii) the driving process;(iii) the way ahead; (iv) the environment and shareholder value; (v) conclusions for business in general and (vi) conclusions for the nuclear industry.(UK)

  3. Fuel cell development for transportation: Catalyst development

    Energy Technology Data Exchange (ETDEWEB)

    Doddapaneni, N. [Sandia National Lab., Albuquerque, NM (United States)

    1996-04-01

    Fuel cells are being considered as alternate power sources for transportation and stationary applications. With proton exchange membrane (PEM) fuel cells the fuel crossover to cathodes causes severe thermal management and cell voltage drop due to oxidation of fuel at the platinized cathodes. The main goal of this project was to design, synthesize, and evaluate stable and inexpensive transition metal macrocyclic catalysts for the reduction of oxygen and be electrochemically inert towards anode fuels such as hydrogen and methanol.

  4. Development of An Advanced JP-8 Fuel

    Science.gov (United States)

    1993-12-01

    included the Microthermal Precipitation Test (MTP), Fuel Reactor Test, Hot Liquid Process Simulator (HLPS), and Isothermal Corrosion Oxidation Test (ICOT... Microthermal Precipitation Test The impetus for this development effort was the need for a screening test that could discriminate between fuels of...varying propensity to produce thermally induced insoluble particulate material in the bulk fuel. The Microthermal Precipitation (MTP) test thermally

  5. Spent fuel storage process equipment development

    International Nuclear Information System (INIS)

    Park, Hyun Soo; Lee, Jae Sol; Yoo, Jae Hyung

    1990-02-01

    Nuclear energy which is a major energy source of national energy supply entails spent fuels. Spent fuels which are high level radioactive meterials, are tricky to manage and need high technology. The objectives of this study are to establish and develop key elements of spent fuel management technologies: handling equipment and maintenance, process automation technology, colling system, and cleanup system. (author)

  6. PWR fuel monitoring: recent progress with hot cells' examination equipment

    International Nuclear Information System (INIS)

    Chenebault, P.

    1989-01-01

    The 'hot' laboratories set up by the French Atomic Energy Authority (CEA) in its nuclear research centers at Saclay and Grenoble, and by the French Electricity Board (EDF) on the Chinon nuclear power station site, are used for dismantling and examining fuel rod assemblies irradiated in PWRs. This article is limited to a description of a number of new or totally updated items of equipment in these laboratories. Nuclear industry companies are also participating in the development of new examination techniques. As an example, the use of wave-guides for remote transmission of signals in a radioactive environment is described. 2 figs

  7. Specialists' meeting on gas-cooled reactor fuel development and spent fuel treatment

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1985-07-01

    Topics covered during the 'Specialists' meeting on gas-cooled reactor fuel development and spent fuel treatment' were as follows: Selection of constructions and materials, fuel element development concepts; Fabrication of spherical coated fuel particles and fuel element on their base; investigation of fuel properties; Spent fuel treatment and storage; Head-end processing of HTGR fuel elements; investigation of HTGR fuel regeneration process; applicability of gas-fluorine technology of regeneration of spent HTGR fuel elements.

  8. Specialists' meeting on gas-cooled reactor fuel development and spent fuel treatment

    International Nuclear Information System (INIS)

    1985-01-01

    Topics covered during the 'Specialists' meeting on gas-cooled reactor fuel development and spent fuel treatment' were as follows: Selection of constructions and materials, fuel element development concepts; Fabrication of spherical coated fuel particles and fuel element on their base; investigation of fuel properties; Spent fuel treatment and storage; Head-end processing of HTGR fuel elements; investigation of HTGR fuel regeneration process; applicability of gas-fluorine technology of regeneration of spent HTGR fuel elements

  9. Globalization of the nuclear fuel cycle impact of developments on fuel management

    Energy Technology Data Exchange (ETDEWEB)

    Van Den Durpel, L.; Bertel, E. [OCDE-NEA, Nuclear Development Div., 92 - Issy-les-Moulineaux (France)

    1999-07-01

    Nuclear energy will have to cope more and more with a rapid changing environment due to economic competitive pressure and the de-regulatory progress. In current economic environment, utilities will have to focus strongly on the reduction of their total generation costs, covering the fuel cycle costs, which are only partly under their control. Developments in the fuel cycle will be in the short-term rather evolutionary addressing the current needs of utilities. However, within the context of sustainable development and more and more inclusion of externalities in energy generation costs, more performing developments in the fuel cycle could become important and feasible. A life-cycle design approach of the fuel cycle will be requested in order to cover all factors in order to decrease significantly the nuclear energy generation cost to compete with other alternative fuels in the long-term. This paper will report on some of the trends one could distinguish in the fuel cycle with emphasis on cost reduction. OECD/NEA is currently conducting a study on the fuel cycle aiming to assess current and future nuclear fuel cycles according the potential for further improvement of the full added-value chain of these cycles from a mainly technological and economical perspective including environmental and social considerations. (authors)

  10. Globalisation of the nuclear fuel cycle - impact of developments on fuel management

    International Nuclear Information System (INIS)

    Durpel, L. van den; Bertel, E.

    2000-01-01

    Nuclear energy will have to cope more and more with a rapid changing environment due to economic competitive pressure and the deregulatory progress. In current economic environment, utilities will have to focus strongly on the reduction of their total generation costs, covering the fuel cycle costs, which are only partly under their control. Developments in the fuel cycle will be in the short-term rather evolutionary addressing the current needs of utilities. However, within the context of sustainable development and more and more inclusion of externalities in energy generation costs, more performing developments in the fuel cycle could become important and feasible. A life-cycle design approach of the fuel cycle will be requested in order to cover all factors in order to decrease significantly the nuclear energy generation cost to complete with other alternative fuels in the long-term. This paper will report on some of the trends one could distinguish in the fuel cycle with emphasis on cost reduction. OECD/NEA is currently conducting a study on the fuel cycle aiming to assess current and future nuclear fuel cycles according to the potential for further improvement of the full added-value chain of these cycles from a mainly technological and economic perspective including environmental and social considerations. (orig.) [de

  11. Consolidated fuel reprocessing. Program progress report, April 1-June 30, 1980

    Energy Technology Data Exchange (ETDEWEB)

    1980-09-01

    This progress report is compiled from major contributions from three programs: (1) the Advanced Fuel Recycle Program at ORNL; (2) the Converter Fuel Reprocessing Program at Savannah River Laboratory; and (3) the reprocessing components of the HTGR Fuel Recycle Program, primarily at General Atomic and ORNL. The coverage is generally overview in nature; experimental details and data are limited.

  12. Progress in development of MOSART concept with Th support

    Energy Technology Data Exchange (ETDEWEB)

    Ignatiev, V.; Feynberg, O.; Merzlyakov, A.; Surenkov, A.; Zagnitko, A. [National Research Center, Kurchatov Inst., Moscow (Russian Federation); Afonichkin, V.; Bovet, A.; Khokhlov, V. [Inst. of High Temperature Electrochemisty, Ekaterinburg (Russian Federation); Subbotin, V.; Fazilov, R.; Gordeev, M.; Panov, A.; Toropov, A. [Inst. of Technical Physics, Snezhinsk (Russian Federation)

    2012-07-01

    A study is under progress to examine the feasibility of Molten Salt Actinide Recycler and Transmuter system with U-Th support (MOSART+Th) fuelled with different compositions of actinide tri-fluorides (AnF{sub 3}) from used LWR fuel. New fast-spectrum design options with homogeneous core and fuel salt with high enough solubility for AnF{sub 3} are being examined because of new goals. Experimental data base created was used for further development of MOSART+Th technology as applied to consumption of AnF{sub 3} while extracting their energy. The flexibility of single fluid MOSART+Th concept fuel cycle is underlined, particularly, possibility of its operation in self-sustainable mode (CR=1) using different loadings and make up. The paper summarizes the current status of the design data for the MOSART+Th concept received within ISTC 3749 Task. (authors)

  13. Spent fuel shipping cask development status

    International Nuclear Information System (INIS)

    Henry, K.H.; Lattin, W.C.

    1989-01-01

    The Nuclear Waste Policy Act of 1982 (NWPA) authorized the US Department of Energy (DOE) to establish a national system for the disposal of spent nuclear fuel and high-level radioactive waste from commercial power generation, and established the Office of Civilian Radioactive Waste Management (OCRWM) within the DOE-Headquarters (DOE-HQ) to carry out these duties. A 1985 presidential decision added the disposal of high-level radioactive waste generated by defense programs to the national disposal system. A primary element of the disposal program is the development and operation of a transportation system to move the waste from its present locations to the facilities that will be included in the waste management system. The primary type of disposal facility to be established is a geologic repository; a Monitored Retrievable Storage (MRS) facility may also be included as an intermediate step in the nuclear waste disposal process. This paper focuses on the progress and status of one facet of the transportation program--the development of a family of shipping casks for transporting spent fuel from nuclear power reactor sites to the repository of MRS facility

  14. Concept and development status of fast breeder reactor fuels in the FaCT project

    International Nuclear Information System (INIS)

    Maeda, S.; Suzuki, M.; Kaito, T.; Tanaka, K.; Abe, T.

    2013-01-01

    The fuel development and the conceptual design study have been progressed in the first phase of the FaCT project in Japan. Significant outcomes of key technologies related to fuel design, fuel properties, core materials, fuel fabrication have been provided. The prospects of these technologies have been identified. After the Fukushima accident, the research and development for reducing the amount and toxic level of radioactive wastes will be promoted more than before. These outcomes will be reflected on the future development

  15. Development of Nuclear Fuel Remote Fabrication Technology

    International Nuclear Information System (INIS)

    Lee, Jung Won; Yang, M. S.; Kim, S. S. and others

    2005-04-01

    The aim of this study is to develop the essential technology of dry refabrication using spent fuel materials in a laboratory scale on the basis of proliferation resistance policy. The emphasis is placed on the assessment and the development of the essential technology of dry refabrication using spent fuel materials. In this study, the remote fuel fabrication technology to make a dry refabricated fuel with an enhanced quality was established. And the instrumented fuel pellets and mini-elements were manufactured for the irradiation testing in HANARO. The design and development technology of the remote fabrication equipment and the remote operating and maintenance technology of the equipment in hot cell were also achieved. These achievements will be used in and applied to the future back-end fuel cycle and GEN-IV fuel cycle and be a milestone for Korea to be an advanced nuclear country in the world

  16. Progress of the DUPIC fuel compatibility analysis (I) - reactor physics

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok; Jeong, Chang Joon; Roh, Gyu Hong; Rhee, Bo Wook; Park, Jee Won

    2003-12-01

    Since 1992, the direct use of spent pressurized water reactor fuel in CANada Deuterium Uranium (CANDU) reactors (DUPIC) has been studied as an alternative to the once-through fuel cycle. The DUPIC fuel cycle study is focused on the technical feasibility analysis, the fabrication of DUPIC fuels for irradiation tests and the demonstration of the DUPIC fuel performance. The feasibility analysis was conducted for the compatibility of the DUPIC fuel with existing CANDU-6 reactors from the viewpoints of reactor physics, reactor safety, fuel cycle economics, etc. This study has summarized the intermediate results of the DUPIC fuel compatibility analysis, which includes the CANDU reactor physics design requirements, DUPIC fuel core physics design method, performance of the DUPIC fuel core, regional overpower trip setpoint, and the CANDU primary shielding. The physics analysis showed that the CANDU-6 reactor can accommodate the DUPIC fuel without deteriorating the physics design requirements by adjusting the fuel management scheme if the fissile content of the DUPIC fuel is tightly controlled.

  17. Development of CANDU advanced fuel bundle

    International Nuclear Information System (INIS)

    Suk, H. C.; Hwang, W.; Rhee, B. W.; Jung, S. H.; Chung, C. H.

    1992-05-01

    This research project is underway in cooperation with AECL to develop the CANDU advanced fuel bundle (so-called, CANFLEX) which can enhance reactor safety and fuel economy in comparison with the current CANDU fuel and which can be used with natural uranium, slightly enriched uranium and other advanced fuel cycle. As the final schedule, the advanced fuel will be verified by carrying out a large scale demonstration of the bundle irradiation in a commercial CANDU reactor for 1996 and 1997, and consequently will be used in the existing and future CANDU reactors in Korea. The research activities during this year include the detail design of CANFLEX fuel with natural enriched uranium (CANFLEX-NU). Based on this design, CANFLEX fuel was mocked up. Out-of-pile hydraulic scoping tests were conducted with the fuel in the CANDU Cold Test Loop to investigate the condition under which maximum pressure drop occurs and the maximum value of the bundle pressure drop. (Author)

  18. Fuel Fabrication Capability Research and Development Plan

    Energy Technology Data Exchange (ETDEWEB)

    Senor, David J.; Burkes, Douglas

    2014-04-17

    The purpose of this document is to provide a comprehensive review of the mission of the Fuel Fabrication Capability (FFC) within the Global Threat Reduction Initiative Convert Program, along with research and development (R&D) needs that have been identified as necessary to ensuring mission success. The design and fabrication of successful nuclear fuels must be closely linked endeavors. Therefore, the overriding motivation behind the FFC R&D program described in this plan is to foster closer integration between fuel design and fabrication to reduce programmatic risk. These motivating factors are all interrelated, and progress addressing one will aid understanding of the others. The FFC R&D needs fall into two principal categories, 1) baseline process optimization, to refine the existing fabrication technologies, and 2) manufacturing process alternatives, to evaluate new fabrication technologies that could provide improvements in quality, repeatability, material utilization, or cost. The FFC R&D Plan examines efforts currently under way in regard to coupon, foil, plate, and fuel element manufacturing, and provides recommendations for a number of R&D topics that are of high priority but not currently funded (i.e., knowledge gaps). The plan ties all FFC R&D efforts into a unified vision that supports the overall Convert Program schedule in general, and the fabrication schedule leading up to the MP-1 and FSP-1 irradiation experiments specifically. The fabrication technology decision gates and down-selection logic and schedules are tied to the schedule for fabricating the MP-1 fuel plates, which will provide the necessary data to make a final fuel fabrication process down-selection. Because of the short turnaround between MP-1 and the follow-on FSP-1 and MP-2 experiments, the suite of specimen types that will be available for MP-1 will be the same as those available for FSP-1 and MP-2. Therefore, the only opportunity to explore parameter space and alternative processing

  19. Pellet fueling development at ORNL

    International Nuclear Information System (INIS)

    Combs, S.K.; Milora, S.L.; Foster, C.A.; Schuresko, D.D.; Foust, C.R.; Simmons, D.W.; Beard, D.S.

    1986-09-01

    Advanced plasma fueling systems for magnetic confinement devices are being developed at the Oak Ridge National Laboratory (ORNL). The general approach is that of producing and accelerating frozen hydrogenic pellets at speeds in the range of 1-2 km/s and higher. Two specific concepts are under development: (1) high-speed pneumatic acceleration; and (2) mechanical (centrifugal) acceleration. Both approaches are being pursued to meet the projected pellet size and delivery rates for major near-term plasma confinement devices, such as the Tokamak Fusion Test Reactor (TFTR), Tore Supra, the Joint European Torus (JET), JT-60, and Doublet III-D (DIII-D), as well as future applications. In addition to these confinement physics related activities, ORNL is pursuing advanced technologies to achieve pellet velocities significantly in excess of the 2-km/s range already attained with pneumatic injectors and has embarked on a development program designed to explore the feasibility of fabricating and accelerating tritium pellets. This paper describes these ongoing activities

  20. Development of a solid oxide fuel cell (SOFC) automotive auxiliary power unit (APU) fueled by gasoline

    International Nuclear Information System (INIS)

    DeMinco, C.; Mukerjee, S.; Grieve, J.; Faville, M.; Noetzel, J.; Perry, M.; Horvath, A.; Prediger, D.; Pastula, M.; Boersma, R.; Ghosh, D.

    2000-01-01

    This paper describes the design and the development progress of a 3 to 5 auxiliary power unit (APU) based on a gasoline fueled solid oxide fuel cell (SOFC). This fuel cell was supplied reformate gas (reactant) by a partial oxidation (POx) catalytic reformer utilizing liquid gasoline and designed by Delphi Automotive Systems. This reformate gas consists mainly of hydrogen, carbon monoxide and nitrogen and was fed directly in to the SOFC stack without any additional fuel reformer processing. The SOFC stack was developed by Global Thermoelectric and operates around 700 o C. This automotive APU produces power to support future 42 volt vehicle electrical architectures and loads. The balance of the APU, designed by Delphi Automotive Systems, employs a packaging and insulation design to facilitate installation and operation on-board automobiles. (author)

  1. Chemical Engineering Division Fuel Cycle Programs. Quarterly progress report, October-December 1981

    International Nuclear Information System (INIS)

    Steindler, M.J.; Bates, J.K.; Cannon, T.F.

    1982-05-01

    Methods of measuring rates of leaching from simulated waste glasses using neutron activation analysis and radiotracers have been developed. Laboratory-scale impact tests of solid alternative waste forms are being performed to obtain a size analysis of the fragments. Logging techniques are being developed to measure the relative amount of residual oil in a depleted oil reservoir by injecting gamma-active solution into it. Work to test the behavior of radionuclides leached from proposed nuclear-waste repositories using laboratory-analog experiments is in progress. High potassium levels in crushed granite from a Northern Illinois drill hole are probably derived by the leaching of potassium feldspar. Testing and development of equipment for the destructive analysis of full-length irradiated fuel rods from the LWBR are in progress. 19 figures, 13 tables

  2. Bio-fuels: results in progress, necessary adaptations

    International Nuclear Information System (INIS)

    2016-02-01

    In this report, the French Court of Auditors examines whether its assessments and recommendations published some years before about the development and use of bio-fuels in France had been taken into account. It shows that the support to bio-fuels has multiple objectives, produced some interesting results, but at high cost. Production processes are described. The authors outline that instruments had not been always coherently implemented, and that the tax system had negative effects. They notice that the objective in terms of bio-diesel share has been reached, whereas that of bio-ethanol has not. They also outline that these results have also been obtained with the help of some palliative measures, and that cost remains high for the consumer. In a second part, the report outlines that the present context calls for adaptations, notably due to its uncertainty (instability of European rules, lack of European ambition, a less promising market, a lower priority for automotive manufacturers), and proposes some perspectives and approaches of adaptation, notably to reach quantitative objectives with a greater transparency for the consumer. The report also contains answers made by the different concerned ministers

  3. General-purpose heat source project and space nuclear safety and fuels program. Progress report

    International Nuclear Information System (INIS)

    Maraman, W.J.

    1980-02-01

    Studies related to the use of 238 PuO 2 in radioisotopic power systems carried out for the Advanced Nuclear Systems and Projects Division of LASL are presented. The three programs involved are: general-purpose heat source development; space nuclear safety; and fuels program. Three impact tests were conducted to evaluate the effects of a high temperature reentry pulse and the use of CBCF on impact performance. Additionally, two 238 PuO 2 pellets were encapsulated in Ir-0.3% W for impact testing. Results of the clad development test and vent testing are noted. Results of the environmental tests are summarized. Progress on the Stirling isotope power systems test and the status of the improved MHW tests are indicated. The examination of the impact failure of the iridium shell of MHFT-65 at a fuel pass-through continued. A test plan was written for vibration testing of the assembled light-weight radioisotopic heater unit. Progress on fuel processing is reported

  4. History of fast reactor fuel development

    Energy Technology Data Exchange (ETDEWEB)

    Kittel, J.H. (Argonne National Lab., IL (United States)); Frost, B.R.T. (Argonne National Lab., IL (United States)); Mustelier, J.P. (COGEMA, Velizy-Villacoublay (France)); Bagley, K.Q. (AEA Reactor Services, Risley (United Kingdom)); Crittenden, G.C. (AEA Reactor Services, Dounreay (United Kingdom)); Dievoet, J. van (Belgonucleaire, Brussels (Belgium))

    1993-09-01

    The first fast breeder eactors, constructed in the 1945-1960 time period, used metallic fuels composed of uranium, plutonium, or their alloys. They were chosen because most existing reactor operating experience had been obtained on metallic fuels and because they provided the highest breeding ratios. Difficulties in obtaining adequate dimensional stability in metallic fuel elements under conditions of high fuel burnup led in the 1960s to the virtual worldwide choice of ceramic fuels. Although ceramic fuels provide lower breeding performance, this objective is no longer an important consideration in most national programs. Mixed uranium and plutonium dioxide became the ceramic fuel that has received the widest use. The more advanced ceramic fuels, mixed uranium and plutonium carbides and nitrides, continue under development. More recently, metal fuel elements of improved design have joined ceramic fuels in achieving goal burnups of 15 to 20 percent. Low-swelling fuel cladding alloys have also been continuously developed to deal with the unexpected problem of void formation in stainless steels subjected to fast neutron irradiation, a phenomenon first observed in the 1960s. (orig.)

  5. Advanced PEFC development for fuel cell powered vehicles

    Science.gov (United States)

    Kawatsu, Shigeyuki

    Vehicles equipped with fuel cells have been developed with much progress. Outcomes of such development efforts include a Toyota fuel cell electric vehicle (FCEV) using hydrogen as the fuel which was developed and introduced in 1996, followed by another Toyota FCEV using methanol as the fuel, developed and introduced in 1997. In those Toyota FCEVs, a fuel cell system is installed under the floor of each RAV4L, to sports utility vehicle. It has been found that the CO concentration in the reformed gas of methanol reformer can be reduced to 100 ppm in wide ranges of catalyst temperature and gas flow rate, by using the ruthenium (Ru) catalyst as the CO selective oxidizer, instead of the platinum (Pt) catalyst known from some time ago. It has been also found that a fuel cell performance equivalent to that with pure hydrogen can be ensured even in the reformed gas with the carbon monoxide (CO) concentration of 100 ppm, by using the Pt-Ru (platinum ruthenium alloy) electrocatalyst as the anode electrocatalyst of a polymer electrolyte fuel cell (PEFC), instead of the Pt electrocatalyst known from some time ago.

  6. Development of metallic fuel fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Young Ho; Lee, Chong Yak; Lee, Myung Ho and others

    1999-03-01

    With the vacuum melting and casting of the U-10wt%Zr alloy which is metallic fuel for liquid metal fast breeder reactor, we studied the microstructure of the alloy and the parameters of the melting and casting for the fuel rods. Internal defects of the U-10wt%Zr fuel by gravity casting, were inspected by non-destructive test. U-10wt%Zr alloy has been prepared for the thermal stability test in order to estimate the decomposition of the lamellar structure with relation to swelling under irradiation condition. (author)

  7. Progress and status of the international project on innovative nuclear reactors and fuel cycles (INPRO) - 5182

    International Nuclear Information System (INIS)

    Ponomarev, A.; Fesenko, G.; Grigoriev, F.G.; Korinny, A.; Phillips, J.R.; Rho, K.

    2015-01-01

    The IAEA's International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) was established in 2000 through IAEA General Conference resolution. INPRO cooperates with Member States to ensure that sustainable nuclear energy is available to help meet the energy needs of the 21. century. INPRO membership has grown to 41 members and 16 observers. The paper presents the current prospectus of the INPRO programme and details the most recent achievements in the following 7 projects: 1) the GAINS project (Global Architecture of Innovative Nuclear Energy Systems with thermal and fast reactors and a closed nuclear fuel cycle); 2) the SYNERGIES project applies and amends the analytical framework developed in GAINS project to examine more specifically the various forms of regional collaboration among nuclear energy suppliers and users; 3) the KIND project (Key Indicators for Innovative Nuclear Energy Systems) has the objective of developing guidance on the evaluation on innovative nuclear technologies; 4) the ROADMAPS project addresses several possible stages toward nuclear energy sustainability; 5) the RISC project aims at demonstrating that the evolution of safety requirements and technical innovations provide continual progress towards the avoidance of evacuation measures outside NPP sites in case of severe accidents; 6) the FANES project has the objective of carrying out feasibility analyses of advanced and innovative fuels for different reactor systems; and 7) the WIRAF project aims at identifying problematic waste from innovative reactor designs and corresponding nuclear fuel cycles

  8. Development of spent fuel dry storage technology

    International Nuclear Information System (INIS)

    Maruoka, Kunio; Matsunaga, Kenichi; Kunishima, Shigeru

    2000-01-01

    The spent fuels are the recycle fuel resources, and it is very important to store the spent fuels in safety. There are two types of the spent fuel interim storage system. One is wet storage system and another is dry storage system. In this study, the dry storage technology, dual purpose metal cask storage and canister storage, has been developed. For the dual purpose metal cask storage, boronated aluminum basket cell, rational cask body shape and shaping process have been developed, and new type dual purpose metal cask has been designed. For the canister storage, new type concrete cask and high density vault storage technology have been developed. The results of this study will be useful for the spent fuel interim storage. Safety and economical spent fuel interim storage will be realized in the near future. (author)

  9. Progress in welding studies for Canadian nuclear fuel waste disposal containers

    International Nuclear Information System (INIS)

    Maak, P.Y.Y.

    1985-11-01

    This report describes the progress in the development of closure-welding technology for Canadian nuclear fuel waste disposal containers. Titanium, copper and Inconel 625 are being investigated as candidate materials for fabrication of these containers. Gas-tungsten-arc welding, gas metal-arc-welding, resistance-heated diffusion bonding and electron beam welding have been evaluated as candidate closure welding processes. Characteristic weldment properties, relative merits of welding techniques, suitable weld joint configurations and fit-up tolerances, and welding parameter control ranges have been identified for various container designs. Furthermore, the automation requirements for candidate welding processes have been assessed. Progress in the development of a computer-controlled remote gas-shielded arc welding system is described

  10. Fast reactor fuel design and development

    International Nuclear Information System (INIS)

    Bishop, J.F.W.; Chamberlain, A.; Holmes, J.A.G.

    1977-01-01

    Fuel design parameters for oxide and carbide fast reactor fuels are reviewed in the context of minimising the total uranium demands for a combined thermal and fast reactor system. The major physical phenomena conditioning fast reactor fuel design, with a target of high burn-up, good breeding and reliable operation, are characterised. These include neutron induced void swelling, irradiation creep, pin failure modes, sub-assembly structural behaviour, behaviour of defect fuel, behaviour of alternative fuel forms. The salient considerations in the commercial scale fabrication and reprocessing of the fuels are reviewed, leading to the delineation of possible routes for the manufacture and reprocessing of Commercial Reactor fuel. From the desiderata and restraints arising from Surveys, Performance and Manufacture, the problems posed to the Designer are considered, and a narrow range of design alternatives is proposed. The paper concludes with a consideration of the development areas and the conceptual problems for fast reactors associated with those areas

  11. Development of a lightweight fuel cell vehicle

    Science.gov (United States)

    Hwang, J. J.; Wang, D. Y.; Shih, N. C.

    This paper described the development of a fuel cell system and its integration into the lightweight vehicle known as the Mingdao hydrogen vehicle (MHV). The fuel cell system consists of a 5-kW proton exchange membrane fuel cell (PEMFC), a microcontroller and other supported components like a compressed hydrogen cylinder, blower, solenoid valve, pressure regulator, water pump, heat exchanger and sensors. The fuel cell not only propels the vehicle but also powers the supporting components. The MHV performs satisfactorily over a hundred-kilometer drive thus validating the concept of a fuel cell powered zero-emission vehicle. Measurements further show that the fuel cell system has an efficiency of over 30% at the power consumption for vehicle cruise, which is higher than that of a typical internal combustion engine. Tests to improve performance such as speed enhancement, acceleration and fuel efficiency will be conducted in the future work. Such tests will consist of hybridizing with a battery pack.

  12. Development of metallic fuel materials

    International Nuclear Information System (INIS)

    Kang, Young Ho; Lee, Chong Tak; Yang, Yeoung Seok; Kim, Ki Hwan; Hwang, Sung Chan; Joo, Keun Sik; Ann, Hyun Suk; Chang, Sae Jung.

    1997-09-01

    Through the control of melting and casting parameters, the sound and homogenous U-10wt.%Zr alloy could be fabricated. The yield and segregation of Zr elements were 85% and ±0.1wt.%, and the density of the alloy was about 16.6 g/cm 3 . The major phase were α-U and δ-UZr 2 . The microstructure showed the laminar structure with fiber morphology which was arranged alternatively with uranium and Zr-rich phase. This alloy will be used for KALIMER fuel material through developing the fabrication technology and the characteristics analysis. And electrorefining study was performed to separate uranium from uranium-neodymium and uranium-zirconium alloy by their different free energy for chloride formation. The liquid cadmium phase becomes the anode of the electrorefining cell. Uranium is electrolytically transported through a molten salt electrolyte to a low carbon steel cathode. The electrolyte is composed of KCl-LiCl eutectic and some UCl 3 , which are installed in the salt to facilitate the electrotransport of uranium. In pyrochemical process the reaction condition of chlorination and the maintenance its purity in preparing UCl 4 by chlorination of UO 2 is strongly dependent on the reaction temperature and time. (author).52 refs., 40 tabs., 129 figs

  13. Key challenges and recent progress in batteries, fuel cells, and hydrogen storage for clean energy systems

    Science.gov (United States)

    Chalk, Steven G.; Miller, James F.

    Reducing or eliminating the dependency on petroleum of transportation systems is a major element of US energy research activities. Batteries are a key enabling technology for the development of clean, fuel-efficient vehicles and are key to making today's hybrid electric vehicles a success. Fuel cells are the key enabling technology for a future hydrogen economy and have the potential to revolutionize the way we power our nations, offering cleaner, more efficient alternatives to today's technology. Additionally fuel cells are significantly more energy efficient than combustion-based power generation technologies. Fuel cells are projected to have energy efficiency twice that of internal combustion engines. However before fuel cells can realize their potential, significant challenges remain. The two most important are cost and durability for both automotive and stationary applications. Recent electrocatalyst developments have shown that Pt alloy catalysts have increased activity and greater durability than Pt catalysts. The durability of conventional fluorocarbon membranes is improving, and hydrocarbon-based membranes have also shown promise of equaling the performance of fluorocarbon membranes at lower cost. Recent announcements have also provided indications that fuel cells can start from freezing conditions without significant deterioration. Hydrogen storage systems for vehicles are inadequate to meet customer driving range expectations (>300 miles or 500 km) without intrusion into vehicle cargo or passenger space. The United States Department of Energy has established three centers of Excellence for hydrogen storage materials development. The centers are focused on complex metal hydrides that can be regenerated onboard a vehicle, chemical hydrides that require off-board reprocessing, and carbon-based storage materials. Recent developments have shown progress toward the 2010 DOE targets. In addition DOE has established an independent storage material testing center

  14. Nuclear Fuel Design Technology Development for the Future Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Yang Hyun; Lee, Byung Ho; Cheon, Jin Sik; Oh, Je Yong; Yim, Jeong Sik; Sohn, Dong Seong; Lee, Byung Uk; Ko, Han Suk; So, Dong Sup; Koo, Dae Seo

    2006-04-15

    The test MOX fuels have been irradiated in the Halden reactor, and their burnup attained 40 GWd/t as of October 2005. The fuel temperature and internal pressure were measured by the sensors installed in the fuels and test rig. The COSMOS code, which was developed by KAERI, well predicted in-reactor behavior of MOX fuel. The COSMOS code was verified by OECD-NEA benchmarks, and the result confirmed the superiority of COSMOS code. MOX in-pile database (IFA-629.3, IFA-610.2 and 4) in Halden was also used for the verification of code. The COSMOS code was improved by introducing Graphic User Interface (GUI) and batch mode. The PCMI analysis module was developed and introduced by the new fission gas behavior model. The irradiation test performed under the arbitrary rod internal pressure could also be analyzed with the COSMOS code. Several presentations were made for the preparation to transfer MOX fuel performance analysis code to the industry, and the transfer of COSMOS code to the industry is being discussed. The user manual and COSMOS program (executive file) were provided for the industry to test the performance of COSMOS code. To envisage the direction of research, the MOX fuel research trend of foreign countries, specially focused on USA's GENP policy, was analyzed.

  15. Nuclear fuel fabrication - developing indigenous capability

    International Nuclear Information System (INIS)

    Gupta, U.C.; Jayaraj, R.N.; Meena, R.; Sastry, V.S.; Radhakrishna, C.; Rao, S.M.; Sinha, K.K.

    1997-01-01

    Nuclear Fuel Complex (NFC), established in early 70's for production of fuel for PHWRs and BWRs in India, has made several improvements in different areas of fuel manufacturing. Starting with wire-wrap type of fuel bundles, NFC had switched over to split spacer type fuel bundle production in mid 80's. On the upstream side slurry extraction was introduced to prepare the pure uranyl nitrate solution directly from the MDU cake. Applying a thin layer of graphite to the inside of the tube was another modification. The Complex has developed cost effective and innovative techniques for these processes, especially for resistance welding of appendages on the fuel elements which has been a unique feature of the Indian PHWR fuel assemblies. Initially, the fuel fabrication plants were set-up with imported process equipment for most of the pelletisation and assembly operations. Gradually with design and development of indigenous equipment both for production and quality control, NFC has demonstrated total self reliance in fuel production by getting these special purpose machines manufactured indigenously. With the expertise gained in different areas of process development and equipment manufacturing, today NFC is in a position to offer know-how and process equipment at very attractive prices. The paper discusses some of the new processes that are developed/introduced in this field and describes different features of a few PLC based automatic equipment developed. Salient features of innovative techniques being adopted in the area Of UO 2 powder production are also briefly indicated. (author)

  16. The development of lower enrichment fuels for Canadian research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Feraday, M A; Belanger, L; Grolway, C M [AECL, Atomic Energy of Canada Limited, Chalk River, ON (Canada); Foo, M T [CRNL, Combustion Engineering Superheater Ltd., Moncton, NB (Canada)

    1983-08-01

    As part of the world wide move to proliferation resistant fuels, new fuels which use reduced enrichment uranium are being developed for use in the NRX and NRU reactors. A fuel consisting of particles of a USiAl alloy dispersed in an Al matrix has been selected for development along with Al-37 wt% U alloy and Al-U{sub 3}O{sub 8} cermet as backup fuels. This report outlines the progress made in the development of the Al-USiAl and Al-37 wt% U. Results show that good quality extruded rods containing either fuel can be made with techniques similar to those used to fabricate the current NRX and NRU fuels. However, the new fuels will be more expensive to make. Although the oxidation behaviour of the Al-USiAl is not as good as that of the Al-U alloys, its corrosion behaviour in high temperature water does not seem much worse. The oxidation and aqueous corrosion of A-37 wt% U are not much different from those of the Al-U alloys currently used. (author)

  17. Romanian concern for advanced fuels development

    International Nuclear Information System (INIS)

    Ohai, Dumitru

    2001-01-01

    The Institute for Nuclear Research (ICN), a subsidiary of Romanian Authority for Nuclear Activities, at Pitesti - Romania, has developed a preliminary design of a fuel bundle with 43 elements named SEU 43 for high burnup in CANDU Reactor. A very high experience in nuclear fuels manufacturing and control has also been accumulated. Additionally, on the nuclear site Pitesti there is the Nuclear Fuel Plant (NFP) qualified to manufacturing CANDU 6 type fuel, the main fuel supplier for NPP Cernavoda. A very good collaboration of ICN with NFP can lead to a low cost upgrading the facilities which ensure at present the CANDU standard fuel fabrication to be able of manufacturing also SEU 43 fuel for extended burnup. The financial founds are allocated by Romanian Authority for Nuclear Activities of the Ministry of Industry and Resources to sustain the departmental R and D program 'Nuclear Fuel'. This Program has the main objective to establish a technology for manufacturing a new CANDU fuel type destined for extended burnup. It is studied the possibility to use the Recovered Uranium (RU) resulted from LWR spent fuel reprocessing facility existing in stockpiles. The International Agency for Atomic Energy (IAEA) sustains also this program. By ROM/4/025/ Model Project, IAEA helps ICN to solve the problems regarding materials (RU, Zircaloy 4 tubes) purchasing, devices' upgrading and personnel training. The paper presents the main actions needing to be create the technical base for SEU 43 fuel bundle manufacturing. First step, the technological experiments and experimental fuel element manufacturing, will be accomplished in ICN installations. Second step, the industrial scale, need thorough studies for each installation from NFP to determine tools and technology modification imposed by the new CANDU fuel bundle manufacturing. All modifications must be done such as to the NFP, standard CANDU and SEU fuel bundles to be manufactured alternatively. (author)

  18. Progress in MELCOR development and assessment

    International Nuclear Information System (INIS)

    Summers, R.M.; Kmetyk, L.N.; Cole, R.K. Jr.; Smith, R.C.; Elsbernd, A.E.; Stuart, D.S.; Thompson, S.L.

    1995-01-01

    MELCOR models the progression of severe accidents in light water reactor nuclear power plants. Recent efforts in MELCOR development to incorporate CORCON-Mod3 models for core-concrete interactions, new models for advanced reactors, and improvements to several other existing models have resulted in release of MELCOR 1.8.3. In addition, continuing efforts to expand the code assessment database have filled in many of the gaps in phenomenological coverage. Efforts are now under way to develop models for chemical interactions of fission products with structural surfaces and for reactions of iodine in the presence of water, and work is also in progress to improve models for the scrubbing of fission products by water pools, the chemical reactions of boron carbide with steam, and the coupling of flow blockages with the hydrodynamics. Several code assessment analyses are in progress, and more are planned

  19. The developments of international hydrogen and fuel cell technology standards and the response strategies in Taiwan

    International Nuclear Information System (INIS)

    Tso, C.

    2009-01-01

    The application of hydrogen and fuel cells has expanded as the technology in international markets has improved. Leading countries have focused on establishing hydrogen and fuel cell technology standards. Both the International Organization for Standardization (ISO) and the International Electrotechnical Commission (IEC) continuously release new hydrogen and fuel cell related standards. Although the government of Taiwan is promoting the development of a hydrogen and fuel cell industry, it may delay the commercialized schedule if there are no hydrogen and fuel cell related standards and regulations in place. Standards and regulations must be established as quickly as possible in order to accelerate the progress of the hydrogen and fuel cell industry. This presentation reviewed the international progress in hydrogen and fuel cell development and explained Taiwan's response strategies regarding the adoption of hydrogen and fuel cell products in niche Taiwanese markets

  20. Development of TVSA VVER-1000 fuel

    International Nuclear Information System (INIS)

    Samoilov, O.; Kaydalov, V.; Romanov, A.; Falkov, A.; Morozkin, O.; Sholin, E.

    2013-01-01

    The TVSA fuel assemblies with a rigid angle-piece skeleton operate at 21 VVER-1000 units of Kalinin NPP, and Ukrainian, and Czech and Bulgarian NPPs. The total of more than 6,000 TVSA fuel assemblies have been fabricated. High lifetime performance has been achieved, namely, the maximum FA burnup is 65 MW∙day/kgU; maximum fuel rod burnup is 72 MW∙day/kgU; the lifetime is 50,000 EFPH. The TVSA fuel assembly is being improved to enhance its technical and economic performance and competitiveness of the Russian fuel for the VVER-1000 reactor: 1) Reliability and safety are being enhanced; repairability is being ensured. 2) High burnup levels in fuel are being ensured. 3) The uranium content in FAs is being increased. 4) The operational life is being extended. 5) Thermal-technical characteristics of FAs are being improved. The basic TVSA fuel assembly design evolved into the TVSA-PLUS with the fuel column elongated by 150 mm. The TVSA-PLUS fuel assembly has been in operation since 2010 at Kalinin NPP power units; an eighteen-month cycle is implemented at the uprated power of 104%. The TVSA-12PLUS fuel assembly has been developed with an elongated fuel column, optimized spacer grid positions (the spacer grid pitch is 340 mm) and with ensuring higher rigidity for the skeleton. It is provided for that fuel rods with the elevated uranium content and mixing intensifier grids will be used. The TVSA-T is developed for VVER-1000 reactor cores at the Temelin NPP. The TVSA-T is characterized by a load-carrying skeleton formed with angle-pieces and combined spacer grids that incorporate mixer grids. The TVSA-T design won the international tender to supply fuel to the Temelin NPP in the Czech Republic, and currently Temelin NPP Unit 1 and 2 are operating with the cores fully loaded with TVSA-Ts

  1. Molybdenum-base cermet fuel development

    International Nuclear Information System (INIS)

    Gurwell, W.E.; Moss, R.W.; Pilger, J.P.; White, G.D.

    1987-07-01

    Development of a multimegawatt (MMW) space nuclear power system requires identification and resolution of several technical feasibility issues before selecting one or more promising system concepts. Demonstration of reactor fuel fabrication technology is required for cermet-fueled reactor concepts. MMW reactor fuel development activity at Pacific Northwest Laboratory (PNL) is focused on producing a molybdenum-matrix uranium-nitride (UN) fueled cermet. This cermet is to have a high matrix density (≥95%) for high strength and high thermal conductance coupled with a high particle (UN) porosity (∼25%) for retention of released fission gas at high burnup. Fabrication process development involves the use of porous TiN microspheres as surrogate fuel material until porous UN microspheres become available. Process development has been conducted in the areas of microsphere synthesis, particle sealing/coating, and high-energy-rate forming (HERF) and vacuum hot press consolidation techniques. This paper summarizes the status of these activities

  2. French development program on fuel cycle

    International Nuclear Information System (INIS)

    Viala, M.; Bourgeois, M.

    1991-01-01

    The need to close the fuel cycle of fast reactors makes the development of the cycle installations (fuel fabrication, irradiated assembly conditioning before reprocessing, reprocessing and waste management) especially independent with the development of the reactor. French experience with the integrated cycle over a period of about 25 years, the tonnage of fuels fabricated (more than 100 t of mixed oxides) for the Rapsodie, Phoenix and SuperPhoenix reactors, and the tonnage of reprocessed fuel (nearly 30 t of plutonium fuel) demonstrate the control of the cycle operations. The capacities of the cycle installations in existence and under construction are largely adequate for presents needs, even including a new European EFR reactor. They include the Cadarache fuel fabrication complex, the La Hague UP2-800 reprocessing plant, and the Marcoule pilot facility. Short- and medium-term R and D programs are connected with fuel developments, with the primary objective of very high burnups. For the longer term and for a specific plant to reprocess fast reactor fuels, the programs could concern new fabrication and reprocessing systems and the study of the consequences of the reduction in fuel out-of-core time

  3. The progressive achievement of a closed fuel cycle in France; La mise en oeuvre progressive d'un cycle ferme en France

    Energy Technology Data Exchange (ETDEWEB)

    Hugelmann, D.; Devezeaux de Lavergne, J.G. [AREVA NC, 78 - Velizy Villacoublay (France)

    2008-03-15

    The author reviews the progressive building of a strong nuclear fuel cycle industry in France. The first major step was the abandon of the graphite-gas reactor system to the PWR system. The government's decision to opt for reactors operating with enriched uranium opened the way to the application at an industrial scale of uranium enrichment technology that was only confined to military purposes. The legal entity 'EURODIF S A' was founded at that time and the different production units of the George-Besse-1 enrichment plant entered into service progressively from 1978 to 1982. The Comurhex company was created in 1969, and was in charge of producing the uranium hexafluoride necessary to the fabrication of nuclear fuels. La-Hague plant entered into service in 1966, its aim was to process spent fuels from graphite-gas reactors. Inside this plant the HAO (High Activity Oxide) dedicated to PWR spent fuels was operating in 1974. The MELOX plant dedicated to the fabrication of mixed oxides fuels (Mox) entered into operation in 1995 (till now more than 5000 Mox assemblies have been fabricated. Another important step was the processing of Mox fuels. During these 30 years, the nuclear industry has made impressing progress concerning: the increase of burn-up rates, the performance of fuels, the increase in the volume being processed, the packaging of radioactive wastes, the development of nuclear transport, and a reduction of the impact on the environment. In order to maintain its level of performance the nuclear industry has made important investments concerning: mining (a global investment of 2.3*10{sup 9} euros), Comurhex-2 (a 610*10{sup 6} euros investment) and Georges-Besse-2 plant (a 3*10{sup 9} euros investment for the enrichment of uranium through centrifugation). (A.C.)

  4. Advanced LWR Nuclear Fuel Cladding Development

    International Nuclear Information System (INIS)

    Bragg-Sitton, S.; Griffith, G.

    2012-01-01

    The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Research and Development (R and D) Pathway encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. To achieve significant operating improvements while remaining within safety boundaries, significant steps beyond incremental improvements in the current generation of nuclear fuel are required. Fundamental enhancements are required in the areas of nuclear fuel composition, cladding integrity, and fuel/cladding interaction to allow improved fuel economy via power uprates and increased fuel burn-up allowance while potentially improving safety margin through the adoption of an 'accident tolerant' fuel system that would offer improved coping time under accident scenarios. In a staged development approach, the LWRS program will engage stakeholders throughout the development process to ensure commercial viability of the investigated technologies. Applying minimum performance criteria, several of the top-ranked materials and fabrication concepts will undergo a rigorous series of mechanical, thermal and chemical characterization tests to better define their properties and operating potential in a relatively low-cost, nonnuclear test series. A reduced number of options will be recommended for test rodlet fabrication and in-pile nuclear testing under steady-state, transient and accident conditions. (author)

  5. Space reactor fuels performance and development issues

    International Nuclear Information System (INIS)

    Wewerka, E.M.

    1984-01-01

    Three compact reactor concepts are now under consideration by the US Space Nuclear Power Program (the SP-100 Program) as candidates for the first 100-kWe-class space reactor. Each of these reactor designs puts unique constraints and requirements on the fuels system, and raises issues of fuel systems feasibility and performance. This paper presents a brief overview of the fuel requirements for the proposed space reactor designs, a delineation of the technical feasibility issues that each raises, and a description of the fuel systems development and testing program that has been established to address key technical issues

  6. Developments in fossil fuel electricity generation

    International Nuclear Information System (INIS)

    Williams, A.; Argiri, M.

    1993-01-01

    A major part of the world's electricity is generated by the combustion of fossil fuels, and there is a significant environmental impact due to the production of fossil fuels and their combustion. Coal is responsible for 63% of the electricity generated from fossil fuels; natural gas accounts for about 20% and fuel oils for 17%. Because of developments in supply and improvements in generating efficiencies there is apparently a considerable shift towards a greater use of natural gas, and by the year 2000 it could provide 25% of the world electricity output. At the same time the amount of fuel oil burned will have decreased. The means to minimize the environmental impact of the use of fossil fuels, particularly coal, in electricity production are considered, together with the methods of emission control. Cleaner coal technologies, which include fluidized bed combustion and an integrated gasification combined cycle (IGCC), can reduce the emissions of NO x , SO 2 and CO 2 . (author)

  7. The conditions of gaseous fuels development

    International Nuclear Information System (INIS)

    Anon.

    1996-01-01

    Face to the actual situation of petrol and gas oil in France, the situation of gaseous fuels appears to be rather modest. However, the aim of gaseous fuels is not to totally supersede the liquid fuels. Such a situation would imply a complete overturn which has not been seriously considered yet. This short paper describes the essential conditions to promote the wider use of gaseous fuels: the intervention of public authorities to adopt a more advantageous tax policy in agreement with the ''Clean Air''law project, a suitable distribution network for gaseous fuels, a choice of vehicles consistent with the urban demand, the development of transformation kits of quality and of dual-fuel vehicles by the car manufacturers. (J.S.)

  8. Strategic Partnerships in Fuel Cell Development

    Science.gov (United States)

    Diab, Dorey

    2006-01-01

    This article describes how forming strategic alliances with universities, emerging technology companies, the state of Ohio, the federal government, and the National Science Foundation, has enabled Stark State College to develop a $5.5 million Fuel Cell Prototyping Center and establish a Fuel Cell Technology program to promote economic development…

  9. Fuel Fabrication Capability Research and Development Plan

    Energy Technology Data Exchange (ETDEWEB)

    Senor, David J.; Burkes, Douglas

    2013-06-28

    The purpose of this document is to provide a comprehensive review of the mission of the Fuel Fabrication Capability (FFC) within the Global Threat Reduction Initiative (GTRI) Convert Program, along with research and development (R&D) needs that have been identified as necessary to ensuring mission success. The design and fabrication of successful nuclear fuels must be closely linked endeavors.

  10. The fuel cell; development and possibilities

    Energy Technology Data Exchange (ETDEWEB)

    Van Rijnsoever, J.W.M.

    Activities on fuel cells and fuel cell development in the USA and Japan are surveyed. Possibilities for large scale application are mentioned. Attention is given to efficiency and environmental aspects. There are no problems about hazardous emissions. Besides electric power some heat is generated, which is not always a disadvantage. In many cases both are useful products. (A.V.)

  11. CANDU fuel performance and development

    International Nuclear Information System (INIS)

    Hardy, D.G.; Wood, J.C.; Bain, A.S.

    1978-12-01

    The fuel defect rate in CANDU (Canada Deuterium Uranium) reactors continues to be very low, 0.06% since 1972. The power ramp defects, which constituted the majority of the early defects, have been virtually eliminated by changed fuelling schemes and through the introduction of graphite CANLUB coatings on the inside of the sheath. Laboratory and loop irradiations have demonstrated that the graphite CANLUB layers increase the tolerance to power ramps, but to obtain the maximum benefit, coating parameters such as thickness, adhesion and wear resistance must be optimized. Siloxane CANLUB coated fuel offers greater tolerance to power ramps than most graphite coatings; quality control appears simpler and no instance of localized sheath hydriding has been seen with cured and irradiated coatings. Limited testing has shown that fuel with graphite discs between fuel pellets also has high tolerance to power ramps, but it is more costly and has lower burnup. The number of defects due to faulty components has been extremely small (0.00014%), but improved quality control and welding procedures can lower this number even further. Defects from causes external to the bundle have also been very few. (author)

  12. Development of nuclear fuel cycle technology

    International Nuclear Information System (INIS)

    Kawahara, Akira; Sugimoto, Yoshikazu; Shibata, Satoshi; Ikeda, Takashi; Suzuki, Kazumichi; Miki, Atsushi.

    1990-01-01

    In order to establish the stable supply of nuclear fuel as an important energy source, Hitachi ltd. has advanced the technical development aiming at the heightening of reliability, the increase of capacity, upgrading and the heightening of performance of the facilities related to nuclear fuel cycle. As for fuel reprocessing, Japan Nuclear Fuel Service Ltd. is promoting the construction of a commercial fuel reprocessing plant which is the first in Japan. The verification of the process performance, the ensuring of high reliability accompanying large capacity and the technical development for recovering effective resources from spent fuel are advanced. Moreover, as for uranium enrichment, Laser Enrichment Technology Research Association was founded mainly by electric power companies, and the development of the next generation enrichment technology using laser is promoted. The development of spent fuel reprocessing technology, the development of the basic technology of atomic process laser enrichment and so on are reported. In addition to the above technologies recently developed by Hitachi Ltd., the technology of reducing harm and solidification of radioactive wastes, the molecular process laser enrichment and others are developed. (K.I.)

  13. Research Establishment progress report 1978 - uranium fuel cycle

    International Nuclear Information System (INIS)

    1978-12-01

    A report of research programs continuing in the following areas is presented: mining and treatment of uranium ores, uranium enrichment, waste treatment, reprocessing and the uranium fuel cycle. Staff responsible for each project are indicated

  14. Hydrogen Fuel Cell: Research Progress and Near-Term Opportunities

    Science.gov (United States)

    2009-04-27

    effort brings together automobile and ener- gy companies , as well as their suppliers and other stakeholders, to evaluate light-duty fuel cell vehicles...emissions compared to conventional power technologies. Grocers, banks, tire and hardware companies , logistics providers, and others in the private sector...Term Direct Hydrogen Proton Exchange Membrane (PEM) Fuel Cell Markets, April 2007. 2. Assumptions: Operate 7 hours/shift, 3 shifts/day, 7 days/week

  15. Development of fuel and energy storage technologies

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    Development of fuel cell power plants is intended of high-efficiency power generation using such fuels with less air pollution as natural gas, methanol and coal gas. The closest to commercialization is phosphoric acid fuel cells, and the high in efficiency and rich in fuel diversity is molten carbonate fuel cells. The development is intended to cover a wide scope from solid electrolyte fuel cells to solid polymer electrolyte fuel cells. For new battery power storage systems, development is focused on discrete battery energy storage technologies of fixed type and mobile type (such as electric vehicles). The ceramic gas turbine technology development is purposed for improving thermal efficiency and reducing pollutants. Small-scale gas turbines for cogeneration will also be developed. Development of superconduction power application technologies is intended to serve for efficient and stable power supply by dealing with capacity increase and increase in power distribution distance due to increase in power demand. In the operations to improve the spread and general promotion systems for electric vehicles, load leveling is expected by utilizing and storing nighttime electric power. Descriptions are given also on economical city systems which utilize wide-area energy. 30 figs., 7 tabs.

  16. Progress report on fast breeder reactor development in Japan

    International Nuclear Information System (INIS)

    1976-05-01

    Following the completion of the in-argon, high temperature test, the in-sodium functional test of Joyo has set in. The fabrication of the equipments for monitoring the flow rate and temperature in the center channel and the power distribution was finished. The modification of design of the prototype fast breeder reactor Monju came into its phase 3. The interim report on the check-up and review of the Monju project by the Government is now ready. Various calculation codes were developed or are in development stage. The mock-up assembly FCA 7-1 has been built, which consists of the Pu-fueled sector region simulating the Monju core and the U-235-fueled driver region. Various reactor physics experiments have been carried out in this assembly. Also, the calculation methods for reactor physics parameters have been developed, and the detailed calculation on the main shield of Joyo was performed. The situation of the developments of the components for Joyo and Monju and the measuring and control systems is shown. Almost all the existing sodium test facilities in Oarai Engineering Center were in service without any trouble, and the new test facility named ''Carbon transfer test loop'' was commissioned. The progress in the fields of sodium technology, fuel and material, safety and steam generators is reported. (Kako, I.)

  17. Safety analysis of DUPIC fuel development facility

    International Nuclear Information System (INIS)

    Lee, H. H.; Park, J. J.; Shin, J. M.; Yang, M. S.; Baek, S. Y.; Ahn, J. Y.

    2001-01-01

    Various experimental facilities are necessary in order to perform experimental verification for development of DUPIC fuel fabrication technology. In special, since highly radioactive material such as spent PWR fuel is used for this experiment, DUPIC fuel fabrication has to be performed in hot cell by remote handling. Therefore, it should be provided with proper engineering requirement and safety. M6 hot cell of IMEF which is to used for DUPIC fuel fabrication experiment was constructed as an α-γ hot cell for material examination of small amount of high-burnup fuel. The characteristics and amount of spent fuel for DUPIC fuel fabrication experiment will be different from the original design criteria. Therefore, the increased amount of spent fuel and different characteristics of experiment result in not only change of shielding and enviornmental evaluation results but new requirement of nuclear criticality evaluation. Therefore, this study includes evaluation of shielding, environmental effect and nuclear criticality in case that IMEF M6 hot cell is used for DUPIC fuel fabrication

  18. Development of PHWR fuel fabrication in Korea

    International Nuclear Information System (INIS)

    Suh, K.S.; Yang, M.S.; Kim, D.H.; Rim, C.S.

    1988-01-01

    Korea Advanced Energy Research Institute (KAERI) started a research project to develop the PHWR (CANDU) nuclear fuel fabrication technology in 1981. Based on the results of the intensive developmental work, several prototype fuel bundles were fabricated and tested in the Hot Test Loop at KAERI continuously in 1983 and 1984. After that, irradiation test and post-irradiation examination were carried out for two KAERI-made fuel bundles at Chalk River Nuclear Laboratories in Canada in 1984. Since the results of in-pile and out-of-pile tests with prototype fuel bundles proved to be satisfactory, 48 additional fuel bundles were loaded in Wolsung reactor (CANDU) in 1984 and 1985, and all of them were discharged without a defect after excellent performance in the power reactor. In 1985, the Korean government decided that KAERI supplies all the fuel necessary for the Wolsung reactor. For the mass production of nuclear fuel bundle, several process equipment, facilities and automation methods have been improved making use of experience accumulated during research. A quality assurance program was also established, and quality inspection technology was reviewed and improved to fit the mass production. This paper deals with the development experience so far obtained with the design and fabrication of the Korean PHWR fuel

  19. Microbial biocatalyst developments to upgrade fossil fuels.

    Science.gov (United States)

    Kilbane, John J

    2006-06-01

    Steady increases in the average sulfur content of petroleum and stricter environmental regulations concerning the sulfur content have promoted studies of bioprocessing to upgrade fossil fuels. Bioprocesses can potentially provide a solution to the need for improved and expanded fuel upgrading worldwide, because bioprocesses for fuel upgrading do not require hydrogen and produce far less carbon dioxide than thermochemical processes. Recent advances have demonstrated that biodesulfurization is capable of removing sulfur from hydrotreated diesel to yield a product with an ultra-low sulfur concentration that meets current environmental regulations. However, the technology has not yet progressed beyond laboratory-scale testing, as more efficient biocatalysts are needed. Genetic studies to obtain improved biocatalysts for the selective removal of sulfur and nitrogen from petroleum provide the focus of current research efforts.

  20. Fuel Cell Development and Test Laboratory | Energy Systems Integration

    Science.gov (United States)

    Facility | NREL Fuel Cell Development and Test Laboratory Fuel Cell Development and Test Laboratory The Energy System Integration Facility's Fuel Cell Development and Test Laboratory supports fuel cell research and development projects through in-situ fuel cell testing. Photo of a researcher running

  1. LOFT advanced fuel rod instrumentation development

    International Nuclear Information System (INIS)

    Billeter, T.R.; Brown, R.L.; Chan, A.I.Y.; Day, C.K.; Meyers, S.C.; Sheen, E.M.; Stringer, J.L.

    1978-01-01

    Advanced fuel rod instrumentation for the Loss of Fluid Test (LOFT) reactor is being developed by the Hanford Engineering Development Laboratory for the Nuclear Regulatory Commission. This effort calls for development of sensors to measure fuel rod axial motion, fuel centerline temperature (to 2200 0 C), fuel rod plenum gas pressure (to 2500 psig), and plenum gas temperature (to 1500 0 F). A parallel test and evaluation of several modified commercial sensors was undertaken and will result in commercial availability of the final qualified sensors. Necessary test facilities were prepared for the development and evaluation effort. Tests to date indicate a three coil Linear Variable Differential Transformer (LVDT), operated from temperature compensating signal source and processing electronics, will meet the desired requirements

  2. VVER fuel cycle development at Slovakia

    International Nuclear Information System (INIS)

    Darilek, P.; Chrapiak, V.; Majerik, J.

    1995-01-01

    Four VVER-440 units are now under exploitation at Bohunice-site in Slovakia. Fuel cycle development of Unit No.3 and No.4 (type 213) is discussed and compared with equilibrium cycles in this paper. (author)

  3. Recent progress in fusion gyrotron development

    International Nuclear Information System (INIS)

    Shively, J.F.; Stone, D.S.

    1981-01-01

    The gyrotron, a microwave tube capable of producing high power output at millimeter wavelengths, has recently found applications for electron cyclotron resonance heating of plasmas in controlled thermonuclear fusion reactor experiments. This paper describes work in progress to develop a gyrotron oscillator to deliver 200 kW CW at 60 GHz (/lambda/sub //. 5 mm). A pulsed oscillator is described which produced over 200 kw peak power. A CW oscillator is under construction. The latest experimental results are presented

  4. Development of a diesel substitute fuel

    Energy Technology Data Exchange (ETDEWEB)

    Reiter, Anton; Mair-Zelenka, Philipp [Graz Univ. of Technology (Austria). Inst. of Chemical Engineering and Environmental Technology; Zeymer, Marc [OMV Refining and Marketing GmbH, Vienna (Austria). MRDI-D Product Development and Innovation

    2013-06-01

    Substitute fuels composed of few real chemical compounds are an alternative characterisation approach for conventional fuels as opposed to the traditional pseudo-component method. With the algorithm proposed in this paper the generation of such substitutes will be facilitated and well-established thermodynamic methods can be applied for physical property-data prediction. Based on some quality criteria like true boiling-point curve, liquid density, C/H ratio, or cloud point of a target fuel a surrogate which meets these properties is determined by fitting its composition. The application and capabilities of the algorithm developed are demonstrated by means of an exemplary diesel substitute fuel. The substitute mixture obtained can be generated and used for evaluation of property-prediction methods. Furthermore this approach can help to understand the effects of mixing fossil fuels with biogenic compounds. (orig.)

  5. Status of progress in IFR development

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1994-01-01

    The author touches on a number of the issues being addressed in work on the IFR reactor concept development at this time. This technology offers the opportunity to achieve a much more efficient utilization of uranium resources in the process of energy generation. Liquid metal coolant is likely to be a design feature of this type reactor, in part due to the passive safety of this design. Fuel reprocessing, and waste management issues are discussed. This reactor design has the advantage of being much better positioned to minimize concerns about nuclear proliferation issues

  6. The low-enrichment fuel development program

    International Nuclear Information System (INIS)

    Stahl, D.

    1993-01-01

    In the 1950s and 1960s, low-power research reactors were built around the world utilized MTR-type fuel elements containing 20% enriched uranium. However, the demand for higher specific power created a need for greater uranium-235 concentrations. Early difficulties in increasing uranium content led to the substitution of highly enriched uranium in place of the 20% enriched fuel previously utilized. The highly enriched material also yielded other benefits including longer core residence time, higher specific reactivity, and somewhat lower cost. Highly enriched material then became readily available and was used for high-power reactors as well as in low-power reactors where 20% enriched material would have sufficed. The trend toward higher and higher specific power also led to the development of the dispersion-type fuels which utilized highly enriched uranium at a concentration of about 40 wt%. In the 1970's, however, concerns were raised about the proliferation resistance of fuels and fuel cycles. As a consequence, the U.S. Department of State has recently prohibited the foreign shipment of highly enriched material, except where prior contractual obligation or special merit exists. This will impact on the availability and utilization of highly enriched uranium for research and test reactor fuel. It has also stimulated development programs on fuels with higher uranium content which would allow the use of uranium of lower enrichment. The purpose of this report is to briefly describe the overall fuel-development program which is coordinated by Argonne National Laboratory for the Department of Energy, and to indicate the current and potential uranium loadings. Other reports will address the individual fuel-development activities in greater detail

  7. Development of Metallic Fuels for Actinide Transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Hayes, Steven Lowe [Idaho National Laboratory; Fielding, Randall Sidney [Idaho National Laboratory; Benson, Michael Timothy [Idaho National Laboratory; Chichester, Heather Jean MacLean [Idaho National Laboratory; Carmack, William Jonathan [Idaho National Laboratory

    2015-09-01

    Research and development activities on metallic fuels are focused on their potential use for actinide transmutation in future sodium fast reactors. As part of this application, there is also a need for a near zero-loss fabrication process and a desire to demonstrate a multifold increase in burnup potential. The incorporation of Am and Np into the traditional U-20Pu-10Zr metallic fuel alloy was demonstrated in the US during the Integral Fast Reactor Program of the 1980’s and early 1990’s. However, the conventional counter gravity injection casting method performed under vacuum, previously used to fabricate these metallic fuel alloys, was not optimized for mitigating loss of the volatile Am constituent in the casting charge; as a result, approximately 40% of the Am casting charge failed to be incorporated into the as-cast fuel alloys. Fabrication development efforts of the past few years have pursued an optimized bottom-pour casting method to increase utilization of the melted charge to near 100%, and a differential pressure casting approach, performed under an argon overpressure, has been demonstrated to result in essentially no loss of Am due to volatilization during fabrication. In short, a path toward zero-loss fabrication of metallic fuels including minor actinides has been shown to be feasible. Irradiation testing of advanced metallic fuel alloys in the Advanced Test Reactor (ATR) has been underway since 2003. Testing in the ATR is performed inside of cadmium-shrouded positions to remove >99% of the thermal flux incident on the test fuels, resulting in an epi-thermal driven fuel test that is free from gross flux depression and producing an essentially prototypic radial temperature profile inside the fuel rodlets. To date, three irradiation test series (AFC-1,2,3) have been completed. Over 20 different metallic fuel alloys have been tested to burnups as high as 30% with constituent compositions of Pu up to 30%, Am up to 12%, Np up to 10%, and Zr between 10

  8. [Progress and prospect of bio-jet fuels industry in domestic and overseas].

    Science.gov (United States)

    Qiao, Kai; Fu, Jie; Zhou, Feng; Ma, Huixia

    2016-10-25

    We reviewed the progress of the bio-jet fuels industry in recent years and systematically analyzed the technical routes that have been approved or in the pipeline for approval by ASTM D7566. In addition, we highlighted a novel pathway to produce drop-in fuel by near-critical hydrolysis of waste cooking oils or algal oils followed by catalytic decarboxylation. Also, we introduced the source of oils and fats feedstock and the domestic bio-jet fuel industry status during the 12th Five-Year-Plan period. Based on our own research, we discussed the prospect of the bio-jet fuel industry and future research needs.

  9. Fuel performance improvement program. Quarterly/annual progress report, October 1977--September 1978

    International Nuclear Information System (INIS)

    Crouthamel, C.E.

    1978-10-01

    This quarterly/annual report reviews and summarizes the activities performed in support of the Fuel Performance Improvement Program (FPIP) during Fiscal Year 1978 with emphasis on those activities that transpired during the quarter ending September 30, 1978. Significant progress has been made in achieving the primary objectives of the program, i.e., to demonstrate commercially viable fuel concepts with improved fuel - cladding interaction (FCI) behavior. This includes out-of-reactor experiments to support the fuel concepts being evaluated, initiation of instrumented test rod experiments in the Halden Boiling Water Reactor (HBWR), and fabrication of the first series of demonstration rods for irradiation in the Big Rock Point Reactor

  10. Development of spent fuel remote handling technology

    International Nuclear Information System (INIS)

    Yoon, J. S.; Hong, H. D.; Kim, Y. H.

    2001-03-01

    Since the amount of the spent fuel rapidly increases, the current R and D activities are focused on the technology development related with the storage and utilization of the spent fuel. In this research, to provide such a technology, the mechanical head-end process has been developed. In detail, the swing and shock-free crane and the RCGLUD(Remote Cask Grappling and Lid Unbolting Device) were developed for the safe transportation of the spent fuel assembly, the LLW drum and the transportation cask. Also, the disassembly devices required for the head-end process were developed. This process consists of an assembly downender, a rod extractor, a rod cutter, a fuel decladding device, a skeleton compactor, a force-rectifiable manipulator for the abnormal spent fuel disassembly, and the gantry type telescopic transporter, etc. To provide reliability and safety of these devices, the 3 dimensional graphic design system is developed. In this system, the mechanical devices are modelled and their operation is simulated in the virtual environment using the graphic simulation tools. So that the performance and the operational mal-function can be investigated prior to the fabrication of the devices. All the devices are tested and verified by using the fuel prototype at the mockup facility

  11. Year One Summary of X-energy Pebble Fuel Development at ORNL

    Energy Technology Data Exchange (ETDEWEB)

    Helmreich, Grant W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hunn, John D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); McMurray, Jake W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hunt, Rodney D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jolly, Brian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Trammell, Michael P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Daniel R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Blamer, Brandon J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Reif, Tyler J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Kim, Howard T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-06-01

    The Advanced Reactor Concepts X-energy (ARC-Xe) Pebble Fuel Development project at Oak Ridge National Laboratory (ORNL) has successfully completed its first year, having made excellent progress in accomplishing programmatic objectives. The primary focus of research at ORNL in support of X-energy has been the training of X-energy fuel fabrication engineers and the establishment of US pebble fuel production capabilities able to supply the Xe-100 pebble-bed reactor. These efforts have been strongly supported by particle fuel fabrication and characterization expertise present at ORNL from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program.

  12. Equipment system for advanced nuclear fuel development

    International Nuclear Information System (INIS)

    Kwon, Hyuk Il; Ji, C. G.; Bae, S. O.

    2002-11-01

    The purpose of the settlement of equipment system for nuclear Fuel Technology Development Facility(FTDF) is to build a seismic designed facility that can accommodate handling of nuclear materials including <20% enriched Uranium and produce HANARO fuel commercially, and also to establish the advanced common research equipment essential for the research on advanced fuel development. For this purpose, this research works were performed for the settlement of radiation protection system and facility special equipment for the FTDF, and the advanced common research equipment for the fuel fabrication and research. As a result, 11 kinds of radiation protection systems such as criticality detection and alarm system, 5 kinds of facility special equipment such as environmental pollution protection system and 5 kinds of common research equipment such as electron-beam welding machine were established. By the settlement of exclusive domestic facility for the research of advanced fuel, the fabrication and supply of HANARO fuel is possible and also can export KAERI-invented centrifugal dispersion fuel materials and its technology to the nations having research reactors in operation. For the future, the utilization of the facility will be expanded to universities, industries and other research institutes

  13. Development of nuclear fuel for integrated reactor

    Energy Technology Data Exchange (ETDEWEB)

    Song, Kee Nam; Kim, H. K.; Kang, H. S.; Yoon, K. H.; Chun, T. H.; In, W. K.; Oh, D. S.; Kim, D. W.; Woo, Y. M

    1999-04-01

    The spacer grid assembly which provides both lateral and vertical support for the fuel rods and also provides a flow channel between the fuel rods to afford the heat transfer from the fuel pellet into the coolant in a reactor, is one of the major structural components of nuclear fuel for LWR. Therefore, the spacer grid assembly is a highly ranked component when the improvement of hardware is pursued for promoting fuel performance. Main objective of this project is to develop the inherent spacer grid assembly and to research relevant technologies on the spacer grid assembly. And, the UO{sub 2}-based SMART fuel is preliminarily designed for the 330MWt class SMART, which is planned to produce heat as well as electricity. Results from this project are listed as follows. 1. Three kinds of spacer grid candidates have been invented and applied for domestic and US patents. In addition, the demo SG(3x3 array) were fabricated, which the mechanical/structural test was carried out with. 2. The mechanical/structural technologies related to the spacer grid development are studied and relevant test requirements were established. 3. Preliminary design data of the UO{sub 2}-based SMART fuel have been produced. The structural characteristics of several components such as the top/bottom end piece and the holddown spring assembly were analysed by consulting the numerical method.

  14. Development of nuclear fuel for integrated reactor

    International Nuclear Information System (INIS)

    Song, Kee Nam; Kim, H. K.; Kang, H. S.; Yoon, K. H.; Chun, T. H.; In, W. K.; Oh, D. S.; Kim, D. W.; Woo, Y. M.

    1999-04-01

    The spacer grid assembly which provides both lateral and vertical support for the fuel rods and also provides a flow channel between the fuel rods to afford the heat transfer from the fuel pellet into the coolant in a reactor, is one of the major structural components of nuclear fuel for LWR. Therefore, the spacer grid assembly is a highly ranked component when the improvement of hardware is pursued for promoting fuel performance. Main objective of this project is to develop the inherent spacer grid assembly and to research relevant technologies on the spacer grid assembly. And, the UO 2 -based SMART fuel is preliminarily designed for the 330MWt class SMART, which is planned to produce heat as well as electricity. Results from this project are listed as follows. 1. Three kinds of spacer grid candidates have been invented and applied for domestic and US patents. In addition, the demo SG(3x3 array) were fabricated, which the mechanical/structural test was carried out with. 2. The mechanical/structural technologies related to the spacer grid development are studied and relevant test requirements were established. 3. Preliminary design data of the UO 2 -based SMART fuel have been produced. The structural characteristics of several components such as the top/bottom end piece and the holddown spring assembly were analysed by consulting the numerical method

  15. Fuel utilization experience in Bohunice NPP and regulatory requirements for implementation of progressive fuel management strategies

    Energy Technology Data Exchange (ETDEWEB)

    Patenyi, V [Nuclear Regulatory Authority, Bratislava (Slovakia); Darilek, P; Majercik, J [Vyskumny Ustav Jadrovych Elektrarni, Trnava (Slovakia)

    1994-12-31

    The experience gained in fuel utilization and the basic requirements for fuel licensing in the Slovak NPPs is described. The original project of WWER-440 reactors supposes 3-year fuel cycle with cycle length of about 320 full power days (FPD). Since 1984 it was reduced to 290 FPD. Based on the experience of other countries, a 4-year fuel cycle utilization started in 1987. It is illustrated with data from the Bohunice NPP units. Among 504 fuel assemblies left for the fourth burnup cycle no leakage was observed. The mean burnup achieved in the different units varied from 33.1 to 38.5 Mwd/kg U. The new fuel assemblies used are different from the recent ones in construction, thermohydraulics, water-uranium ratio, enrichment and material design. To meet the safety criteria, regulatory requirements for exploitation of new fuel in WWER-440 were formulated by the Nuclear Regulatory Authority of Slovak Republic. 1 tab., 5 refs.

  16. Development and progress of the South African uranium enrichment project

    International Nuclear Information System (INIS)

    Roux, A.J.A.; Grant, W.L.; Barbour, R.A.; Loubser, R.S.; Wannenburg, J.J.

    1977-01-01

    The earlier development of the project is briefly reviewed, and some of the salient features of the South African process are touched upon. Development of the separation element in the last 18 months is discussed, as well as further work on the helikon cascade process. A brief description of the helikon cascade operation is given by means of diagrams. Because of time limitations, the complete helikon theory is not presented, but only some examples shown. Experimental work done to verify the helikon concept, as well as theoretical treatment, is presented. A brief report of the progress made on the experimental module of 6 t/a separative work capacity is given. This module, known as Mini-Z, is well advanced and details of its features and construction are shown. A short discussion of progress on the full-scale prototype module, known as Proto-Z, is next presented. The flexibility of such a design to fit a wide range of cascade sizes is considered, as well as cost implications of various approaches to design. Apart from progress on the development of the commercial plant, a brief review is given of the present state of the pilot plant at Valindaba. Some of the information obtained is mentioned. In conclusion, some information is given in regard to further planning and other work on the commercial plant at present being undertaken. Projected operation of the plant and some nuclear fuel service aspects are touched on

  17. Progress in irradiation performance of experimental uranium - Molybdenum dispersion fuel

    International Nuclear Information System (INIS)

    Hofman, Gerard L.; Meyer, Mitchell K.

    2002-01-01

    High-density dispersion fuel experiment, RERTR-4, was removed from the Advanced Test Reactor (ATR) after reaching a peak U-235 burnup of ∼80% and is presently undergoing postirradiation examination at the ANL alpha-gamma hot cells. This test consists of 32 mini fuel plates of which 27 were fabricated with nominally 6 and 8 g cm -3 atomized and machined uranium alloy powders containing 7 wt% and 10 wt% molybdenum. In addition, two miniplates containing solid U-10 wt% Mo foils and three containing 6 g cm -3 U 3 Si 2 are part of the test. The results of the postirradiation examination and analysis of RERTR-4 in conjunction with data from previous tests performed to lower burnup will be presented. (author)

  18. Progress with Scottish Nuclear Limited's dry fuel store proposals

    International Nuclear Information System (INIS)

    Cathro, I.S.

    1994-01-01

    At present Scottish Nuclear plc's largest operating cost, associated with the Advanced Gas Cooled Reactors that it manages, is the reprocessing of spent fuels. Looking to reduce the costs, the company has considered alternative disposal options. Dry vault storage has emerged as a clear leader. An adaption of the United States Modular Vault Dry Storage design is being studied in order to examine the feasibility of a store of this type at each of its power stations. (UK)

  19. Progress in FMIT test assembly development

    International Nuclear Information System (INIS)

    Opperman, E.K.; Vogel, M.A.; Shen, E.J.; Trego, A.L.

    1983-08-01

    Research and development supporting the completed design of the Fusion Materials Irradiation Test (FMIT) Facility is continuing at the Hanford Engineering Development Laboratory (HEDL) in Richland, Washington. The FMIT, a deuteron accelerator based (d + Li) neutron source, will produce an intense flux of high energy neutrons for use in radiation damage studies of fusion reactor materials. The most intense flux magnitude of greater than 10 15 n/cm 2 -s is located close to the neutron producing lithium target and is distributed within a volume about the size of an American football. The conceptual design and development of FMIT experiments called Test Assemblies has progressed over the past five years in parallel with the design of the FMIT. The paper will describe the recent accomplishments made in developing test assemblies appropriate for use in the limited volume close to the FMIT target where high neutron flux and heating rates and the associated spacial gradients significantly impact design considerations

  20. Development of CANFLEX fuel fabrication technology

    Energy Technology Data Exchange (ETDEWEB)

    Kang, M. S.; Choi, C. B.; Park, C. H.; Kwon, W. J.; Kim, C. H.; Kim, B. J.; Koo, C. H.; Cho, D. S.; So, D. Y.; Suh, S. W.; Park, C. J.; Chang, D. H.; Yun, S. H. [KEPCO Nuclear Fuel Company, Taejeon (Korea)

    2000-04-01

    Wolsong Unit 1 as the first heavy water reactor in Korea has been in service for 17 years since 1983. It would be about the time to prepare a plan for the solution of problems due to aging of the reactor. The aging of CANDU reactor could lead especially to the steam generator cruding and pressure tube sagging and creep and then decreases the operation margin to make some problems on reactor operations and safety. The counterplan could be made in two ways. One is to repair or modify reactor itself. The other is to develop new advanced fuel to increase of CANDU operation margin effectively, so as to compensate the reduced operation margin. Therefore, the first objectives in the present R and D is to develop the CANFLEX-NU(CANDU Flexible fuelling-Natural Uranium) fuel as a CANDU advanced fuel. One of the improvements in CANDU fuel fabrication technology, and advanced method of Zr-Be brazing was developed. For the formation of Zr-Be alloy, preheating and main heating temperature in the furnace is 700 deg C, 1200 deg C respectively. In order to find an appropriate material for the brazing joints in the CANDU fuel, the composition of Zr based amorphous metals were designed. And, the effect of hydrogen on the mechanical properties of cladding sheath and feasibility of the eddy current test to evaluate quality of end cap weld were also studied for the fundamental research purpose. As a preliminary study to suggest optimal way for the mass production of CANFLEX-NU fuel at KNFC the existing CANDU fuel facilities and fabrication/inspection processes were reviewed. The best way is that the current CANDU facility shall be modified to produce small diametrial CANFLEX elements and a new facility shall be constructed to produce large diametrial CANFLEX fuel elements. 46 refs., 99 figs., 10 tabs. (Author)

  1. GCRA review and appraisal of fuel material development programs

    International Nuclear Information System (INIS)

    1980-09-01

    The Fuel material Development Program has as its principal objective and responsibility the development of a fuel that is both economical and licensable and that, at the same time, will fulfill the required performance criteria. To accomplish this, the program is broken down into the following major fuel development task areas: development of the experimental and analytical data base for selecting, qualifying, and verifying the reference fuel design; providing the data base and developing models for evaluating fuel performance under upset and accident conditions; and developing and justifying fuel fabrication specifications which are consistent with the overall fuel performance criteria and with the fuel fabrication process capabilities

  2. Progress and controversies in developing cancer vaccines

    Directory of Open Access Journals (Sweden)

    Speiser Daniel E

    2005-04-01

    Full Text Available Abstract Immunotherapy has become a standard approach for cancer management, through the use of cytokines (eg: interleukin-2 and monoclonal antibodies. Cancer vaccines hold promise as another form of immunotherapy, and there has been substantial progress in identifying shared antigens recognized by T cells, in developing vaccine approaches that induce antigen-specific T cell responses in cancer patients, and in developing new technology for monitoring immune responses in various human tissue compartments. Dramatic clinical regressions of human solid tumors have occurred with some cancer vaccines, but the rate of those responses remains low. This article is part of a 2-part point:counterpoint series on peptide vaccines and adoptive therapy approaches for cancer. The current status of cancer vaccination, and associated challenges, are discussed. Emphasis is placed on the need to increase our knowledge of cancer immunobiology, as well as to improve monitoring of cellular immune function after vaccination. Progress in both areas will facilitate development of effective cancer vaccines, as well as of adoptive therapy. Effective cancer vaccines promise to be useful for treatment and prevention of cancer at low cost and with low morbidity.

  3. Development of spent fuel remote handling technology

    International Nuclear Information System (INIS)

    Yoon, J. S.; Hong, H. D.; Kim, S. H.

    2004-02-01

    In this research, the remote handling technology is developed for the advanced spent fuel conditioning process which gives a possible solution to deal with the rapidly increasing spent fuels. In detail, a fuel rod slitting device is developed for the decladding of the spent fuel. A series of experiments has been performed to find out the optimal condition of the spent fuel voloxidation which converts the UO 2 pellet into U 3 O 8 powder. The design requirements of the ACP equipment for hot test is established by analysing the modular requirement, radiation hardening and thermal protection of the process equipment, etc. The prototype of the servo manipulator is developed. The manipulator has an excellent performance in terms of the payload to weight ratio that is 30 % higher than that of existing manipulators. To provide reliability and safety of the ACP, the 3 dimensional graphic simulator is developed. Using the simulator the remote handling operation is simulated and as a result, the optimal layout of ACP is obtained. The supervisory control system is designed to control and monitor the several different unit processes. Also the failure monitoring system is developed to detect the possible accidents of the reduction reactor

  4. Design and development of PWR fuel

    International Nuclear Information System (INIS)

    Dehon, C.; Leclercq, J.; Watteau, M.

    1982-06-01

    After a brief description of the FRAGEMA fuel assembly which equips at the present time the pressurized water reactors of EdF (Electricite de France), and a presentation of the experience obtained on this fuel, one reviews the main aims and trends of the research and development program carried out by FRAGEMA to improve the design of fuels and to propose to the national customer, but also on the foreign markets, new products adapted to the demands of operators. One insists more particularly on new products that are on one hand the AFA fuel and on the other hand the burnable poison UO 2 -Gd 2 O 3 ; their description is presented and their advantages are given. To conclude, one insists on the importance of the collaboration that have to be kept between the designer and the operator, the manufacturer, the R and D groups and the boiler specialist [fr

  5. BR-100 spent fuel shipping cask development

    International Nuclear Information System (INIS)

    McGuinn, E.J.; Childress, P.C.

    1990-01-01

    Continued public acceptance of commercial nuclear power is contingent to a large degree on the US Department of Energy (DOE) establishing an integrated waste management system for spent nuclear fuel. As part of the from-reactor transportation segment of this system, the B ampersand W Fuel Company (BWFC) is under contract to the DOE to develop a spent-fuel cask that is compatible with both rail and barge modes of transportation. Innovative design approaches were the keys to achieving a cask design that maximizes payload capacity and cask performance. The result is the BR-100, a 100-ton rail/barge cask with a capacity of 21 PWR or 52 BWR ten-year cooled, intact fuel assemblies. 3 figs

  6. Fuel Rod Melt Progression Simulation Using Low-Temperature Melting Metal Alloy

    International Nuclear Information System (INIS)

    Seung Dong Lee; Suh, Kune Y.; GoonCherl Park; Un Chul Lee

    2002-01-01

    The TMI-2 accident and various severe fuel damage experiments have shown that core damage is likely to proceed through various states before the core slumps into the lower head. Numerous experiments were conducted to address when and how the core can lose its original geometry, what geometries are formed, and in what processes the core materials are transported to the lower plenum of the reactor pressure vessel. Core degradation progresses along the line of clad ballooning, clad oxidation, material interaction, metallic blockage, molten pool formation, melt progression, and relocation to the lower head. Relocation into the lower plenum may occur from the lateral periphery or from the bottom of the core depending upon the thermal and physical states of the pool. Determining the quantities and rate of molten material transfer to the lower head is important since significant amounts of molten material relocated to the lower head can threaten the vessel integrity by steam explosion and thermal and mechanical attack of the melt. In this paper the focus is placed on the melt flow regime on a cylindrical fuel rod utilizing the LAMDA (Lumped Analysis of Melting in Degrading Assemblies) facility at the Seoul National University. The downward relocation of the molten material is a combination of the external film flow and the internal pipe flow. The heater rods are 0.8 m long and are coated by a low-temperature melting metal alloy. The electrical internal heating method is employed during the test. External heating is adopted to simulate the exothermic Zircaloy-steam reaction. Tests are conducted in several quasi-steady-state conditions. Given the variable boundary conditions including the heat flux and the water level, observation is made for the melting location, progression, and the mass of molten material. Finally, the core melt progression model is developed from the visual inspection and quantitative analysis of the experimental data. As the core material relocates

  7. Development of nuclear fuel microsphere handling techniques and equipment

    International Nuclear Information System (INIS)

    Mack, J.E.; Suchomel, R.R.; Angelini, P.

    1979-01-01

    Considerable progress has been made in the development of microsphere handling techniques and equipment for nuclear applications. Work at Oak Ridge National Laboratory with microspherical fuel forms dates back to the early sixties with the development of the sol-gel process. Since that time a number of equipment items and systems specifically related to microsphere handling and characterization have been identified and developed for eventual application in a remote recycle facility. These include positive and negative pressure transfer systems, samplers, weighers, a blender-dispenser, and automated devices for particle size distribution and crushing strength analysis. The current status of these and other components and systems is discussed

  8. Progress in researches on MOX fuel pellet producing technology in China

    International Nuclear Information System (INIS)

    Hu Xiaodan

    2010-01-01

    Being the key section of nuclear-fuel cycle, the producing technology of MOX(UO 2 -PuO 2 ) fuel had driven to maturity in France, England, Russia, Belgium, etc. MOX fuel had been applied in FBR and LWR successfully in those countries. With the rapidly developing of nuclear-generated power, the MOX fuel for FBR and LWR was active demanded in China. However, the producing technology of MOX fuel developed slowly. During the period of 'the seventh five year's project', MOX fuel pellet was produced by mechanically mixed method and oxalate deposited method, respectively. Parts of cool performance of MOX fuel pellet produced by oxalate deposited method reached the qualification of fuel for FBR. During the period of 'the ninth five year's project' and 'the tenth five year's project', the technical route of producing MOX fuel was determined, and the test line of producing MOX fuel was built preliminarily. In the same time, the producing technology and analyzing technology of MOX fuel pellet by mechanically mixed was studied roundly, and the representative analogue pellet(UO 2 -CeO 2 ) was produced. That settled the supporting technology for the commercial process and research of MOX fuel rod and MOX fuel module. (authors)

  9. Progress in Development of I2S-LWR Concept

    International Nuclear Information System (INIS)

    Petrovic, Bojan

    2014-01-01

    The paper will present the progress in developing the Integral Inherently Safe Light Water Reactor (12S-LWR) concept. This new concept aims to combine the competitive economics of a large nuclear power plant, with enhanced safety achieved by the integral primary circuit configuration (previously considered only for PWRs with power levels not exceeding several hundred MWc), and with enhanced accident tolerance (to address concerns after the Fukushima Dai-lchi accidents). Several new technologies are being developed to enable this concept, including novel silicide fuel and micro-channel primary heat exchangers. This project is performed by a multi-disciplinary multi-organization team led by Georgia Tech, including academia, a national laboratory, nuclear industry, and a power utility, wit expected participation of the University of Zagreb. (author)

  10. Nuclear Fuel Cycle Strategy For Developing Countries

    International Nuclear Information System (INIS)

    Kim, Chang Hyo

    1987-01-01

    The world's uranium market is very uncertain at the moment while other front-end fuel cycle services including enrichment show a surplus of supply. Therefore, a current concern of developing countries is how to assure a long-term stable supply of uranium, so far as front-end fuel cycle operation is concerned. So, as for the front-end fuel cycle strategy, I would like to comment only on uranium procurement strategy. I imagine that you are familiar with, yet let me begin my talk by having a look at, the nuclear power development program and current status of fuel cycle technology of developing countries. It is a nice thing to achieve the full domestic control of fuel cycle operation. The surest way to do so is localization of related technology. Nevertheless, developing at a time due to enormous capital requirements, not to mention the non-proliferation restrictions. Therefore, the important which technology to localize prior to other technology and how to implement. The non-proliferation restriction excludes the enrichment and reprocessing technology for the time being. As for the remaining technology the balance between the capital costs and benefits must dictate the determination of the priority as mentioned previously. As a means to reduce the commercial risk and heavy financial burdens, the multi-national joint venture of concerned countries is desirable in implementing the localization projects

  11. Coordinated irradiation plan for the Fuel Refabrication and Development Program

    International Nuclear Information System (INIS)

    Barner, J.O.

    1979-04-01

    The Department of Energy's Fuel Refabrication and Development (FRAD) Program is developing a number of proliferation-resistant fuel systems and forms for alternative use in nuclear reactors. A major portion of the program is the development of irradiation behavioral information for the fuel system/forms with the ultimate objective of qualifying the design for licensing and commercial utilization. The nuclear fuel systems under development include denatured thoria--urania fuels and spiked urania--plutonia or thoria--plutonia fuels. The fuel forms being considered include pellet fuel produced from mechanically mixed or coprecipitated feed materials, pellet fuel fabricated from partially calcined gel-derived or freeze-dried spheres (hybrid fuel) and packed-particle fuel produced from sintered gel-derived spheres (sphere-pac). This document describes the coordinated development program that will be used to test and demonstrate the irradiation performance of alternative fuels

  12. Development of advanced spent fuel management process

    International Nuclear Information System (INIS)

    Park, Seong Won; Shin, Y. J.; Cho, S. H.

    2004-03-01

    The research on spent fuel management focuses on the maximization of the disposal efficiency by a volume reduction, the improvement of the environmental friendliness by the partitioning and transmutation of the long lived nuclides, and the recycling of the spent fuel for an efficient utilization of the uranium source. In the second phase which started in 2001, the performance test of the advanced spent fuel management process consisting of voloxidation, reduction of spent fuel and the lithium recovery process has been completed successfully on a laboratory scale. The world-premier spent fuel reduction hot test of a 5 kgHM/batch has been performed successfully by joint research with Russia and the valuable data on the actinides and FPs material balance and the characteristics of the metal product were obtained with experience to help design an engineering scale reduction system. The electrolytic reduction technology which integrates uranium oxide reduction in a molten LiCl-Li 2 O system and Li 2 O electrolysis is developed and a unique reaction system is also devised. Design data such as the treatment capacity, current density and mass transfer behavior obtained from the performance test of a 5 kgU/batch electrolytic reduction system pave the way for the third phase of the hot cell demonstration of the advanced spent fuel management technology

  13. Progress performance report of clean uses of fossil fuels

    Energy Technology Data Exchange (ETDEWEB)

    Todd, Jr., Lee T.; Boggess, Ronald J.; Carson, Ronald J.; Falkenberg, Virginia P.; Flanagan, Patrick; Hettinger, Jr., William P.; Kimel, Kris; Kupchella, Charles E.; Magid, Lee J.; McLaughlin, Barbara; Royster, Wimberly C.; Streepey, Judi L.; Wells, James H.; Stencel, John; Derbyshire, Frank J.; Hanley, Thomas R.; Magid, Lee J.; McEllistrem, Marc T.; Riley, John T.; Steffen, Joseph M.

    1992-01-01

    A one-year USDOE/EPSCOR Traineeship Grant, entitled Clean Uses of Fossil Fuels.'' was awarded to the Kentucky EPSCoR Committee in September 1991 and administered through the the DOE/EPSCoR Subcommittee. Ten Traineeships were awarded to doctoral students who are enrolled or accepted into Graduate Programs at either the University of Kentucky or the University of Louisville. The disciplines of these students include Biology, Chemical Engineering, Chemistry, Geological Sciences, and Physics. The methods used for a statewide proposal solicitation and to award the Traineeships are presented. The review panel and Kentucky DOE/EPSCoR Subcommittee involved in awarding the Traineeships are described. A summary of the proposed research to be performed within these awards is presented, along with a description of the qualifications of the faculty and students who proposed projects. Future efforts to increase participation in Traineeship proposals for the succeeding funding period are outlined.

  14. Progress performance report of clean uses of fossil fuels

    Energy Technology Data Exchange (ETDEWEB)

    1992-09-01

    A one-year USDOE/EPSCOR Traineeship Grant, entitled ``Clean Uses of Fossil Fuels.`` was awarded to the Kentucky EPSCoR Committee in September 1991 and administered through the the DOE/EPSCoR Subcommittee. Ten Traineeships were awarded to doctoral students who are enrolled or accepted into Graduate Programs at either the University of Kentucky or the University of Louisville. The disciplines of these students include Biology, Chemical Engineering, Chemistry, Geological Sciences, and Physics. The methods used for a statewide proposal solicitation and to award the Traineeships are presented. The review panel and Kentucky DOE/EPSCoR Subcommittee involved in awarding the Traineeships are described. A summary of the proposed research to be performed within these awards is presented, along with a description of the qualifications of the faculty and students who proposed projects. Future efforts to increase participation in Traineeship proposals for the succeeding funding period are outlined.

  15. Progress in fuel-motion-diagnostics instrumentation evaluation at Parka

    International Nuclear Information System (INIS)

    Evans, A.E. Jr.; Orndoff, J.D.

    1979-01-01

    Hodoscope neutron and gamma-ray imaging of bundles containing from 1 to 127 pins have been studied. Ability to image a single missing pin from assemblies of up to 127 pins has been established. The image of a single-pin void in a test assembly was found to vary with the depth of the void in the assembly, for both 37-pin and 127-pin bundles. The degree to which a thick steel test casing will impede image quality has been studied. A 21-mm-thick casing was found to reduce the sensitivity of fast-neutron hodoscope images to fuel defects by 30% and of gamma-ray images by 50%. 9 refs

  16. Progress and challenges in utilization of palm oil biomass as fuel for decentralized electricity generation

    Energy Technology Data Exchange (ETDEWEB)

    Bazmi, Aqeel Ahmed [Process Systems Engineering Centre (PROSPECT), Department of Chemical Engineering, Faculty of Chemical and Natural Resources Engineering, University Technology Malaysia, Skudai 81310, Johor Bahru, JB (Malaysia); Biomass Conversion Research Center (BCRC), Department of Chemical Engineering, COMSATS Institute of Information Technology, Lahore (Pakistan); Zahedi, Gholamreza; Hashim, Haslenda [Process Systems Engineering Centre (PROSPECT), Department of Chemical Engineering, Faculty of Chemical and Natural Resources Engineering, University Technology Malaysia, Skudai 81310, Johor Bahru, JB (Malaysia)

    2011-01-15

    It has been broadly accepted worldwide that global warming, indeed, is the greatest threat of the time to the environment. Renewable energy (RE) is expected as a perfect solution to reduce global warming and to endorse sustainable development. Progressive release of greenhouse gases (GHG) from increasing energy-intensive industries has eventually caused human civilization to suffer. Realizing the exigency of reducing emissions and simultaneously catering to needs of industries, researchers foresee the RE as the perfect entrant to overcome these challenges. RE provides an effective option for the provision of energy services from the technical point of view while biomass, a major source of energy in the world until before industrialization when fossil fuels become dominant, appears an important renewable source of energy and researches have proven from time to time its viability for large-scale production. Being a widely spread source, biomass offers the execution of decentralized electricity generation gaining importance in liberalized electricity markets. The decentralized power is characterized by generation of electricity nearer to the demand centers, meeting the local energy needs. Researchers envisaged an increasing decentralization of power supply, expected to make a particular contribution to climate protection. This article investigates the progress and challenges for decentralized electricity generation by palm oil biomass according to the overall concept of sustainable development. (author)

  17. Progress and challenges in utilization of palm oil biomass as fuel for decentralized electricity generation

    International Nuclear Information System (INIS)

    Bazmi, Aqeel Ahmed; Zahedi, Gholamreza; Hashim, Haslenda

    2011-01-01

    It has been broadly accepted worldwide that global warming, indeed, is the greatest threat of the time to the environment. Renewable energy (RE) is expected as a perfect solution to reduce global warming and to endorse sustainable development. Progressive release of greenhouse gases (GHG) from increasing energy-intensive industries has eventually caused human civilization to suffer. Realizing the exigency of reducing emissions and simultaneously catering to needs of industries, researchers foresee the RE as the perfect entrant to overcome these challenges. RE provides an effective option for the provision of energy services from the technical point of view while biomass, a major source of energy in the world until before industrialization when fossil fuels become dominant, appears an important renewable source of energy and researches have proven from time to time its viability for large-scale production. Being a widely spread source, biomass offers the execution of decentralized electricity generation gaining importance in liberalized electricity markets. The decentralized power is characterized by generation of electricity nearer to the demand centers, meeting the local energy needs. Researchers envisaged an increasing decentralization of power supply, expected to make a particular contribution to climate protection. This article investigates the progress and challenges for decentralized electricity generation by palm oil biomass according to the overall concept of sustainable development. (author)

  18. Status of spent fuel shipping cask development

    International Nuclear Information System (INIS)

    Hall, I.K.; Hinschberger, S.T.

    1989-01-01

    This paper discusses how several new-generation shopping cask systems are being developed for safe and economical transport of commercial spent nuclear fuel and other radioactive wastes for the generating sites to a federal geologic repository or monitored retrievable storage (MRS) facility. Primary objectives of the from-reactor spent fuel cask development work are: to increase cask payloads by taking advantage of the increased at-reactor storage time under the current spent fuel management scenario, to facilitate more efficient cask handling operations with reduced occupational radiation exposure, and to promote standardization of the physical interfaces between casks and the shipping and receiving facilities. Increased cask payloads will significantly reduce the numbers of shipments, with corresponding reductions in transportation costs and risks to transportation workers, cask handling personnel, and the general public

  19. Development of advanced spent fuel management process

    International Nuclear Information System (INIS)

    Ro, Seung Gy; Shin, Y. J.; Do, J. B.; You, G. S.; Seo, J. S.; Lee, H. G.

    1998-03-01

    This study is to develop an advanced spent fuel management process for countries which have not yet decided a back-end nuclear fuel cycle policy. The aims of this process development based on the pyroreduction technology of PWR spent fuels with molten lithium, are to reduce the storage volume by a quarter and to reduce the storage cooling load in half by the preferential removal of highly radioactive decay-heat elements such as Cs-137 and Sr-90 only. From the experimental results which confirm the feasibility of metallization technology, it is concluded that there are no problems in aspects of reaction kinetics and equilibrium. However, the operating performance test of each equipment on an engineering scale still remain and will be conducted in 1999. (author). 21 refs., 45 tabs., 119 figs

  20. Baseline gas turbine development program. Eighteenth quarterly progress report

    Energy Technology Data Exchange (ETDEWEB)

    Schmidt, F W; Wagner, C E [comps.

    1977-04-30

    Progress is reported for a program whose goals are to demonstrate an experimental upgraded gas turbine powered automobile which meets the 1978 Federal Emissions Standards, has significantly improved fuel economy, and is competitive in performance, reliability, and potential manufacturing cost with the conventional piston engine powered, compact-size American automobile. Initial running of the upgraded engine took place on July 13, 1976. The engine proved to be mechanically sound, but was also 43% deficient in power. A continuing corrective development effort has to date reduced the power deficiency to 32%. Compressor efficiency was increased 2 points by changing to a 28-channel diffuser and tandem deswirl vanes; improved processing of seals has reduced regenerator leakage from about 5 to 2.5% of engine flow; a new compressor turbine nozzle has increased compressor turbine stage efficiency by about 1 point; and adjustments to burner mixing ports has reduced pressure drop from 2.8 to 2.1% of engine pressure. Key compressor turbine component improvements are scheduled for test during the next quarterly period. During the quarter, progress was also made on development of the Upgraded Vehicle control system; and instrumentation of the fourth program engine was completed by NASA. The engine will be used for development efforts at NASA LeRC.

  1. Developments in MOX fuel pellet fabrication technology: Indian experience

    International Nuclear Information System (INIS)

    Kamath, H.S.; Majumdar, S.; Purusthotham, D.S.C.

    1998-01-01

    India is interested in mixed oxide (MOX) fuel technology for better utilisation of its nuclear fuel resources. In view of this, a programme involving MOX fuel design, fabrication and irradiation in research and power reactors has been taken up. A number of experimental irradiations in research reactors have been carried out and a few MOX assemblies of ''All Pu'' type have been loaded in our commercial BWRs at Tarapur. An island type of MOX fuel design is under study for use in PHWRs which can increase the burn-up of the fuel by more than 30% compared to natural UO 2 fuel. The MOX fuel pellet fabrication technology for the above purpose and R and D efforts in progress for achieving better fuel performance are described in the paper. The standard MOX fuel fabrication route involves mechanical mixing and milling of UO 2 and PuO 2 powders. After detailed investigations with several types of mixing and milling equipments, dry attritor milling has been found to be the most suitable for this operation. Neutron Coincident Counting (NCC) technique was found to be the most convenient and appropriate technique for quick analysis of Pu content in milled MOX powder and to know Pu mixing is homogenous or not. Both mechanical and hydraulic presses have been used for powder compaction for green pellet production although the latter has been preferred for better reproducibility. Low residue admixed lubricants have been used to facilitate easy compaction. The normal sintering temperature used in Nitrogen-Hydrogen atmosphere is between 1600 deg. C to 1700 deg. C. Low temperature sintering (LTS) using oxidative atmospheres such as carbon dioxide, Nitrogen and coarse vacuum have also been investigated on UO 2 and MOX on experimental scale and irradiation behaviour of such MOX pellets is under study. Ceramic fibre lined batch furnaces have been found to be the most suitable for MOX pellet production as they offer very good flexibility in sintering cycle, and ease of maintainability

  2. Strategies for fuel cell product development. Developing fuel cell products in the technology supply chain

    International Nuclear Information System (INIS)

    Hellman, H.L.

    2004-01-01

    Due to the high cost of research and development and the broad spectrum of knowledge and competences required to develop fuel cell products, many product-developing firms outsource fuel cell technology, either partly or completely. This article addresses the inter-firm process of fuel cell product development from an Industrial Design Engineering perspective. The fuel cell product development can currently be characterised by a high degree of economic and technical uncertainty. Regarding the technology uncertainty: product-developing firms are more often then not unfamiliar with fuel cell technology technology. Yet there is a high interface complexity between the technology supplied and the product in which it is to be incorporated. In this paper the information exchange in three current fuel cell product development projects is analysed to determine the information required by a product designer to develop a fuel cell product. Technology transfer literature suggests that transfer effectiveness is greatest when the type of technology (technology uncertainty) and the type of relationship between the technology supplier and the recipient are carefully matched. In this line of thinking this paper proposes that the information required by a designer, determined by the design strategy and product/system volume, should be met by an appropriate level of communication interactivity with a technology specialist. (author)

  3. [Landscape classification: research progress and development trend].

    Science.gov (United States)

    Liang, Fa-Chao; Liu, Li-Ming

    2011-06-01

    Landscape classification is the basis of the researches on landscape structure, process, and function, and also, the prerequisite for landscape evaluation, planning, protection, and management, directly affecting the precision and practicability of landscape research. This paper reviewed the research progress on the landscape classification system, theory, and methodology, and summarized the key problems and deficiencies of current researches. Some major landscape classification systems, e. g. , LANMAP and MUFIC, were introduced and discussed. It was suggested that a qualitative and quantitative comprehensive classification based on the ideology of functional structure shape and on the integral consideration of landscape classification utility, landscape function, landscape structure, physiogeographical factors, and human disturbance intensity should be the major research directions in the future. The integration of mapping, 3S technology, quantitative mathematics modeling, computer artificial intelligence, and professional knowledge to enhance the precision of landscape classification would be the key issues and the development trend in the researches of landscape classification.

  4. Corrosion studies on retrievable spent fuel containers: a progress report

    International Nuclear Information System (INIS)

    Ludemann, W.D.; Abrego, L.; McCright, R.D.

    1978-12-01

    Spent fuel canisters stored in halite (NaCl) deposits (salt beds) are subject to a severely corrosive environment when the hot brine inclusions, rich in calcium and magnesium chlorides, migrate to the canister. Since no data base exists on corrosion in halite brines, a survey was made of the corrosion resistance of potential canister materials in other concentrated brine environments. Corrosion-resistant metals include Ta, Ti Code 12, TiPd Alloy, Inconel 625, Hastelloy C-276, and Fe-base 29-4 Alloy. Although carbon steels have cost and availability advantages, they suffer from excessive corrosion rates in brines. Corrosion-resistant nonmetals include carbon, Teflon-type fluorocarbons, epoxide coatings, and polymer cements. While these materials are not suitable for constructing the canister, they could be used as a protective coating on a carbon steel canister. On the basis of this survey, we recommend a coated carbon steel canister, used with cathodic protection. It is important to start a test program to gather a data base on the corrosion of materials in halite brines and to verify the suitability of canister materials

  5. Technical development of fuel alcohol

    Energy Technology Data Exchange (ETDEWEB)

    1988-06-01

    Research and development of a technology for biologically manufacturing alcohol from agricultural and forestry wastes has been conducted according to an eight year-program beginning in 1983. This paper presents the findings in FY 1987 and the future schedule. Exploration and breeding of superior bacteria are the basic subject through the eight years. In FY 1987, preparation and evaluation of hybrid plasmids into which Zymomonas, BETA-glucosidase gene and CM case gene are inserted, improvement of variation to enhance the salt resistance of Zymomonas and screening of Cm-and Sm-resistant bacteria to develop thermophilic, anaerobic cellulose were made. In addition, the total process combining the cell adhesion method as the immobilization technique with the flash technique is continuously studied. Improvement of the salt-resistance of Zymomonas by increasing the density with photosetting resin, the upper concentration of alcohol and effect of pulverzing treatment in a small apparatus were investigated. A test plant was designed and constructed. (3 photos.)

  6. Review of monitoring instruments for transuranics in fuel fabrication and reprocessing plants. A progress report to the physical and technological programs, Division of Biomedical and Environmental Research, U.S. Energy Research and Development Administration

    International Nuclear Information System (INIS)

    Kordas, J.F.; Phelps, P.L.

    A comprehensive review of the monitoring instruments for transuranic elements released from nuclear fuel fabrication and reprocessing plants has been compiled. The extent of routine operational releases has been reviewed for the light water reactor (LWR) fuel cycle (including plutonium recycle), the breeder reactor fuel cycle, and the high-temperature gas cooled reactor (HTGR) fuel cycle. The stack monitoring instrumentation presently in use at the various fabrication and reprocessing plants around the country is discussed. Sampling difficulties and the effectiveness of the entire sampling system are reviewed, as are the measurement problems for alpha-emitting, long-lived, transuranic aerosols, 129 I, 106 Ru, and tritium oxide. The potential problems in the HTGR fuel cycle such as the measurement of releases of alpha-emitting aerosols and of gaseous releases of 220 Rn and 14 C are also considered

  7. FuelPHP application development blueprints

    CERN Document Server

    Drouyer, Sébastien

    2015-01-01

    This book is for intermediary to seasoned web developers who want to learn how to use the FuelPHP framework and build complex projects using it. You should be familiar with PHP, HTML, CSS, and JavaScript, but no prior knowledge about MVC frameworks is required.

  8. Development of spent fuel remote handling technology

    Energy Technology Data Exchange (ETDEWEB)

    Park, B. S.; Yoon, J. S.; Hong, H. D. (and others)

    2007-02-15

    In this research, the remote handling technology was developed for the ACP application. The ACP gives a possible solution to reduce the rapidly cumulative amount of spent fuels generated from the nuclear power plants in Korea. The remote technologies developed in this work are a slitting device, a voloxidizer, a modified telescopic servo manipulator and a digital mock-up. A slitting device was developed to declad the spent fuel rod-cuts and collect the spent fuel UO{sub 2} pellets. A voloxidizer was developed to convert the spent fuel UO{sub 2} pellets obtained from the slitting process in to U{sub 3}O{sub 8} powder. Experiments were performed to test the capabilities and remote operation of the developed slitting device and voloxidizer by using simulated rod-cuts and fuel in the ACP hot cell. A telescopic servo manipulator was redesigned and manufactured improving the structure of the prototype. This servo manipulator was installed in the ACP hot cell, and the target module for maintenance of the process equipment was selected. The optimal procedures for remote operation were made through the maintenance tests by using the servo manipulator. The ACP digital mockup in a virtual environment was established to secure a reliability and safety of remote operation and maintenance. The simulation for the remote operation and maintenance was implemented and the operability was analyzed. A digital mockup about the preliminary conceptual design of an enginnering-scale ACP was established, and an analysis about a scale of facility and remote handling was accomplished. The real-time diagnostic technique was developed to detect the possible fault accidents of the slitting device. An assessment of radiation effect for various sensors was also conducted in the radiation environment.

  9. Process development and fabrication for sphere-pac fuel rods

    International Nuclear Information System (INIS)

    Welty, R.K.; Campbell, M.H.

    1981-06-01

    Uranium fuel rods containing sphere-pac fuel have been fabricated for in-reactor tests and demonstrations. A process for the development, qualification, and fabrication of acceptable sphere-pac fuel rods is described. Special equipment to control fuel contamination with moisture or air and the equipment layout needed for rod fabrication is described and tests for assuring the uniformity of the fuel column are discussed. Fuel retainers required for sphere-pac fuel column stability and instrumentation to measure fuel column smear density are described. Results of sphere-pac fuel rod fabrication campaigns are reviewed and recommended improvements for high throughput production are noted

  10. Progress in quantitative GPR development at CNDE

    Energy Technology Data Exchange (ETDEWEB)

    Eisenmann, David; Margetan, F. J.; Chiou, C.-P.; Roberts, Ron; Wendt, Scott [Center for Nondestructive Evaluation, Iowa State University, 1915 Scholl Road, Ames, IA 50011-3042 (United States)

    2014-02-18

    Ground penetrating radar (GPR) uses electromagnetic (EM) radiation pulses to locate and map embedded objects. Commercial GPR instruments are generally geared toward producing images showing the location and extent of buried objects, and often do not make full use of available absolute amplitude information. At the Center for Nondestructive Evaluation (CNDE) at Iowa State University efforts are underway to develop a more quantitative approach to GPR inspections in which absolute amplitudes and spectra of measured signals play a key role. Guided by analogous work in ultrasonic inspection, there are three main thrusts to the effort. These focus, respectively, on the development of tools for: (1) analyzing raw GPR data; (2) measuring the EM properties of soils and other embedding media; and (3) simulating GPR inspections. This paper reviews progress in each category. The ultimate goal of the work is to develop model-based simulation tools that can be used assess the usefulness of GPR for a given inspection scenario, to optimize inspection choices, and to determine inspection reliability.

  11. Progress in quantitative GPR development at CNDE

    Science.gov (United States)

    Eisenmann, David; Margetan, F. J.; Chiou, C.-P.; Roberts, Ron; Wendt, Scott

    2014-02-01

    Ground penetrating radar (GPR) uses electromagnetic (EM) radiation pulses to locate and map embedded objects. Commercial GPR instruments are generally geared toward producing images showing the location and extent of buried objects, and often do not make full use of available absolute amplitude information. At the Center for Nondestructive Evaluation (CNDE) at Iowa State University efforts are underway to develop a more quantitative approach to GPR inspections in which absolute amplitudes and spectra of measured signals play a key role. Guided by analogous work in ultrasonic inspection, there are three main thrusts to the effort. These focus, respectively, on the development of tools for: (1) analyzing raw GPR data; (2) measuring the EM properties of soils and other embedding media; and (3) simulating GPR inspections. This paper reviews progress in each category. The ultimate goal of the work is to develop model-based simulation tools that can be used assess the usefulness of GPR for a given inspection scenario, to optimize inspection choices, and to determine inspection reliability.

  12. Progress in quantitative GPR development at CNDE

    International Nuclear Information System (INIS)

    Eisenmann, David; Margetan, F. J.; Chiou, C.-P.; Roberts, Ron; Wendt, Scott

    2014-01-01

    Ground penetrating radar (GPR) uses electromagnetic (EM) radiation pulses to locate and map embedded objects. Commercial GPR instruments are generally geared toward producing images showing the location and extent of buried objects, and often do not make full use of available absolute amplitude information. At the Center for Nondestructive Evaluation (CNDE) at Iowa State University efforts are underway to develop a more quantitative approach to GPR inspections in which absolute amplitudes and spectra of measured signals play a key role. Guided by analogous work in ultrasonic inspection, there are three main thrusts to the effort. These focus, respectively, on the development of tools for: (1) analyzing raw GPR data; (2) measuring the EM properties of soils and other embedding media; and (3) simulating GPR inspections. This paper reviews progress in each category. The ultimate goal of the work is to develop model-based simulation tools that can be used assess the usefulness of GPR for a given inspection scenario, to optimize inspection choices, and to determine inspection reliability

  13. Low enrichment fuel development at INEL

    International Nuclear Information System (INIS)

    Newton, D.G.

    1993-01-01

    EG and G Idaho, Inc. is under contract to the Department of Energy to operate the Idaho National Engineering Laboratory (INEL). The INEL is located in southeastern Idaho. This facility has been operating since 1949 and was originally called the National Reactor Testing Station. Several contractors manage projects on this facility. Most projects at INEL are concerned with either reactor safety or irradiation testing. At Test Area North, for example, experiments are being conducted on the effects of loss of coolant. At the Test Reactor Area the ATR (Advanced Test Reactor) and ETR (Engineering Test Reactor) are used for irradiation testing and, of course, those of you working at Argonne will recognize the Experimental Breeder Reactors I and II. SPERT is an acronym for Special Power Excursion Reactor Test. A part of this former reactor facility has been converted into a fuel fabrication laboratory facility. At SPERT IV a miniature fabrication facility has been set up to duplicate the aluminide plate fuel processing line at Atomics International. In other words, a model of the supplier's processing has been created, so that what process changes are developed here can then be scaled up to production. The process is described showing: making UAI x powder, making compact for fuel core, making experimental fuel plate and compact assembly, inspection and testing the fuel plate. Main concern was related to possible swelling

  14. MOX fuel fabrication: Technical and industrial developments

    International Nuclear Information System (INIS)

    Lebastard, G.; Bairiot, H.

    1990-01-01

    The plutonium available in the near future is generally estimated rather precisely on the basis of the reprocessing contracts and the performance of the reprocessing plants. A few years ago, decision makers were convinced that a significant share of this fissile material would be used as the feed material for fast breeder reactors (FBRs) or other advanced reactors. The facts today are that large reprocessing plants are coming into commercial operations: UP3 and soon UP2-800 and THORP, but that FBR deployment is delayed worldwide. As a consequence, large quantities of plutonium will be recycled in light water reactors as mixed oxide (MOX) fuels. MOX fuel technology has been properly demonstrated in the past 25 years. All specific problems have been addressed, efficient fabrication processes and engineering background have been implemented to a level of maturity which makes MOX fuel behaving as well as Uranium fuel. The paper concentrates on todays MOX fabrication expertise and presents the technical and industrial developments prepared by the MOX fuel fabrication industry for this last decade of the century

  15. Fuel development at CERCA. Status of development - September 1984

    International Nuclear Information System (INIS)

    Fanjas, Y.; Dewez, Ph.; Savornin, B.

    1985-01-01

    Since 1978, CERCA has developed high density aluminide (UAl x ), oxide (U 3 O 8 ) and silicides (U 3 Si 2 , U 3 Si) fuels allowing the use of 19.75 enriched uranium in research and test reactors. An extensive irradiation program has been carried out to test the full size fuel plates and fuel elements fabricated by CERCA. So far, all the irradiation tests have given satisfactory results whatever the uranium density, the burn-up level and the type of fuel. In particular, silicides which cover the whole density range from 1 to 7 g U/cm 3 appear more and more as the standard fuels for the future. (author)

  16. Progress on Fast Reactor Development in Japan

    International Nuclear Information System (INIS)

    Ohira, Hiroaki; Uto, Nariaki

    2012-01-01

    Situation of National Policy Making and FaCT Project: On July 19, 2011, JAEC decided to continue the FR cycle technology development program in the limited range of activities to contribute to international standardization (ex. safety criteria) and to maintain the technology base level until the determination of new nuclear energy policy. On Sept. 27, 2011, JAEC restarted the deliberation process for new Framework for Nuclear Energy Policy. - The process was suspended after the Fukushima NPS accidents. - Major issues: Safety, Cost, Nuclear Power and Fuel Cycle Options, Waste Management, International Perspectives, R&D planning, etc. - The process has been carried out to determine the new Framework with the relationship to new governmental energy/environmental policy making. On Dec. 21, 2011, Energy and Environment Council complied Basic guideline toward Presentation of Alternatives regarding the Strategy for Energy and the Environment. In the FaCT project, focus has been on further improvement on safety of next generation SFRs based on lessons learned from the Fukushima NPS accidents

  17. Progress on development of nuclear power in Japan

    International Nuclear Information System (INIS)

    Anon.

    2000-01-01

    Since three Laws on the nuclear power were published 45 years has passed. Now, development on nuclear power in Japan is at an emergent state. In Japan, 51 units of commercial nuclear reactors with 44.917 GW are in operation, occupy about 37% of total electric power generation, and is positioned at an essential basic energy source supporting economical society in Japan. However, an accident occurred at Tokai Works of the JCO Co., Ltd., one of the uranium reconversion company, on September 30, 1999, was the first critical accident in Japan, and became the worst case in history on development of nuclear power in Japan, because of forming three heavy radiation disabled persons (One of them was dead) in its operators. This was a big crisis with relation to existence on development of nuclear power in Japan, by which anxiety and distrust of the Japanese against the nuclear power were amplified rapidly. On the other side, for Japan short in energy sources and of a big energy consumption, in order to intend for a long term to carry out energy security, global environmental conservation, and sustainable maintenance of essential growth, it remains to be one of important optional methods to further promote nuclear power generation and to establish nuclear fuel cycle. Here were described on progress on peaceful applications of nuclear power in Japan, progress on the field of nuclear power in Japan (from 1955 to 1999), progress on Tokai nuclear power station, introduction of nuclear power generation and effort on its domestic production. (G.K.)

  18. Developing safety in the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Brown, M.L.

    1996-01-01

    The nuclear fuel cycle had its origins in the new technology developed in the 1940s and 50s involving novel physical and chemical processes. At the front end of the cycle, mining, milling and fuel fabrication all underwent development, but in general the focus of process development and safety concerns was the reprocessing stage, with radiation, contamination and criticality the chief hazards. Safety research is not over and there is still work to be done in advancing technical knowledge to new generation nuclear fuels such as Mixed Oxide Fuel and in refining knowledge of margins and of potential upset conditions. Some comments are made on potential areas for work. The NUCEF facility will provide many useful data to aid safety analysis and accident prevention. The routine operations in such plants, basically chemical factories, requires industrial safety and in addition the protection of workers against radiation or contamination. The engineering and management measures for this were novel and the early operation of such plants pioneering. Later commissioning and operating experience has improved routine operating safety, leading to a new generation of factories with highly developed worker protection, engineering safeguards and safety management systems. Ventilation of contamination control zones, remote operation and maintenance, and advanced neutron shielding are engineering examples. In safety management, dose control practices, formally controlled operating procedures and safety cases, and audit processes are comparable with, or lead, best industry practice in other hazardous industries. Nonetheless it is still important that the knowledge and experience from operating plants continue to be gathered together to provide a common basis for improvement. The NEA Working Group on Fuel Cycle Safety provides a forum for much of this interchange. Some activities in the Group are described in particular the FINAS incident reporting system. (J.P.N.)

  19. Clean uses of fossil fuels. Progress performance report, September 29, 1991--January 25, 1994

    Energy Technology Data Exchange (ETDEWEB)

    Stencel, J.M.

    1994-01-25

    Science and engineering doctoral students performing energy related research were supported by a USDOE/ESPCoR Traineeship grant awarded to the Kentucky EPSCoR Committee. The grant, administered by the KY DOE/EPSCoR Subcommittee, focused on research having the general description of {open_quotes}Clean Uses of Fossil Fuels{close_quotes}. The value of the grant was $500,000 for three years duration, beginning September 30, 1991 and ending September 29, 1994. Ten PhD students were selected for support during the first year of the Traineeship. Upon reviewing coursework and research progress of the students at the end of the first year, the KY DOE/EPSCoR Subcommittee awarded a second year of support at the same $25,000/year funding level. A total of 12 students will have been supported during the duration of the grant as a consequence of one student completing his degree during the support period and of one student deciding that she wanted to complete only a Masters rather than a PhD degree. The students supported were at either the University of Kentucky or the University of Louisville - the two PhD, science and engineering granting universities within the Commonwealth of Kentucky. The disciplines of these students included Biology, Chemical Engineering, Chemistry, Geological Sciences, and Physics. The methods used for the initial statewide solicitation for student support, the annual review of the students progress for support renewal, and a summary of progress and impact of the awards after two years are presented. It is shown that the Traineeships presented opportunities to: perform high quality research; initiate interactions between different scientific disciplines and departments; develop collaborations at national DOE laboratories, universities outside of Kentucky and industries; and establish research ideas for submittal to funding agencies.

  20. Irradiated fuel performance evaluation technology development

    International Nuclear Information System (INIS)

    Koo, Yang Hyun; Bang, J. G.; Kim, D. H.

    2012-01-01

    Alpha version performance code for dual-cooled annular fuel under steady state operation, so called 'DUOS', has been developed applying performance models and proposed methodology. Furthermore, nonlinear finite element module which could be integrated into transient/accident fuel performance code was also developed and evaluated using commercial FE code. The first/second irradiation and PIE test of annular pellet for dual-cooled annular fuel in the world have been completed. In-pile irradiation test DB of annular pellet up to burnup of 10,000 MWd/MTU through the 1st test was established and cracking behavior of annular pellet and swelling rate at low temperature were studied. To do irradiation test of dual-cooled annular fuel under PWR's simulating steady-state conditions, irradiation test rig/rod design/manufacture of mock-up/performance test have been completed through international collaboration program with Halden reactor project. The irradiation test of large grain pellets has been continued from 2002 to 2011 and completed successfully. Burnup of 70,000 MWd/MTU which is the highest burnup among irradiation test pellets in domestic was achieved

  1. Canadian fuel development program in 1997/98

    International Nuclear Information System (INIS)

    Lau, J.H.; Kohn, E.; Sejnoha, R.; Cox, D.S.; Macici, N.N.; Steed, R.G.

    1997-01-01

    This paper describes the CANDU fuel development activities in Canada during 1997 through 1998. The activities include those of the Fuel Technology Program sponsored by the CANDU Owners Group. The goal of the Fuel Technology Program is to maintain and improve the reliability, economics and safety of CANDU fuel in operating reactors. These activities, therefore, concentrate on the present designs of 28-element and 37-element fuel bundles. The Canadian fuel development activities also include those of the Advanced Fuel and Fuel Cycle Technology Program at AECL. These activities concentrate on the development of advanced fuel designs and advanced fuel cycles, which among other advantages, can reduce the capital and fuelling costs, maintain operating margins in aging reactors, improve natural-uranium utilization, and reduce the amount of spent fuel. (author)

  2. Development of Coated Particle Fuel Technology

    International Nuclear Information System (INIS)

    Lee, Young Woo; Kim, B. G.; Kim, S. H.

    2007-06-01

    Uranium kernel fabrication technology using a wet chemical so-gel method, a key technology in the coated particle fuel area, is established up to the calcination step and the first sintering of UO2 kernel was attempted. Experiments on the parametric study of the coating process using the surrogate ZrO2 kernel give the optimum conditions for the PyC and SiC coating layer and ZrC coating conditions were obtained for the vaporization of the ZrCl4 precursor and coating condition from ZrC coating experiments using plate-type graphite substrate. In addition, by development of fuel performance analysis code a part of the code system is completed which enables the participation to the benchmark calculation and comparison in the IAEA collaborated research program. The technologies for irradiation and post irradiation examination, which are important in developing the HTGR fuel technology of its first kind in Korea was started to develop and, through a feasibility study and preliminary analysis, the technologies required to be developed are identified for further development as well as the QC-related basic technologies are reviewed, analyzed and identified for the own technology development. Development of kernel fabrication technology can be enhanced for the remaining sintering technology and completed based on the technologies developed in this phase. In the coating technology, the optimum conditions obtained using a surrogate ZrO2 kernel material can be applied for the uranium kernel coating process development. Also, after completion of the code development in the next phase, more extended participation to the international collaboration for benchmark calculation can be anticipated which will enable an improvement of the whole code system. Technology development started in this phase will be more extended and further focused on the detailed technology development to be required for the related technology establishment

  3. State of development progress of advanced boiling water reactors

    International Nuclear Information System (INIS)

    Tomono, Katsuya

    1982-01-01

    Advanced BWRs being developed at present are those aiming at the improvement of reliability and safety, the reduction of radiation exposure, the improvement of operation performance and capacity ratio of plants, and the heightening of economical efficiency by concentrating the experience and excellent technology of BWR manufacturers in the world. Now in Japan, the independence with Japanese technology is possible in almost all fields of nuclear power generation, and the improvement and standardization project is in progress to obtain the steady results. However, in order to pursue the most desirable BWRs conceivable at present, five BWR manufacturers in the world organized the Advanced Engineering Team in July, 1978, and performed the feasibility study of advanced BWRs for more than one year. Tokyo Electric Power Co., Inc., evaluated the report on the results, and judged that it is desirable to advance into the next stage aiming at the practical use of advanced BWRs. For the purpose, the electric power common research on advanced BWRs has been in progress, and the A-BWR project is to be examined in the third improvement and standardization project of MITI. The main technical features such as the coolant recirculation system of internal pump type, reinforced concrete containment vessels, fine motion control rod drive, improved core and fuel and others are explained. (Kako, I.)

  4. Fuel-cycle programs. Quarterly progress report, January-March 1981

    International Nuclear Information System (INIS)

    Steindler, M.J.; Vogler, S.; Vandegrift, G.F.

    1982-03-01

    A program continues for the development of an interim waste form that can be transported from facilities where waste is generated to terminal waste processing. Measurements of rates of leaching from simulated nuclear waste forms ar continuing. The data suggest that leaching from SRL glasses is inhibited by the formation of a Si/Ca/Fe/Al-rich surface layer. Impact tests of solid alternative waste forms (glass and ceramic) are being performed to assess source terms of airborne release (from particle size distributions) and to assess the increases in source terms of leach rates (from increases in surface area). Logging techniques are being developed to measure the relative amount of residual oil in a depleted oil reservoir. A simple technique for the elution of pore fluid (groundwater) from igneous rocks is described, using a special coreholder, as is the elution of brine (in the same apparatus) from a core of Precambrian granite. Testing and development of equipment for the destructive analysis of full-length irradiated fuel rods from the LWBR is in progress

  5. Development of coated particle fuel technology

    International Nuclear Information System (INIS)

    Cho, Moonsung; Kim, B. G.; Kim, D. J.

    2011-06-01

    Ammonia contacting method for prehardenning the surfaces of ADU liquid droplets and the ageing/washing/drying method and equipment for spherical dried-ADU particles were improved and tested with laboratory sacle. After the improvement of fabrication process, the sphericity of UO 2 kernel obtained to 1.1, and the sintered density and O/U ratio of final UO 2 kernel were above 10.60g/cm 3 . 2.01 respectively. Defects of SiC coating layer could be minimized by optimization of gas flow rate. The fracture strength of SiC layer increased from 450 MPa to 530 MPa by controlling the coating defects. An effort was made to develop the fundamental technology for the fuel element compact for use in High Temperature Gas-cooled Reactor(HTGR) through an establishment of fabrication process, required materials and process equipment as well as performing experiments to identify the basic process conditions and optimize them. Thermal load simulation and verification experiments were carried out for an assesment of the design feasibility of the irradiation rod. Out-of-pile testing of irradiation device such as measurement of pressure drop and vibration, endurance test was performed and the validity of its design was confirmed. A fuel performance analysis code, COPA has been developed to calculate the fuel temperature, the failure fractions of coated fuel particles, the release of fission products. The COPA code can be used to evaluate the performance of the high temperature reactor fuel under the reactor operation, irradiation, heating conditions. KAERI participated in the round robin test of IAEA CRP-6 program to characterize the diameter, sphericity, coating thickness, density and anisotropy of coated particles provided by Korea, USA and South Africa. QC technology was established for TRISO-coated fuel particle. A method for accurate measurement of the optical anisotropy factor for PyC layers of coated particles was developed. Technology and inspection procedures for density

  6. Advanced Fuel Cycle Initiative University Fellowship Program. Final Progress Report

    International Nuclear Information System (INIS)

    Dixon, Cathy

    2012-01-01

    2004-2011 Final Report for AFCI University Fellowship Program. The goal of this effort was to be supportive of university students and university programs - particularly those students and programs that will help to strengthen the development of nuclear-related fields. The program also supported the stability of the nuclear infrastructure and developed research partnerships that are helping to enlarge the national nuclear science technology base. In this fellowship program, the U.S. Department of Energy sought master's degree students in nuclear, mechanical, or chemical engineering, engineering/applied physics, physics, chemistry, radiochemistry, or fields of science and engineering applicable to the AFCI/Gen IV/GNEP missions in order to meet future U.S. nuclear program needs. The fellowship program identified candidates and selected full time students of high-caliber who were taking nuclear courses as part of their degree programs. The DOE Academic Program Managers encouraged fellows to pursue summer internships at national laboratories and supported the students with appropriate information so that both the fellows and the nation's nuclear energy objectives were successful.

  7. Review of monitoring instruments for transuranics in fuel fabrication and reprocessing plants. A progress report to the physical and technological programs, Division of Biomedical and Environmental Research, U.S. Energy Research and Development Administration

    International Nuclear Information System (INIS)

    Kordas, J.F.; Phelps, P.L.

    1977-01-01

    A comprehensive review of the monitoring instruments for transuranic elements released from nuclear fuel fabrication and reprocessing plants has been compiled. The extent of routine operational releases has been reviewed for the light water reactor (LWR) fuel cycle (including plutonium recycle), the breeder reactor fuel cycle, and the high-temperature gas cooled reactor (HTGR) fuel cycle. The stack monitoring instrumentation that is presently in use at the various fabrication and reprocessing plants around the country is examined. Sampling difficulties including the inlet-probe arrangement and the effectiveness of the entire sampling system are discussed, as are the measurement problems for alpha-emitting, long-lived, transuranic aerosols, 129 I, 106 Ru, and tritium oxide. The potential problems in the HTGR fuel cycle such as the measurement of releases of alpha-emitting aerosols and of gaseous releases of 220 Rn and 14 C are also considered. Monitoring requirements range from the detection of low-level, routine releases to high-level accidental releases. Both first and second kinds of detection errors are considered in a discussion of adequate detection limits. The presently deployed monitors are critically examined in this light and the drawbacks and limitations of each are noted. Prototype instrumentation is studied, including Argonne's mechanical separation technique, Battelle's mass separation by surface ionization method, and in particular, LLL's transuranic aerosol measurement system. The potentials, sensitivities, advantages, and limitations of each system are enumerated. The additional potential uses of the LLL system are also discussed

  8. Development of cutting device for irradiated fuel rod

    International Nuclear Information System (INIS)

    Lee, E. P.; Jun, Y. B.; Hong, K. P.; Min, D. K.; Lee, H. K.; Su, H. S.; Kim, K. S.; Kwon, H. M.; Joo, Y. S.; Yoo, K. S.; Joo, J. S.; Kim, E. K.

    2004-01-01

    Post Irradiation Examination(PIE) on irradiated fuel rods is essential for the evaluation of integrity and irradiation performance of fuel rods of commercial reactor fuel. For PIE, fuel rods should be cut very precisely. The cutting positions selected from NDT data are very important for further destructive examination and analysis. A fuel rod cutting device was developed witch can cut fuel rods longitudinal very precisely and can also cut the fuels into the same length rod cuts repeatedly. It is also easy to remove the fuel cutting powder after cutting works and it can extend the life time of cutting device and lower the contamination level of hot cell

  9. Development of advanced spent fuel management process

    International Nuclear Information System (INIS)

    Shin, Young Joon; Cho, S. H.; You, G. S.

    2001-04-01

    Currently, the economic advantage of any known approach to the back end fuel cycle of a nuclear power reactor has not been well established. Thus the long term storage of the spent fuel in a safe manner is one of the important issues to be resolved in countries where the nuclear power has a relatively heavy weight in power production of that country. At KAERI, as a solution to this particular issue midterm storage of the spent fuel, an alternative approach has been developed. This approach includes the decladding and pulverization process of the spent PWR fuel rod, the reducing process from the uranium oxide to a metallic uranium powder using Li metal in a LiCl salt, the continuous casting process of the reduced metal, and the recovery process of Li from mixed salts by the electrolysis. We conducted the laboratory scale tests of each processes for the technical feasibility and determination for the operational conditions for this approach. Also, we performed the theoretical safety analysis and conducted integral tests for the equipment integration through the Mock-up facility with non-radioactive samples. There were no major issues in the approach, however, material incompatibility of the alkaline metal and oxide in a salt at a high temperature and the reactor that contains the salt became a show stopper of the process. Also the difficulty of the clear separation of the salt with metals reduced from the oxide became a major issue

  10. Hydrogen Fuel Cell development in Columbia (SC)

    Energy Technology Data Exchange (ETDEWEB)

    Reifsnider, Kenneth [Univ. of South Carolina, Columbia, SC (United States); Chen, Fanglin [Univ. of South Carolina, Columbia, SC (United States); Popov, Branko [Univ. of South Carolina, Columbia, SC (United States); Chao, Yuh [Univ. of South Carolina, Columbia, SC (United States); Xue, Xingjian [Univ. of South Carolina, Columbia, SC (United States)

    2012-09-15

    This is an update to the final report filed after the extension of this program to May of 2011. The activities of the present program contributed to the goals and objectives of the Fuel Cell element of the Hydrogen, Fuel Cells and Infrastructure Technologies Program of the Department of Energy through five sub-projects. Three of these projects have focused on PEM cells, addressing the creation of carbon-based metal-free catalysts, the development of durable seals, and an effort to understand contaminant adsorption/reaction/transport/performance relationships at low contaminant levels in PEM cells. Two programs addressed barriers in SOFCs; an effort to create a new symmetrical and direct hydrocarbon fuel SOFC designs with greatly increased durability, efficiency, and ease of manufacturing, and an effort to create a multiphysics engineering durability model based on electrochemical impedance spectroscopy interpretations that associate the micro-details of how a fuel cell is made and their history of (individual) use with specific prognosis for long term performance, resulting in attendant reductions in design, manufacturing, and maintenance costs and increases in reliability and durability.

  11. An update of ERC's carbonate fuel cell development program

    International Nuclear Information System (INIS)

    Bernard, R.; Doyon, J.; Paetsch, L.; Patel, P.; Skok, A.; Yuh, C.; Steinfeld, G.; O'Shea, T.

    1992-01-01

    ERC has made significant accomplishments in stack height scale-up, resolved issues relevant to attainment of a long useful life for the carbonate fuel cell, and progressed towards addressing organizational and financial aspects of power plant demonstration

  12. Highlights of 50 years of nuclear fuel development

    International Nuclear Information System (INIS)

    Simnad, M.T.

    1989-01-01

    The development of nuclear fuels since the discovery of nuclear fission is briefly surveyed in this paper. The fabrication of the uranium fuel for the first nuclear pile, CP-1, is described. The research and development studies and fabrication of the different types of nuclear fuels for the variety of research and power reactors are reviewed. The important factors involved to achieve low fuel-cycle costs and reliable performance in the fuel elements are discussed in the historical context. 10 refs

  13. Highlights of 50 years of nuclear fuels developments

    International Nuclear Information System (INIS)

    Simnad, M.T.

    1989-01-01

    The development of nuclear fuels since the discovery of nuclear fission is briefly surveyed in this paper. The fabrication of the uranium fuel for the first nuclear pile, CP-1, is described. The research and development studies and fabrication of the different types of nuclear fuels for the variety of research and power reactors are reviewed. The important factors involved to achieve low fuel cycle costs and reliable performance in the fuel elements are discussed in the historical context

  14. Fuel cycle programs. Quarterly progress report, January-March 1980

    International Nuclear Information System (INIS)

    Steindler, M.J.; Bates, J.K.; Brock, R.E.

    1981-02-01

    Studies were continued on the encapsulation of radioactive waste in metal. The possibility of interactions between simulated waste glass and lead in air and argon atmospheres was explored. No reactions were observed at 350 and 500 0 C; however, a potential reaction was indicated at 685 0 C in air. A comparative economic analysis of four waste solidification processes subcontracted to TERA Corporation was completed. Within the uncertainties of the cost estimates, the cost of the four processes (two yielding metal-matrix encapsulated waste forms and two yielding glass monoliths) was about the same, with the major cost component being the cost of the building structures. Results of ten drop-weight brittle-fracture tests with Pyrex cylinders indicated that, for the same impact energy density, threefold variations in either drop velocity or cylinder size had no significant effect on the principal parameters of the test model. The kinetics of fracture is being studied by means of high-speed motion pictures. Tests indicated that evaporation losses of solutions collected during infiltration experiments were too small to significantly affect the experimental results. Studies were initiated to determine the cause of the discrepancy between the experimental Cs + velocity in kaolinite columns during 0.1M NaHCO 3 flow and the velocity calculated from the partitioning coefficient. A mathematical model was developed in an effort to explain why the experimentally determined isotherm for IO 3 sorption on Fe 2 O 3 differed from the Langmuir isotherm. Developmental studies of the neutron activation method and the use of tracers for determining very low leach rates were continued. Other studies showed that hydrothermal leaching of glass produced a large increase in surface area

  15. Review of fuel element development for nuclear rocket engines

    International Nuclear Information System (INIS)

    Taub, J.M.

    1975-06-01

    The Los Alamos Scientific Laboratory (LASL) entered the nuclear propulsion field in 1955 and began work on all aspects of a nuclear propulsion program involving uranium-loaded graphite fuels, hydrogen propellant, and a target exhaust temperature of approximately 2500 0 C. A very extensive uranium-loaded graphite fuel element technology evolved from the program. Selection and composition of raw materials for the extrusion mix had to be coupled with heat treatment studies to give optimum element properties. The highly enriched uranium in the element was incorporated as UO 2 , pyrocarbon-coated UC 2 , or solid solution UC . ZrC particles. An extensive development program resulted in successful NbC or ZrC coatings on elements to withstand hydrogen corrosion at elevated temperatures. Hot gas, thermal shock, thermal stress, and NDT evaluation procedures were developed to monitor progress in preparation of elements with optimum properties. Final evaluation was made in reactor tests at NRDS. Aerojet-General, Westinghouse Astronuclear Laboratory, and the Oak Ridge Y-12 Plant of Union Carbide Nuclear Company entered the program in the early 1960's, and their activities paralleled those of LASL in fuel element development. (U.S.)

  16. Nuclear fuels and development of nuclear fuel elements

    International Nuclear Information System (INIS)

    Sundaram, C.V.; Mannan, S.L.

    1989-01-01

    Safe, reliable and economic operation of nuclear fission reactors, the source of nuclear power at present, requires judicious choice, careful preparation and specialised fabrication procedures for fuels and fuel element structural materials. These aspects of nuclear fuels (uranium, plutonium and their oxides and carbides), fuel element technology and structural materials (aluminium, zircaloy, stainless steel etc.) are discussed with particular reference to research and power reactors in India, e.g. the DHRUVA research reactor at BARC, Trombay, the pressurised heavy water reactors (PHWR) at Rajasthan and Kalpakkam, and the Fast Breeder Test Reactor (FBTR) at Kalpakkam. Other reactors like the gas-cooled reactors operating in UK are also mentioned. Because of the limited uranium resources, India has opted for a three-stage nuclear power programme aimed at the ultimate utilization of her abundant thorium resources. The first phase consists of natural uranium dioxide-fuelled, heavy water-moderated and cooled PHWR. The second phase was initiated with the attainment of criticality in the FBTR at Kalpakkam. Fast Breeder Reactors (FBR) utilize the plutonium and uranium by-products of phase 1. Moreover, FBR can convert thorium into fissile 233 U. They produce more fuel than is consumed - hence, the name breeders. The fuel parameters of some of the operating or proposed fast reactors in the world are compared. FBTR is unique in the choice of mixed carbides of plutonium and uranium as fuel. Factors affecting the fuel element performance and life in various reactors e.g. hydriding of zircaloys, fuel pellet-cladding interaction etc. in PHWR and void swelling; irradiation creep and helium embrittlement of fuel element structural materials in FBR are discussed along with measures to overcome some of these problems. (author). 15 refs., 9 tabs., 23 figs

  17. Progress report on fast breeder reactor development in Japan, July - September 1977

    International Nuclear Information System (INIS)

    1978-06-01

    As for the experimental fast breeder reactor ''Joyo'', the low power performance tests have been continued, and the measurements of reactor noise, the reactivity of fuel assemblies, power distribution and Na-void effect have been made. Efforts have been exerted to develop the required maintenance equipments, to manufacture the transfer rotor maintenance facilities, and to construct the spent fuel storing and cooling facilities. The analysis and calculation of the core characteristics have been in progress. The design work on the prototype fast breeder reactor ''Monju'' has been continued, and the development of the computer codes for the design has progressed. Informations have been gathered regarding the technological developments of LMFBRs overseas. The surveys on the site for ''Monju'' have been carried out. The design and research works on the demonstration reactor were started, and the general design factors such as the steam condition and the plant layout have been studied. As for the research and development of reactor physics, structural components, instrumentation and control, sodium technology, fuel materials, structural materials, safety, and steam generators, the progresses are reported in detail. High performance neutron detectors for nuclear instrumentation have been under development, and the tagging gas method for fuel failure detection and location system has been tested. (Kako, I.)

  18. Chemical Engineering Division fuel cycle programs. Quarterly progress report, July-September 1978

    Energy Technology Data Exchange (ETDEWEB)

    Steindler, M.J.; Ader, M.; Barletta, R.E.

    1980-01-01

    Fuel cycle work included hydraulic performance and extraction efficiency of eight-stage centrifugal contactors, flowsheet for the Aralex process, Ru and Zr extraction in a miniature centrifugal contactor, study of Zr aging in the organic phase and its effect on Zr extraction and hydraulic testing of the 9-cm-ID contactor. Work for predicting accident consequences in LWR fuel processing covered the relation between energy input (to subdivide a solid) and the modes of particle size frequency distribution. In the pyrochemical and dry processing program corrosion-testing materials for containment vessels and equipment for studying carbide reactions in bismuth is under way. Analytical studies have been made of salt-transport processes; efforts to spin tungsten crucibles 13 cm dia continue, and other information on tungsten fabrication is being assembled; the process steps of the chloride volatility process have been demonstrated and the thoria powder product used to produce oxide pellets; solubility of UO/sub 2/, PuO/sub 2/, and fission products in molten alkali nitrates is being investigated; work was continued on reprocessing actinide oxides by extracting the actinides into ammonium chloroaluminate from bismuth; the preparation of thorium-uranium carbide from the oxide is being studied as a means of improving the oxide reactivity; studies are in progress on producing uranium metal and decontaminated ThO/sub 2/ by the reaction of (Th,U)O/sub 2/ solid solution in molten salts containing ThCl/sub 4/ and thorium metal chips. In the molten tin process, no basic thermodynamic or kinetic factors have been found that may limit process development.

  19. Development of spent fuel remote handling technology

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Ji Sup; Park, B S; Park, Y S; Oh, S C; Kim, S H; Cho, M W; Hong, D H

    1997-12-01

    Since the nation`s policy on spent fuel management is not finalized, the technical items commonly required for safe management and recycling of spent fuel - remote technologies of transportation, inspection, maintenance, and disassembly of spent fuel - are selected and pursued. In this regards, the following R and D activities are carried out : collision free transportation of spent fuel assembly, mechanical disassembly of spent nuclear fuel and graphical simulation of fuel handling / disassembly process. (author). 36 refs., 16 tabs., 77 figs

  20. Development of spent fuel remote handling technology

    International Nuclear Information System (INIS)

    Yoon, Ji Sup; Park, B. S.; Park, Y. S.; Oh, S. C.; Kim, S. H.; Cho, M. W.; Hong, D. H.

    1997-12-01

    Since the nation's policy on spent fuel management is not finalized, the technical items commonly required for safe management and recycling of spent fuel - remote technologies of transportation, inspection, maintenance, and disassembly of spent fuel - are selected and pursued. In this regards, the following R and D activities are carried out : collision free transportation of spent fuel assembly, mechanical disassembly of spent nuclear fuel and graphical simulation of fuel handling / disassembly process. (author). 36 refs., 16 tabs., 77 figs

  1. Progress in Understanding of Fuel-Cladding Chemical interaction in Metal Fuel

    International Nuclear Information System (INIS)

    Inagaki, Okenta; Nakamura, Kinya; Ogata, Takanari

    2013-01-01

    Conclusion: Representative phases formed in FCCI were identified: • The reaction between lanthanide elements and cladding; • The reaction between U-PU-Zr and cladding (Fe). Characteristics of the wastage layer were clarified: • Time and temperature dependency of the growth ratio of the wastage layer formed by lanthanide elements; • Threshold temperature of the liquid phase formation in the reaction between U-Pu-Zr and Fe. These results are used: - as a basis for the FCCI modeling; - as a reference data in post-irradiation examination of irradiated metallic fuels

  2. Research reactor fuel development at AECL

    International Nuclear Information System (INIS)

    Sears, D.F.; Wang, N.

    2000-09-01

    This paper reviews recent U 3 Si 2 and U-Mo dispersion fuel development activities at AECL. The scope of work includes fabrication development, irradiation testing, post-irradiation examination and performance qualification. U-Mo alloys with a variety of compositions, ranging from 6 to 10 wt % Mo, have been fabricated with high purity and homogeneity in the product. The alloys and powders were characterized using optical and scanning electron microscopy, chemical analysis, and X-ray diffraction and neutron diffraction analysis. U-Mo powder samples have been supplied to the Argonne National Laboratory for irradiation testing in the ATR reactor. Low-enriched uranium fuel elements containing U-7 wt % Mo and U-10 wt % Mo with loadings up to 4.5 gU/cm 3 have been fabricated at CRL for irradiation testing in the NRU reactor. The U-Mo fuel elements will be tested in NRU at linear powers up to 145 kW/m, and to 85 atom % 235 U burnup. (author)

  3. Prospects for development of fuel cells

    Directory of Open Access Journals (Sweden)

    В. М. Шабер

    2017-10-01

    Full Text Available The article is devoted to the solution of a complex of problems that arise in small and medium-scale treatment complexes, gas production plants and small and medium-capacity power plants associated with the processing of crude methane and the possibility of reducing the greenhouse effect.The economic feasibility of the development of fuel cells (FC on raw biomethane was demonstrated by the authors in previous publications.The specificity of the solution of problems is focused on small and medium-scale treatment complexes, gas production plants and small and medium power plants.The aim of the study is to show the possibility of solving a multicomponent task of developing fuel cells, including the experimental determination of the actual use of sodium formate as a reducing agent for the production of electricity in a fuel cell (FC.Results are the following: the possibility of solving the issues of reducing greenhouse gas emissions into the atmosphere during processing of waste products of human vital activity is proved. A method for converting methane and carbon dioxide emissions into useful products is shown.

  4. Research reactor fuel development at AECL

    International Nuclear Information System (INIS)

    Sears, D.F.; Wang, N.

    2000-01-01

    This paper reviews recent U 3 Si 2 and U-Mo dispersion fuel development activities at AECL. The scope of work includes fabrication development, irradiation testing, postirradiation examination and performance qualification. U-Mo alloys with a variety of compositions, ranging from 6 to 10 wt % Mo, have been fabricated with high purity and homogeneity in the product. The alloys and powders were characterized using optical and scanning electron microscopy, chemical analysis, and X-ray diffraction and neutron diffraction analysis. U-Mo powder samples have been supplied to the Argonne National Laboratory for irradiation testing in the ATR reactor. Low-enriched uranium fuel elements containing U-7 wt % Mo and U-10 wt % Mo with loadings up to 4.5 gU/cm 3 have been fabricated at CRL for irradiation testing in the NRU reactor. The U-Mo fuel elements will be tested in NRU at linear powers up to 145 kW/m, and to 85 atom % 235 U burnup. (author)

  5. Progress and Overview on Neutronics Modelling Development in RTP

    International Nuclear Information System (INIS)

    Mohamad Hairie Rabir; Muhammad Rawi Mohamed Zin; Julia Abdul Karim

    2016-01-01

    Reactor calculation and simulation are significantly important to ensure safety and better utilization of a research reactor. The Malaysian PUSPATI TRIGA Reactor (RTP) achieved initial criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes. Since early 90s, neutronics modelling were used as part of its routine in-core fuel management activities. The are several computer codes have been used in RTP since then, based on 1D neutron diffusion, 2D neutron diffusion and 3D Monte Carlo neutron transport method. This paper describes current progress and overview on neutronics modelling development in RTP. Several important parameters were analysed such as k_e_f_f, reactivity, neutron flux, power distribution, B_e_f_f, and fission product build-up for the latest core configuration. The developed core neutronics model was validated by means of comparison with experimental and measurement data. Along with the RTP core model, the calculation procedure also developed to establish better prediction capability of RTP behaviour. (author)

  6. Progress in integrated energy-economy-environment model system development

    International Nuclear Information System (INIS)

    Yasukawa, Shigeru; Mankin, Shuichi; Sato, Osamu; Tadokoro, Yoshihiro; Nakano, Yasuyuki; Nagano, Takao

    1987-11-01

    The Integrated Energy-Economy-Environment Model System has been developed for providing analytical tools for the system analysis and technology assessments in the field of nuclear research and development. This model system consists of the following four model groups. The first model block installs 5 models and can serve to analyze and generate long-term scenarios on economy-energy-environment evolution. The second model block installs 2 models and can serve to analyze the structural transition phenomena in energy-economy-environment interactions. The third model block installs 2 models and can handle power reactor installation strategy problem and long-term fuel cycle analysis. The fourth model block installs 5 models and codes and can treats cost-benefit-risk analysis and assessments. This report describes mainly the progress and the outlines of application of the model system in these years after the first report on the research and development of the model system (JAERI-M 84 - 139). (author)

  7. Thirty years of transport package development for spent fuels

    International Nuclear Information System (INIS)

    Cory, A.R.

    2005-01-01

    By June 2005, when shipments of spent fuel for reprocessing from Germany are concluded, BNFL flask types will have been responsible for transporting more than 2000 tonnes of heavy metal in Europe in the form of spent fuel. Several thousand more tonnes of spent fuel have been transported by sea from Japan over the last thirty years. The design of spent fuel packages has not stood still for that time. In order to anticipate the changing needs of the nuclear power generation industry, advances have been made both in package design and analysis. Thirty years ago spent fuel burnup and initial enrichment were considerably lower, which was reflected in the different demands placed on the shielding design of packages, and in the design of the internal basket to separate the fuel assemblies. Technical development of both 'wet' (water-filled cavity) and 'dry' packages has progressed in parallel, and the relative merits and peculiarities of each type is explored. BNFL has considerable experience in the operation of both types, and is well placed to comment on practical and functional issues associated with both types. While there have been certain evolutionary changes affecting package design, there have also been more significant changes in the Design Safety Case. These have sometimes been necessary to meet changes in IAEA Regulations, or the challenges posed by the regulators themselves. In other cases advantage has been taken of improvements in analytical techniques to demonstrate increased margins of operational safety. Where possible these margins have also been increased by other means, such as taking advantage of commercial trends to reduce package thermal loads. A key factor over the last thirty years has been the increasing influence of the Regulating Authorities and the development of the IAEA Regulations. The various Competent Authorities now tend to have a higher proportion of technical experts, often recruited from the nuclear industry, and are thus more able to

  8. Subchannel analysis code development for CANDU fuel channel

    International Nuclear Information System (INIS)

    Park, J. H.; Suk, H. C.; Jun, J. S.; Oh, D. J.; Hwang, D. H.; Yoo, Y. J.

    1998-07-01

    Since there are several subchannel codes such as COBRA and TORC codes for a PWR fuel channel but not for a CANDU fuel channel in our country, the subchannel analysis code for a CANDU fuel channel was developed for the prediction of flow conditions on the subchannels, for the accurate assessment of the thermal margin, the effect of appendages, and radial/axial power profile of fuel bundles on flow conditions and CHF and so on. In order to develop the subchannel analysis code for a CANDU fuel channel, subchannel analysis methodology and its applicability/pertinence for a fuel channel were reviewed from the CANDU fuel channel point of view. Several thermalhydraulic and numerical models for the subchannel analysis on a CANDU fuel channel were developed. The experimental data of the CANDU fuel channel were collected, analyzed and used for validation of a subchannel analysis code developed in this work. (author). 11 refs., 3 tabs., 50 figs

  9. Development of Coated Particle Fuel Technology

    International Nuclear Information System (INIS)

    Cho, Moon Sung; Kim, B. G.; Kim, Y. K.

    2009-04-01

    UO 2 kernel fabrication technology was developed at the lab sacle(20∼30g-UO 2 /batch). The GSP technique, modified method of sol-gel process, was used in the preparation of spherical ADU gel particle and these particles were converted to UO 3 and UO 2 phases in calcination furnace and sintering furnace respectively. Based on the process variables optimized using simulant kernels in 1-2 inch beds, SiC TRISO-coated particles were fabricated using UO 2 kernel. Power densities of TRISO coated particle fuels and gamma heat of the tubes are calculated as functions of vertical location of the fuel specimen in the irradiation holes by using core physics codes, MCNP and Helios. A finite model was developed for the calculations of temperatures and stresses of the specimen and the irradiation tubes. Dimensions of the test tubes are determined based on the temperatures and stresses as well as the gamma heat generated at the given condition. 9 modules of the COPA code (MECH, FAIL, TEMTR, TEMBL, TEMPEB, FPREL, MPRO, BURN, ABAQ), the MECH, FAIL, TEMTR, TEMBL, TEMPEB, and FPREL were developed. The COPA-FPREL was verified through IAEA CRP-6 accident benchmarking problems. KAERI participated in the round robin test of IAEA CRP-6 program to characterize the diameter, sphericity, coating thickness, density and anisotropy of coated particles provided by Korea, USA and South Africa. The inspection and test plan describing specifications and inspection method of coated particles was developed to confirm the quality standard of coated particles. The quality inspection instructions were developed for the inspection of coated particles by particle size analyzer, density inspection of coating layers by density gradient column, coating thickness inspection by X-ray, and inspection of optical anistropy factor of PyC layer. The quality control system for the TRISO-coated particle fuel was derived based on the status of quality control systems of other countries

  10. CANDU RU fuel manufacturing basic technology development and advanced fuel verification tests

    International Nuclear Information System (INIS)

    Chung, Chang Hwan; Chang, S.K.; Hong, S.D.

    1999-04-01

    A PHWR advanced fuel named the CANFLEX fuel has been developed through a KAERI/AECL joint Program. The KAERI made fuel bundle was tested at the KAERI Hot Test Loop for the performance verification of the bundle design. The major test activities were the fuel bundle cross-flow test, the endurance fretting/vibration test, the freon CHF test, and the fuel bundle heat-up test. KAERI also has developing a more advanced PHWR fuel, the CANFLEX-RU fuel, using recovered uranium to extend fuel burn-up in the CANDU reactors. For the purpose of proving safety of the RU handling techniques and appraising feasibility of the CANFLEX-RU fuel fabrication in near future, a physical, chemical and radiological characterization of the RU powder and pellets was performed. (author). 54 refs., 46 tabs., 62 figs

  11. CANDU RU fuel manufacturing basic technology development and advanced fuel verification tests

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Chang Hwan; Chang, S.K.; Hong, S.D. [and others

    1999-04-01

    A PHWR advanced fuel named the CANFLEX fuel has been developed through a KAERI/AECL joint Program. The KAERI made fuel bundle was tested at the KAERI Hot Test Loop for the performance verification of the bundle design. The major test activities were the fuel bundle cross-flow test, the endurance fretting/vibration test, the freon CHF test, and the fuel bundle heat-up test. KAERI also has developing a more advanced PHWR fuel, the CANFLEX-RU fuel, using recovered uranium to extend fuel burn-up in the CANDU reactors. For the purpose of proving safety of the RU handling techniques and appraising feasibility of the CANFLEX-RU fuel fabrication in near future, a physical, chemical and radiological characterization of the RU powder and pellets was performed. (author). 54 refs., 46 tabs., 62 figs.

  12. Development and engineering plan for graphite spent fuels conditioning program

    International Nuclear Information System (INIS)

    Bendixsen, C.L.; Fillmore, D.L.; Kirkham, R.J.; Lord, D.L.; Phillips, M.B.; Pinto, A.P.; Staiger, M.D.

    1993-09-01

    Irradiated (or spent) graphite fuel stored at the Idaho Chemical Processing Plant (ICPP) includes Fort St. Vrain (FSV) reactor and Peach Bottom reactor spent fuels. Conditioning and disposal of spent graphite fuels presently includes three broad alternatives: (1) direct disposal with minimum fuel packaging or conditioning, (2) mechanical disassembly of spent fuel into high-level waste and low-level waste portions to minimize geologic repository requirements, and (3) waste-volume reduction via burning of bulk graphite and other spent fuel chemical processing of the spent fuel. A multi-year program for the engineering development and demonstration of conditioning processes is described. Program costs, schedules, and facility requirements are estimated

  13. Status of SFR Metal Fuel Development

    International Nuclear Information System (INIS)

    Lee, Chan Bock; Lee, Byoung Oon; Kim, Ki Hwan; Kim, Sung Ho

    2013-01-01

    Conclusion: • Metal fuel recycling in SFR: - Enhanced utilization of uranium resource; - Efficient transmutation of minor actinides; - Inherent passive reactor safety; - Proliferation resistance with pyro-electrochemical fuel recycling. • Demonstration of technical feasibility of recycling TRU metal fuel by 2020: - Remote fuel fabrication; - Irradiation performance up to high burnup

  14. Reaction progress pathways for glass and spent fuel under unsaturated conditions

    International Nuclear Information System (INIS)

    Bates, J.; Finn, P.; Bourcier, W.; Stout, R.

    1994-10-01

    The source term for the release of radionuclides from a nuclear waste repository is the waste form. In order to assess the performance of the repository and the engineered barrier system (EBS) compared to regulations established by the Nuclear Regulatory Commission and the Environmental Protection Agency it is necessary (1) to use available data to place bounding limits on release rates from the EBS, and (2) to develop a mechanistic predictive model of the radionuclide release and validate the model against tests done under a variety of different potential reaction conditions. The problem with (1) is that there is little experience to use when evaluating waste form reaction under unsaturated conditions such that errors in applying expert judgment to the problem may be significant. The second approach, to test and model the waste form reaction, is a more defensible means of providing input to the prediction of radionuclide release. In this approach, information related to the source term has a technical basis and provides a starting point to make reasonable assumptions for long-term behavior. Key aspects of this approach are an understanding of the reaction progress mechanism and the ability to model the tests using a geochemical code such as EQ3/6. Current knowledge of glass, UO 2 , and spent fuel reactions under different conditions are described below

  15. MOX fuel development: Experience in Argentina

    International Nuclear Information System (INIS)

    Marchi, D.E.; Adelfang, P.; Menghini, J.E.

    1999-01-01

    Since 1973, when a laboratory conceived for the safe manipulation of a few hundred grams of plutonium was built, the CNEA (Argentinean Atomic Energy Commission) has been involved in the small-scale development of MOX fuel technology. The plutonium laboratory consists in a glove box facility (α Facility) featuring the necessary equipment to prepare MOX fuel rods for experimental irradiations and to carry out studies on preparative processes development and chemical and physical characterization. The irradiation of the first prototypes of (U,Pu)O 2 fuels fabricated in Argentina began in 1986. These experiments were carried out in the HFR (High Flux Reactor)- Petten , Holland. The rods were prepared and controlled in the CNEA's a Facility. The post-irradiation examinations (PIE) were performed in the KFK (Kernforschungszentrum Karlsruhe), Germany and the JRC (Joint Research Center), Petten. In the period 1991-1995, the development of new laboratory methods of co-conversion of uranium and plutonium were carried out: reverse strike co-precipitation of ADU-Pu(OH) 4 and direct denitration using microwaves. The reverse strike process produced pellets with a high sintered density, excellent micro-homogeneity and good solubility in nitric acid. Liquid wastes showed a very low content of actinides and the process is easy to operate in a glove box environment. The microwave direct denitration was optimized with uranium alone and the conditions to obtain high density pellets, with a good microstructure, without using a milling step, have been developed. At present, new experiments are being carried out to improve the reverse strike co-precipitation process and direct microwave denitration. A new glove box is being installed at the plutonium laboratory, this glove box has process equipment designed to recover scrap from previous fabrication campaigns, and to co-convert mixed U-Pu solutions by direct microwave denitration. (author)

  16. Development of indigenous irradiator - current progress and challenges

    International Nuclear Information System (INIS)

    Anwar A Rahman; Mohd Arif Hamzah; Muhd Nor Atan; Aznor Hassan; Fadil Ismail; Julia A Karim; Rosli Darmawan

    2009-01-01

    The development of indigenous irradiator is one of Prototype Development Center main project to support Nuclear Malaysia services. Three (3) projects have been identified and currently the status is in final stage of design. There are some issues and challenges encountered, which delayed the project progress. The paper will discuss the current progress of development and challenges faced in designing the irradiator. (Author)

  17. Development of CH{sub 3}OH fueled PEMFC power plants for hybrid transit buses

    Energy Technology Data Exchange (ETDEWEB)

    Baumert, R; Cooper, R; Feasey, G [DBB Fuel Cell Engines Corp., Poway, CA (United States)

    1999-12-31

    An overview of the methanol fuel cell power system was provided, identifying improved efficiency and reduced emissions as the principal advantages. Four critical tasks regarding on-board fuel processing were described: (1) efficient methanol conversion (steam reforming), (2) effective reformate purification (selective catalytic oxidation), (3) optimized heat integration, and (4) rapid response to transients. A description of a 100 kW PEM fuel cell bus engine package was also presented. As far as a development time table is concerned, the DBB Fuel Cell Engines Corp. of Poway California has completed two methanol fueled PEMFC power plants, fabrication of the initial 100 kW PEMFC engine is in progress and scheduled for delivery by 1998. The two methanol fueled commercial products which are in the planning stages are the 100 and 200 kW class FCPS for hybrid and non-hybrid buses and other applications. tabs., figs.

  18. MOX fuel design and development consideration

    International Nuclear Information System (INIS)

    Yamate, K.; Abeta, S.; Suzuki, K.; Doi, S.

    1997-01-01

    Pu thermal utilization in Japan will be realized in several plants in late 1990's, and will be expanded gradually. For this target, adequacy of methods for MOX fuel design, nuclear design, and safety analysis has been evaluated by the committee of competent authorities organized by government in advance of the licensing application. There is no big difference of physical properties and irradiation behaviors between MOX fuel and UO 2 fuel, because Pu content of MOX fuel for Pu thermal utilization is low. The fuel design code for UO 2 fuel will be applied with some modifications, taking into account of characteristic of MOX fuel. For nuclear design, new code system is to be applied to treat the heterogeneity in MOX fuel assembly and the neutron spectrum interaction with UO 2 fuel more accurately. For 1/3 MOX fueled core in three loop plant, it was confirmed that the fuel rod mechanical design could meet the design criteria, with slight reduction of initial back-fitting pressure, and with appropriate fuel loading patterns in the core to match power with UO 2 fuel. With the increase of MOX fuel fraction in the core, control rod worth and boron worth decrease. Compensating the decrease by adding control rod and utilizing enriched B-10 in safety injection system, 100% MOX fueled core could be possible. Up to 1/3 MOX fueled core in three loop plant, no such modifications of the plant is necessary. The fraction of MOX fuel in PWR is designed to less than 1/3 in the present program. In order to improve Pu thermal utilization in future, various R and D program on fuel design and nuclear design are being performed, such as the irradiation program of MOX fuel manufactured through new process to the extent of high burnup. (author). 8 refs, 9 figs, 2 tabs

  19. Development of advanced fuels in the Swiss Federal Institute for Reactor Research (EIR)

    International Nuclear Information System (INIS)

    Stratton, R.W.

    1984-02-01

    The work of the project Fuel Development over the three years 1981-83 is reported. In this period virtually all of the development, demonstration and preparatory work for the fabrication of mixed carbide sphere-pac fuel pins for the FFTF experiment was completed. As well as describing the background to and the progress of the work, selected details are given of some of the results achieved in all areas of activity - fuel fabrication, pin manufacture, quality assurance, pin behaviour and modelling. Names of all principle contributers to each activity are given and in addition to references the complete list of publications over the period is provided. (Auth.)

  20. Automotive Stirling engine development program. [fuel economy assessment

    Science.gov (United States)

    Kitzner, E. W.

    1978-01-01

    The Ford/DOE automotive Stirling engine development program is directed towards establishing the technological and developmental base that would enable a decision on whether an engineering program should be directed at Stirling engine production. The fuel economy assessment aims to achieve, with a high degree of confidence, the ERDA proposal estimate of 20.6 MPG (gasoline) for a 4500 lb 1WC Stirling engine passenger car. The current M-H fuel economy projection for the 170 HP Stirling engine is 15.7 MPG. The confidence level for this projection is 32%. A confidence level of 29% is projected for a 22.1 MPG estimate. If all of the planned analyses and test work is accomplished at the end of the one year effort, and the projected improvements are substantiated, the confidence levels would rise to 59% for the 20.6 MPG projection and 54% for the 22.1 MPG projection. Progress achieved thus far during the fuel economy assessment is discussed.

  1. Development of Spent Fuel Examination Technology

    International Nuclear Information System (INIS)

    Kim, Ho Dong; Park, K. J.; Shin, H. S.

    2007-04-01

    For the official operation of ACPF Facility Attachment based on facility declared DIQ was issued by IAEA and officialized upon ROK government approval. This procedure gives an essential ground to negotiate Joint Determination between governments of ROK and US. For ACPF process material accountability a neutron coincidence counting system was developed and calibrated with Cf-252 source. Its performance test demonstrated that over-all counting efficiency was about 21% with random error, 1.5% against calibration source, which found to be satisfactory to the expected design specification. A calibration curve derived by MCNP code with relationship between ASNC doublet counts vs. neutron activity of Cm-244 showed calibration constant to be 2.78x10E5 counts/s.g which would be used for initial ACP hot operation test. Nuclear material transportation and temporary storage system was established for active demonstration of advanced spent fuel management process line and would be directly applied to the effective management of wastes arising from active demonstration and would later contribute as a base data to development of inter hot-cell movement system in pyro-processing line. In addition, an optimal spent fuel for the ACP demonstration was selected and a computer code was developed as a tool to estimate the expected source term at each key measurement point of ACP

  2. Development of Spent Fuel Examination Technology

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ho Dong; Park, K. J.; Shin, H. S. (and others)

    2007-04-15

    For the official operation of ACPF Facility Attachment based on facility declared DIQ was issued by IAEA and officialized upon ROK government approval. This procedure gives an essential ground to negotiate Joint Determination between governments of ROK and US. For ACPF process material accountability a neutron coincidence counting system was developed and calibrated with Cf-252 source. Its performance test demonstrated that over-all counting efficiency was about 21% with random error, 1.5% against calibration source, which found to be satisfactory to the expected design specification. A calibration curve derived by MCNP code with relationship between ASNC doublet counts vs. neutron activity of Cm-244 showed calibration constant to be 2.78x10E5 counts/s.g which would be used for initial ACP hot operation test. Nuclear material transportation and temporary storage system was established for active demonstration of advanced spent fuel management process line and would be directly applied to the effective management of wastes arising from active demonstration and would later contribute as a base data to development of inter hot-cell movement system in pyro-processing line. In addition, an optimal spent fuel for the ACP demonstration was selected and a computer code was developed as a tool to estimate the expected source term at each key measurement point of ACP.

  3. Research and development of nitride fuel cycle technology in Japan

    International Nuclear Information System (INIS)

    Minato, Kazuo; Arai, Yasuo; Akabori, Mitsuo; Tamaki, Yoshihisa; Itoh, Kunihiro

    2004-01-01

    The research on the nitride fuel was started for an advanced fuel, (U, Pn)N, for fast reactors, and the research activities have been expanded to minor actinide bearing nitride fuels. The fuel fabrication, property measurements, irradiation tests and pyrochemical process experiments have been made. In 2002 a five-year-program named PROMINENT was started for the development of nitride fuel cycle technology within the framework of the Development of Innovative Nuclear Technologies by the Ministry of Education, Culture, Sports, Science and Technology of Japan. In the research program PROMINENT, property measurements, pyrochemical process and irradiation experiments needed for nitride fuel cycle technology are being made. (author)

  4. Status of research reactor fuel development in KAERI

    International Nuclear Information System (INIS)

    Kim, Chang-Kyu; Ryu, Woo-Seok; Park, Jong-Man; Lee, Don-Bae; Kim, Ki-Hwan; Kuk, Il-Hyun

    1996-01-01

    The development of uranium silicide dispersion fuel fabrication technology has been carried out in KAERI. LEU fuel bundle was prepared for irradiation test. In order to compare the performance of atomized and comminuted U 3 Si dispersed fuels, the bundle of two kinds of fuel elements were prepared. Irradiation test will be performed in the OR-hole of HANARO in the near future. U 3 Si 2 atomization technology has been improved by using ceramic crucible and nozzle. Irradiation test for atomized U 3 Si 2 plate type fuel will be carried out in cooperation with ANL by using HANARO in connection with RERTR advanced fuel development. (author)

  5. Strategies in development of advanced fuels for LMFBR

    International Nuclear Information System (INIS)

    Handa, Muneo

    1976-12-01

    Overseas strategies in development of advanced fuels for LMFBR are reviewed. Recent irradiation experiment and out-of-pile test data of the fuels are given in detail. The present status of development of oxide fueled LMFBR is also treated. (auth.)

  6. Development of on-site spent fuel transfer system designs

    International Nuclear Information System (INIS)

    Lambert, R.W.; Pennington, C.W.; Guerra, G.V.

    1993-01-01

    The Electric Power Research Institute (EPRI) of the United States has sponsored development of conceptual designs for accomplishing spent fuel transfer from spent fuel pools to casks and from one cask to another. Under an EPRI research contract, transnuclear has developed several concepts for spent fuel transfer systems. (J.P.N.)

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

    International Nuclear Information System (INIS)

    Snelgrove, J.L.; Hofman, G.L.; Hayes, S.L.; Meyer, M.K.

    2001-01-01

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

  8. Development of quality assurance methods for low enriched fuel assemblies

    International Nuclear Information System (INIS)

    Woolstenhulme, N.E.; Moore, G.A.; Perez, D.M.; Wachs, D.M.

    2010-01-01

    As the Reduced Enrichment for Research and Test Reactors (RERTR) fuel development program has furthered the technology of low enriched uranium fuels, much effort has been expended to specify requirements, perform appropriate inspections, and to qualify experimental fuel plates and assemblies for irradiation. A great deal of consideration has been given to generate examinations and criteria that are both applicable to the unique fuel types being developed and consistent with industry practices for inspecting plate-type reactor fuel. Recent developments in quality assurance (QA) methodologies have given a heightened confidence in satisfactory fuel plate performance. At the same time, recommendations are given to further develop a system suitable for the testing and acceptance of production fuel elements containing low enriched uranium fuels. (author)

  9. International development within the spent nuclear fuel cycle

    International Nuclear Information System (INIS)

    Aggeryd, I.; Broden, K.; Gelin, R.

    1990-06-01

    The report gives a survey of the newest international development of the fuel processing and the spent nuclear fuel cycle. The transmutation technology of long lived nuclides is discussed in more details. (K.A.E)

  10. Recent progress in development of radioisotope production

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Byung Mok [HANARO Center, Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-10-01

    The Korea multipurpose research reactor, HANARO(Hi-flux Advanced Neutron Application Reactor) is designed and constructed to obtain high density neutron flux (max. 5x10{sup 14} n/cm{sup 2}{center_dot}sec) with relatively low thermal output (30 MW) in order to utilize for various studies such as fuel and material test, radioisotope production, neutron activation analysis, neutron beam experiment, neutron transmutation doping, etc. HANARO has 32 vertical channels (3 in-core, 4 out-core, 25 reflector) and 7 horizontal channels. KAERI has constructed 4 concrete hot cells for production of Co-60, Ir-192, etc. and 6 lead hot cells for production of medical RIs(I-131, Mo-99, etc.). Other 11 lead hot cells will be completed by Feb. 1998 for production of Sm-153, Dy-165, Ho-166, etc. Clean room facilities were installed for production of radiopharmaceuticals. (author)

  11. Development of failed fuel detection and location system in sodium-cooled large reactor. Sampling method of failed fuels under the slit

    International Nuclear Information System (INIS)

    Aizawa, Kousuke; Fujita, Kaoru; Kamide, Hideki; Kasahara, Naoto

    2010-01-01

    A conceptual design study of Japan Sodium-cooled Fast Reactor (JSFR) is in progress as an issue of the 'Fast Reactor Cycle Technology Development (FaCT)' project in Japan. JSFR adopts a Selector-Valve mechanism for the failed fuel detection and location (FFDL) system. The Selector-Valve FFDL system identifies failed fuel subassemblies by sampling sodium from each fuel subassembly outlet and detecting fission product. One of the JSFR design features is employing an upper internal structure (UIS) with a radial slit, in which an arm of fuel handling machine can move and access the fuel assemblies under the UIS. Thus, JSFR cannot place sampling nozzles right above the fuel subassemblies located under the slit. In this study, the sampling method for indentifying under-slit failed fuel subassemblies has been demonstrated by water experiments. (author)

  12. Status and development of the thorium fuel cycle

    International Nuclear Information System (INIS)

    Yi Weijing; Wei Renjie

    2003-01-01

    A perspective view of the thorium fuel cycle is provided in this paper. The advantages and disadvantages of the thorium fuel cycle are given and the development of thorium fuel cycle in several types of reactors is introduced. The main difficulties in developing the thorium fuel cycle lie in the reprocessing and disposal of the waste and its economy, and the ways tried by foreign countries to solve the problems are presented in the paper

  13. Baseline gas turbine development program. Seventeenth quarterly progress report

    Energy Technology Data Exchange (ETDEWEB)

    Schmidt, F W; Wagner, C E

    1977-01-31

    Progress is reported for a program whose goals are to demonstrate an experimental upgraded gas turbine powered automobile which meets the 1978 Federal Emissions Standards, has significantly improved fuel economy, and is competitive in performance, reliability, and potential manufacturing cost with the conventional piston engine powered, compact-size American automobile. Initial running of the upgraded engine took place on July 13, 1976. The engine has proved to be mechanically sound, but has also been seriously deficient in power. Principal program effort has therefore been in the area of diagnostic testing and corrective development. To date, three upgraded engines have been assembled and run in the test cell. Engine 2 was installed in an upgraded vehicle and became operational on January 25, 1977. Special diagnostic instrumentation was installed on Engine 3 to evaluate the compressor, turbine, and hot engine leakage. It was determined that the power deficiency was principally due to problems in the compressor and first stage turbine areas and during this quarter several corrective changes have been initiated. Parts for a fourth engine being built for NASA Lewis have been shipped to NASA for installation of special instrumentation.

  14. Progress report on fast breeder reactor development in Japan

    International Nuclear Information System (INIS)

    1979-01-01

    The experimental fast reactor ''Joyo'' will be tested at 75 MW output, starting in April, 1980. In connection with the accident in the Three Mile Island plant, the reexamination of the plant safety and the rechecking-up of the maintenance control system were carried out, and the special inspection by the Science and Technology Agency was executed from May 21 to 23, 1979. Thereafter, the preparation for raising the power output was completed. The periodical inspection after the completion of 50 MW operation is being carried out. The state of progress of various equipments and the codes for core characteristic analysis is reported. The construction preliminary design (2) of the prototype reactor ''Monju'' is examined, and the same design (3) is prepared. The analysis of the decay heat in the prototype reactor is carried on for the safety licensing. The technological investigation of LMFBRs in foreign countries is under way. The preliminary design (4) of the demonstration reactor is under examination, and the technical specifications of the conceptual design (1) are prepared. The researches and developments of reactor physics, structural components, instrumentation and control, sodium technology, fuel materials, structures and materials, safety and steam generators are reported. (Kako, I.)

  15. Recent progress in electrocatalysts with mesoporous structures for application in polymer electrolyte membrane fuel cells

    OpenAIRE

    Xing, Wei; Wu, Zucheng; Tao, Shanwen

    2016-01-01

    Recently mesoporous materials have drawn great attention in fuel cell related applications, such as preparation of polymer electrolyte membranes and catalysts, hydrogen storage and purification. In this mini-review, we focus on recent developments in mesoporous electrocatalysts for polymer electrolyte membrane fuel cells, including metallic and metal-free catalysts for use as either anode or cathode catalysts. Mesoporous Pt-based metals have been synthesized as anode catalysts with improved a...

  16. High Performance Fuel Technology Development(I)

    International Nuclear Information System (INIS)

    Song, Kun Woo; Kim, Keon Sik; Bang, Jeong Yong; Park, Je Keon; Chen, Tae Hyun; Kim, Hyung Kyu

    2010-04-01

    The dual-cooled annular fuel has been investigated for the purpose of achieving the power uprate of 20% and decreasing pellet temperature by 30%. The 12x12 rod array and basic design was developed, which is mechanically compatible with the OPR-1000. The reactor core analysis has been performed using this design, and the results have shown that the criteria of nuclear, thermohydraulic and safety design are satisfied and pellet temperature can be lowered by 40% even in 120% power. The basic design of fuel component was developed and the cladding thickness was designed through analysis and experiments. The solutions have been proposed and analyzed to the technical issues such as 'inner channel blockage' and 'imbalance between inner and outer coolant'. The annular pellet was fabricated with good control of shape and size, and especially, a new sintering technique has been developed to control the deviation of inner diameter within ±5μm. The irradiation test of annular pellets has been conducted up to 10 MWD/kgU to find out the densification and swelling behaviors. The 11 types of materials candidates have developed for the PCI-endurance pellet, and the material containing the Mn-Al additive showed its creep performance of much better than UO2 material. The HANA cladding has been irradiated up to 61 MWD/kgU, and the results have shown that its oxidation resistance is better by 40% than that of Zircaloy. The 30 types of candidate materials for next generation have been developed through alloy design and property tests

  17. Progress on Plant-Level Components for Nuclear Fuel Recycling: Commonality

    International Nuclear Information System (INIS)

    De Almeida, Valmor F.

    2011-01-01

    Progress made in developing a common mathematical modeling framework for plant-level components of a simulation toolkit for nuclear fuel recycling is summarized. This ongoing work is performed under the DOE Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, which has an element focusing on safeguards and separations (SafeSeps). One goal of this element is to develop a modeling and simulation toolkit for used nuclear fuel recycling. The primary function of the SafeSeps simulation toolkit is to enable the time-dependent coupling of separation modules and safeguards tools (either native or third-party supplied) that simulate and/or monitor the individual separation processes in a separations plant. The toolkit integration environment will offer an interface for the modules to register in the toolkit domain based on the commonality of diverse unit operations. This report discusses the source of this commonality from a combined mathematical modeling and software design perspectives, and it defines the initial basic concepts needed for development of application modules and their integrated form, that is, an application software. A unifying mathematical theory of chemical thermomechanical network transport for physicochemical systems is proposed and outlined as the basis for developing advanced modules. A program for developing this theory from the underlying first-principles continuum thermomechanics will be needed in future developments; accomplishment of this task will enable the development of a modern modeling approach for plant-level models. Rigorous, advanced modeling approaches at the plant-level can only proceed from the development of reduced (or low-order) models based on a solid continuum field theory foundation. Such development will pave the way for future programmatic activities on software verification, simulation validation, and model uncertainty quantification on a scientific basis; currently, no satisfactory foundation exists for

  18. Progress on Plant-Level Components for Nuclear Fuel Recycling: Commonality

    Energy Technology Data Exchange (ETDEWEB)

    de Almeida, Valmor F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2011-08-15

    Progress made in developing a common mathematical modeling framework for plant-level components of a simulation toolkit for nuclear fuel recycling is summarized. This ongoing work is performed under the DOE Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, which has an element focusing on safeguards and separations (SafeSeps). One goal of this element is to develop a modeling and simulation toolkit for used nuclear fuel recycling. The primary function of the SafeSeps simulation toolkit is to enable the time-dependent coupling of separation modules and safeguards tools (either native or third-party supplied) that simulate and/or monitor the individual separation processes in a separations plant. The toolkit integration environment will offer an interface for the modules to register in the toolkit domain based on the commonality of diverse unit operations. This report discusses the source of this commonality from a combined mathematical modeling and software design perspectives, and it defines the initial basic concepts needed for development of application modules and their integrated form, that is, an application software. A unifying mathematical theory of chemical thermomechanical network transport for physicochemical systems is proposed and outlined as the basis for developing advanced modules. A program for developing this theory from the underlying first-principles continuum thermomechanics will be needed in future developments; accomplishment of this task will enable the development of a modern modeling approach for plant-level models. Rigorous, advanced modeling approaches at the plant-level can only proceed from the development of reduced (or low-order) models based on a solid continuum field theory foundation. Such development will pave the way for future programmatic activities on software verification, simulation validation, and model uncertainty quantification on a scientific basis; currently, no satisfactory foundation exists for

  19. Cell module and fuel conditioner development

    Science.gov (United States)

    Hoover, D. Q., Jr.

    1980-01-01

    Components for the first 5 cell stack (no cooling plates) of the MK-2 design were fabricated. Preliminary specfications and designs for the components of a 23 cell MK-1 stack with four DIGAS cooling plates were developed. The MK-2 was selected as a bench mark design and a preliminary design of the facilities required for high rate manufacture of fuel cell modules was developed. Two stands for testing 5 cell stacks were built and design work for modifying existing stands and building new stands for 23 and 80 cell stacks was initiated. Design and procurement of components and materials for the catalyst test stand were completed and construction initiated. Work on the specifications of pipeline gas, tap water and recovered water and definition of equipment required for treatment was initiated. An innovative geometry for the reformer was conceived and modifications of the computer program to be used in its design were stated.

  20. Planning developments in British Nuclear Fuels Ltd

    Energy Technology Data Exchange (ETDEWEB)

    Roper, D A [British Nuclear Fuels Ltd., Risley

    1978-10-01

    The state of the corporate planning art in British Nuclear Fuels Ltd. was described by N.R.Geary (Long Range Planning, September (1973)) just 2 years after Company formation. This article discusses more recent planning developments over the period to date during which the Company adopted a Divisionalized structure (from October 1974) and has been required to submit an annual Company plan to the Department of Energy (from November 1975). Background information on the origin and nature of the BNFL and its business, and the particular features of the Company which reflect into the nature and method of its planning were given in the 1973 article and only a brief introductory updating of the Company position is included here. Subsequently the features and problems of BNFL's operating and development planning system are described. Finally, messages arising from BNFL's planning experience to date which may be of general application and therefore of value to other practitioners of planning are listed.

  1. The development of microfabricated biocatalytic fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Sasaki, Satoshi; Karube, Isao [University of Tokyo (Japan). Research Center for Advanced Science and Technology

    1999-02-01

    The production of electricity by biocatalytic fuel cells has been feasible for almost two decades and can produce electric power at a practical level. These fuel cells use immobilized microorganisms or enzymes as catalysts, and glucose as a fuel. A microfabricated enzyme battery has recently been made that is designed to function as a power supply for microsurgery robots or artificial organs. (author)

  2. The status of nuclear fuel cycle system analysis for the development of advanced nuclear fuel cycles

    Energy Technology Data Exchange (ETDEWEB)

    Ko, Won Il; Kim, Seong Ki; Lee, Hyo Jik; Chang, Hong Rae; Kwon, Eun Ha; Lee, Yoon Hee; Gao, Fanxing [KAERI, Daejeon (Korea, Republic of)

    2011-11-15

    The system analysis has been used with different system and objectives in various fields. In the nuclear field, the system can be applied from uranium mining to spent fuel reprocessing or disposal which is called the nuclear fuel cycle. The analysis of nuclear fuel cycle can be guideline for development of advanced fuel cycle through integrating and evaluating the technologies. For this purpose, objective approach is essential and modeling and simulation can be useful. In this report, several methods which can be applicable for development of advanced nuclear fuel cycle, such as TRL, simulation and trade analysis were explained with case study

  3. Development of dynamic simulation code for fuel cycle fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Aoki, Isao; Seki, Yasushi [Department of Fusion Engineering Research, Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka, Ibaraki (Japan); Sasaki, Makoto; Shintani, Kiyonori; Kim, Yeong-Chan

    1999-02-01

    A dynamic simulation code for fuel cycle of a fusion experimental reactor has been developed. The code follows the fuel inventory change with time in the plasma chamber and the fuel cycle system during 2 days pulse operation cycles. The time dependence of the fuel inventory distribution is evaluated considering the fuel burn and exhaust in the plasma chamber, purification and supply functions. For each subsystem of the plasma chamber and the fuel cycle system, the fuel inventory equation is written based on the equation of state considering the fuel burn and the function of exhaust, purification, and supply. The processing constants of subsystem for steady states were taken from the values in the ITER Conceptual Design Activity (CDA) report. Using this code, the time dependence of the fuel supply and inventory depending on the burn state and subsystem processing functions are shown. (author)

  4. Evolutionary developments of advanced PWR nuclear fuels and cladding materials

    International Nuclear Information System (INIS)

    Kim, Kyu-Tae

    2013-01-01

    Highlights: • PWR fuel and cladding materials development processes are provided. • Evolution of PWR advanced fuel in U.S.A. and in Korea is described. • Cutting-edge design features against grid-to-rod fretting and debris are explained. • High performance data of advanced grids, debris filters and claddings are given. -- Abstract: The evolutionary developments of advanced PWR fuels and cladding materials are explained with outstanding design features of nuclear fuel assembly components and zirconium-base cladding materials. The advanced PWR fuel and cladding materials development processes are also provided along with verification tests, which can be used as guidelines for newcomers planning to develop an advanced fuel for the first time. The up-to-date advanced fuels with the advanced cladding materials may provide a high level of economic utilization and reliable performance even under current and upcoming aggressive operating conditions. To be specific, nuclear fuel vendors may achieve high fuel burnup capability of between 45,000 and 65,000 MWD/MTU batch average, overpower thermal margin of as much as 15% and longer cycle length up to 24 months on the one hand and fuel failure rates of around 10 −6 on the other hand. However, there is still a need for better understanding of grid-to-rod fretting wear mechanisms leading to major PWR fuel defects in the world and subsequently a driving force for developing innovative spacer grid designs with zero fretting wear-induced fuel failure

  5. Diversion-resistant nuclear-fuels processing. Progress report, 1980-1981

    International Nuclear Information System (INIS)

    Tomlinson, R.E.; Campbell, M.H.; Hansen, L.E.; Jaech, J.L.; Merker, L.G.; Malody, C.W.; Nilson, R.; Schneider, R.A.

    1983-01-01

    Design objectives for the projected colocated facilities were formulated. An assessment methodology, was developed. As a baseline, the modified Delphi procedure was used to evaluate the most recent US designs of a fuel reprocessing plant and a fuels refabrication plant against the identified regulations and goals. An upgraded design concept was synthesized, using the baseline fuel reprocessing plant design as a starting point but using a new design concept for fuel conversion and refabrication. The modified Delphi procedure was used to evaluate the upgrading design concepts against identified regulations and goals. The upgraded portions, product conversion, fuel fabrication, and laboratory received ratings of 95% or higher compared with ratings of about 60% for the baseline designs. Alternative reprocessing and refabrication processes were evaluated to determine if any process could offer an inherent safeguards advantage over the combination included in the upgraded design concept. Tentative conclusions reached are: A combination of a modified Purex solvent extraction fuel reprocessing and a Sphere-Pac fuel fabrication flowsheet, coupled with an improved measurement system and a rapid draw-down inventory procedure, can provide the means for meeting most NRC and IAEA Goals. Given industry and DOE support, fuels fabricated by the Sphere-Pac process can probably be licensed by 1990. With a modest demonstration effort, the processes and equipment modifications envisioned can be ready for incorporation in a detailed design by 1985. Practical techniques and equipment are available for the assured control of the movement of plutonium and personnel into and out of the plant and between plant segments. The incremental cost of facilities and procedures needed to provide the above capabilities would probably increase the unit cost of fuel reprocessing and conversion by 5 to 10%

  6. Development of System Engineering Technology for Nuclear Fuel Cycle

    International Nuclear Information System (INIS)

    Kim, Hodong; Choi, Iljae

    2013-04-01

    The development of efficient process for spent fuel and establishment of system engineering technology to demonstrate the process are required to develop nuclear energy continuously. The demonstration of pyroprocess technology which is proliferation resistance nuclear fuel cycle technology can reduce spent fuel and recycle effectively. Through this, people's trust and support on nuclear power would be obtained. Deriving the optimum nuclear fuel cycle alternative would contribute to establish a policy on back-end nuclear fuel cycle in the future, and developing the nuclear transparency-related technology would contribute to establish amendments of the ROK-U. S. Atomic Energy Agreement scheduled in 2014

  7. Development of FR fuel cycle in japan (1) development scope of fuel cycle technology

    International Nuclear Information System (INIS)

    Nakamura, H.; Funasaka, H.; Namekawa, T.

    2008-01-01

    A fast reactor (FR) cycle has a potential to realize a sustainable energy supply system that is harmonized with environment by fully recycling both uranium (U) and transuranium (TRU) elements. In Japan, a Feasibility Study on Commercialized FR Cycle Systems (FS) was launched in July 1999, and through two different study phases, a final report was presented in 2006. As a result of FS, a combined system of sodium-cooled FR with mixed-oxide (MOX) fuel, advanced aqueous reprocessing and simplified pelletizing fuel fabrication was considered to be most promising for commercialization. The advanced aqueous reprocessing system, which is called the New Extraction system for TRU recovery (NEXT), consists of a U crystallization process for the bulk of U recovery, a simplified solvent extraction process for residual U, plutonium (Pu) and neptunium (Np) without Pu partitioning and purification, and a process for recovering americium (Am) and curium (Cm) from the raffinate. The ratio of Pu/U concentration in the mother solution after crystallization is adequate for MOX fuel fabrication, and thus complicated powder mixing processes for adjusting Pu content in MOX fuel can be eliminated in the subsequent simplified fuel fabrication system. In this system, lubricant-mixing process can also be eliminated by adopting the advanced technology in which lubricant is coated on the inner surface of a die before fuel powder supply. Such a simplification could help us overcoming the difficulty to treat MA bearing fuel powders in a hot cell. Ministry of Education, Culture, Sports, Science and Technology (MEXT) reviewed these results of FS in 2006 and identified the most promising FR cycle concept proposed in the FS phase II study as a mainline choice for commercialization. According to such a governmental assessment, R and D activities of FR cycle systems were decided to be concentrated mainly to the innovative technology development for the mainline concept. The stage of R and D project was

  8. Technology readiness levels for advanced nuclear fuels and materials development

    Energy Technology Data Exchange (ETDEWEB)

    Carmack, W.J., E-mail: jon.carmack@inl.gov [Idaho National Laboratory, Idaho Falls, ID (United States); Braase, L.A.; Wigeland, R.A. [Idaho National Laboratory, Idaho Falls, ID (United States); Todosow, M. [Brookhaven National Laboratory, Upton, NY (United States)

    2017-03-15

    Highlights: • Definition of nuclear fuels system technology readiness level. • Identification of evaluation criteria for nuclear fuel system TRLs. • Application of TRLs to fuel systems. - Abstract: The Technology Readiness process quantitatively assesses the maturity of a given technology. The National Aeronautics and Space Administration (NASA) pioneered the process in the 1980s to inform the development and deployment of new systems for space applications. The process was subsequently adopted by the Department of Defense (DoD) to develop and deploy new technology and systems for defense applications. It was also adopted by the Department of Energy (DOE) to evaluate the maturity of new technologies in major construction projects. Advanced nuclear fuels and materials development is needed to improve the performance and safety of current and advanced reactors, and ultimately close the nuclear fuel cycle. Because deployment of new nuclear fuel forms requires a lengthy and expensive research, development, and demonstration program, applying the assessment process to advanced fuel development is useful as a management, communication, and tracking tool. This article provides definition of technology readiness levels (TRLs) for nuclear fuel technology as well as selected examples regarding the methods by which TRLs are currently used to assess the maturity of nuclear fuels and materials under development in the DOE Fuel Cycle Research and Development (FCRD) Program within the Advanced Fuels Campaign (AFC).

  9. Definition of Technology Readiness Levels for Transmutation Fuel Development

    International Nuclear Information System (INIS)

    Jon Carmack; Kemal O. Pasamehmetoglu

    2008-01-01

    To quantitatively assess the maturity of a given technology, the Technology Readiness Level (TRL) process is used. The TRL process has been developed and successfully used by the Department of Defense (DOD) for development and deployment of new technology and systems for defense applications. In addition, NASA has also successfully used the TRL process to develop and deploy new systems for space applications. Transmutation fuel development is a critical technology needed for closing the nuclear fuel cycle. Because the deployment of a new nuclear fuel forms requires a lengthy and expensive research, development, and demonstration program, applying the TRL concept to the transmutation fuel development program is very useful as a management and tracking tool. This report provides definition of the technology readiness level assessment process as defined for use in assessing nuclear fuel technology development for the Transuranic Fuel Development Campaign

  10. Development of fuel cell systems for aircraft applications based on synthetic fuels

    Energy Technology Data Exchange (ETDEWEB)

    Pasel, J.; Samsun, R.C.; Doell, C.; Peters, R.; Stolten, D. [Forschungszentrum Juelich GmbH (Germany)

    2010-07-01

    At present, in the aviation sector considerable scientific project work deals with the development of fuel cell systems based on synthetic fuels to be integrated in future aircraft. The benefits of fuel cell systems in aircraft are various. They offer the possibility to simplify the aircraft layout. Important systems, i.e. the gas turbine powered auxiliary power unit (APU) for electricity supply, the fuel tank inserting system and the water tank, can be substituted by one single system, the fuel cell system. Additionally, the energy demand for ice protection can be covered assisted by fuel cell systems. These measures reduce the consumption of jet fuel, increase aircraft efficiency and allow the operation at low emissions. Additionally, the costs for aircraft related investments, for aircraft maintenance and operation can be reduced. On the background of regular discussions about environmental concerns (global warming) of kerosene Jet A-1 and its availability, which might be restricted in a few years, the aircraft industry is keen to employ synthetic, sulfur-free fuels such as Fischer-Tropsch fuels. These comprise Bio-To-Liquid and Gas-To-Liquid fuels. Within this field of research the Institute of Energy Research (IEF-3) in Juelich develops complete and compact fuel cell systems based on the autothermal reforming of these kinds of fuels in cooperation with industry. This paper reports about this work. (orig.)

  11. Development of new membrane materials for direct methanol fuel cells

    NARCIS (Netherlands)

    Yildirim, M.H.

    2009-01-01

    Development of new membrane materials for direct methanol fuel cells Direct methanol fuel cells (DMFCs) can convert the chemical energy of a fuel directly into electrical energy with high efficiency and low emission of pollutants. DMFCs can be used as the power sources to portable electronic devices

  12. Trend of fuel for light water reactors and development hereafter

    International Nuclear Information System (INIS)

    Ichikawa, Michio; Maru, Akira; Shimoshige, Takanori

    1993-01-01

    Recently, the heightening of fuel burnup has been actively advanced internationally. Its degree is different according to the policy and the economical factors in respective countries. The extension of the period of operation cycle urges high burnup in view of economy. The circumstances in USA, Europe and Japan are explained. The corrosion of zircaloy cladding is the factor of limiting fuel life. The state of corrosion in reactors is different in BWRs and PWRs, and both cases are explained. The emission of FP gas from pellets to fuel rods raises the internal pressure of the fuel rods, and affects the gap conductance between pellets and cladding tubes. In the fuel for LWRs, plutonium is formed locally and burns in pellet rim part. This rim effect is discussed. The irradiation growth of fuel rods, creep down and pellet-cladding interaction are explained. The MOX fuel for LWRs and the trend of development of new type fuel are reported. The fuel for BWRs of Hitachi Ltd. and Toshiba Corp. and Nuclear Fuel Industries Ltd., the fuel for PWRs of Mitsubishi Heavy Industries Ltd. and Nuclear fuel Industries Ltd., and the recent development of the fuel cladding tubes for LWRs are described. (K.I.)

  13. Status and development of RBMK fuel rods and reactor materials

    International Nuclear Information System (INIS)

    Bibilashvili, Yu.K.; Reshetnikov, F.G.; Ioltukhovsky, A.G.

    1998-01-01

    The paper presents current status and development of RBMK fuel rods and reactor materials. With regard to fuel rod cladding the following issues have been discussed: corrosion, tensile properties, welding technology and testing of an alternative cladding alloy with a composition of Zr-Nb-Sn-Fe. Erbium doped fuel has been suggested for safety improvement. Also analysis of fuel reliability is presented in the paper. (author)

  14. Interim development report: engineering-scale HTGR fuel particle crusher

    International Nuclear Information System (INIS)

    Baer, J.W.; Strand, J.B.

    1978-09-01

    During the reprocessing of HTGR fuel, a double-roll crusher is used to fracture the silicon carbide coatings on the fuel particles. This report describes the development of the roll crusher used for crushing Fort-St.Vrain type fissile and fertile fuel particles, and large high-temperature gas-cooled reactor (LHTGR) fissile fuel particles. Recommendations are made for design improvements and further testing

  15. Training development in Juzbado's Fuel Cycle Facility

    International Nuclear Information System (INIS)

    Perez, A.; Cunado, E.; Ortiz, D.

    2003-01-01

    In Juzbado's fuel cycle facility, because of the special activities developed, training is a very important issues. Training has been evolved, due to changes on the standards applicable each moment, and also due to the technological resources available. Both aspects have resulted in an evolution of the documents referred to training, such as training programs procedures, Radiation Protection Manual as well as the teaching methods. In the report we are going to present, we will show more precisely the changes that take place, referring to the different training methods used, present training sanitations, and the improvements already planned in training subjects as well as tools used, accomplishing with the legislation and improving in our effort of a better assimilation of the necessary knowledge. (Author)

  16. Further developments of PWR and BWR fuel elements

    International Nuclear Information System (INIS)

    Sofer, G.A.; Busselman, G.J.; Federico, L.J.

    1988-01-01

    The performance, safety, and economy of nuclear power plants in inluenced very decisively by the quality of their fuel elements. This is why quality assurance in fuel fabrication has been a factor of great importance from the outset. Operating experince and more stringent performance requirements have resulted in a continuous process of further development of fuel elements, which has been reflected also in lower and lower failure rates and increasingly higher burn-ups. Next to further development also innovation has been an important factor contributing to the present high quality level of fuel elements, which also has allowed fuel cycle costs to be decreased quite considerably. (orig.) [de

  17. Development on nuclear fuel cycle business in Japan

    International Nuclear Information System (INIS)

    Usami, Kogo

    2002-01-01

    The Japan Nuclear Fuel Co., Ltd. (JNF) develops five businesses on nuclear fuel cycle such as uranium concentration, storage and administration of high level radioactive wastes, disposition of low level radioactive wastes, used fuel reprocessing, MOX fuel, at Rokkasho-mura in Aomori prefecture. Here were introduced on outline, construction and operation in reprocessing and MOX fuel works, outline, present state and future subjects on technical development of uranium concentration, outline and safety of disposition center on low level radioactive wastes, and storage and administration of high level radioactive wastes. (G.K.)

  18. International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO). 2008 progress report

    International Nuclear Information System (INIS)

    2009-02-01

    The purpose of the work is to review the progress of the IAEA international project for innovative reactors and fuel cycle technologies (INPRO). The publication reports about the recognition of INPRO and on general Information on INPRO, its strengths, memberships, collaboration with other international initiatives, the INPRO organization and management and the history of INPRO. The section on the progress of INPRO in 2008 contains task 1: INPRO Methodology, task 2: Assessment Studies, task 3: Nuclear Energy Visions for the 21st Century, task 4: Infrastructure and Institutional Innovation, task 5: Common User Considerations and task 6: Collaborative Projects. Conclusions and New Trends are followed by a bibliography. Annex I deals with the INPRO project management in 2008 and Annex II provides a selection of photographs from 2008. Finally a list of acronyms is provided

  19. Studies in development immunogenetics. Annual progress report

    Energy Technology Data Exchange (ETDEWEB)

    Owen, R D

    1975-03-26

    This contract provides the research support for a group concerned with a relatively large range of problems. The integrating thread that runs through it is that of an interest in development and its genetic regulation, mainly in complex organisms and with an emphasis on the immune system as a model for developmental analysis and as a tool for following the development of other systems, especially the brain. It includes studies of biochemical genetics, primarily from a developmental viewpoint and with particular regard to defense mechanisms, and cellular aspects of the immune system. It extends into the area of cancer immunology and cell specificities as related to tumor systems, primarily from an immunogenetic viewpoint and with particular reference to leukemias in the mouse, and to disruptions of genetic control mechanisms in tumor development, especially as approached through the reappearance of fetal antigens associated with tumor development.

  20. Interim report spent nuclear fuel retrieval system fuel handling development testing

    Energy Technology Data Exchange (ETDEWEB)

    Ketner, G.L.; Meeuwsen, P.V.; Potter, J.D.; Smalley, J.T.; Baker, C.P.; Jaquish, W.R.

    1997-06-01

    Fuel handling development testing was performed in support of the Fuel Retrieval System (FRS) Sub-Project at the Hanford Site. The project will retrieve spent nuclear fuel, clean and remove fuel from canisters, repackage fuel into baskets, and load fuel into a multi-canister overpack (MCO) for vacuum drying and interim dry storage. The FRS is required to retrieve basin fuel canisters, clean fuel elements sufficiently of uranium corrosion products (or sludge), empty fuel from canisters, sort debris and scrap from whole elements, and repackage fuel in baskets in preparation for MCO loading. The purpose of fuel handling development testing was to examine the systems ability to accomplish mission activities, optimization of equipment layouts for initial process definition, identification of special needs/tools, verification of required design changes to support performance specification development, and validation of estimated activity times/throughput. The test program was set up to accomplish this purpose through cold development testing using simulated and prototype equipment; cold demonstration testing using vendor expertise and systems; and graphical computer modeling to confirm feasibility and throughput. To test the fuel handling process, a test mockup that represented the process table was fabricated and installed. The test mockup included a Schilling HV series manipulator that was prototypic of the Schilling Hydra manipulator. The process table mockup included the tipping station, sorting area, disassembly and inspection zones, fuel staging areas, and basket loading stations. The test results clearly indicate that the Schilling Hydra arm cannot effectively perform the fuel handling tasks required unless it is attached to some device that can impart vertical translation, azimuth rotation, and X-Y translation. Other test results indicate the importance of camera locations and capabilities, and of the jaw and end effector tool design. 5 refs., 35 figs., 3 tabs.

  1. Future fuel cycle development for CANDU reactors

    International Nuclear Information System (INIS)

    Hatcher, S.R.; McDonnell, F.N.; Griffiths, J.; Boczar, P.G.

    1987-01-01

    The CANDU reactor has proven to be safe and economical and has demonstrated outstanding performance with natural uranium fuel. The use of on-power fuelling, coupled with excellent neutron economy, leads to a very flexible reactor system with can utilize a wide variety of fuels. The spectrum of fuel cycles ranges from natural uranium, through slightly enriched uranium, to plutonium and ultimately thorium fuels which offer many of the advantages of the fast breeder reactor system. CANDU can also burn the recycled uranium and/or the plutonium from fuel discharged from light water reactors. This synergistic relationship could obviate the need to re-enrich the reprocessed uranium and allow a simpler reprocessing scheme. Fule management strategies that will permit future fuel cycles to be used in existing CANDU reactors have been identified. Evolutionary design changes will lead to an even greater flexibility, which will guarantee the continued success of the CANDU system. (author)

  2. Development and progress: advancing towards environmental crisis

    International Nuclear Information System (INIS)

    Gomez G, Luis Jair

    2011-01-01

    Physical, biological and social evolution is doubtless. One of its first manifestations is the arrival of technique when hominids emerge from pre-hominids. Those first technical developments implied a new relation man/environment that was expressed in three components that appeared successively and pushed each other in time and space like this: dominion over nature, population concentration (urbanism), and population growth. Techniques are to generate three notorious effects on the relation man/nature: 1. Deep intervention on the physical environment: mining and industrial transformation processes; 2. Deep intervention on the biological environment: development of agriculture with a decrease in biodiversity; and 3. Deep intervention on the social environment: going from a pre-modern communitarian world, to the individualism of modernity; and from the agrarian field to the big city. All these technical developments boosted dominion of the technosphere over the ecosphere, which led to the appearance of the Environmental Crisis, whose most notable manifestation is Climatic Change.

  3. Tularemia vaccine development: paralysis or progress?

    Directory of Open Access Journals (Sweden)

    Sunagar R

    2016-05-01

    Full Text Available Raju Sunagar, Sudeep Kumar, Brian J Franz, Edmund J Gosselin Center for Immunology and Microbial Disease, Albany Medical College, Albany, NY, USA Abstract: Francisella tularensis (Ft is a gram-negative intercellular pathogen and category A biothreat agent. However, despite 15 years of strong government investment and intense research focused on the development of a US Food and Drug Administration-approved vaccine against Ft, the primary goal remains elusive. This article reviews research efforts focused on developing an Ft vaccine, as well as a number of important factors, some only recently recognized as such, which can significantly impact the development and evaluation of Ft vaccine efficacy. Finally, an assessment is provided as to whether a US Food and Drug Administration-approved Ft vaccine is likely to be forthcoming and the potential means by which this might be achieved. Keywords: Sex bias, media impact, differential protection, cellular immunity, humoral immunity

  4. Development of the Fuel Element Database of PUSPATI TRIGA Reactor

    International Nuclear Information System (INIS)

    Nurhayati Ramli; Naim Syauqi Hamzah; Nurfazila Husain; Yahya Ismail; Mat Zin Mat Husin; Mohd Fairus Abd Farid

    2015-01-01

    Since June 28th, 1982, the PUSPATI TRIGA Reactor (RTP) operates safely with an accumulated energy release of about 17,200 MWhr, which corresponds to about 882 g of uranium burn-up. The reactor core has been reconfigured 15th times. Presently, there are 111 TRIGA fuel elements in the core, which 66 of the fuel elements are from the initial criticality while the rest of the fuel elements have been added to compensate the uranium consumption. As 59 % of the fuel elements are older than 30 years old, it is necessary to put the history of every fuel element in a database for easy access of the fuel element movement, inspection results history and integrity status. This paper intends to describe how the fuel element database is developed and related formulae used in determining the RTP fuel element elongation. (author)

  5. Development of High Performance Hybrid Fuels

    Data.gov (United States)

    National Aeronautics and Space Administration — NASA's strategic goals call for innovation in space technology for our nation's explorative future. Early phase paraffin fuel technology could enable practical...

  6. Experience in the development of metal uranium-base nuclear fuel for heavy-water gas-cooled reactors

    International Nuclear Information System (INIS)

    Ashikhmin, V.P.; Vorob'ev, M.A.; Gusarov, M.S.; Davidenko, A.S.; Zelenskij, V.F.; Ivanov, V.E.; Krasnorutskij, V.S.; Petel'guzov, I.A.; Stukalov, A.I.

    1978-01-01

    Investigations were carried out to solve the problem of making the development of radiation-resistant uranium fuel for power reactors including the heavy-water gas-cooled KS-150 reactor. Factors are considered that limit the lifetime of uranium fuel elements, and the ways of suppressing them are discussed. Possible reasons of the insufficient radiation resistance of uranium rod fuel element and the progress attained are analyzed. Some general problems on the fuel manufacture processes are discussed. The main results are presented on the operation of the developed fuel in research reactor loops and the commercial heavy-water KS-150 reactor. The results confirm an exceptionally high radiation resistance of fuel to burn-ups of 1.5-2%. The successful solution of a large number of problems associated with the development of metal uranium fuel provides for new possibilities of using metal uranium in power reactors

  7. UCLA accelerator research ampersand development. Progress report

    International Nuclear Information System (INIS)

    1997-01-01

    This report discusses work on advanced accelerators and beam dynamics at ANL, BNL, SLAC, UCLA and Pulse Sciences Incorporated. Discussed in this report are the following concepts: Wakefield acceleration studies; plasma lens research; high gradient rf cavities and beam dynamics studies at the Brookhaven accelerator test facility; rf pulse compression development; and buncher systems for high gradient accelerator and relativistic klystron applications

  8. Development of fuel number reader by ultrasonic imaging techniques

    International Nuclear Information System (INIS)

    Omote, T.; Yoshida, T.

    1991-01-01

    This paper reports on a spent fuel ID number reader using ultrasonic imaging techniques that has been developed to realize efficient and automatic verification of fuel numbers, thereby to reduce mental load and radiation exposure for operators engaged in the verification task. The ultrasonic imaging techniques for automatic fuel number recognition are described. High-speed and high reliability imaging of the spent fuel ID number are obtained by using linear array type ultrasonic probe. The ultrasonic wave is scanned by switching array probe in vertical direction, and scanned mechanically in horizontal direction. Time for imaging of spent fuel ID number on assembly was confirmed less than three seconds by these techniques. And it can recognize spent fuel ID number even if spent fuel ID number can not be visualized by an optical method because of depositing fuel number regions by soft card. In order to recognize spent fuel ID number more rapidly and more reliably, coded fuel number expressed by plural separate recesses form is developed. Every coded fuel number consists of six small holes (about 1 mm dia.) and can be marked adjacent to the existing fuel number expressed by letters and numbers

  9. Developing clean fuels: Novel techniques for desulfurization

    Science.gov (United States)

    Nehlsen, James P.

    The removal of sulfur compounds from petroleum is crucial to producing clean burning fuels. Sulfur compounds poison emission control catalysts and are the source of acid rain. New federal regulations require the removal of sulfur in both gasoline and diesel to very low levels, forcing existing technologies to be pushed into inefficient operating regimes. New technology is required to efficiently produce low sulfur fuels. Two processes for the removal of sulfur compounds from petroleum have been developed: the removal of alkanethiols by heterogeneous reaction with metal oxides; and oxidative desulfurization of sulfides and thiophene by reaction with sulfuric acid. Alkanethiols, common in hydrotreated gasoline, can be selectively removed and recovered from a hydrocarbon stream by heterogeneous reaction with oxides of Pb, Hg(II), and Ba. The choice of reactive metal oxides may be predicted from simple thermodynamic considerations. The reaction is found to be autocatalytic, first order in water, and zero order in thiol in the presence of excess oxide. The thiols are recovered by reactive extraction with dilute oxidizing acid. The potential for using polymer membrane hydrogenation reactors (PEMHRs) to perform hydrogenation reactions such as hydrodesulfurization is explored by hydrogenating ketones and olefins over Pt and Au group metals. The dependence of reaction rate on current density suggests that the first hydrogen addition to the olefin is the rate limiting step, rather than the adsorption of hydrogen, for all of the metals tested. PEMHRs proved unsuccessful in hydrogenating sulfur compounds to perform HDS. For the removal of sulfides, a two-phase reactor is used in which concentrated sulfuric acid oxidizes aromatic and aliphatic sulfides present in a hydrocarbon solvent, generating sulfoxides and other sulfonated species. The polar oxidized species are extracted into the acid phase, effectively desulfurizing the hydrocarbon. A reaction scheme is proposed for this

  10. Recent Progress on the Key Materials and Components for Proton Exchange Membrane Fuel Cells in Vehicle Applications

    Directory of Open Access Journals (Sweden)

    Cheng Wang

    2016-07-01

    Full Text Available Fuel cells are the most clean and efficient power source for vehicles. In particular, proton exchange membrane fuel cells (PEMFCs are the most promising candidate for automobile applications due to their rapid start-up and low-temperature operation. Through extensive global research efforts in the latest decade, the performance of PEMFCs, including energy efficiency, volumetric and mass power density, and low temperature startup ability, have achieved significant breakthroughs. In 2014, fuel cell powered vehicles were introduced into the market by several prominent vehicle companies. However, the low durability and high cost of PEMFC systems are still the main obstacles for large-scale industrialization of this technology. The key materials and components used in PEMFCs greatly affect their durability and cost. In this review, the technical progress of key materials and components for PEMFCs has been summarized and critically discussed, including topics such as the membrane, catalyst layer, gas diffusion layer, and bipolar plate. The development of high-durability processing technologies is also introduced. Finally, this review is concluded with personal perspectives on the future research directions of this area.

  11. Recent development and progress of IBA cyclotrons

    Energy Technology Data Exchange (ETDEWEB)

    Kleeven, W., E-mail: Willem.Kleeven@iba-group.com [Ion Beam Applications s.a. Chemin du Cyclotron 3, Louvain-la-Neuve (Belgium); Abs, M., E-mail: Michel.Abs@iba-group.com [Ion Beam Applications s.a. Chemin du Cyclotron 3, Louvain-la-Neuve (Belgium); Delvaux, J.L., E-mail: Jean-Luc.Delvaux@iba-group.com [Ion Beam Applications s.a. Chemin du Cyclotron 3, Louvain-la-Neuve (Belgium); Forton, E., E-mail: Eric.Forton@iba-group.com [Ion Beam Applications s.a. Chemin du Cyclotron 3, Louvain-la-Neuve (Belgium); Jongen, Y., E-mail: Yves.Jongen@iba-group.com [Ion Beam Applications s.a. Chemin du Cyclotron 3, Louvain-la-Neuve (Belgium); Medeiros Romao, L., E-mail: Luis.MedeirosRomao@iba-group.com [Ion Beam Applications s.a. Chemin du Cyclotron 3, Louvain-la-Neuve (Belgium); Nactergal, B., E-mail: Benoit.Nactergal@iba-group.com [Ion Beam Applications s.a. Chemin du Cyclotron 3, Louvain-la-Neuve (Belgium); Nuttens, V., E-mail: Vincent.Nuttens@iba-group.com [Ion Beam Applications s.a. Chemin du Cyclotron 3, Louvain-la-Neuve (Belgium); Servais, T., E-mail: Thomas.Servais@iba-group.com [Ion Beam Applications s.a. Chemin du Cyclotron 3, Louvain-la-Neuve (Belgium); Vanderlinden, T., E-mail: Thierry.Vanderlinden@iba-group.com [Ion Beam Applications s.a. Chemin du Cyclotron 3, Louvain-la-Neuve (Belgium); Zaremba, S., E-mail: Simon.Zaremba@iba-group.com [Ion Beam Applications s.a. Chemin du Cyclotron 3, Louvain-la-Neuve (Belgium)

    2011-12-15

    Several cyclotron development projects were recently realized by Ion Beam Applications S.A. (IBA). This contribution presents three of them: (i) the intensity enhancement of the Cyclone 30 cyclotron, a machine mainly used for the production of SPECT isotopes. This project is related with the increased demand for {sup 201}Tl because of the shortage of Mo/Tc generators from nuclear reactors, (ii) development of a new versatile multiple-particle K = 30 isotope-production cyclotron (the Cyclone 30XP) being able to accelerate H{sup -}, D{sup -} and also {alpha}-particles. The {alpha}-beam of this cyclotron will allow the production of new therapeutic isotopes (e.g. {sup 211}At) and (iii) commissioning of the Cyclone 70 cyclotron installed for Arronax in France. This machine is similar to the C30XP but provides higher energy (K = 70) and allows research on new types of medical isotopes.

  12. Development of an extended-burnup Mark B design. Second semiannual progress report, January-June 1979

    International Nuclear Information System (INIS)

    1979-11-01

    The immediate goal of the DOE/AP and L/B and W project is to extend the burnup of light water reactor fuel assemblies beyond present limits to 50,000 MWd/mtU batch average burnup. Fuel management plans and fuel designs are being directed to attain the increased burnup limits. Lead-test assemblies of extended-burnup designs will be manufactured, irradiated in a commercial pressurized water reactor, and examined to support extended-burnup fuel cycles. This report, covering the period from January through June 1979, is the second semiannual progress report for the program. Efforts have included analyses of extended-burnup fuel cycles, developed of both annular fuel pellet and segmented rod designs, and design of a nondestructive post-irradiation examination system

  13. Progress in CPI Microwave Tube Development

    Science.gov (United States)

    Wright, Edward L.; Bohlen, Heinz

    2006-01-01

    CPI continues its role as a leading supplier of state-of-the-art, high-power microwave tubes; from linear beam, velocity- and density-modulated devices, to high frequency gyro-devices. Klystrons are the device-of-choice for many high-power microwave applications, and can provide multi-megawatts to multi-kilowatts of power from UHF to W-band, respectively. A number of recent and on-going developments will be described. At UHF frequencies, the inductive output tube (IOT) has replaced the klystron for terrestrial NTSC and HDTV broadcast, due to its high efficiency and linearity, and is beginning to see use in scientific applications requiring 300 kW or less. Recent advances have enabled use well into L-band. CPI has developed a number of multiple-beam amplifiers. The VKL-8301 multiple-beam klystron (MBK) was built for the TESLA V/UV and x-ray FEL projects, and is a candidate RF source for the International Linear Collider (ILC). We have also contributed to the development of the U.S. Naval Research Laboratory (NRL) high-power fundamental-mode S-band MBK. The VHP-8330B multiple-beam, high-order mode (HOM) IOT shows great promise as a compact, CW UHF source for high power applications. These topics will be discussed, along with CPI's development capabilities for new and novel applications. Most important is our availability to provide design and fabrication services to organizations requiring CPI's manufacturing and process control infrastructure to build and test state-of-the-art devices.

  14. Progress in development of iron base alloys

    International Nuclear Information System (INIS)

    Zackay, V.V.; Parker, E.R.

    1980-01-01

    The ways of development of new iron base high-strength alloys are considered. Perspectiveness of ferritic steel strengthening with intermetallides (TaFe 2 , for instance) is shown. Favourable combination of plasticity, strength and fracture toughness in nickel-free iron-manganese alloys (16-20%) is also pointed out. A strength level of alloyed maraging steels can be achieved by changes in chemical composition and by proper heat treatments of low- and medium-alloyed steels

  15. Recent Progress in Perennial Buckwheat Development

    Directory of Open Access Journals (Sweden)

    Qing-Fu Chen

    2018-02-01

    Full Text Available Grains in the genus Fagopyrum have benefits to human health and are an excellent gluten-free raw material. Of all cereal foods, this genus has the highest total content of amino-acid nutrients necessary for humans; nutrients that are resistant to digestion (protein and starch resulting in their sustained release; higher dietary fiber content than key cereals, and is rich in a special healthy ingredient (flavonoids. Fagopyrum includes 24 species of which five are perennial. Among them, golden buckwheat (F.cymosum complex is the most important perennial buckwheat, which is not only used in Chinese medicine, but also has great potential in healthy food crop. In order to provide some clues for perennial crop studies and their industry development, this paper presents the state of perennial buckwheat research in terms of taxonomy; natural chemical products and pharmacological and health functions; genetics and evolution; breeding; and product development and utilization. The great advances such as successful interspecific crossing and its subsequent new perennial buckwheat varieties will speed up the development of the perennial buckwheat industry.

  16. Progress on development of SPIDER diagnostics

    Science.gov (United States)

    Pasqualotto, R.; Agostini, M.; Barbisan, M.; Bernardi, M.; Brombin, M.; Cavazzana, R.; Croci, G.; Palma, M. Dalla; Delogu, R. S.; Gorini, G.; Lotto, L.; Muraro, A.; Peruzzo, S.; Pimazzoni, A.; Pomaro, N.; Rizzolo, A.; Serianni, G.; Spolaore, M.; Tardocchi, M.; Zaniol, B.; Zaupa, M.

    2017-08-01

    SPIDER experiment, the full size prototype of the beam source for the ITER heating neutral beam injector, has to demonstrate extraction and acceleration to 100 kV of a large negative ion hydrogen or deuterium beam with co-extracted electron fraction e-/D- SPIDER plant systems are being installed, the different diagnostic systems are in the procurement phase. Their final design is described here with a focus on some key solutions and most original and cost effective implementations. Thermocouples used to measure the power load distribution in the source and over the beam dump front surface will be efficiently fixed with proven technique and acquired through commercial and custom electronics. Spectroscopy needs to use well collimated lines of sight and will employ novel design spectrometers with higher efficiency and resolution and filtered detectors with custom built amplifiers. The electrostatic probes will be operated through electronics specifically developed to cope with the challenging environment of the RF source. The instrumented calorimeter STRIKE will use new CFC tiles, still under development. Two linear cameras, one built in house, have been tested as suitable for optical beam tomography. Some diagnostic components are off the shelf, others are custom developed: some of these are being prototyped or are under test before final production and installation, which will be completed before start of SPIDER operation.

  17. Developing Spent Fuel Assembly for Advanced NDA Instrument Calibration - NGSI Spent Fuel Project

    Energy Technology Data Exchange (ETDEWEB)

    Hu, Jianwei [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauld, Ian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Banfield, James [GE Hitachi Nuclear Energy, Wilmington, NC (United States); Skutnik, Steven [Univ. of Tennessee, Knoxville, TN (United States)

    2014-02-01

    This report summarizes the work by Oak Ridge National Laboratory to investigate the application of modeling and simulation to support the performance assessment and calibration of the advanced nondestructive assay (NDA) instruments developed under the Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) Project. Advanced NDA instrument calibration will likely require reference spent fuel assemblies with well-characterized nuclide compositions that can serve as working standards. Because no reference spent fuel standard currently exists, and the practical ability to obtain direct measurement of nuclide compositions using destructive assay (DA) measurements of an entire fuel assembly is prohibitive in the near term due to the complexity and cost of spent fuel experiments, modeling and simulation will be required to construct such reference fuel assemblies. These calculations will be used to support instrument field tests at the Swedish Interim Storage Facility (Clab) for Spent Nuclear Fuel.

  18. Development of nuclear fuel materials for research reactor

    International Nuclear Information System (INIS)

    Kim, Chang Kyu; Park, H. D.; Kim, K. H.; Lee, J. T.; Ryu, W. S.; Hwang, W.; Kim, H. N.; Kim, H. I.; Kwon, H. I.; Park, C.; Lee, B. C.; Park, J. M.; Lee, C. S.; Chae, H. T.; Im, N. J.; Cho, M. S.; Im, I. C.; Nam, C.; Lee, D. B.; Goh, Y. M.; Kim, J. D.; Ahn, H. S.; Woo, Y. M.; Chang, S. J.; Cho, H. D.

    1997-09-01

    This project has aimed at the development of U 3 Si dispersion fuel for the localization of HANARO fuel and the application of atomization process to advanced RERTR fuel development. The design criteria were established through the modified computer codes. Design documents were prepared and issued. The acceptable co-extrusion cladding was achieved. The electron beam welding technology has been developed and the sealing of the end plug and cladding was accomplished without defects. The atomization fuel meats have about 200% higher elongation and about 20% higher than comminution fuel meats. The thermal compatibility test showed that atomization fuel have about 30% higher stability that the comminution fuel. The pressure drops of 18 rods fuel assembly and 36 rods fuel assembly were measured to have 213 kPa and 205 kPa respectively. Apparent wear was not found in endurance test. The irradiation fuel was designed and fabricated by using low enriched uranium metal following the developed Q/A system. The safety analysis of irradiation fuel assembly was performed through linear power calculation by using MCNP4A code and centerline temperature calculation by using DIFAIR code. The quality assurance system has been established. The quality inspection technologies were developed. By acquiring the license, low enriched uranium of 100 kg as well as depleted uranium can be used. U 3 Si 2 -Al fuel swelled less than comminution fuel irrespective of temperature and fuel fraction in a compatibility test. The atomized U-10wt.%Mo powder were found to have gamma phase of isotropic structure. Gamma structure remained with a little swelling without any structure change at 400 deg C for 100 hours. Irradiation miniplate and test rig were designed preliminary manufactured. Thermal hydraulic and linear power calculations were performed by using PLTEMP and MCNP4A computer codes respectively. The hydraulic test showed that the pressure drop met the HANARO requirement. The vibration

  19. Nuclear fuel

    International Nuclear Information System (INIS)

    D Hondt, P.

    1998-01-01

    The research and development programme on nuclear fuel at the Belgian Nuclear Research Centre SCK/CEN is described. The objective of this programme is to enhance the quantitative prediction of the operational limits of nuclear fuel and to assess the behaviour of fuel under incidental and accidental conditions. Progress is described in different domains including the modelling of fission gas release in LWR fuel, thermal conductivity, basic physical phenomena, post-irradiation examination for fuel performance assessment, and conceptual studies of incidental and accidental fuel experiments

  20. RU fuel development program for an advanced fuel cycle in Korea

    International Nuclear Information System (INIS)

    Suk, Hochum; Sim, Kiseob; Kim, Bongghi; Inch, W.W.; Page, R.

    1998-01-01

    reactors in Korea. The RU fuel development is an international collaboration between KAERI, AECL and BNFL. It is expected that the work will be completed before 2005, and there should be no impediment to the use of RU fuel in the CANDU 6 reactors on the Wolsong site in Korea, if RU is available and competitive in price with NU and SEU. (author)

  1. The further development of WWER-440 fuel design performance

    International Nuclear Information System (INIS)

    Lushin, V.; Vasilchenko, I.; Ananjev, J.; Abashina, G.

    2011-01-01

    The most distinguished stages in VVER-440 fuel development of the latest ten years are: designing of second generation FA complex; and designing of sheathless working fuel assembly of the third generation (RK-3) which are presented in this report. Designing of fuel assemblies of the second generation and RK-3 is characterized by the tendency to power increase of VVER-440 operating units with V-213-type reactor, that, in turn, has given a stimulus to further design enhancement of fuel assemblies specified. The further development of the second generation fuel assembly design and the change-over to the third generation working assemblies will allow for fuel utilization to be considerably increased under the conditions of application the more long-term fuel cycles for VVER-440 reactors and operation of the Units at the increased power

  2. Development of LWR fuel performance code FEMAXI-6

    International Nuclear Information System (INIS)

    Suzuki, Motoe

    2006-01-01

    LWR fuel performance code: FEMAXI-6 (Finite Element Method in AXIs-symmetric system) is a representative fuel analysis code in Japan. Development history, background, design idea, features of model, and future are stated. Characteristic performance of LWR fuel and analysis code, what is model, development history of FEMAXI, use of FEMAXI code, fuel model, and a special feature of FEMAXI model is described. As examples of analysis, PCMI (Pellet-Clad Mechanical Interaction), fission gas release, gap bonding, and fission gas bubble swelling are reported. Thermal analysis and dynamic analysis system of FEMAXI-6, function block at one time step of FEMAXI-6, analytical example of PCMI in the output increase test by FEMAXI-III, analysis of fission gas release in Halden reactor by FEMAXI-V, comparison of the center temperature of fuel in Halden reactor, and analysis of change of diameter of fuel rod in high burn up BWR fuel are shown. (S.Y.)

  3. Recent developments in the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Wunderer, A.

    1984-01-01

    There is a description of the present situation in each individual area of the nuclear fuel cycle. Further topics are: risk and safety factors and emissions from the fuel cycle, availability and disruptions, waste disposal and the storage of radioactive waste. (UA) [de

  4. HPLC method for rapidly following biodiesel fuel transesterification reaction progress using a core-shell column.

    Science.gov (United States)

    Allen, Samuel J; Ott, Lisa S

    2012-07-01

    There are a wide and growing variety of feedstocks for biodiesel fuel. Most commonly, these feedstocks contain triglycerides which are transesterified into the fatty acid alkyl esters (FAAEs) which comprise biodiesel fuel. While the tranesterification reaction itself is simple, monitoring the reaction progress and reaction products is not. Gas chromatography-mass spectrometry is useful for assessing the FAAE products, but does not directly address either the tri-, di-, or monoglycerides present from incomplete transesterification or the free fatty acids which may also be present. Analysis of the biodiesel reaction mixture is complicated by the solubility and physical property differences among the components of the tranesterification reaction mixture. In this contribution, we present a simple, rapid HPLC method which allows for monitoring all of the main components in a biodiesel fuel transesterification reaction, with specific emphasis on the ability to monitor the reaction as a function of time. The utilization of a relatively new, core-shell stationary phase for the HPLC column allows for efficient separation of peaks with short elution times, saving both time and solvent.

  5. Recent advances in fuel product and manufacturing process development

    International Nuclear Information System (INIS)

    Slember, R.J.; Doshi, P.K.

    1987-01-01

    This paper discusses advancements in commercial nuclear fuel products and manufacturing made by the Westinghouse Electric Corporation in response to the commercial nuclear fuel industry's demand for high reliability, increased plant availability and improved operating flexibility. The features and benefits of Westinghouse's most advanced fuel products--VANTAGE 5 for PWR plants and QUAD+ for BWR plants--are described, as well as 'high performance' fuel concepts now under development for delivery in the late 1980s. The paper also disusses the importance of in-process quality control throughout manufacturing towards reducing product variability and improving fuel reliability. (author)

  6. Siemens fuel gasification technology - solutions and developments

    Energy Technology Data Exchange (ETDEWEB)

    Hannemann, F.; Schingnitz, M.; Schmid, C. [Siemens Fuel Gasification Technology GmbH, Freiberg (Germany)

    2007-07-01

    In 2006, Siemens Power Generation Group acquired the GSP Gasification technology, and renamed it SFGT. The presentation reviews the technology and provides an update on current projects. The future plans for the development of the technology based on extensive experience and comprehensive development work gathered over many years and proven in a number of gasification plants is covered. SFGT operates, at its Freiberg facility, a 5 MWth pilot plant which was built to test prototype designs and to determine process conditions for various feed streams. An overview is given of the results of tests completed on a wide range of carbonaceous materials including all types of solid fuels from lignite to anthracite, as well as brown coal, oil, sludge or biomass, and low-temperature coke or petcoke. The technical focus of the paper is on the unique design features such as the cooling screen and alternative refractory lining, as well as the dense flow feeding system that allows the preferable use of lignite applications.

  7. Development of advanced zirconium fuel cladding

    International Nuclear Information System (INIS)

    Jeong, Young Hwan; Park, S. Y.; Lee, M. H.

    2007-04-01

    This report includes the manufacturing technology developed for HANA TM claddings, a series of their characterization results as well as the results of their in-pile and out-of pile performances tests which were carried out to develop some fuel claddings for a high burn-up (70,000MWd/mtU) which are competitive in the world market. Some of the HANA TM claddings, which had been manufactured based on the results from the 1st and 2nd phases of the project, have been tested in a research reactor in Halden of Norway for an in-pile performance qualification. The results of the in-pile test showed that the performance of the HANA TM claddings for corrosion and creep was better than 50% compared to that of Zircaloy-4 or A cladding. It was also found that the out-of pile performance of the HANA TM claddings for such as LOCA and RIA in some accident conditions corrosion creep, tensile, burst and fatigue was superior or equivalent to that of the Zircaloy-4 or A cladding. The project also produced the other many data which were required to get a license for an in-pile test of HANA TM claddings in a commercial reactor. The data for the qualification or characterization were provided for KNFC to assist their activities to get the license for the in-pile test of HANA TM Lead Test Rods(LTR) in a commercial reactor

  8. Development of Advanced Spent Fuel Management Process

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Chung Seok; Choi, I. K.; Kwon, S. G. (and others)

    2007-06-15

    As a part of research efforts to develop an advanced spent fuel management process, this project focused on the electrochemical reduction technology which can replace the original Li reduction technology of ANL, and we have successfully built a 20 kgHM/batch scale demonstration system. The performance tests of the system in the ACPF hot cell showed more than a 99% reduction yield of SIMFUEL, a current density of 100 mA/cm{sup 2} and a current efficiency of 80%. For an optimization of the process, the prevention of a voltage drop in an integrated cathode, a minimization of the anodic effect and an improvement of the hot cell operability by a modulation and simplization of the unit apparatuses were achieved. Basic research using a bench-scale system was also carried out by focusing on a measurement of the electrochemical reduction rate of the surrogates, an elucidation of the reaction mechanism, collecting data on the partition coefficients of the major nuclides, quantitative measurement of mass transfer rates and diffusion coefficients of oxygen and metal ions in molten salts. When compared to the PYROX process of INL, the electrochemical reduction system developed in this project has comparative advantages in its application of a flexible reaction mechanism, relatively short reaction times and increased process yields.

  9. Development of Advanced Spent Fuel Management Process

    International Nuclear Information System (INIS)

    Seo, Chung Seok; Choi, I. K.; Kwon, S. G.

    2007-06-01

    As a part of research efforts to develop an advanced spent fuel management process, this project focused on the electrochemical reduction technology which can replace the original Li reduction technology of ANL, and we have successfully built a 20 kgHM/batch scale demonstration system. The performance tests of the system in the ACPF hot cell showed more than a 99% reduction yield of SIMFUEL, a current density of 100 mA/cm 2 and a current efficiency of 80%. For an optimization of the process, the prevention of a voltage drop in an integrated cathode, a minimization of the anodic effect and an improvement of the hot cell operability by a modulation and simplization of the unit apparatuses were achieved. Basic research using a bench-scale system was also carried out by focusing on a measurement of the electrochemical reduction rate of the surrogates, an elucidation of the reaction mechanism, collecting data on the partition coefficients of the major nuclides, quantitative measurement of mass transfer rates and diffusion coefficients of oxygen and metal ions in molten salts. When compared to the PYROX process of INL, the electrochemical reduction system developed in this project has comparative advantages in its application of a flexible reaction mechanism, relatively short reaction times and increased process yields

  10. Measurement of thermal neutron distribution of flux in the fuel element cluster - Progress report

    International Nuclear Information System (INIS)

    Krcevinac, S.B.; Takac, S.M.

    1966-12-01

    Relative distribution of thermal neutron flux in the fuel element cluster (19 UO 2 rods with Zr-II cladding, D 2 O moderator and coolant) was measured by newly developed cell perturbation method. The obtained values of mean density ratios are compared to the results obtained by TER-I code using Amouyal - Benoist model [sr

  11. Advanced thermally stable jet fuels. Technical progress report, October 1993--December 1993

    Energy Technology Data Exchange (ETDEWEB)

    Schobert, H.H.; Eser, S.; Song, C.; Hatcher, P.G.; Walsh, P.M.; Coleman, M.M.

    1994-01-01

    The Penn State program in advancd thermally stable coal-based jet fuels has five broad objectives: (1) development of mechanisms of degradation and solids formation; (2) quantitative measurement of growth of sub-micrometer and micrometer-sized particles suspended in fuels during thermal stressing; (3) characterization of carbonaceous deposits by various instrumental and microscopic methods; (4) elucidation of the role of additives in retarding them formation of vcarbonaceous solids; and, (5) assessment of the potential of production of high yields of cycloalkanes by direct liquefaction of coal.

  12. Spent nuclear fuel retrieval system fuel handling development testing. Final report

    International Nuclear Information System (INIS)

    Jackson, D.R.; Meeuwsen, P.V.

    1997-09-01

    Fuel handling development testing was performed in support of the Fuel Retrieval System (FRS) Sub-Project, a subtask of the Spent Nuclear Fuel Project at the Hanford Site in Richland, Washington. The FRS will be used to retrieve and repackage K-Basin Spent Nuclear Fuel (SNF) currently stored in old K-Plant storage basins. The FRS is required to retrieve full fuel canisters from the basin, clean the fuel elements inside the canister to remove excessive uranium corrosion products (or sludge), remove the contents from the canisters and sort the resulting debris, scrap, and fuel for repackaging. The fuel elements and scrap will be collected in fuel storage and scrap baskets in preparation for loading into a multi canister overpack (MCO), while the debris is loaded into a debris bin and disposed of as solid waste. This report describes fuel handling development testing performed from May 1, 1997 through the end of August 1997. Testing during this period was mainly focused on performance of a Schilling Robotic Systems' Conan manipulator used to simulate a custom designed version, labeled Konan, being fabricated for K-Basin deployment. In addition to the manipulator, the camera viewing system, process table layout, and fuel handling processes were evaluated. The Conan test manipulator was installed and fully functional for testing in early 1997. Formal testing began May 1. The purposes of fuel handling development testing were to provide proof of concept and criteria, optimize equipment layout, initialize the process definition, and identify special needs/tools and required design changes to support development of the performance specification. The test program was set up to accomplish these objectives through cold (non-radiological) development testing using simulated and prototype equipment

  13. CANFLEX-RU fuel development programs as one option of advanced fuel cycles in Korea

    International Nuclear Information System (INIS)

    Suk, Ho Chun; Sim, Ki-Seob; Chung, Jang Hwan

    1999-01-01

    development is an international collaboration between KAERI, AECL and BNFL. It is expected that the work will be completed before 2005, and there should be no impediment to the use of RU fuel in the CANDU-6 reactors in Korea, if the RU in the world is available and competitive with NU and SEU on price. (author)

  14. Predictive assessment of the Ukrainian power industry fuel balance structure progress

    Directory of Open Access Journals (Sweden)

    Alexander P. Voinov

    2014-12-01

    Full Text Available The existing structure of the fuel balance in Ukraine does largely determine high prices of such energy products as the electricity and the heat. The main reason lies in the fact that approximately 40% of the fuel balance represents the expensive imported liquid and gaseous fuels. Of course, such balance structure significantly affects the energy security of Ukraine. The article analyzes the three main directions of this situation possible normalization: replacing the imported gas with relatively inexpensive domestic coal burned at boilers and innovative coal-fired burner systems, especially in the low-temperature boilers with fluidized bed furnaces; replacement of imported fuel with respectively cheap pyrolysis gas obtained from the domestic low-grade brown coal; organization of domestic production of shale gas, which industrial production is scheduled for 2020. These promising directions of the national energetics’ fuel balance development can be considered as a priority part of the national tasks of increasing the country’s energy independence.

  15. Development of oxygen scavenger additives for jet fuels

    Energy Technology Data Exchange (ETDEWEB)

    Beaver, B.D.; Demunshi, R.; Sharief, V.; Tian, D.; Teng, Y. [Duquesne Univ., Pittsburgh, PA (United States)

    1995-05-01

    Our current research program is in response to the US Air Force`s FY93 New Initiative entitled {open_quotes}Advanced Fuel Composition and Use.{close_quotes} The critical goal of this initiative is to develop aircraft fuels which can operate at supercritical conditions. This is a vital objective since future aircraft designs will transfer much higher heat loads into the fuel as compared with current heat loads. In this paper it is argued that the thermal stability of most jet fuels would be dramatically improved by the efficient in flight-removal of a fuel`s dissolved oxygen. It is proposed herein to stabilize the bulk fuel by the addition of an additive which will be judiciously designed and programmed to react with oxygen and produce an innocuous product. It is envisioned that a thermally activated reaction will occur, between the oxygen scavenging additive and dissolved oxygen, in a controlled and directed manner. Consequently formation of insoluble thermal degradation products will be limited. It is believed that successful completion of this project will result in the development of a new type of jet fuel additive which will enable current conventional jet fuels to obtain sufficient thermal stability to function in significantly higher temperature regimes. In addition, it is postulated that the successful development of thermally activated oxygen scavengers will also provide the sub-critical thermal stability necessary for future development of endothermic fuels.

  16. Critical experiments supporting underwater storage of tightly packed configurations of spent fuel pins. Technical progress report, January 1-March 31, 1981

    International Nuclear Information System (INIS)

    Hoovler, G.S.; Baldwin, M.N.

    1981-04-01

    Critical experiments are in progress on arrays of 2 1/2% enriched UO 2 fuel pins simulating underwater pin storage of spent power reactor fuel. Pin storage refers to a spent fuel storage concept in which the fuel assemblies are dismantled and the fuel pins are tightly packed into specially designed canisters. These experiments are providing benchmark data with which to validate nuclear codes used to design spent fuel pin storage racks

  17. Canadian fuel development program and recent operational experience

    International Nuclear Information System (INIS)

    Cox, D.S.; Kohn, E.; Lau, J.H.K.; Dicke, G.J.; Macici, N.N.; Sancton, R.W.

    1995-01-01

    This paper provides an overview of the current Canadian CANDU fuel R and D programs and operational experience. The details of operational experience for fuel in Canadian reactors are summarized for the period 1991-1994; excellent fuel performance has been sustained, with steady-state bundle defect rates currently as low as 0.02%. The status of introducing long 37-element bundles, and bundles with rounded bearing pads is reviewed. These minor changes in fuel design have been selectively introduced in response to operational constraints (end-plate cracking and pressure-tube fretting) at Ontario Hydro's Bruce-B and Darlington stations. The R and D programs are generating a more complete understanding of CANDU fuel behaviour, while the CANDU Owners Group (COG) Fuel Technology Program is being re-aligned to a more exclusive focus on the needs of operating stations. Technical highlights and realized benefits from the COG program are summarized. Re-organization of AECL to provide a one-company focus, with an outward looking view to new CANDU markets, has strengthened R and D in advanced fuel cycles. Progress in AECL's key fuel cycle programs is also summarized. (author)

  18. A Development of Ethanol/Percarbonate Membraneless Fuel Cell

    Directory of Open Access Journals (Sweden)

    M. Priya

    2014-01-01

    Full Text Available The electrocatalytic oxidation of ethanol on membraneless sodium percarbonate fuel cell using platinum electrodes in alkaline-acidic media is investigated. In this cell, ethanol is used as the fuel and sodium percarbonate is used as an oxidant for the first time in an alkaline-acidic media. Sodium percarbonate generates hydrogen peroxide in aqueous medium. At room temperature, the laminar-flow-based microfluidic membraneless fuel cell can reach a maximum power density of 18.96 mW cm−2 with a fuel mixture flow rate of 0.3 mL min−2. The developed fuel cell features no proton exchange membrane. The simple planar structured membraneless ethanol fuel cell presents with high design flexibility and enables easy integration of the microscale fuel cell into actual microfluidic systems and portable power applications.

  19. Development of solid oxide fuel cell technology

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Dae Kab; Kim, Sun Jae; Jung, Choong Hwan; Kim, Kyung Hoh; Park, Ji Yun; Oh, Suk Jin [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-01-01

    Solid Oxide Fuel Cell (SOFC) technologies that use zirconium oxide as the electrolyte material were studied in this present report. SOFC exhibits a very high power generation efficiency of over 50 %, and does not discharge pollution materials such as dusts, sulfur dioxide, and nitrogen oxide. Zirconia, Ni/YSZ (yttria stabilized zirconia), and La-Sr-Mn-Oxide materials were developed for the electrolyte material, for the anode, and for the cathode, respectively. After making thin zirconia plate using tape casting process, anode and cathode powders were screen printed on the zirconia plate for fabricating unit cells. A test system composed of a vertical tube furnace, digital multimeter, DC current supplier, and measuring circuit was constructed for testing the unit cell performance. This system was controlled by a home-made computer program. Founded on this unit cell technology and system, a multi-stack SOFC system was studied. This system was composed of 10 unit cells each of them had an electrode area of 40 x 40 mm. Based on this system design, large and thin zirconia plates of 70 x 70 mm in area was fabricated for the electrolyte. Different from in the unit cell system, interconnectors are needed in the multi-stack system for connecting unit cells electrically. For this interconnectors, Inconel 750 alloy was selected, sliced into wafers, machined, surface finished, and then Pt-plated. 55 figs, 8 tabs, 51 refs. (Author).

  20. Development of solid oxide fuel cell technology

    International Nuclear Information System (INIS)

    Kang, Dae Kab; Kim, Sun Jae; Jung, Choong Hwan; Kim, Kyung Hoh; Park, Ji Yun; Oh, Suk Jin

    1995-01-01

    Solid Oxide Fuel Cell (SOFC) technologies that use zirconium oxide as the electrolyte material were studied in this present report. SOFC exhibits a very high power generation efficiency of over 50 %, and does not discharge pollution materials such as dusts, sulfur dioxide, and nitrogen oxide. Zirconia, Ni/YSZ (yttria stabilized zirconia), and La-Sr-Mn-Oxide materials were developed for the electrolyte material, for the anode, and for the cathode, respectively. After making thin zirconia plate using tape casting process, anode and cathode powders were screen printed on the zirconia plate for fabricating unit cells. A test system composed of a vertical tube furnace, digital multimeter, DC current supplier, and measuring circuit was constructed for testing the unit cell performance. This system was controlled by a home-made computer program. Founded on this unit cell technology and system, a multi-stack SOFC system was studied. This system was composed of 10 unit cells each of them had an electrode area of 40 x 40 mm. Based on this system design, large and thin zirconia plates of 70 x 70 mm in area was fabricated for the electrolyte. Different from in the unit cell system, interconnectors are needed in the multi-stack system for connecting unit cells electrically. For this interconnectors, Inconel 750 alloy was selected, sliced into wafers, machined, surface finished, and then Pt-plated. 55 figs, 8 tabs, 51 refs. (Author)

  1. Innovative membrane development for fuel cells

    CSIR Research Space (South Africa)

    Vaivars, G

    2011-10-01

    Full Text Available The innovative membranes for alternative energy devices will be presented. An electrical car is long waited solution to environmental and fuel supply problems in transport. Most probably, the shift from a combustion engine to an electrical car...

  2. Indoor fuel exposure and the lung in both developing and developed countries: An update

    OpenAIRE

    Sood, Akshay

    2012-01-01

    Almost 3 billion people worldwide burn solid fuels indoors. These fuels include biomass and coal. Although indoor solid fuel smoke is likely a greater problem in developing countries, wood burning populations in developed countries may also be at risk from these exposures. Despite the large population at risk worldwide, the effect of exposure to indoor solid fuel smoke has not been adequately studied. Indoor air pollution from solid fuel use is strongly associated with COPD (both emphysema an...

  3. Interfuel: development of fuel fabrication techniques

    International Nuclear Information System (INIS)

    Anon.

    1975-01-01

    On July 5 1972, an understanding was reached between Rijn-Schelde-Verolme NV(RSV), Shell Kernenergie NV, Gemeenschappelijke Kernenergiecentrale Nederland NV (GKN), Comprimo BV and Stichting Reactor Centrum Nederland (RCN), regarding the formation of a company to co-operate concerning the fuel cycle for nuclear reactors-special emphasis being given to the production of fuel elements for light water reactor systems. (Auth.)

  4. Development of fuel cycle technology for molten-salt reactor systems

    International Nuclear Information System (INIS)

    Uhlir, J.

    2006-01-01

    applicable as a starting material for the production of liquid fluoride fuel for Molten-Salt Transmutation Reactor. R and D on Electrochemical separation processes from fluoride melt media is aimed to the final chemical partitioning of selected actinides from lanthanides, passed from the Fluoride Volatility Process and to the 'on-line' reprocessing (partitioning) of the circulating fuel in Molten-Salt Reactors. Besides the two main experimental partitioning activities, the flow-sheeting research is in the focus of interest as well. Current progress in pyrochemical partitioning development allows us to design conceptual flow-sheets dedicated to MSR fresh transuranium fuel processing as well as the MSR spent fuel on-line reprocessing

  5. The Canadian CANDU fuel development program and recent fuel operating experience

    International Nuclear Information System (INIS)

    Lau, J.H.K.; Inch, W.W.R.; Cox, D.S.; Steed, R.G.; Kohn, E.; Macici, N.N.

    1999-01-01

    This paper reviews the performance of the CANDU fuel in the Canadian CANDU reactors in 1997 and 1998. The operating experience demonstrates that the CANDU fuel has performed very well. Over the two-year period, the fuel-bundle defect rate for all bundles irradiated in the Canadian CANDU reactors has remained very low, at between 0.006% to 0.016%. On a fuel element basis, this represents an element defect rate of less than about 0.0005%. One of the reasons for the good fuel performance is the support provided by the Canadian fuel research and development programs. These programs address operational issues and provide evolutionary improvements to the fuel products. The programs consist of the Fuel Technology Program, funded by the CANDU Owners Group, and the Advanced Fuel and Fuel Cycles Technology Program, funded by Atomic Energy of Canada Ltd. These two programs, which have been in place for many years, complement each other by sharing expert resources and experimental facilities. This paper describes the programs in 1999/2000, to provide an overview of the scope of the programs and the issues that these programs address. (author)

  6. Development of advanced LWR fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Yong Hwan; Park, S. Y.; Lee, M. H. [and others

    2000-04-01

    This report describes the results from evaluating the preliminary Zr-based alloys to develop the advanced Zr-based alloys for the nuclear fuel claddings, which should have good corrosion resistance and mechanical properties at high burn-up over 70,000MWD/MTU. It also includes the results from the basic studies for optimizing the processes which are involved in the development of the advanced Zr-based alloys. Ten(10) kinds of candidates for the alloys of which performance is over that of the existing Zircaloy-4 or ZIRLO alloy were selected out of the preliminary alloys of 150 kinds which were newly designed and repeatedly manufactured and evaluated to find out the promising alloys. First of all, the corrosion tests on the preliminary alloys were carried out to evaluate their performance in both pure water and LiOH solution at 360 deg C and in steam at 400 deg C. The tensile tests were performed on the alloys which proved to be good in the corrosion resistance. The creep behaviors were tested at 400 deg C for 10 days with the application of constant load on the samples which showed good performance in the corrosion resistance and tensile properties. The effect of the final heat treatment and A-parameters as well as Sn or Nb on the corrosion resistance, tensile properties, hardness, microstructures of the alloys was evaluated for some alloys interested. The other basic researches on the oxides, electrochemical properties, corrosion mechanism, and the establishment of the phase diagrams of some alloys were also carried out.

  7. Development of advanced LWR fuel cladding

    International Nuclear Information System (INIS)

    Jeong, Yong Hwan; Park, S. Y.; Lee, M. H.

    2000-04-01

    This report describes the results from evaluating the preliminary Zr-based alloys to develop the advanced Zr-based alloys for the nuclear fuel claddings, which should have good corrosion resistance and mechanical properties at high burn-up over 70,000MWD/MTU. It also includes the results from the basic studies for optimizing the processes which are involved in the development of the advanced Zr-based alloys. Ten(10) kinds of candidates for the alloys of which performance is over that of the existing Zircaloy-4 or ZIRLO alloy were selected out of the preliminary alloys of 150 kinds which were newly designed and repeatedly manufactured and evaluated to find out the promising alloys. First of all, the corrosion tests on the preliminary alloys were carried out to evaluate their performance in both pure water and LiOH solution at 360 deg C and in steam at 400 deg C. The tensile tests were performed on the alloys which proved to be good in the corrosion resistance. The creep behaviors were tested at 400 deg C for 10 days with the application of constant load on the samples which showed good performance in the corrosion resistance and tensile properties. The effect of the final heat treatment and A-parameters as well as Sn or Nb on the corrosion resistance, tensile properties, hardness, microstructures of the alloys was evaluated for some alloys interested. The other basic researches on the oxides, electrochemical properties, corrosion mechanism, and the establishment of the phase diagrams of some alloys were also carried out

  8. Coordination of ministerial actions regarding the use of liquefied natural gas as marine fuel. Progress report at 31 August 2014

    International Nuclear Information System (INIS)

    Maler, Philippe; Erhardt, Jean-Bernard; Ourliac, Jean-Paul

    2014-09-01

    This report is the second of a series dealing with the coordination of ministerial actions in favor of the use of liquefied natural gas (LNG) as marine fuel. Tougher sulfur oxides pollution regulations will lead to the progressive abandonment of heavy fuels in maritime propulsion. LNG can meet the future environmental imperatives but its introduction as marine fuel implies important naval and infrastructure investments. This report presents, first, a summary of the report's recommendations and the aim of this coordination study, and, then, treats more thoroughly of the different coordination aspects: 1 - Progresses made by the coordination mission between February 2013 and July 2014 (multiplicity of intervening actors and communication problems); 2 - situation and perspectives of member countries policy having an impact on marine bunker fuels (fuel substitution directive project, marine CO 2 pollution monitoring project, EU's air quality policy and ships emissions, energy and environment policies by 2030, maritime transport and environmental pollution); 3 - rules and standards for LNG-fueled ships supply and exploitation (LNG-fueled ships, fuel supply, regulations, personnel training, European framework); 4 - ships and LNG facilities financing (European programs, financing); 5 - pilot project's situation (Dunkerque harbour, Brittany Ferries project, Montoir and Fos LNG terminals, big and decentralized harbours, Ministry's actions, French projects with European participation)

  9. Development of molecular dynamics potential for uranium silicide fuels

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Jianguo; Zhang, Yongfeng; Hales, Jason D.

    2016-09-01

    Use of uranium–silicide (U-Si) in place of uranium dioxide (UO2) is one of the promising concepts being proposed to increase the accident tolerance of nuclear fuels. This is due to a higher thermal conductivity than UO2 that results in lower centerline temperatures. U-Si also has a higher fissile density, which may enable some new cladding concepts that would otherwise require increased enrichment limits to compensate for their neutronic penalty. However, many critical material properties for U-Si have not been determined experimentally. For example, silicide compounds (U3Si2 and U3Si) are known to become amorphous under irradiation. There was clear independent experimental evidence to support a crystalline to amorphous transformation in those compounds. However, it is still not well understood how the amorphous transformation will affect on fuel behavior. It is anticipated that modeling and simulation may deliver guidance on the importance of various properties and help prioritize experimental work. In order to develop knowledge-based models for use at the engineering scale with a minimum of empirical parameters and increase the predictive capabilities of the developed model, inputs from atomistic simulations are essential. First-principles based density functional theory (DFT) calculations will provide the most reliable information. However, it is probably not possible to obtain kinetic information such as amorphization under irradiation directly from DFT simulations due to size and time limitations. Thus, a more feasible way may be to employ molecular dynamics (MD) simulation. Unfortunately, so far no MD potential is available for U-Si to discover the underlying mechanisms. Here, we will present our recent progress in developing a U-Si potential from ab initio data. This work is supported by the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program funded by the U.S. Department of Energy, Office of Nuclear Energy.

  10. Advances in nuclear fuel technology. 3. Development of advanced nuclear fuel recycle systems

    International Nuclear Information System (INIS)

    Arie, Kazuo; Abe, Tomoyuki; Arai, Yasuo

    2002-01-01

    Fast breeder reactor (FBR) cycle technology has a technical characteristics flexibly easy to apply to diverse fuel compositions such as plutonium, minor actinides, and so on and fuel configurations. By using this characteristics, various feasibilities on effective application of uranium resources based on breeding of uranium of plutonium for original mission of FBR, contribution to radioactive wastes problems based on amounts reduction of transuranium elements (TRU) in high level radioactive wastes, upgrading of nuclear diffusion resistance, extremely upgrading of economical efficiency, and so on. In this paper, were introduced from these viewpoints, on practice strategy survey study on FBR cycle performed by cooperation of the Japan Nuclear Cycle Development Institute (JNC) with electric business companies and so on, and on technical development on advanced nuclear fuel recycle systems carried out at the Central Research Institute of Electric Power Industry, Japan Atomic Energy Research Institute, and so on. Here were explained under a vision on new type of fuels such as nitride fuels, metal fuels, and so on as well as oxide fuels, a new recycle system making possible to use actinides except uranium and plutonium, an 'advanced nuclear fuel cycle technology', containing improvement of conventional wet Purex method reprocessing technology, fuel manufacturing technology, and so on. (G.K.)

  11. Status of the atomized uranium silicide fuel development at KAERI

    Energy Technology Data Exchange (ETDEWEB)

    Kim, C.K.; Kim, K.H.; Park, H.D.; Kuk, I.H. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-08-01

    While developing KMRR fuel fabrication technology an atomizing technique has been applied in order to eliminate the difficulties relating to the tough property of U{sub 3}Si and to take advantage of the rapid solidification effect of atomization. The comparison between the conventionally comminuted powder dispersion fuel and the atomized powder dispersion fuel has been made. As the result, the processes, uranium silicide powdering and heat treatment for U{sub 3}Si transformation, become simplified. The workability, the thermal conductivity and the thermal compatibility of fuel meat have been investigated and found to be improved due to the spherical shape of atomized powder. In this presentation the overall developments of atomized U{sub 3}Si dispersion fuel and the planned activities for applying the atomizing technique to the real fuel fabrication are described.

  12. Thermal fuel research and development facilities in BNFL

    International Nuclear Information System (INIS)

    Roberts, V.A.; Vickers, J.

    1996-01-01

    BNFL is committed to providing high quality, cost effective nuclear fuel cycle services to customers on a National and International level. BNFL's services, products and expertise span the complete fuel cycle; from fuel manufacture through to fuel reprocessing, transport, waste management and decommissioning and the Company maintains its technical and commercial lead by investment in continued research and development (R and D). This paper discusses BNFL's involvement in R and D and gives an account of the current facilities available together with a description of the advanced R and D facilities constructed or planned at Springfields and Sellafield. It outlines the work being carried out to support the company fuel technology business, to (1) develop more cost effective routes to existing fuel products; (2) maximize the use of recycled uranium, plutonium and tails uranium and (3) support a successful MOX business

  13. Development of on-board fuel metering and sensing system

    Science.gov (United States)

    Hemanth, Y.; Manikanta, B. S. S.; Thangaraja, J.; Bharanidaran, R.

    2017-11-01

    Usage of biodiesel fuels and their blends with diesel fuel has a potential to reduce the tailpipe emissions and reduce the dependence on crude oil imports. Further, biodiesel fuels exhibit favourable greenhouse gas emission and energy balance characteristics. While fossil fuel technology is well established, the technological implications of biofuels particularly biodiesel is not clearly laid out. Hence, the objective is to provide an on-board metering control in selecting the different proportions of diesel and bio-diesel blends. An on-board fuel metering system is being developed using PID controller, stepper motors and a capacitance sensor. The accuracy was tested with the blends of propanol-1, diesel and are found to be within 1.3% error. The developed unit was tested in a twin cylinder diesel engine with biodiesel blended diesel fuel. There was a marginal increase (5%) in nitric oxide and 14% increase in smoke emission with 10% biodiesel blended diesel at part load conditions.

  14. The French development program for a UMo fuel

    International Nuclear Information System (INIS)

    Romano, R.; Nigon, J.L.; Languille, A.; Le Borgne, E.; Freslon, H.

    1999-01-01

    Until now high density U 3 Si 2 fuels were satisfactory for LEU conversion of certain reactors, but their use is limited because their density is physically limited to 5,8 gU/cm3 and they have very poor reprocessing capacities. After the end of the present US return policy in may 2006, the reactor operators will be indeed in a very difficult position with silicides. The international community is thus interested in a very high density fuel with good reprocessing capacities in order to convert most reactors and to find a back end solution. In France, CEA, CERCA, and COGEMA have thus launched an important program in order to sort potential candidates of uranium alloys. UMo is one of the most interesting candidates. After the selection of UMo alloys, France has pooled different skills to start an important program on UMo fuels: CEA has started an important project for a new reactor (Jules Horowitz); CERCA is the main manufacturer for MTR fuel; TECHNICATOME is the design expert for research reactors and associated cores; FRAMATOME is the parent company of CERCA and is interested in the development of new reactors; COGEMA is interested in reprocessing spent fuels. This new fuel has three aims: to allow reactors to benefit from a high performing fuel; to have a reprocessable fuel to limit the fuel storage period and the associate safety problem, and solve the back end issue; to support the international effort for non proliferation involving the end of the use of HEU. This high density fuel will decrease the number of fuel assemblies needed to run the reactors and decrease the global cost of the fuel cycle as the back end management cost is in proportion with the quantity of fuel. Reactor operators will thus derive an advantage from this new fuel, in terms of economy

  15. Endplug Welding Techniques developed for SFR Metallic Fuel Elements

    International Nuclear Information System (INIS)

    Lee, Jung Won; Kim, Soo Sung; Woo, Yoon Myeng; Kim, Hyung Tae; Lee, Ho Jin; Kim, Ki Hwan

    2013-01-01

    In Korea, the R and D on SFR has been begun since 1997, as one of the national long-term nuclear R and D programs. The international collaborative research is under way on fuel developments within Advanced Fuel Project for Gen-IV SFR with the closed fuel cycle of full actinide recycling, while TRU bearing metallic fuel, U-TRU-Zr alloy fuel, was selected and is being developed. For the fabrication of SFR metallic fuel elements, the endplug welding is a crucial process. The sealing of endplug to cladding tube should be hermetically perfect to prevent a leakage of fission gases and to maintain a good reactor performance. In this study, the welding technique, welding equipment, welding conditions and parameters were developed to make SFR metallic fuel elements. The TIG welding technique was adopted and the welding joint design was developed. And the optimal welding conditions and parameters were also established. In order to make SFR metallic fuel elements, the welding technique, welding equipment, welding conditions and parameters were developed. The TIG welding technique was adopted and the welding joint design was developed. And the optimal welding conditions and parameters were also established

  16. Endplug Welding Techniques developed for SFR Metallic Fuel Elements

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jung Won; Kim, Soo Sung; Woo, Yoon Myeng; Kim, Hyung Tae; Lee, Ho Jin; Kim, Ki Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    In Korea, the R and D on SFR has been begun since 1997, as one of the national long-term nuclear R and D programs. The international collaborative research is under way on fuel developments within Advanced Fuel Project for Gen-IV SFR with the closed fuel cycle of full actinide recycling, while TRU bearing metallic fuel, U-TRU-Zr alloy fuel, was selected and is being developed. For the fabrication of SFR metallic fuel elements, the endplug welding is a crucial process. The sealing of endplug to cladding tube should be hermetically perfect to prevent a leakage of fission gases and to maintain a good reactor performance. In this study, the welding technique, welding equipment, welding conditions and parameters were developed to make SFR metallic fuel elements. The TIG welding technique was adopted and the welding joint design was developed. And the optimal welding conditions and parameters were also established. In order to make SFR metallic fuel elements, the welding technique, welding equipment, welding conditions and parameters were developed. The TIG welding technique was adopted and the welding joint design was developed. And the optimal welding conditions and parameters were also established.

  17. Site safety progress review of spent fuel central interim storage facility. Final report

    International Nuclear Information System (INIS)

    Gurpinar, A.; Serva, L.; Giuliani

    1995-01-01

    Following the request of the Czech Power Board (CEZ) and within the scope of the Technical Cooperation Project CZR/9/003, a progress review of the site safety of the Spent Fuel Central Interim Storage Facility (SFCISF) was performed. The review involved the first two stages of the works comprising the regional survey and identification of candidate sites for the underground and surface storage options. Five sites have been identified as a result of the previous works. The following two stages will involved the identification of the preferred candidate sites for the two options and the final site qualification. The present review had the purpose of assessing the work already performed and making recommendations for the next two stages of works

  18. Conceptual development of a test facility for spent fuel management

    Energy Technology Data Exchange (ETDEWEB)

    Park, S.W.; Lee, H.H.; Lee, J.Y.; Lee, J.S.; Ro, S.G. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    Spent fuel management is an important issue for nuclear power program, requiring careful planning and implementation. With the wait-and-see policy on spent fuel management in Korea, research efforts are directed at KAERI to develop advanced technologies for safer and more efficient management of the accumulating spent fuels. In support of these research perspectives, a test facility of pilot scale is being developed with provisions for integral demonstration of a multitude of technical functions required for spent fuel management. The facility, baptized SMART (Spent fuel MAnagement technology Research and Test facility), is to be capable of handling full size assembly of spent PWR fuel (as well as CANDU fuel) with a maximum capacity of 10 MTU/y (about 24 assemblies of PWR type). Major functions of the facility are consolidation of spent PWR fuel assembly into a half-volume package and optionally transformation of the fuel rod into a fuel of CANDU type (called DUPIC). Objectives of these functions are to demonstrate volume reduction of spent fuel (for either longer-term dry storage or direct disposal ) in the former case and direct refabrication of the spent PWR fuel into CANDU-type DUPIC fuel for reuse in CANDU reactors in the latter case, respectively. In addition to these major functions, there are other associated technologies to be demonstrated : such as waste treatment, remote maintenance, safeguards, etc. As the facility is to demonstrate not only the functional processes but also the safety and efficiency of the test operations, engineering criteria equivalent to industrial standards are incorporated in the design concept. The hot cell structure enclosing the radioactive materials is configured in such way to maximize costs within the given functional and operational requirements. (author). 3 tabs., 4 figs.

  19. Conceptual development of a test facility for spent fuel management

    International Nuclear Information System (INIS)

    Park, S.W.; Lee, H.H.; Lee, J.Y.; Lee, J.S.; Ro, S.G.

    1997-01-01

    Spent fuel management is an important issue for nuclear power program, requiring careful planning and implementation. With the wait-and-see policy on spent fuel management in Korea, research efforts are directed at KAERI to develop advanced technologies for safer and more efficient management of the accumulating spent fuels. In support of these research perspectives, a test facility of pilot scale is being developed with provisions for integral demonstration of a multitude of technical functions required for spent fuel management. The facility, baptized SMART (Spent fuel MAnagement technology Research and Test facility), is to be capable of handling full size assembly of spent PWR fuel (as well as CANDU fuel) with a maximum capacity of 10 MTU/y (about 24 assemblies of PWR type). Major functions of the facility are consolidation of spent PWR fuel assembly into a half-volume package and optionally transformation of the fuel rod into a fuel of CANDU type (called DUPIC). Objectives of these functions are to demonstrate volume reduction of spent fuel (for either longer-term dry storage or direct disposal ) in the former case and direct refabrication of the spent PWR fuel into CANDU-type DUPIC fuel for reuse in CANDU reactors in the latter case, respectively. In addition to these major functions, there are other associated technologies to be demonstrated : such as waste treatment, remote maintenance, safeguards, etc. As the facility is to demonstrate not only the functional processes but also the safety and efficiency of the test operations, engineering criteria equivalent to industrial standards are incorporated in the design concept. The hot cell structure enclosing the radioactive materials is configured in such way to maximize costs within the given functional and operational requirements. (author). 3 tabs., 4 figs

  20. Technology developments for Japanese BWR MOX fuel utilization

    International Nuclear Information System (INIS)

    Oguma, M.; Mochida, T.; Nomata, T.; Asahi, K.

    1997-01-01

    The Long-Term Program for Research, Development and Utilization of Nuclear Energy established by the Atomic Energy Commission of Japan asserts that Japan will promote systematic utilization of MOX fuel in LWRs. Based on this Japanese nuclear energy policy, we have been pushing development of MOX fuel technology aimed at future full scale utilization of this fuel in BWRs. In this paper, the main R and D topics are described from three subject areas, MOX core and fuel design, MOX fuel irradiation behaviour, and MOX fuel fabrication technology. For the first area, we explain the compatibility of MOX fuel with UO 2 core, the feasibility of the full MOX core, and the adaptability of MOX design methods based on a mock-up criticality experiment. In the second, we outline the Tsuruga MOX irradiation program and the DOMO program, and suggest that MOX fuel behaviour is comparable to ordinary BWR UO 2 fuel behaviour. In the third, we examine the development of a fully automated MOX bundle assembling apparatus and its features. (author). 14 refs, 11 figs, 3 tabs

  1. A state of the art on metallic fuel technology development

    International Nuclear Information System (INIS)

    Hwang, Woan; Kang, Hee Young; Nam, Cheol; Kim, Jong Oh

    1997-01-01

    Since worldwide interest turned toward ceramic fuels before the full potential of metallic fuel could be achieved in the late 1960's, the development of metallic fuels continued throughout the 1970's at ANL's experimental breeder reactor II (EBR-II) because EBR-II continued to be fueled with the metallic uranium-fissium alloy, U-5Fs. During this decade the performance limitations of metallic fuel were satisfactorily resolved resolved at EBR-II. The concept of the IFR developed at ANL since 1984. The technical feasibility had been demonstrated and the technology database had been established to support its practicality. One key features of the IFR is that the fuel is metallic, which brings pronounced benefits over oxide in improved inherent safety and lower processing costs. At the outset of the 1980's, it appeared that metallic fuels are recognized as a professed viable option with regard to safety, integral fuel cycle, waste minimization and deployment economics. This paper reviews the key advances in the last score and summarizes the state-of the art on metallic fuel technology development. (author). 29 refs., 1 tab

  2. A state of the art on metallic fuel technology development

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Woan; Kang, Hee Young; Nam, Cheol; Kim, Jong Oh [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    Since worldwide interest turned toward ceramic fuels before the full potential of metallic fuel could be achieved in the late 1960`s, the development of metallic fuels continued throughout the 1970`s at ANL`s experimental breeder reactor II (EBR-II) because EBR-II continued to be fueled with the metallic uranium-fissium alloy, U-5Fs. During this decade the performance limitations of metallic fuel were satisfactorily resolved resolved at EBR-II. The concept of the IFR developed at ANL since 1984. The technical feasibility had been demonstrated and the technology database had been established to support its practicality. One key features of the IFR is that the fuel is metallic, which brings pronounced benefits over oxide in improved inherent safety and lower processing costs. At the outset of the 1980`s, it appeared that metallic fuels are recognized as a professed viable option with regard to safety, integral fuel cycle, waste minimization and deployment economics. This paper reviews the key advances in the last score and summarizes the state-of the art on metallic fuel technology development. (author). 29 refs., 1 tab.

  3. Development of a fresh MOX fuel transport package for disposition of weapons plutonium

    International Nuclear Information System (INIS)

    Ludwig, S.B.; Pope, R.B.; Shappert, L.B.; Michelhaugh, R.D.; Chae, S.M.

    1998-01-01

    The US Department of Energy announced its Record of Decision on January 14, 1997, to embark on a dual-track approach for disposition of surplus weapons-usable plutonium using immobilization in glass or ceramics and burning plutonium as mixed-oxide (MOX) fuel in reactors. In support of the MOX fuel alternative, Oak Ridge National Laboratory initiated development of conceptual designs for a new package for transporting fresh (unirradiated) MOX fuel assemblies between the MOX fabrication facility and existing commercial light-water reactors in the US. This paper summarizes progress made in development of new MOX transport package conceptual designs. The development effort has included documentation of programmatic and technical requirements for the new package and development and analysis of conceptual designs that satisfy these requirements

  4. Workforce Training and Economic Development Fund: 2015 Annual Progress Report

    Science.gov (United States)

    Iowa Department of Education, 2015

    2015-01-01

    The Department of Education, Division of Community Colleges, will annually provide the State Board of Education with The Workforce Training and Economic Development (WTED) Fund Annual Progress Report. Administration and oversight responsibility for the fund was transferred from the Iowa Economic Development Authority to the Iowa Department of…

  5. The Progressive Development of Environmental Education in Sweden and Denmark

    Science.gov (United States)

    Breiting, Soren; Wickenberg, Per

    2010-01-01

    Our paper traces the history and progressive development of environmental education and education for sustainable development (ESD) in Sweden and Denmark. Our main focus is on work in primary and lower secondary schools as part of a search for trends of international interest related to the conceptualisation and practice of environmental education…

  6. Chemistry research and development. Progress report, December 1978-May 1979

    International Nuclear Information System (INIS)

    Miner, F.J.

    1980-01-01

    Progress and activities are reported on component development, pilot plant development, and instrumentation and statistical systems. Specific items studied include processing of pond sludge, transport of radioactive materials and wastes, corrosion, decontamination and cleaning, fluidized-bed incineration, Pu contamination of soils, chemical analysis, radiometric analysis, security

  7. Liquid fuels production from biomass. Progress report No. 5, July 1-September 30, 1978

    Energy Technology Data Exchange (ETDEWEB)

    Sanderson, J.E.; Wise, D.L.

    1978-01-01

    The current program to convert biomass into liquid hydrocarbon fuels is an extension of the previous program to ferment marine algae to acetic acid. In that study, it was found that marine algae could be converted to higher aliphatic organic acids and that these acids could be readily removed from the fermentation both by membrane or liquid-liquid extraction. It was then proposed to convert these higher organic acids to aliphatic hydrocarbons via Kolbe Electrolysis, which may be used as a diesel fuel. The specific goals for the current program are: (1) Establish conditions under which substrates other than marine algae may be converted in good yield to organic acids. The primary task in this regard is methane suppression. (2) Modify the current 300 liter fixed packed bed batch fermenter to operate in a continuous mode. (3) Change from membrane extraction of organic acids to liquid-liquid extraction. (4) Optimize the energy balance of the electrolytic oxidation process. The primary task in this regard is to reduce the working potential required for the electrolysis while maintaining an adequate current density. (5) Scale the entire process up to match the output of the 300 liter fermenter. The accomplishments in this program are on schedule. Substantial progress has been made on the problem of methane suppression through the use of sulfide addition and the identification of bromoethane-sulfonic acid as a specific inhibitor of methanogenesis. A conceptual design of a continuously fed fixed packed bed fermenter is presented. Experimental results show that the electrolysis of organic acids produced by fermentation to liquid hydrocarbon fuels already have a favorable energy balance of 6/1 based on the applied potential and over 10/1 based on the working potential.

  8. Development of information management system on LWR spent fuel

    International Nuclear Information System (INIS)

    Lee, B. D.; Lee, S. H.; Song, D. Y.; Jeon, I.; Park, S. J.; Seo, D. S.

    2002-01-01

    LWRs in Korea should manage all the information of spent fuel to implement the obligations under Korea-IAEA safeguards agreement and to perform the nuclear material accountancy work at the facility level. The information management system on LWR spent fuel was developed to manage all movement records from receipt to shipment of LWR fuels, and to get the necessary information such as nuclear fuel inventory lists and status, maps of fresh fuel storage, reactor and spent fuel pool, receipt and shipment records and so on. This information management system has a function to setup the system environments to cover the various kinds of storage types for all LWRs ; reactor, spent fuel pool and fresh fuel storage. The movements of nuclear fuel between the storages can be easily done by double click of the mouse to the destination. It also has a several error checking routines for maintaining the correct accounting data. Using this information management system of LWR spent fuel, facility operators can perform efficiently and effectively the safeguards related works including nuclear material accountancy at each facility

  9. Status of LEU fuel development and conversion of NRU

    International Nuclear Information System (INIS)

    Sears, D.F.; Herbert, L.N.; Vaillancourt, K.D.

    1989-11-01

    The status of the low-enrichment uranium (LEU) fuel development and NRU conversion program at Chalk River Nuclear Laboratories is reviewed. Construction of a new fuel fabrication facility is essentially completed and installation of LEW fuel manufacturing equipment has begun. The irradiation of 31 prototype Al-61 wt% U 3 Si dispersion fuel rods, approximately one third of a full NRU core, is continuing without incident. Recent post-irradiation examination of spent fuel rods revealed that the prototype LEU fuel achieved the design burnup (80 at%) in excellent condition, confirming that the Al-U 3 Si 2 dispersion fuel to complement out Al-U 3 Si capability. Three full-size NRU rods containing Al-U 3 Si 2 dispersion fuel have been fabricated for a qualification irradiation in NRU. Post-irradiation examinations of mini-elements containing Al-U 3 Si 2 fuel revealed that the U 3 Si 2 behaved similarly to U 3 Si 2 fuel revealed that the U 3 Si 2 particles and the aluminum matrix, and fission gas bubbles up to 10 μm in diameter, could be seen in the particles after 60 at% and 80 at% burnup. The mini-elements contained a variety of silicide particle sizes; however, no significant swelling dependence on particle size distribution was observed

  10. Development of information management system on LWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, B. D.; Lee, S. H.; Song, D. Y.; Jeon, I.; Park, S. J.; Seo, D. S. [KAERI, Taejon (Korea, Republic of)

    2002-10-01

    LWRs in Korea should manage all the information of spent fuel to implement the obligations under Korea-IAEA safeguards agreement and to perform the nuclear material accountancy work at the facility level. The information management system on LWR spent fuel was developed to manage all movement records from receipt to shipment of LWR fuels, and to get the necessary information such as nuclear fuel inventory lists and status, maps of fresh fuel storage, reactor and spent fuel pool, receipt and shipment records and so on. This information management system has a function to setup the system environments to cover the various kinds of storage types for all LWRs ; reactor, spent fuel pool and fresh fuel storage. The movements of nuclear fuel between the storages can be easily done by double click of the mouse to the destination. It also has a several error checking routines for maintaining the correct accounting data. Using this information management system of LWR spent fuel, facility operators can perform efficiently and effectively the safeguards related works including nuclear material accountancy at each facility.

  11. Learning FuelPHP for effective PHP development

    CERN Document Server

    Tweedie, Ross

    2013-01-01

    The book follows a standard tutorial approach, which will enable readers to use the FuelPHP framework efficiently while developing PHP applications.If you are a PHP developer who is looking to learn more about using the FuelPHP framework for effective PHP development, this book is ideal for you. If you are interested in this book, you should already have a basic understanding of general PHP development.

  12. General-purpose heat source project and space nuclear safety and fuels program. Progress report

    International Nuclear Information System (INIS)

    Maraman, W.J.

    1979-12-01

    This formal monthly report covers the studies related to the use of 238 PuO 2 in radioisotopic power systems carried out for the Advanced Nuclear Systems and Projects Division of the Los Alamos Scientific Laboratory. The two programs involved are general-purpose heat source development and space nuclear safety and fuels. Most of the studies discussed hear are of a continuing nature. Results and conclusions described may change as the work continues

  13. Design and operation of equipment used to develop remote coating capability for HTGR fuel particles

    International Nuclear Information System (INIS)

    Suchomel, R.R.; Stinton, D.P.; Preston, M.K.; Heck, J.L.; Bolfing, B.J.; Lackey, W.J.

    1978-12-01

    Refabrication of HTGR fuels is a manufacturing process that consists of preparation of fuel kernels, application of multiple layers of pyrolytic carbon and silicon carbide, preparation of fuel rods, and assembly of fuel rods into fuel elements. All the equipment for refabrication of 233 U-containing fuel must be designed for completely remote operation and maintenance in hot-cell facilities. Equipment to remotely coated HTGR fuel particles has been designed and operated. Although not all of the equipment development needed for a fully remote coating system has been completed, significant progress has been made. The most important component of the coating furnace is the gas distributor, which must be simple, reliable, and easily maintainable. Techniques for loading and unloading the coater and handling microspheres have been developed. An engineering-scale system, currently in operation, is being used to verify the workability of these concepts. Coating crucible handling components are used to remove the crucible from the furnace, remove coated particles, and exchange the crucible, if necessary. After the batch of particles has been unloaded, it is transferred, weighed, and sampled. The components used in these processes have been tested to ensure that no particle breakage or holdup occurs. Tests of the particle handling system have been very encouraging because no major problems have been encountered. Instrumentation that controls the equipment performed very smoothly and reliably and can be operated remotely

  14. Fuel tax incidence in developing countries. The case of Costa Rica

    Energy Technology Data Exchange (ETDEWEB)

    Blackman, Allen [Resources for the Future, 1616 P Street, N.W. Washington, DC (United States); Environment for Development Center for Central America, Turriabla (Costa Rica); Osakwe, Rebecca; Alpizar, Francisco [Environment for Development Center for Central America, Turriabla (Costa Rica)

    2010-05-15

    Although fuel taxes are a practical means of curbing vehicular air pollution, congestion, and accidents in developing countries - all of which are typically major problems - they are often opposed on distributional grounds. Yet few studies have investigated fuel tax incidence in a developing country context. We use household survey data and income-outcome coefficients to analyze fuel tax incidence in Costa Rica. We find that the effect of a 10% fuel price hike through direct spending on gasoline would be progressive, its effect through spending on diesel - both directly and via bus transportation - would be regressive (mainly because poorer households rely heavily on buses), and its effect through spending on goods other than fuel and bus transportation would be relatively small, albeit regressive. Finally, we find that the overall effect of a 10% fuel price hike through all types of direct and indirect spending would be neutral and the magnitude of this combined effect would be modest. We conclude that distributional concerns need not rule out using fuel taxes to address pressing public health and safety problems, particularly if gasoline and diesel taxes can be differentiated. (author)

  15. Fuel tax incidence in developing countries. The case of Costa Rica

    International Nuclear Information System (INIS)

    Blackman, Allen; Osakwe, Rebecca; Alpizar, Francisco

    2010-01-01

    Although fuel taxes are a practical means of curbing vehicular air pollution, congestion, and accidents in developing countries - all of which are typically major problems - they are often opposed on distributional grounds. Yet few studies have investigated fuel tax incidence in a developing country context. We use household survey data and income-outcome coefficients to analyze fuel tax incidence in Costa Rica. We find that the effect of a 10% fuel price hike through direct spending on gasoline would be progressive, its effect through spending on diesel - both directly and via bus transportation - would be regressive (mainly because poorer households rely heavily on buses), and its effect through spending on goods other than fuel and bus transportation would be relatively small, albeit regressive. Finally, we find that the overall effect of a 10% fuel price hike through all types of direct and indirect spending would be neutral and the magnitude of this combined effect would be modest. We conclude that distributional concerns need not rule out using fuel taxes to address pressing public health and safety problems, particularly if gasoline and diesel taxes can be differentiated. (author)

  16. Fuel tax incidence in developing countries: The case of Costa Rica

    Energy Technology Data Exchange (ETDEWEB)

    Blackman, Allen, E-mail: blackman@rff.or [Resources for the Future, 1616 P Street, N.W. Washington, DC (United States); Environment for Development Center for Central America, Turriabla (Costa Rica); Osakwe, Rebecca; Alpizar, Francisco [Environment for Development Center for Central America, Turriabla (Costa Rica)

    2010-05-15

    Although fuel taxes are a practical means of curbing vehicular air pollution, congestion, and accidents in developing countries-all of which are typically major problems-they are often opposed on distributional grounds. Yet few studies have investigated fuel tax incidence in a developing country context. We use household survey data and income-outcome coefficients to analyze fuel tax incidence in Costa Rica. We find that the effect of a 10% fuel price hike through direct spending on gasoline would be progressive, its effect through spending on diesel-both directly and via bus transportation-would be regressive (mainly because poorer households rely heavily on buses), and its effect through spending on goods other than fuel and bus transportation would be relatively small, albeit regressive. Finally, we find that the overall effect of a 10% fuel price hike through all types of direct and indirect spending would be neutral and the magnitude of this combined effect would be modest. We conclude that distributional concerns need not rule out using fuel taxes to address pressing public health and safety problems, particularly if gasoline and diesel taxes can be differentiated.

  17. Development of an instantaneous local fuel-concentration measurement probe: an engine application

    Science.gov (United States)

    Guibert, P.; Boutar, Z.; Lemoyne, L.

    2003-11-01

    This work presents a new tool which can deliver instantaneous local measurements of fuel concentration in an engine cylinder with a high temporal resolution, particularly during compression strokes. Fuel concentration is represented by means of equivalence fuel-air ratio, i.e. the real engine mass ratio of fuel to air divided by the same ratio in ideal stoichiometry conditions. Controlling the mixture configuration for any strategy in a spark ignition engine and for auto-ignition combustion has a dominant effect on the subsequent processes of ignition, flame propagation and auto-ignition combustion progression, pollutant formation under lean or even stoichiometric operating conditions. It is extremely difficult, under a transient operation, to control the equivalence air/fuel ratio precisely at a required value and at the right time. This requires the development of a highly accurate equivalence air/fuel ratio control system and a tool to measure using crank angle (CA) resolution. Although non-intrusive laser techniques have considerable advantages, they are most of the time inappropriate due to their optical inaccessibility or the complex experimental set-up involved. Therefore, as a response to the demand for a relatively simple fuel-concentration measurement system a probe is presented that replaces a spark plug and allows the engine to run completely normally. The probe is based on hot-wire like apparatus, but involves catalytic oxidation at the wire surface. The development, characteristics and calibration of the probe are presented followed by applications to in-cylinder engine measurements.

  18. Selection and development of advanced nuclear fuel products

    International Nuclear Information System (INIS)

    Stucker, David L.; Miller, Richard S.; Arnsberger, Peter L.

    2004-01-01

    The highly competitive international marketplace requires a continuing product development commitment, short development cycle times and timely, on-target product development to assure customer satisfaction and continuing business. Westinghouse has maintained its leadership position within the nuclear fuel industry with continuous developments and improvements to fuel assembly materials and design. This paper presents a discussion of the processes used by Westinghouse in the selection and refinement of advanced concepts for deployment in the highly competitive US and international nuclear fuel fabrication marketplace. (author)

  19. Development of fabrication technology for CANDU advanced fuel -Development of the advanced CANDU technology-

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Chang Beom; Kim, Hyeong Soo; Kim, Sang Won; Seok, Ho Cheon; Shim, Ki Seop; Byeon, Taek Sang; Jang, Ho Il; Kim, Sang Sik; Choi, Il Kwon; Cho, Dae Sik; Sheo, Seung Won; Lee, Soo Cheol; Kim, Yoon Hoi; Park, Choon Ho; Jeong, Seong Hoon; Kang, Myeong Soo; Park, Kwang Seok; Oh, Hee Kwan; Jang, Hong Seop; Kim, Yang Kon; Shin, Won Cheol; Lee, Do Yeon; Beon, Yeong Cheol; Lee, Sang Uh; Sho, Dal Yeong; Han, Eun Deok; Kim, Bong Soon; Park, Cheol Joo; Lee, Kyu Am; Yeon, Jin Yeong; Choi, Seok Mo; Shon, Jae Moon [Korea Atomic Energy Res. Inst., Taejon (Korea, Republic of)

    1994-07-01

    The present study is to develop the advanced CANDU fuel fabrication technologies by means of applying the R and D results and experiences gained from localization of mass production technologies of CANDU fuels. The annual portion of this year study includes following: 1. manufacturing of demo-fuel bundles for out-of-pile testing 2. development of technologies for the fabrication and inspection of advanced fuels 3. design and munufacturing of fuel fabrication facilities 4. performance of fundamental studies related to the development of advanced fuel fabrication technology.

  20. Development of a proton exchange membrane fuel cell cogeneration system

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Jenn Jiang; Zou, Meng Lin [Department of Greenergy, National University of Tainan, Tainan 700 (China)

    2010-05-01

    A proton exchange membrane fuel cell (PEMFC) cogeneration system that provides high-quality electricity and hot water has been developed. A specially designed thermal management system together with a microcontroller embedded with appropriate control algorithm is integrated into a PEM fuel cell system. The thermal management system does not only control the fuel cell operation temperature but also recover the heat dissipated by FC stack. The dynamic behaviors of thermal and electrical characteristics are presented to verify the stability of the fuel cell cogeneration system. In addition, the reliability of the fuel cell cogeneration system is proved by one-day demonstration that deals with the daily power demand in a typical family. Finally, the effects of external loads on the efficiencies of the fuel cell cogeneration system are examined. Results reveal that the maximum system efficiency was as high as 81% when combining heat and power. (author)

  1. Dry refabrication technology development of spent nuclear fuel

    International Nuclear Information System (INIS)

    Park, Geun Il; Lee, J. W.; Song, K. C.

    2012-04-01

    Key technologies highly applicable to the development of advanced nuclear fuel cycle for the spent fuel recycling were developed using spent fuel and simulated spent fuel (SIMFUEL). In the frame work of dry process oxide products fabrication and the property characteristics of dry process products, hot cell experimental data for decladding, powdering and oxide product fabrication from low and high burnup spent fuel have been produced, basic technology for fabrication of spent fuel standard material has been developed, and remotely modulated welding equipment has been designed and fabricated. Also, fabrication technology of simulated dry process products was established and property models were developed based on reproducible property measurement data. In the development of head-end technology for dry refabrication of spent nuclear fuel and key technologies for volume reduction of head-end process waste which are essential in back-end fuel cycle field including pyro-processing, advanced head-end unit process technology development includes the establishment of experimental conditions for synthesis of porous fuel particles using a granulating furnace and for preparation of UO2 pellets, and fabrication and performance demonstration of engineering scale equipment for off-gas treatment of semi-volatile nuclides, and development of phosphate ceramic technology for immobilization of used filters. Radioactivation characterization and treatment equipment design of metal wastes from pretreatment process was conducted, and preliminary experiments of chlorination/electrorefining techniques for the treatment of hull wastes were performed. Based on the verification of the key technologies for head-end process via the hot-cell tests using spent nuclear fuel, pre-conceptual design for the head-end equipments was performed

  2. Dry refabrication technology development of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Park, Geun Il; Lee, J. W.; Song, K. C.; and others

    2012-04-15

    Key technologies highly applicable to the development of advanced nuclear fuel cycle for the spent fuel recycling were developed using spent fuel and simulated spent fuel (SIMFUEL). In the frame work of dry process oxide products fabrication and the property characteristics of dry process products, hot cell experimental data for decladding, powdering and oxide product fabrication from low and high burnup spent fuel have been produced, basic technology for fabrication of spent fuel standard material has been developed, and remotely modulated welding equipment has been designed and fabricated. Also, fabrication technology of simulated dry process products was established and property models were developed based on reproducible property measurement data. In the development of head-end technology for dry refabrication of spent nuclear fuel and key technologies for volume reduction of head-end process waste which are essential in back-end fuel cycle field including pyro-processing, advanced head-end unit process technology development includes the establishment of experimental conditions for synthesis of porous fuel particles using a granulating furnace and for preparation of UO2 pellets, and fabrication and performance demonstration of engineering scale equipment for off-gas treatment of semi-volatile nuclides, and development of phosphate ceramic technology for immobilization of used filters. Radioactivation characterization and treatment equipment design of metal wastes from pretreatment process was conducted, and preliminary experiments of chlorination/electrorefining techniques for the treatment of hull wastes were performed. Based on the verification of the key technologies for head-end process via the hot-cell tests using spent nuclear fuel, pre-conceptual design for the head-end equipments was performed.

  3. Used fuel disposition research and development roadmap - FY10 status.

    Energy Technology Data Exchange (ETDEWEB)

    Nutt, W. M. (Nuclear Engineering Division)

    2010-10-01

    Since 1987 the U.S. has focused research and development activities relevant to the disposal of commercial used nuclear fuel and U.S. Department of Energy (DOE) owned spent nuclear fuel and high level waste on the proposed repository at Yucca Mountain, Nevada. At the same time, the U.S. successfully deployed a deep geologic disposal facility for defense-related transuranic waste in bedded salt at the Waste Isolation Pilot Plant. In 2009 the DOE established the Used Fuel Disposition Campaign (UFDC) within the Office of Nuclear Energy. The Mission of the UFDC is to identify alternatives and conduct scientific research and technology development to enable storage, transportation and disposal of used nuclear fuel and wastes generated by existing and future nuclear fuel cycles. The U.S. national laboratories have participated on these programs and has conducted research and development related to these issues to a limited extent. However, a comprehensive research and development (R&D) program investigating a variety of geologic media has not been a part of the U.S. waste management program since the mid 1980s. Such a comprehensive R&D program is being developed in the UFDC with a goal of meeting the UFDC Grand Challenge to provide a sound technical basis for absolute confidence in the safety and security of long-term storage, transportation, and disposal of used nuclear fuel and wastes from the nuclear energy enterprise. The DOE has decided to no longer pursue the development of a repository at Yucca Mountain, Nevada. Since a repository site will ultimately have to be selected, sited, characterized, designed, and licensed, other disposal options must now be considered. In addition to the unsaturated volcanic tuff evaluated at Yucca Mountain, several different geologic media are under investigation internationally and preliminary assessments indicate that disposal of used nuclear fuel and high level waste in these media is feasible. Considerable progress has been made in

  4. Used fuel disposition research and development roadmap - FY10 status

    International Nuclear Information System (INIS)

    Nutt, W.M.

    2010-01-01

    Since 1987 the U.S. has focused research and development activities relevant to the disposal of commercial used nuclear fuel and U.S. Department of Energy (DOE) owned spent nuclear fuel and high level waste on the proposed repository at Yucca Mountain, Nevada. At the same time, the U.S. successfully deployed a deep geologic disposal facility for defense-related transuranic waste in bedded salt at the Waste Isolation Pilot Plant. In 2009 the DOE established the Used Fuel Disposition Campaign (UFDC) within the Office of Nuclear Energy. The Mission of the UFDC is to identify alternatives and conduct scientific research and technology development to enable storage, transportation and disposal of used nuclear fuel and wastes generated by existing and future nuclear fuel cycles. The U.S. national laboratories have participated on these programs and has conducted research and development related to these issues to a limited extent. However, a comprehensive research and development (R and D) program investigating a variety of geologic media has not been a part of the U.S. waste management program since the mid 1980s. Such a comprehensive R and D program is being developed in the UFDC with a goal of meeting the UFDC Grand Challenge to provide a sound technical basis for absolute confidence in the safety and security of long-term storage, transportation, and disposal of used nuclear fuel and wastes from the nuclear energy enterprise. The DOE has decided to no longer pursue the development of a repository at Yucca Mountain, Nevada. Since a repository site will ultimately have to be selected, sited, characterized, designed, and licensed, other disposal options must now be considered. In addition to the unsaturated volcanic tuff evaluated at Yucca Mountain, several different geologic media are under investigation internationally and preliminary assessments indicate that disposal of used nuclear fuel and high level waste in these media is feasible. Considerable progress has been

  5. Oxide fuel fabrication technology development of the FaCT project (1). Overall review of fuel technology development of the FaCT project

    International Nuclear Information System (INIS)

    Abe, Tomoyuki; Namekawa, Takashi; Tanaka, Kenya

    2011-01-01

    The FaCT project is in progress in Japan for the commercialization of fast reactor cycle system. The development goal of the fuel in the FaCT project is a low-decontaminated TRU homo-recycling in a closed cycle and extension in average discharge burn-up to 150 GWd/t. Research and development on innovative technologies concerning the short process, remote maintenance and cooling system of automatic fuel production equipments, long life cladding material and control of oxygen potential have been conducted in phase I of the FaCT project. As the result of various test including 600 g batch MOX tests, it is concluded that the short process is available to fuel pellet fabrication of the FaCT project. Although cold mock-up tests on test model of some typical process equipments suggest possibilities of remote maintenance of automatic fuel fabrication equipment, it is concluded that it still needs further efforts to judge the operability of the completely remote fabrication for low-decontaminated TRU fuel. A cold mock-up test on fuel pin assembling equipment show that influence of decay heat of MA can be managed by cooling system. Irradiation tests in BOR-60 indicate that 9Cr-ODS possess the satisfactory in-reactor performance as the long life cladding material if homogeneity of alloy element is adequately controlled. Modification of cladding tube fabrication process to ensure homogeneity and further development of measures to control oxygen potential inside the fuel pin are necessary to reach the burn-up target of the FaCT project. (author)

  6. Development of alternative fuel for pressurized water reactors

    International Nuclear Information System (INIS)

    Cardoso, P.E.; Ferreira, R.A.N.; Ferraz, W.B.; Lameiras, F.S.; Santos, A.; Assis, G. de; Doerr, W.O.; Wehner, E.L.

    1984-01-01

    The utilization of alternative fuel cycles in Pressurized Water Reactors (PWR) such as Th/U and Th/Pu cycles can permit a better utilization of uranium reserves without the necessity of developing new power reactor concepts. The development of the technology of alternative fuels for PWR is one of the objectives of the 'Program on Thorium Utilization in Pressurized Water Reactors' carried out jointly by Empresas Nucleares Brasileiras S.A. (NUCLEBRAS), through its Centro de Desenvolvimento da Tecnologia Nuclear (CDTN) and by German institutions, the Julich Nuclear Research Center (KFA), the Kraftwerk Union A.G. (KWU) and NUKEM GmbH. This paper summarizes the results so far obtained in the fuel technology. The development of a fabrication process for PWR fuel pellets from gel-microspheres is reported as well as the design, the specification, and the fabrication of prototype fuel rods for irradiation tests. (Author) [pt

  7. Development of wire wrapping technology for FBR fuel pin

    International Nuclear Information System (INIS)

    Nogami, Tetsuya; Seki, Nobuo; Sawayama, Takeo; Ishibashi, Takashi

    1991-01-01

    For the FBR fuel assembly, the spacer wire is adopted to maintain the space between fuel pins. The developments have been carried out to achieve automatically wire wrapping with high precision. Based on the fundamental technology developed through the mock-up test operation, Joyo 'MK-I', fuel pin fabrication was started using partially mechanized wire wrapping machine in 1973. In 1978, an automated wire wrapping machine for Joyo 'MK-II' was developed by the adoption of some improvements for the wire inserting system to end plug hole and the precision of wire pitch. On the bases of these experiences, fully automated wire wrapping machine for 'Monju' fuel pin was installed at Plutonium Fuel Production Facility (PFPF) in 1987. (author)

  8. Deep-Burn Modular Helium Reactor Fuel Development Plan

    Energy Technology Data Exchange (ETDEWEB)

    McEachern, D

    2002-12-02

    This document contains the workscope, schedule and cost for the technology development tasks needed to satisfy the fuel and fission product transport Design Data Needs (DDNs) for the Gas Turbine-Modular Helium Reactor (GT-MHR), operating in its role of transmuting transuranic (TRU) nuclides in spent fuel discharged from commercial light-water reactors (LWRs). In its application for transmutation, the GT-MHR is referred to as the Deep-Burn MHR (DB-MHR). This Fuel Development Plan (FDP) describes part of the overall program being undertaken by the U.S. Department of Energy (DOE), utilities, and industry to evaluate the use of the GT-MHR to transmute transuranic nuclides from spent nuclear fuel. The Fuel Development Plan (FDP) includes the work on fuel necessary to support the design and licensing of the DB-MHR. The FDP is organized into ten sections. Section 1 provides a summary of the most important features of the plan, including cost and schedule information. Section 2 describes the DB-MHR concept, the features of its fuel and the plan to develop coated particle fuel for transmutation. Section 3 describes the knowledge base for fabrication of coated particles, the experience with irradiation performance of coated particle fuels, the database for fission product transport in HTGR cores, and describes test data and calculations for the performance of coated particle fuel while in a repository. Section 4 presents the fuel performance requirements in terms of as-manufactured quality and performance of the fuel coatings under irradiation and accident conditions. These requirements are provisional because the design of the DB-MHR is in an early stage. However, the requirements are presented in this preliminary form to guide the initial work on the fuel development. Section 4 also presents limits on the irradiation conditions to which the coated particle fuel can be subjected for the core design. These limits are based on past irradiation experience. Section 5 describes

  9. DUPIC nuclear fuel manufacturing and process technology development

    International Nuclear Information System (INIS)

    Yang, Myung Seung; Park, J. J.; Lee, J. W.

    2000-05-01

    In this study, DUPIC fuel fabrication technology and the active fuel laboratory were developed for the study of spent nuclear fuel. A new nuclear fuel using highly radioactive nuclear materials can be studied at the active fuel laboratory. Detailed DUPIC fuel fabrication process flow was developed considering the manufacturing flow, quality control process and material accountability. The equipment layout of about twenty DUPIC equipment at IMEF M6 hot cell was established for the minimization of the contamination during DUPIC processes. The characteristics of the SIMFUEL powder and pellets was studied in terms of milling conditions. The characteristics of DUPIC powder and pellet was studied by using 1 kg of spent PWR fuel at PIEF nr.9405 hot cell. The results were used as reference process conditions for following DUPIC fuel fabrication at IMEF M6. Based on the reference fabrication process conditions, the main DUPIC pellet fabrication campaign has been started at IMEF M6 using 2 kg of spent PWR fuel since 2000 January. As of March 2000, about thirty DUPIC pellets were successfully fabricated

  10. DUPIC nuclear fuel manufacturing and process technology development

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Myung Seung; Park, J. J.; Lee, J. W. [and others

    2000-05-01

    In this study, DUPIC fuel fabrication technology and the active fuel laboratory were developed for the study of spent nuclear fuel. A new nuclear fuel using highly radioactive nuclear materials can be studied at the active fuel laboratory. Detailed DUPIC fuel fabrication process flow was developed considering the manufacturing flow, quality control process and material accountability. The equipment layout of about twenty DUPIC equipment at IMEF M6 hot cell was established for the minimization of the contamination during DUPIC processes. The characteristics of the SIMFUEL powder and pellets was studied in terms of milling conditions. The characteristics of DUPIC powder and pellet was studied by using 1 kg of spent PWR fuel at PIEF nr.9405 hot cell. The results were used as reference process conditions for following DUPIC fuel fabrication at IMEF M6. Based on the reference fabrication process conditions, the main DUPIC pellet fabrication campaign has been started at IMEF M6 using 2 kg of spent PWR fuel since 2000 January. As of March 2000, about thirty DUPIC pellets were successfully fabricated.

  11. Preliminary investigation study of code of developed country for developing Korean fuel cycle code

    International Nuclear Information System (INIS)

    Jeong, Chang Joon; Ko, Won Il; Lee, Ho Hee; Cho, Dong Keun; Park, Chang Je

    2012-01-01

    In order to develop Korean fuel cycle code, the analyses has been performed with the fuel cycle codes which are used in advanced country. Also, recommendations were proposed for future development. The fuel cycle codes are AS FLOOWS: VISTA which has been developed by IAEA, DANESS code which developed by ANL and LISTO, and VISION developed by INL for the Advanced Fuel Cycle Initiative (AFCI) system analysis. The recommended items were proposed for software, program scheme, material flow model, isotope decay model, environmental impact analysis model, and economics analysis model. The described things will be used for development of Korean nuclear fuel cycle code in future

  12. Development of fuel performance and thermal hydraulic technology

    International Nuclear Information System (INIS)

    Jung, Youn Ho; Song, K. N.; Kim, H. K. and others

    2000-03-01

    Space grid in LWR fuel assembly is a key structural component to support fuel rods and to enhance heat transfer from fuel rod to the coolant. Therefore, the original spacer grid has been developed. In addition, new phenomena in fuel behavior occurs at the high burnup, so that models to analyze those new phenomena were developed. Results of this project can be summarized as follows. - Seven different spacer grid candidates have been invented and submitted for domestic and US patents. Spacer grid test specimen(3x3 array and 5x5 array) were fabricated for each candidate and the mechanical tests were performed. - Basic technologies in the mechanical and thermal hydraulic behavior in the spacer grid development are studied and relevant test facilities were established - Fuel performance analysis models and programs were developed for the high burnup pellet and cladding, and fuel performance data base were compiled - Procedures of fuel characterization and in-/out of-pile tests were prepared - Conceptual design of fuel rod for integral PWR was carried out. (author)

  13. New developments in dry spent fuel storage

    International Nuclear Information System (INIS)

    Bonnet, C.; Chevalier, Ph.

    2001-01-01

    As shown in various new examples, HABOG facility (Netherlands), CERNAVODA (Candu - Romania), KOZLODUY (WWER - Bulgaria), CHERNOBYL ( RMBK - Ukraine), MAYAK (Spent Fuel from submarine and Icebreakers - Russia), recent studies allow to confirm the flexibility and performances of the CASCAD system proposed by SGN, both in safety and operability, for the dry storage of main kinds of spent fuel. The main features are: A multiple containment barrier system: as required by international regulation, 2 independent barriers are provided (tight canister and storage pit); Passive cooling, while the Fuel Assemblies are stored in an inert atmosphere and under conditions of temperature preventing from degradation of rod cladding; Sub-criticality controlled by adequate arrangements in any conditions; Safe facility meeting ICPR 60 Requirements as well as all applicable regulations (including severe weather conditions and earthquake); Safe handling operations; Retrievability of the spent fuel either during storage period or at the end of planned storage period (100 years); Future Decommissioning of the facility facilitated through design optimisation; Construction and operating cost-effectiveness. (author)

  14. Numerical simulation of progressive inlet orifices in boiling water reactor fuel

    International Nuclear Information System (INIS)

    Lundgren, Sara

    2004-07-01

    This thesis was carried out at Forsmark Nuclear Power Plant. The power plant in Forsmark consists of three boiling water reactors (BWR) which produce about 17% of Swedish electricity. In a BWR the nuclear reactions are used to boil water inside the reactor vessel. The water works both as a coolant and as a moderator and the resulting steam is used directly to run the turbines. A problem when running a BWR at low flow conditions is the density wave oscillations that might occur to the water flow inside the fuel assemblies. These oscillations arise due to the connection between power and flow rate in a heated channel with two-phase flow. In order to improve the stability performance of the channel an orifice plate is placed at the inlet of each fuel assembly. Today these orifice plates have sharp edges and a constant resistance coefficient. Experimental work has been done with progressive orifices, the edge of which is half-oval in shape. The advantage of progressive orifices is the lower pressure losses with an increase of the Reynolds number, a similar phenomenon that appears in external flow around curved bodies. Since there are high costs associated with experimental generation of high- temperature and high-pressure data, it is of some interest to be able to reproduce and generate data using Computational Fluid Dynamics (CFD). This work deals with the possibility to use the CFD-code Fluent to do numerical simulations of the flow through progressive orifices. The following conclusions may be drawn from the numerical results: All simulations using Reynolds-Averaged Navier-Stokes (RANS) turbulence models, two-dimensional and three-dimensional, capture an abrupt decrease of the resistance coefficient at higher Reynolds numbers. Two-equation models seem to under-predict the critical Reynolds number. The five-equation Reynolds Stress Model (RSM) gives a critical Reynolds number of the same order of magnitude of that measured in experiments. No major differences have

  15. Full size U-10Mo monolithic fuel foil and fuel plate fabrication-technology development

    International Nuclear Information System (INIS)

    Moore, G.A.; Jue, J-F.; Rabin, B.H.; Nilles, M.J.

    2010-01-01

    Full-size U-10Mo foils are being developed for use in high density LEU monolithic fuel plates. The application of a zirconium barrier layer to the foil is performed using a hot co-rolling process. Aluminium clad fuel plates are fabricated using Hot Isostatic Pressing (HIP) or a Friction Bonding (FB) process. An overview is provided of ongoing technology development activities, including: the co-rolling process, foil shearing/slitting and polishing, cladding bonding processes, plate forming, plate-assembly swaging, and fuel plate characterization. Characterization techniques being employed include, Ultrasonic Testing (UT), radiography, and microscopy. (author)

  16. Development of advanced spent fuel management process. System analysis of advanced spent fuel management process

    International Nuclear Information System (INIS)

    Ro, S.G.; Kang, D.S.; Seo, C.S.; Lee, H.H.; Shin, Y.J.; Park, S.W.

    1999-03-01

    The system analysis of an advanced spent fuel management process to establish a non-proliferation model for the long-term spent fuel management is performed by comparing the several dry processes, such as a salt transport process, a lithium process, the IFR process developed in America, and DDP developed in Russia. In our system analysis, the non-proliferation concept is focused on the separation factor between uranium and plutonium and decontamination factors of products in each process, and the non-proliferation model for the long-term spent fuel management has finally been introduced. (Author). 29 refs., 17 tabs., 12 figs

  17. AFR spent fuel storage program. Technical progress report, January 1981-March 1981

    International Nuclear Information System (INIS)

    1981-01-01

    Work on this project is focused on developing design and licensing information for the model facility. Deliverables were submitted to DOE for facility modification, security design, and licensing documentation. Work is essentially complete on the high-density rack design with deliverable planned for April 30, 1981. A continuing program for completion of design and licensing information development was initiated by AGNS in February 1981. Progress to date is satisfactory

  18. Advances in carbide fuel element development for fast reactor application

    International Nuclear Information System (INIS)

    Dienst, W.; Kleykamp, H.; Muehling, G.; Reiser, H.; Steiner, H.; Thuemmler, F.; Wedermeyer, H.; Weimar, P.

    1977-01-01

    The features of the carbide fuel development programme are reviewed and evaluated. Single pin and bundle irradiations are carried out under thermal, epithermal and fast flux conditions, the latter in the DFR and KNK-II reactors. Several fuel concepts in the region of representative SNR clad temperatures are compared by parameter and performance tests. A conservative concept is based on He-bonded 8 mm pins with (U,Pu)C pellets and a smear density of 75% TD, operating at 800 W/cm rod power and burnup to 70 MWd/kg. The preparation of mixed carbide fuels is carried out by carbothermic reduction of the oxides in different methods supported by equivalent carbon content, grain size and phase distribution analysis. The fuel for subassembly performance tests is produced in a pilot plant of 0,5 t/year capacity. Compatibility studies reveal that cladding carburization is the only chemical interaction with carbide fuels. This effect leads to a reduction in ductility of the stainless steel. Fission products apparently play no role in the compatibility behaviour. Comprehensive studies lead to reliable information on the chemical and thermodynamic state of the fuel under irradiation. The swelling of carbide fuels and the fission gas release are examined and analysed. Cladding plastic strain by fuel swelling occurs during steady-state operation because the irradiation creep is rather slow compared to oxide fuels. The cladding strain observed depends on the fuel porosity and the cladding strength. The development of carbide fuel pins is complemented by the application of comprehensive computer models. In addition to the steady-state tests power cycling and safety tests are under performance. Up to 1980 the results are summarized for the final design and specification. The development target of the present program is to fabricate several subassemblies for test operation in the SNR 300 by 1981

  19. Development of high performance hybrid rocket fuels

    Science.gov (United States)

    Zaseck, Christopher R.

    In this document I discuss paraffin fuel combustion and investigate the effects of additives on paraffin entrainment and regression. In general, hybrid rockets offer an economical and safe alternative to standard liquid and solid rockets. However, slow polymeric fuel regression and low combustion efficiency have limited the commercial use of hybrid rockets. Paraffin is a fast burning fuel that has received significant attention in the 2000's and 2010's as a replacement for standard fuels. Paraffin regresses three to four times faster than polymeric fuels due to the entrainment of a surface melt layer. However, further regression rate enhancement over the base paraffin fuel is necessary for widespread hybrid rocket adoption. I use a small scale opposed flow burner to investigate the effect of additives on the combustion of paraffin. Standard additives such as aluminum combust above the flame zone where sufficient oxidizer levels are present. As a result no heat is generated below the flame itself. In small scale opposed burner experiments the effect of limited heat feedback is apparent. Aluminum in particular does not improve the regression of paraffin in the opposed burner. The lack of heat feedback from additive combustion limits the applicability of the opposed burner. In turn, the results obtained in the opposed burner with metal additive loaded hybrid fuels do not match results from hybrid rocket experiments. In addition, nano-scale aluminum increases melt layer viscosity and greatly slows the regression of paraffin in the opposed flow burner. However, the reactive additives improve the regression rate of paraffin in the opposed burner where standard metals do not. At 5 wt.% mechanically activated titanium and carbon (Ti-C) improves the regression rate of paraffin by 47% in the opposed burner. The mechanically activated Ti C likely reacts in or near the melt layer and provides heat feedback below the flame region that results in faster opposed burner regression

  20. A study on KMRR utilization for fuel development

    International Nuclear Information System (INIS)

    Kang, Young Hwan; Ryu, Woo Seog; Park, Ji Yeon; Joo, Kee Nam; Park, Jong Man; Park, Se Jin

    1991-01-01

    The most effective utilization scheme of the KMRR was studied in the field of nuclear fuel development through reviewing literatural documents on irradiation facilities and in-pile test. It is suggested that the KMRR should be used for verification tests of advanced fuels and for power ramping / cycling tests of fuel rods. In addition, the characterization tests for fuel development and the basic material research should be also performed. In-pile loops for fuel verification and/or power ramping / cycling tests are proposed to be installed in advance, and capsules are necessary for power ramping / cycling tests, fuel characterization tests and / or material tests. Instrumentation technologies for thermocouple, SPND (Self-Powered Neutron Detector) and pressure transducer, and the in-situ dimensional measuring systems have to be developed to obtain the useful and various results from irradiation tests in the KMRR. A mock-up test rod for characterizing fuel thermal response was manufactured and the related technologies as well as the design specification were set up. An equipment for microdrilling and grooving of fuel pellets and an apparatus for diffusion-bonding between zircaloy-4 and stainless steel were made. A study to verify the integrity of test rod weldments is presented using out-of pile corrosion test. (Author)

  1. Development of a fissile particle for HTGR fuel recycle

    International Nuclear Information System (INIS)

    Homan, F.J.; Long, E.L. Jr.; Lindemer, T.B.; Beatty, R.L.; Tiegs, T.N.

    1976-12-01

    Recycle fissile fuel particles for high-temperature gas-cooled reactors (HTGRs) have been under development since the mid-1960s. Irradiation performance on early UO 2 and Th 0 . 8 U 0 . 2 O 2 kernels is described in this report, and the performance limitations associated with the dense oxide kernels are presented. The development of the new reference fuel kernel, the weak-acid-resin-derived (WAR) UO 2 --UC 2 , is discussed in detail, including an extensive section on the irradiation performance of this fuel in HFIR removable beryllium capsules HRB-7 through -10. The conclusion is reached that the irradiation performance of the WAR fissile fuel kernel is better than that of any coated particle fuel yet tested. Further, the present fissile kernel is adequate for steam cycle HTGRs as well as for many advanced applications such as gas turbine and process heat HTGRs

  2. Development of probabilistic fast reactor fuel design method

    International Nuclear Information System (INIS)

    Ozawa, Takayuki

    1997-01-01

    Under the current method of evaluating fuel robustness in FBR fuel rod design, a variety of uncertain quantities including fuel production tolerance and power density are estimated conservatively. In the future, in order to proceed with improvements in the FBR core's performance and optimize the fuel's specifications, a rationalization of fuel design tolerance is required. Among the measures aimed at realizing this rationalization, the introduction of a probabilistic fast reactor fuel design method is currently under consideration. I have developed a probabilistic fast reactor fuel design code named BORNFREE, in order to make use of this method in FBR fuel design. At the same time, I have carried out a trial calculation of the cladding stress using this code and made a study and an evaluation of the possibility of employing tolerance rationalization in fuel rod design. In this paper, I provide an outline description of BORNFREE and report the results of the above study and evaluation. After performing cladding stress trial calculations using the probabilistic method, I was able to confirm that this method promises more rational design evaluation results than the conventional deterministic method. (author)

  3. The development and localization of nuclear fuel technology for KMRR

    International Nuclear Information System (INIS)

    Kim, Seong Yun; Lee, Ji Bok; Suk, Ho Chun; Kuk, Il Hyun; Hwang, Woan; Kim, Bong Goo; Park, Joo Hwan; Kim, Young Jin; Kang, Thae Khapp; Lee, Jae Choon

    1988-05-01

    This project was implemented aiming at localizing the fabrication of the KMRR fuel by october 1993. The contents of this project were divided into three parts: fuel design, fuel fabrication and process criticality analysis. In the fuel design, the radial power distribution in the fuel core was modeled and formulated taking account of the neutron flux depression in the radial direction. It was also performed to model and formulate the thermal characteristics such as the thermal conductivity and specific heat of the fuel core, U3Si-Al, the swelling and the film coefficient of heat transfer between the aluminum clad and light water coolant. The two dimensional heat transfer in the finned fuel element was equated based on the general equation governing the heat transfer in materials in order to develope a computer code, TEMP2D. TEMP2D solves finite differenced equations to calculate a two dimensional fuel temperature distribution under the steady and transient states. In the fuel fabrication, the technologies of fabricating uranium silicide fuel meat were tried by using depleted uranium as a raw material. These were extended to find the problems in technologies and to establish the ways of approach. The end product, so called fuel meat, was a metallic powder compound, U3Six(1≤x≤2), dispersed in Al matrix. The fuel meat was fabricated by the horizontal extrusion technique, and powder extrusion technique. Fabrication technologies comprise five different continuous processes: melting and casting of metallic uranium with silicon and aluminum, heat treatment, chipping and crushing, pulverizing, and extrusion. In the process criticality analysis, AMPX-KENO benchmark calculation was performed and calculational error of AMPX-KENO system was established. (Author)

  4. Development status of irradiation devices and instrumentation for material and nuclear fuel irradiation tests in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Bong Goo; Sohn, Jae Min; Choo, Kee Nam [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-04-15

    The High flux Advanced Neutron Application ReactOr (HANARO), an open-tank-in-pool type reactor, is one of the multi-purpose research reactors in the world. Since the commencement of HANARO's operations in 1995, a significant number of experimental facilities have been developed and installed at HANARO, and continued efforts to develop more facilities are in progress. Owing to the stable operation of the reactor and its frequent utilization, more experimental facilities are being continuously added to satisfy various fields of study and diverse applications. The irradiation testing equipment for nuclear fuels and materials at HANARO can be classified into capsules and the Fuel Test Loop (FTL). Capsules for irradiation tests of nuclear fuels in HANARO have been developed for use under the dry conditions of the coolant and materials at HANARO and are now successfully utilized to perform irradiation tests. The FTL can be used to conduct irradiation testing of a nuclear fuel under the operating conditions of commercial nuclear power plants. During irradiation tests conducted using these capsules in HANARO, instruments such as the thermocouple, Linear Variable Differential Transformer (LVDT), small heater, Fluence Monitor (F/M) and Self-Powered Neutron Detector (SPND) are used to measure various characteristics of the nuclear fuel and irradiated material. This paper describes not only the status of HANARO and the status and perspective of irradiation devices and instrumentation for carrying out nuclear fuel and material tests in HANARO but also some results from instrumentation during irradiation tests

  5. Manufacturing experience and perspectives of WWER nuclear fuel development

    International Nuclear Information System (INIS)

    Aksenov, P.; Kolosovskiy, V.

    2011-01-01

    The purposes of new shroudless working fuel assembly (PK-3) development, basic design peculiarities of working fuel assembly (PK-3) and the results of PK-3 implementation are presented in this paper. Values of 440.19.000-02 working fuel assembly with debris filter Burnup at Kola NPP unit 2 are given. The main issues settled in the course of TVSA-T implementation like: The development of the design and fabrication method of mixing grids; The development of the design and fabrication method of basic assemblies and components of TVSA-T, including fuel rods of new generation; and The obtainment of specified pellet microstructure with average grain size more than 25μm are listed. The development of the design and fabrication method of removable uprated headpiece of shortened length as well as the development of the design and fabrication method of a tailpiece equipped with a debris filter are also illustrated

  6. Dry Refabrication Technology Development of Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    Lee, Jung Won; Park, G. I.; Park, C. J.

    2010-04-01

    Key technical data on advanced nuclear fuel cycle technology development for the spent fuel recycling have been produced in this study. In the frame work of DUPIC, dry process oxide products fabrication, hot cell experimental data for decladding, powdering and oxide product fabrication from low and high burnup spent fuel have been produced, basic technology for fabrication of spent fuel standard material has been developed, and remote modulated welding equipment has been designed and fabricated. In the area of advanced pre-treatment process development, a rotary-type oxidizer and spherical particle fabrication process were developed by using SIMFUEL and off-gas treatment technology and zircalloy tube treatment technology were studied. In the area of the property characteristics of dry process products, fabrication technology of simulated dry process products was established and property models were developed based on reproducible property measurement data

  7. New developments in tire derived fuels (TDF)

    Energy Technology Data Exchange (ETDEWEB)

    Hawkins, G. [Portland Cement Association, Skokie, IL (United States)

    2006-07-01

    Portland cement is a mixture of finely ground raw materials that are processed in a rotary kiln heated to 1500 degrees C by various fuels. Cooled clinker from the kiln is ground into portland cement. This presentation provided an overview of the role that tire-derived fuels is now playing in cement manufacturing. Data on the extent of TDF use in cement manufacturing processes were presented, as well as the results of air emissions monitoring programs conducted to assess the environmental impacts of TDF. During the cement manufacturing process, shredded and chipped tires are blown into the kiln to provide fuel, while whole tires are placed in the midpoint of long dry or wet kilns. For preheater and precalciner kilns, whole tires are added to raw material entering the kiln. The kiln's controlled combustion environment ensures the complete destruction of scrap tires. Approximately 130 million scrap tires were recycled in Portland cement plants in 2003. Forty-four portland cement plants are currently using TDF in North America. The United States Department of Energy has estimated that TDF produces less carbon dioxide (CO{sub 2}) per unit of energy than coal. Monitoring programs at Portland cement plants have indicated that the use of TDF in place of coal has resulted in moderate reductions of nitrogen oxide (NO{sub x}) emissions, slightly lower sulfuric oxide (SO{sub 2}) emissions and moderate reductions in carbon monoxide (CO). Some plants have reported slight reductions in particulate matter (PM) emissions. It was concluded that the use of scrap tires as a fuel reduces a waste stream and contributes to energy and raw material requirements. Moreover, plant air emissions are either reduced or are not impacted by TDF use. refs., tabs., figs.

  8. Development of incident progress prediction technologies for nuclear emergency preparedness. Current status and future subjects

    International Nuclear Information System (INIS)

    Yoshida, Yoshitaka; Yamamoto, Yasunori; Kusunoki, Takayoshi; Kawasaki, Ikuo; Yanagi, Chihiro; Kinoshita, Ikuo; Iwasaki, Yoshito

    2014-01-01

    Nuclear licensees are required to maintain a prediction system during normal condition for using a nuclear emergency by the Basic Plan for Disaster Prevention of government. With prediction of the incident progress, if the present condition of nuclear power plant is understood appropriately and it grows more serious with keeping the present situation, it is in predicting what kind of situation will be occurred in the near future, choosing the effective countermeasures against the coming threat, and understanding the time available of intervention time. Following the accident on September 30 1999 in the nuclear fuel fabrication facility in Tokai Village of Ibaraki Prefecture, the Institute of Nuclear Safety System started development of incident progress prediction technologies for nuclear emergency preparedness. We have performed technical applications and made improvements in nuclear emergency exercises and verified the developed systems using the observed values of the Fukushima Daiichi Nuclear Power Plant accident. As a result, our developed Incident Progress Prediction System was applied to nuclear emergency exercises and we accumulated knowledge and experience by which we improved the system to make predictions more rapidly and more precisely, including for example, the development of a prediction method for leak size of reactor coolant. On the other hand, if a rapidly progressing incident occurs, since end users need simple and quick predictions about the public's protection and evacuation areas, we developed the Radioactive Materials Release, Radiation Dose and Radiological Protection Area Prediction System which changed solving an inverse problem into a forward problem solution. In view of the water-level-decline incident of the spent fuel storage facility at the Fukushima Daiichi Nuclear Power Plant, the spent fuel storage facility water level and the water temperature evaluation tool were improved. Such incident progress prediction technologies were

  9. Development and verifications of fast reactor fuel design code ''Ceptar''

    International Nuclear Information System (INIS)

    Ozawa, T.; Nakazawa, H.; Abe, T.

    2001-01-01

    The annular fuel is very beneficial for fast reactors, because it is available for both high power and high burn-up. Concerning the irradiation behavior of the annular fuel, most of annular pellets irradiated up to high burn-up showed shrinkage of the central hole due to deformation and restructuring of the pellets. It is needed to predict precisely the shrinkage of the central hole during irradiation, because it has a great influence on power-to-melt. In this paper, outline of CEPTAR code (Calculation code to Evaluate fuel pin stability for annular fuel design) developed to meet this need is presented. In this code, the radial profile of fuel density can be computed by using the void migration model, and law of conservation of mass defines the inner diameter. For the mechanical analysis, the fuel and cladding deformation caused by the thermal expansion, swelling and creep is computed by the stress-strain analysis using the approximation of plane-strain. In addition, CEPTAR can also take into account the effect of Joint-Oxide-Gain (JOG) which is observed in fuel-cladding gap of high burn-up fuel. JOG has an effect to decrease the fuel swelling and to improve the gap conductance due to deposition of solid fission product. Based on post-irradiation data on PFR annular fuel, we developed an empirical model for JOG. For code verifications, the thermal and mechanical data obtained from various irradiation tests and post-irradiation examinations were compared with the predictions of this code. In this study, INTA (instrumented test assembly) test in JOYO, PTM (power-to-melt) test in JOYO, EBR-II, FFTF and MTR in Harwell laboratory, and post-irradiation examinations on a number of PFR fuels, were used as verification data. (author)

  10. Fuel cells:

    DEFF Research Database (Denmark)

    Sørensen, Bent

    2013-01-01

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

  11. The Chip-Scale Atomic Clock - Recent Development Progress

    Science.gov (United States)

    2004-09-01

    35th Annual Precise Time and Time Interval (PTTI) Meeting 467 THE CHIP-SCALE ATOMIC CLOCK – RECENT DEVELOPMENT PROGRESS R. Lutwak ...1] R. Lutwak , et al., 2003, “The Chip-Scale Atomic Clock – Coherent Population Trapping vs. Conventional Interrogation,” in

  12. Proposed plan for critical experiments supporting thorium fuel cycle development

    International Nuclear Information System (INIS)

    Gore, B.F.

    1978-09-01

    A preliminary plan is proposed for critical experiments to provide data needed for the recycle of thorium based nuclear fuels. The sequence of experimentation starts with well moderated solutions followed by highly concentrated low moderated solutions. It then progresses through lattices moderated by water, by water plus soluble poisons, and by fissile solutions, to solutions poisoned by raschig rings and soluble poisons. Final experiments would treat lattices moderated by poisoned fissile solution, and arrays of stored fissile units

  13. Feasibility study on the development of advanced LWR fuel technology

    International Nuclear Information System (INIS)

    Jung, Youn Ho; Sohn, D. S.; Jeong, Y. H.; Song, K. W.; Song, K. N.; Chun, T. H.; Bang, J. G.; Bae, K. K.; Kim, D. H. and others.

    1997-07-01

    Worldwide R and D trends related to core technology of LWR fuels and status of patents have been surveyed for the feasibility study. In addition, various fuel cycle schemes have been studied to establish the target performance parameters. For the development of cladding material, establishment of long-term research plan for alloy development and optimization of melting process and manufacturing technology were conducted. A work which could characterize the effect of sintering additives on the microstructure of UO 2 pellet has been experimentally undertaken, and major sintering variables and their ranges have been found in the sintering process of UO 2 -Gd 2 O 3 burnable absorber pellet. The analysis of state of the art technology related to flow mixing device for spacer grid and debris filtering device for bottom nozzle and the investigation of the physical phenomena related to CHF enhancement and the establishment of the data base for thermal-hydraulic performance tests has been done in this study. In addition, survey on the documents of the up-to-date PWR fuel assemblies developed by foreign vendors have been carried out to understand their R and D trends and establish the direction of R and D for these structural components. And, to set the performance target of the new fuel, to be developed, fuel burnup and economy under the extended fuel cycle length scheme were estimated. A preliminary study on the failure mechanism of CANDU fuel, key technology and advanced coating has been performed. (author). 190 refs., 31 tabs., 129 figs

  14. LG Solid Oxide Fuel Cell (SOFC) Model Development

    Energy Technology Data Exchange (ETDEWEB)

    Haberman, Ben [LG Fuel Cell Systems Inc., North Canton, OH (United States); Martinez-Baca, Carlos [LG Fuel Cell Systems Inc., North Canton, OH (United States); Rush, Greg [LG Fuel Cell Systems Inc., North Canton, OH (United States)

    2013-05-31

    This report presents a summary of the work performed by LG Fuel Cell Systems Inc. during the project LG Solid Oxide Fuel Cell (SOFC) Model Development (DOE Award Number: DE-FE0000773) which commenced on October 1, 2009 and was completed on March 31, 2013. The aim of this project is for LG Fuel Cell Systems Inc. (formerly known as Rolls-Royce Fuel Cell Systems (US) Inc.) (LGFCS) to develop a multi-physics solid oxide fuel cell (SOFC) computer code (MPC) for performance calculations of the LGFCS fuel cell structure to support fuel cell product design and development. A summary of the initial stages of the project is provided which describes the MPC requirements that were developed and the selection of a candidate code, STAR-CCM+ (CD-adapco). This is followed by a detailed description of the subsequent work program including code enhancement and model verification and validation activities. Details of the code enhancements that were implemented to facilitate MPC SOFC simulations are provided along with a description of the models that were built using the MPC and validated against experimental data. The modeling work described in this report represents a level of calculation detail that has not been previously available within LGFCS.

  15. Development of Green Fuels From Algae - The University of Tulsa

    Energy Technology Data Exchange (ETDEWEB)

    Crunkleton, Daniel; Price, Geoffrey; Johannes, Tyler; Cremaschi, Selen

    2012-12-03

    The general public has become increasingly aware of the pitfalls encountered with the continued reliance on fossil fuels in the industrialized world. In response, the scientific community is in the process of developing non-fossil fuel technologies that can supply adequate energy while also being environmentally friendly. In this project, we concentrate on green fuels which we define as those capable of being produced from renewable and sustainable resources in a way that is compatible with the current transportation fuel infrastructure. One route to green fuels that has received relatively little attention begins with algae as a feedstock. Algae are a diverse group of aquatic, photosynthetic organisms, generally categorized as either macroalgae (i.e. seaweed) or microalgae. Microalgae constitute a spectacularly diverse group of prokaryotic and eukaryotic unicellular organisms and account for approximately 50% of global organic carbon fixation. The PI's have subdivided the proposed research program into three main research areas, all of which are essential to the development of commercially viable algae fuels compatible with current energy infrastructure. In the fuel development focus, catalytic cracking reactions of algae oils is optimized. In the species development project, genetic engineering is used to create microalgae strains that are capable of high-level hydrocarbon production. For the modeling effort, the construction of multi-scaled models of algae production was prioritized, including integrating small-scale hydrodynamic models of algae production and reactor design and large-scale design optimization models.

  16. FY2009 Annual Progress Report for Advanced Combustion Engine Research and Development

    Energy Technology Data Exchange (ETDEWEB)

    none,

    2009-12-01

    Fiscal Year 2009 Annual Progress Report for the Advanced Combustion Engine Research and Development (R&D) subprogram. The Advanced Combustion Engine R&D subprogram supports the mission of the VTP program by removing the critical technical barriers to commercialization of advanced internal combustion engines (ICEs) for passenger and commercial vehicles that meet future Federal emissions regulations. Dramatically improving the efficiency of ICEs and enabling their introduction in conventional as well as hybrid electric vehicles is the most promising and cost-effective approach to increasing vehicle fuel economy over the next 30 years.

  17. FY 2014 Annual Progress Report - Advanced Combustion Engine Research and Development (Book)

    Energy Technology Data Exchange (ETDEWEB)

    2014-11-01

    In the past year, the DOE Hydrogen Program (the Program) made substantial progress toward its goals and objectives. The Program has conducted comprehensive and focused efforts to enable the widespread commercialization of hydrogen and fuel cell technologies in diverse sectors of the economy. With emphasis on applications that will effectively strengthen our nation's energy security and improve our stewardship of the environment, the Program engages in research, development, and demonstration of critical improvements in the technologies. Highlights of the Program's accomplishments can be found in the sub-program chapters of this report.

  18. Development of code SFINEL (Spent fuel integrity evaluator)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yong Soo; Min, Chin Young; Ohk, Young Kil; Yang, Yong Sik; Kim, Dong Ju; Kim, Nam Ku [Hanyang University, Seoul (Korea)

    1999-01-01

    SFINEL code, an integrated computer program for predicting the spent fuel rod integrity based on burn-up history and major degradation mechanisms, has been developed through this project. This code can sufficiently simulate the power history of a fuel rod during the reactor operation and estimate the degree of deterioration of spent fuel cladding using the recently-developed models on the degradation mechanisms. SFINEL code has been thoroughly benchmarked against the collected in-pile data and operating experiences: deformation and rupture, and cladding oxidation, rod internal pressure creep, then comprehensive whole degradation process. (author). 75 refs., 51 figs., 5 tabs.

  19. Battery and Fuel Cell Development for NASA's Constellation Missions

    Science.gov (United States)

    Manzo, Michelle A.

    2009-01-01

    NASA's return to the moon will require advanced battery, fuel cell and regenerative fuel cell energy storage systems. This paper will provide an overview of the planned energy storage systems for the Orion Spacecraft and the Aries rockets that will be used in the return journey to the Moon. Technology development goals and approaches to provide batteries and fuel cells for the Altair Lunar Lander, the new space suit under development for extravehicular activities (EY A) on the Lunar surface, and the Lunar Surface Systems operations will also be discussed.

  20. Battery and Fuel Cell Development for NASA's Exploration Missions

    Science.gov (United States)

    Manzo, Michelle A.; Reid, Concha M.

    2009-01-01

    NASA's return to the moon will require advanced battery, fuel cell and regenerative fuel cell energy storage systems. This paper will provide an overview of the planned energy storage systems for the Orion Spacecraft and the Aries rockets that will be used in the return journey to the Moon. Technology development goals and approaches to provide batteries and fuel cells for the Altair Lunar Lander, the new space suit under development for extravehicular activities (EVA) on the Lunar surface, and the Lunar Surface Systems operations will also be discussed.

  1. Development of machine vision system for PHWR fuel pellet inspection

    Energy Technology Data Exchange (ETDEWEB)

    Kamalesh Kumar, B.; Reddy, K.S.; Lakshminarayana, A.; Sastry, V.S.; Ramana Rao, A.V. [Nuclear Fuel Complex, Hyderabad, Andhra Pradesh (India); Joshi, M.; Deshpande, P.; Navathe, C.P.; Jayaraj, R.N. [Raja Ramanna Centre for Advanced Technology, Indore, Madhya Pradesh (India)

    2008-07-01

    Nuclear Fuel Complex, a constituent of Department of Atomic Energy; India is responsible for manufacturing nuclear fuel in India . Over a million Uranium-di-oxide pellets fabricated per annum need visual inspection . In order to overcome the limitations of human based visual inspection, NFC has undertaken the development of machine vision system. The development involved designing various subsystems viz. mechanical and control subsystem for handling and rotation of fuel pellets, lighting subsystem for illumination, image acquisition system, and image processing system and integration. This paper brings out details of various subsystems and results obtained from the trials conducted. (author)

  2. Development of a Reliable Fuel Depletion Methodology for the HTR-10 Spent Fuel Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Kiwhan [Los Alamos National Laboratory; Beddingfield, David H. [Los Alamos National Laboratory; Geist, William H. [Los Alamos National Laboratory; Lee, Sang-Yoon [unaffiliated

    2012-07-03

    A technical working group formed in 2007 between NNSA and CAEA to develop a reliable fuel depletion method for HTR-10 based on MCNPX and to analyze the isotopic inventory and radiation source terms of the HTR-10 spent fuel. Conclusions of this presentation are: (1) Established a fuel depletion methodology and demonstrated its safeguards application; (2) Proliferation resistant at high discharge burnup ({approx}80 GWD/MtHM) - Unfavorable isotopics, high number of pebbles needed, harder to reprocess pebbles; (3) SF should remain under safeguards comparable to that of LWR; and (4) Diversion scenarios not considered, but can be performed.

  3. Baseline Gas Turbine Development Program. Fourteenth quarterly progress report

    Energy Technology Data Exchange (ETDEWEB)

    Schmidt, F W; Wagner, C E

    1976-04-30

    Progress is reported for a Baseline Gas Turbine Development Program sponsored by the Heat Engine Systems Branch, Division of Transportation Energy Conservation (TEC) of the Energy Research and Development Administration (ERDA). Structurally, this program is made up of three parts: (1) documentation of the existing automotive gas turbine state-of-the-art; (2) conduction of an extensive component improvement program; and (3) utilization of the improvements in the design, and building of an Upgraded Engine capable of demonstrating program goals.

  4. Development of alternative materials for BWR fuel springs

    International Nuclear Information System (INIS)

    Uruma, Y.; Osato, T.; Yamazaki, K.

    2002-01-01

    Major sources of radioactivity introduced into reactor water of BWR were estimated fuel crud and in-core materials (especially, fuel springs). Fuel springs are used for fixation of fuel cladding tubes with spacer grid. Those are small parts (total length is only within 25 mm) and so many numbers are loaded simultaneously and then total surfaces area are calculated up to about 200 m 2 . Fuel springs are located under high radiation field and high oxidative environment. Conventional fuel spring is made of alloy-X750 which is one of nickel-based alloy and is reported to show relatively higher corrosion release rate. 58 Co and 60 Co will be released directly into reactor water from intensely radio-activated fuel springs surface and increase radioactivity concentrations in primary coolant. Corrosion release control from fuel springs is an important technical item and a development of alternative material instead of alloy-X750 for fuel spring is a key subject to achieve ultra low man-rem exposure BWR plant. In present work, alloy-X718 which started usage for PWR fuel springs and stainless steel type 316L which has many mechanical property data are picked up for alternative materials and compared their corrosion behaviors with conventional material. Corrosion experiment was conducted under vapor-water two phases flow which is simulated fuel cladding surface boiling condition. After exposure, corrosion film formed under corrosion test was analyzed in detail and corrosion film amount and corrosion release amount are estimated among three materials. (authors)

  5. EURATOM safeguards efforts in the development of spent fuel verification methods by non-destructive assay

    Energy Technology Data Exchange (ETDEWEB)

    Matloch, L.; Vaccaro, S.; Couland, M.; De Baere, P.; Schwalbach, P. [Euratom, Communaute europeenne de l' energie atomique - CEEA (European Commission (EC))

    2015-07-01

    The back end of the nuclear fuel cycle continues to develop. The European Commission, particularly the Nuclear Safeguards Directorate of the Directorate General for Energy, implements Euratom safeguards and needs to adapt to this situation. The verification methods for spent nuclear fuel, which EURATOM inspectors can use, require continuous improvement. Whereas the Euratom on-site laboratories provide accurate verification results for fuel undergoing reprocessing, the situation is different for spent fuel which is destined for final storage. In particular, new needs arise from the increasing number of cask loadings for interim dry storage and the advanced plans for the construction of encapsulation plants and geological repositories. Various scenarios present verification challenges. In this context, EURATOM Safeguards, often in cooperation with other stakeholders, is committed to further improvement of NDA methods for spent fuel verification. In this effort EURATOM plays various roles, ranging from definition of inspection needs to direct participation in development of measurement systems, including support of research in the framework of international agreements and via the EC Support Program to the IAEA. This paper presents recent progress in selected NDA methods. These methods have been conceived to satisfy different spent fuel verification needs, ranging from attribute testing to pin-level partial defect verification. (authors)

  6. Design and development of a cathode processor for electrometallurgical treatment of spent nuclear fuel

    International Nuclear Information System (INIS)

    Brunsvold, A. R.; Roach, P. D.; Westphal, B. R.

    1999-01-01

    The electrometallurgical processing of spent fuel developed at Argonne National Laboratory produces a cathode which contains dendrites of heavy metal (principally U), salts, and residual cadmium. The cathode requires further treatment which is accomplished by loading it into a cathode processor to first purify and then consolidate the heavy metal. The principal steps in cathode processing are: the cathode is loaded into a crucible and both loaded into the cathode processor; the crucible is heated under vacuum to an intermediate temperature to distill the salt and cadmium from the crucible; the crucible is heated further to melt and consolidate the heavy metal; the crucible and charge are then cooled forming a heavy metal ingot in the crucible mold. The cathode processor development program has progressed through the design, fabrication, qualification, and demonstration phases. Two identical units were built. One (a prototype unit) has been installed at Argonne's site in Illinois and the other (the production unit) has been installed in the Fuel Conditioning Facility (FCF) at Argonne's Idaho site. Both units are presently in operation. The most recent activities completed in the FCF fuel processing project were the EBR-II driver fuel and blanket fuel demonstration phases. All of the cathode processor success criteria were met during these demonstration phases. These included finalizing the operation conditions applicable to irradiated fuel and process throughput criteria

  7. Development of compaction technique for spent fuel skeletons

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Ji Sup; Kim, Young Hwan; Jung, Jae Hoo [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-12-01

    To increase the utilization of uranium resources contained in the spent fuel, the spent fuel is reused. For this, the spent fuel is dismantled or spent fuel rod is extracted from the spent fuel assembly. When the rod is extracted, the remaining components of spent fuel assembly, so called a NFBC(Non-Fuel Bearing Components), should be compacted for the final disposal. To this end, several companies developed the NFBC compactors. German company, named as GNS has developed the direct compression devices of the NFBCs for the rod consolidation and installed it at the PKA(2) of pilot conditioning plant. B and W (Babcock and Wilcox) in USA adopted cutting method rather than the compression method and developed the special cutting devices of NFBC which can be applied underwater environment. In this study the characteristics of these two methods was investigated, in terms of fabrication cost of devices, maintainability in a high radioactive environment, required power and work volume for operation. Also, the optimal power source is selected by comparing the maximum power versus the work volume for operation. In addition to these, the reduction ratio of the bulk volume is obtained while varying the cutting length of the NFBC through a series of experiments. Based on the results of analysis and experiments, the cutting method after compression is selected as an optimal volume reduction method and its design specification is obtained. 8 refs., 62 figs., 32 tabs. (Author)

  8. General-purpose heat source project and space nuclear safety fuels program. Progress report, February 1980

    International Nuclear Information System (INIS)

    Maraman, W.J.

    1980-05-01

    This formal monthly report covers the studies related to the use of 238 PuO 2 in radioisotopic power systems carried out for the Advanced Nuclear Systems and Projects Division of the Los Alamos Scientific Laboratory. The two programs involved are: General-Purpose Heat Source Development and Space Nuclear Safety and Fuels. Most of the studies discussed here are of a continuing nature. Results and conclusions described may change as the work continues. Published reference to the results cited in this report should not be made without the explicit permission of the person in charge of the work

  9. LSDS Development for Isotopic Fissile Assay in Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yong Deok; Park, Chang Je; Park, Geun Il; Lee, Jung Won; Song, Kee Chan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-07-01

    As an option to reduce a spent fuel and reuse an existing fissile material in spent fuel, sodium fast reactor SFR program linked with pyro-processing is under development in KAERI. A uranium-TRU mixture through a pyro-process is used to fabricate SFR fuel. An assay of isotopic fissile content plays an important role in an optimum design of storage site and reuse of fissile materials of spent fuel. Lead slowing down spectrometer LSDS is being developed in KAERI to analyze isotopic fissile material content. LSDS has several features: direct fissile assay, near real time fissile assay, no influence from radiation background, fissile isotopic assay and applicable to spent fuel and recycled fuel. Based on the designed geometry, neutron energy resolution was investigated. The neutron energy spectrum was analyzed as well. Spent fuel emits large number of neutrons by spontaneous fission. Neutron generator must overcome the neutron background to get the pure fission signals from fissile materials. Neutron generator is planned to have compact system with one section electron linac which is easy maintenance, less cost and high neutron yield. The LSD has the power to resolve the fission characteristics from each fissile material. This feature can analyze the content of isotopic fissile. From 1keV to 0.1eV energy range, the energy resolution is enough to get the individual fissile fission signatures. The dominant fission signature is shown below 1eV for each fissile isotope. The neutron generation system with target was designed to get fission signals by fissile materials. The system was decided to overcome neutron backgrounds and to get good counting statistics. Finally, an accurate fissile material content will contribute to safety of spent fuel reuse in future nuclear energy system and optimum design of spent fuel storage site. Additionally, an accurate fissile material content will increase international transparence and credibility for the reuse of PWR spent fuel.

  10. LSDS Development for Isotopic Fissile Assay in Spent Fuel

    International Nuclear Information System (INIS)

    Lee, Yong Deok; Park, Chang Je; Park, Geun Il; Lee, Jung Won; Song, Kee Chan

    2011-01-01

    As an option to reduce a spent fuel and reuse an existing fissile material in spent fuel, sodium fast reactor SFR program linked with pyro-processing is under development in KAERI. A uranium-TRU mixture through a pyro-process is used to fabricate SFR fuel. An assay of isotopic fissile content plays an important role in an optimum design of storage site and reuse of fissile materials of spent fuel. Lead slowing down spectrometer LSDS is being developed in KAERI to analyze isotopic fissile material content. LSDS has several features: direct fissile assay, near real time fissile assay, no influence from radiation background, fissile isotopic assay and applicable to spent fuel and recycled fuel. Based on the designed geometry, neutron energy resolution was investigated. The neutron energy spectrum was analyzed as well. Spent fuel emits large number of neutrons by spontaneous fission. Neutron generator must overcome the neutron background to get the pure fission signals from fissile materials. Neutron generator is planned to have compact system with one section electron linac which is easy maintenance, less cost and high neutron yield. The LSD has the power to resolve the fission characteristics from each fissile material. This feature can analyze the content of isotopic fissile. From 1keV to 0.1eV energy range, the energy resolution is enough to get the individual fissile fission signatures. The dominant fission signature is shown below 1eV for each fissile isotope. The neutron generation system with target was designed to get fission signals by fissile materials. The system was decided to overcome neutron backgrounds and to get good counting statistics. Finally, an accurate fissile material content will contribute to safety of spent fuel reuse in future nuclear energy system and optimum design of spent fuel storage site. Additionally, an accurate fissile material content will increase international transparence and credibility for the reuse of PWR spent fuel

  11. Development of long-life low enrichment fuel

    International Nuclear Information System (INIS)

    Gietzen, A.J.; West, G.B.

    1978-01-01

    With only a few exceptions, TRIGA reactors have always used low-enriched-uranium (LEU) fuel with an enrichment of 19.9%. The exceptions have either been converted from the standard low-enriched fuel to the 70% enriched FLIP fuel in order to achieve extended lifetime, or are higher powered reactors which were designed for long life using 93%-enriched uranium during the time when the use and export of highly enriched uranium (HEU) was not restricted. The advent of international policies focusing attention on non-proliferation and safeguards made the HEU fuels obsolete. General Atomic immediately undertook a development effort (nearly two years ago) in order to be in a position to comply with these policies for all future export sales and also to provide a low-enriched alternative to fully enriched plate-type fuels. This important work was subsequently partially supported by the U. S. Department of Energy. The laboratory and production tests have shown that higher uranium densities can be achieved to compensate for reducing the enrichment to 20%, and that the fuels maintain the characteristics of the very thoroughly proven standard TRIGA fuels. In May of this year, General Atomic announced that these fuels were available for TRIGA reactors and for plate-type reactors with power levels up to 15 MW with GA's standard commercial warranty

  12. Progress in developing the concept for the irradiation research facility

    International Nuclear Information System (INIS)

    Lee, A.G.; Bishop, W.E.; Gillespie, G.E.; Zeng, Y.

    1996-04-01

    At the 16th annual Canadian Nuclear Society conference, AECL presented the case for replacing the NRU reactor with an Irradiation Research Facility (IRF) to test CANDU fuels and materials and to perform advanced materials research using neutrons. AECL developed a cost estimate of $500 million for the reference IRF concept, and estimated that it would require 87 months to complete. AECL has initiated a pre-project program to develop the IRF concept and to minimize uncertainties related to feasibility and licensability, and to examine options for reducing the overall project cost before project implementation begins. (author) 10 refs., 2 figs

  13. Design and development of progressive tool for manufacturing washer

    Science.gov (United States)

    Annigeri, Ulhas K.; Raghavendra Ravi Kiran, K.; Deepthi, Y. P.

    2017-07-01

    In a progressive tool the raw material is worked at different station to finally fabricate the component. A progressive tool is a lucrative tool for mass production of components. A lot of automobile and other transport industries develop progressive tool for the production of components. The design of tool involves lot of planning and the same amount of skill of process planning is required in the fabrication of the tool. The design also involves use of thumb rules and standard elements as per experience gained in practice. Manufacturing the press tool is a laborious task as special jigs and fixtures have to be designed for the purpose. Assembly of all the press tool elements is another task where use of accurate measuring instruments for alignment of various tool elements is important. In the present study, design and fabrication of progressive press tool for production of washer has been developed and the press tool has been tried out on a mechanical type of press. The components produced are to dimensions.

  14. Chemical Engineering Division fuel cycle programs. Progress report, January--March 1978

    Energy Technology Data Exchange (ETDEWEB)

    Steindler, M.J.; Ader, M.; Barletta, R.E.

    1979-04-01

    Fuel cycle studies reported for this period include studies of advanced solvent extraction techniques focussed on the development of centrifugal contactors for use in Purex processes. Miniature single-stage and eight-stage centrifugal contactors are being employed in studies of contactor performance and the kinetics of extraction. A 9-cm-ID centrifugal contactor has been completed, and fabrication drawings are being prepared for a plant-scale contactor. In other work, tricaprylmethyl-ammonium nitrate and di-n-amyl n-amylphosphonate are being evaluated as extractants in the Thorex process. Literature on the dispersion of liquids by explosions is being reviewed. A process was developed for extracting TBP degradation products from TBP-Na/sub 2/CO/sub 3/ scrub solutions while the actinides remain with the raffinate. In the program on pyrochemical and dry processing of nuclear fuel, the literature is being reviewed for acceptable materials for containment vessels, decladding methods are being evaluated, salt transport processes are being studied, a candidate flow sheet (based upon the Dow Aluminum Pyrometallurgical process) for reprocessing spent uranium metal fuel was prepared, work was begun on the use of molten salts for reprocessing actinide oxides, and the reprocessing of (Th,U)O/sub 2/ solid solution in a KCl-LiCl salt containing ThCl/sub 4/ and thorium chips was studied. Work on the encapsulation of solidified radioactive waste in a metal matrix includes study of (1) chemical interactions between simulated waste forms and matrix metals, (2) the leach rates of simulated encapsulated waste forms, and (3) the corrosion of candidate matrix metals and canister materials in brine solutions.Work to establish criteria for the handling of waste cladding hulls is continuing. The transport properties of nuclear waste in geologic media are being studied to estimate leaching of radionuclides from deep repositories by groundwater.

  15. 2010 national progress report on R and D on LEU fuel and target technology in Argentina

    International Nuclear Information System (INIS)

    Balart, S.; Blaumann, H.; Cristini, P.; Gonzalez, A.G.; Gonzalez, R.; Hermida, J.D.; Lopez, M.; Mirandou, M.; Taboada, H.

    2010-01-01

    Since last RRFM meeting, CNEA has deployed several related tasks. The RA-6 MTR type reactor, converted its core from HEU to a new LEU silicide one is scaling up the power, according to a protocol requested by the national regulatory body, ARN. CNEA is deploying an intense R and D activity to fabricate both dispersed U-Mo (Al-Si matrix and Al cladding) and monolithic (Zry-4 cladding) miniplates to develop possible solutions to VHD dispersed and monolithic fuels technical problems. Some monolithic 58% enrichment U8%Mo and U10%Mo are being delivered to INL-DoE to be irradiated in ATR reactor core. A conscientious study on compound interphase formation in both cases is being carried out. CNEA, a worldwide leader on LEU technology for fission radioisotope production is providing Brazil with these radiopharmaceutical products and Egypt and Australia with the technology through INVAP SE. CNEA is also committed to improve the diffusion of LEU target and radiochemical technology for radioisotope production and target and process optimization. Future plans include: 1) Fabrication of a LEU dispersed U-Mo fuel prototype following the recommendations of the IAEA's Good Practices document, to be irradiated in a high flux reactor in the frame of the ARG/4/092 IAEA's Technical Cooperation project. 2) Development of LEU very high density monolithic and dispersed U-Mo fuel plates with Zry-4 or Al cladding as a part of the RERTR program. 3) Optimization of LEU target and radiochemical techniques for radioisotope production. (author)

  16. Current developments of fuel fabrication technologies at the plutonium fuel production facility, PFPF

    International Nuclear Information System (INIS)

    Asakura, K.; Aono, S.; Yamaguchi, T.; Deguchi, M.

    2000-01-01

    The Japan Nuclear Cycle Development Institute, JNC, designed, constructed and has operated the Plutonium Fuel Production Facility, PFPF, at the JNC Tokai Works to supply MOX fuels to the proto-type Fast Breeder Reactor, FBR, 'MONJU' and the experimental FBR 'JOYO' with 5 tonMOX/year of fabrication capability. Reduction of personal radiation exposure to a large amount of plutonium is one of the most important subjects in the development of MOX fabrication facility on a large scale. As the solution of this issue, the PFPF has introduced automated and/or remote controlled equipment in conjunction with computer controlled operation scheme. The PFPF started its operation in 1988 with JOYO reload fuel fabrication and has demonstrated MOX fuel fabrication on a large scale through JOYO and MONJU fuel fabrication for this decade. Through these operations, it has become obvious that several numbers of equipment initially installed in the PFPF need improvements in their performance and maintenance for commercial utilization of plutonium in the future. Furthermore, fuel fabrication of low density MOX pellets adopted in the MONJU fuel required a complete inspection because of difficulties in pellet fabrication compared with high density pellet for JOYO. This paper describes new pressing equipment with a powder recovery system, and pellet finishing and inspection equipment which has multiple functions, such as grinding measurements of outer diameter and density, and inspection of appearance to improve efficiency in the pellet finishing and inspection steps. Another development of technology concerning an annular pellet and an innovative process for MOX fuel fabrication are also described in this paper. (author)

  17. IMHEX{sup {reg_sign}} fuel cells progress toward commercialization

    Energy Technology Data Exchange (ETDEWEB)

    Scroppo, J.A.; Laurens, R.M.; Petraglia, V.J.

    1995-12-31

    The overall goal of M-C Power is the development and subsequent commercialization of Molten Carbonate Fuel Cell (MCFC) stacks. More specifically, MCFC`s Manifolded Heat Exchange (IMHEX{sup {reg_sign}}) plate design created by the Institute of Technology. In order to achieve the aforementioned goal, M-C Power assembled a formidable team of industry leaders. This group, refered to as the (IHMEX{sup {reg_sign}}) Team, has developed a strategy to move decisively through the stages of Technology Development and Product Design and Improvement through commercialization. This paper is to review the status of the overall commercialization program and activities. It will also provide an overview of the market entry product. Furthermore, we will evaluate the opportunities and benefits this product brings to a competitive power industry.

  18. 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)

  19. The development of fuel cell systems for mobile applications

    Energy Technology Data Exchange (ETDEWEB)

    Van den Oosterkamp, P.F.; Kraaij, G.J.; Van der Laag, P.C.; Stobbe, E.R.; Wouters, D.A.J.

    2006-09-15

    The ECN fuel cell related R and D program on fuel cells is linked to the stationary market and the automotive market. This paper will summarize our R and D activities for the automotive market. The role of fuels cells in two transport application area's will be described: the development of dedicated hydrogen based platforms in combination with advanced electricity storage for special logistic applications and the APU (auxiliary power unit) market for passenger cars and trucks, as well as for ships and airplanes. The associated aspects of hydrogen transport and storage, as well as the reforming of logistic fuels and bio-fuels to hydrogen will be described with some illustrative examples. These examples show that an integrated approach using applied catalysis, chemical reactor design and engineering, process simulation, control modelling and electrical engineering is required to address all aspects of the development of fuel cell technology for automotive applications. The paper concludes with a summary of the important environmental and economic drivers that influence the fuel cell market application.

  20. What Happens Inside a Fuel Cell? Developing an Experimental Functional Map of Fuel Cell Performance

    KAUST Repository

    Brett, Daniel J. L.

    2010-08-20

    Fuel cell performance is determined by the complex interplay of mass transport, energy transfer and electrochemical processes. The convolution of these processes leads to spatial heterogeneity in the way that fuel cells perform, particularly due to reactant consumption, water management and the design of fluid-flow plates. It is therefore unlikely that any bulk measurement made on a fuel cell will accurately represent performance at all parts of the cell. The ability to make spatially resolved measurements in a fuel cell provides one of the most useful ways in which to monitor and optimise performance. This Minireview explores a range of in situ techniques being used to study fuel cells and describes the use of novel experimental techniques that the authors have used to develop an \\'experimental functional map\\' of fuel cell performance. These techniques include the mapping of current density, electrochemical impedance, electrolyte conductivity, contact resistance and CO poisoning distribution within working PEFCs, as well as mapping the flow of reactant in gas channels using laser Doppler anemometry (LDA). For the high-temperature solid oxide fuel cell (SOFC), temperature mapping, reference electrode placement and the use of Raman spectroscopy are described along with methods to map the microstructural features of electrodes. The combination of these techniques, applied across a range of fuel cell operating conditions, allows a unique picture of the internal workings of fuel cells to be obtained and have been used to validate both numerical and analytical models. © 2010 Wiley-VCH Verlag GmbH& Co. KGaA, Weinheim.