WorldWideScience

Sample records for fuels progress performance

  1. 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

  2. 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

  3. 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.

  4. 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.

  5. 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

  6. 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

  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. 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.

  9. 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.

  10. 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)

  11. 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.

  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. 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.

  14. 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.

  15. 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.

  16. The METEOR/TRANSURANUS fuel performance code

    International Nuclear Information System (INIS)

    Struzik, C.; Guerin, Y.

    1996-01-01

    The first calculations for the FUMEX exercise were performed using version 1.1 of the METEOR/TRANSURANUS code. Since then, important improvements have been implemented on several models. In its present state, the code describes fuel rod behaviour in standard PWR conditions. Its validity extends to UO 2 and MOX fuels clad in Zircaloy-4. Power transient calculations for UO 2 and Gd doped fuel calculations are possible, but further developments are in progress, and the applications will be fully qualified in version 2.0. A considerable effort is made to replace semi-empirical models with models that have a sounder physical basis. (authors). 14 refs

  17. 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.

  18. 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.

  19. 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)

  20. 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.)

  1. 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)

  2. 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

  3. BWR fuel performance

    International Nuclear Information System (INIS)

    Baily, W.E.; Armijo, J.S.; Jacobson, J.; Proebstle, R.A.

    1979-01-01

    The General Electric experience base on BWR fuel includes over 29,000 fuel assemblies which contain 1,600,000 fuel rods. Over the last five years, design, process and operating changes have been introduced which have had major effects in improving fuel performance. Monitoring this fuel performance in BWRs has been accomplished through cooperative programs between GE and utilities. Activities such as plant fission product monitoring, fuel sipping and fuel and channel surveillance programs have jointly contributed to the value of this extensive experience base. The systematic evaluation of this data has established well-defined fuel performance trends which provide the assurance and confidence in fuel reliability that only actual operating experience can provide

  4. 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.)

  5. 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)

  6. The improvement of performances for PWR fuels

    International Nuclear Information System (INIS)

    Debes

    2001-01-01

    UO 2 fuels used in French nuclear power plants are authorized for values of burn-ups up to 52 GWj/t. Constant technological progress concerning pellets, cladding, and the design of the assembly has led to better performance and a broader safety margin. EDF is gathering all the elements to qualify and back its demand to increase the limit burn-up to 65 GWj/t in 2004 and to 70 GWj/t in 2008. For the same amount of energy produced, this policy of higher burn-ups will allow: - a reduction of the number of spent fuel assemblies, - a direct economic spare by using less fuel assemblies, - a reduction of personnel dosimetry because of longer irradiation campaigns, and - less quantity of residual plutonium produced. (A.C.)

  7. CANDU fuel performance

    International Nuclear Information System (INIS)

    Ivanoff, N.V.; Bazeley, E.G.; Hastings, I.J.

    1982-01-01

    CANDU fuel has operated successfully in Ontario Hydro's power reactors since 1962. In the 19 years of experience, about 99.9% of all fuel bundles have performed as designed. Most defects occurred before 1979 and subsequent changes in fuel design, fuel management, reactor control, and manufacturing quality control have reduced the current defect rate to near zero. Loss of power production due to defective fuel has been negligible. The outstanding performance continues while maintaining a low unit energy cost for fuel

  8. 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

  9. 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

  10. Utility industry evaluation of the metal fuel facility and metal fuel performance for liquid metal reactors

    International Nuclear Information System (INIS)

    Burstein, S.; Gibbons, J.P.; High, M.D.; O'Boyle, D.R.; Pickens, T.A.; Pilmer, D.F.; Tomonto, J.R.; Weinberg, C.J.

    1990-02-01

    A team of utility industry representatives evaluated the liquid metal reactor metal fuel process and facility conceptual design being developed by Argonne National Laboratory (ANL) under Department of Energy sponsorship. The utility team concluded that a highly competent ANL team was making impressive progress in developing high performance advanced metal fuel and an economic processing and fabrication technology. The utility team concluded that the potential benefits of advanced metal fuel justified the development program, but that, at this early stage, there are considerable uncertainties in predicting the net overall economic benefit of metal fuel. Specific comments and recommendations are provided as a contribution towards enhancing the development program. 6 refs

  11. Improving the service life and performance of CANDU fuel channels

    International Nuclear Information System (INIS)

    Coleman, C.E.; Cheadle, B.A.; Causey, A.R.; Doubt, G.L.; Fong, R.W.L.; Venkatapathi, S.

    1996-03-01

    The development objective for CANDU fuel channels is to produce a design that can operate for 40 years at 90% capacity. Steady progress toward this objective is being made. The factors that determine the life of a CANDU fuel channel are reviewed and the processes necessary to achieve the objectives are identified. Performance of future fuel channels will be enhanced by reduced operating costs and increased safety margins to postulated accident conditions compared with those for current channels. The approaches to these issues are discussed briefly in this report. (author)

  12. 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

  13. Performance of Fragema fuel in pressurized water reactors

    International Nuclear Information System (INIS)

    Dumont, A.; Ravier, G.; Ballot, B.

    1986-06-01

    FRAGEMA fuel operating experience in power reactors is very extensive. Performance over a range of power and burnup levels for various operating conditions is quite satisfactory. However significant development programs are presently in progress to further extend our knowledge under increasingly severe operating conditions. In particular, upcoming data acquisition programs (1985-1988) will cover site and hot cell measurements on Gd poison rods, 4.5 % overenriched fuel rods over four operating cycles, 17 x 17 AFA fuel assemblies. For these products the same surveillance strategy as the one used for the standard assembly has been adopted, in order to continuously provide more data which can be used to upgrade design models and pave the way for the development of future products

  14. High Performance Fuel Cell and Electrolyzer Membrane Electrode Assemblies (MEAs) for Space Energy Storage Systems

    Science.gov (United States)

    Valdez, Thomas I.; Billings, Keith J.; Kisor, Adam; Bennett, William R.; Jakupca, Ian J.; Burke, Kenneth; Hoberecht, Mark A.

    2012-01-01

    Regenerative fuel cells provide a pathway to energy storage system development that are game changers for NASA missions. The fuel cell/ electrolysis MEA performance requirements 0.92 V/ 1.44 V at 200 mA/cm2 can be met. Fuel Cell MEAs have been incorporated into advanced NFT stacks. Electrolyzer stack development in progress. Fuel Cell MEA performance is a strong function of membrane selection, membrane selection will be driven by durability requirements. Electrolyzer MEA performance is catalysts driven, catalyst selection will be driven by durability requirements. Round Trip Efficiency, based on a cell performance, is approximately 65%.

  15. 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

  16. 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

  17. Improving the service life and performance of CANDU fuel channels

    International Nuclear Information System (INIS)

    Causey, A.R.; Cheadle, B.A.; Coleman, C.E.; Price, E.G.

    1996-02-01

    The development objective for CANDU fuel channels is to produce a design that can operate for 40 years at 90% capacity. Steady progress toward this objective is being made. The factors that determine the life of a CANDU fuel channel are reviewed and the processes necessary to achieve the objectives identified. Performance of future fuel channels will be enhanced by reduced operating costs, increased safety margins to postulated accident conditions, and reduced retubing costs compared with those for current channels. The approaches to these issues are discussed briefly in this report. (author). 14 refs., 1 tab., 8 figs

  18. Dry Process Fuel Performance Evaluation

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-04-15

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

  19. Dry Process Fuel Performance Evaluation

    International Nuclear Information System (INIS)

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

    2005-04-01

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

  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. 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.

  2. Safety analysis of MOX fuels by fuel performance code

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-12-01

    Performance of plutonium rick mixed oxide fuels specified for the Reduced-Moderation Water Reactor (RMWR) has been analysed by modified fuel performance code. Thermodynamic properties of these fuels up to 120 GWd/t burnup have not been measured and estimated using existing uranium fuel models. Fission product release, pressure rise inside fuel rods and mechanical loads of fuel cans due to internal pressure have been preliminarily assessed based on assumed axial power distribution history, which show the integrity of fuel performance. Detailed evaluation of fuel-cladding interactions due to thermal expansion or swelling of fuel pellets due to high burnup will be required for safety analysis of mixed oxide fuels. Thermal conductivity and swelling of plutonium rich mixed oxide fuels shall be taken into consideration. (T. Tanaka)

  3. 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.)

  4. Experience in the manufacture and performance of CANDU fuel for KANUPP

    International Nuclear Information System (INIS)

    Salim, M.; Ahmed, I.; Butt, P.

    1995-01-01

    Karachi Nuclear Power Plant (KANUPP) a 137 MWe CANDU unit is In operation since 1971. Initially, it was fueled with Canadian fuel bundles. In July 1980 Pakistani manufactured fuel was introduced in the reactor core, irradiated to a burnup of about 7500 MWd-teU -1 and successfully discharged in May 1984. The core was progressively fuelled with Pakistani fuel and in August 1990 the reactor core contained all Pakistani made fuel. As of the present, 3 core equivalent Pakistani fuel bundles have been successfully discharged at an average bumup of 6500 MWd-teU -1 . with a maximum burnup of ∼ 10,200 MWd-teU -1 . No fuel failure of Pakistani bundles has been observed so far. This paper presents the indigenous efforts towards manufacture and operational aspects of KANUPP fuel and compares its behaviour with that of Canadian supplied fuel. The Pakistani fuel has performed well and is as good as the Canadian fuel. (author)

  5. MOX and UOX PWR fuel performances EDF operating experience

    International Nuclear Information System (INIS)

    Provost, Jean-Luc; Debes, Michel

    2005-01-01

    Based on a large program of experimentations implemented during the 90s, the industrial achievement of new FAs designs with increased performances opens up new prospects. The currently UOX fuels used on the 58 EDF PWR units are now authorized up to a maximum FA burn-up of 52 GWd/t with a large experience from 45 to 50 GWd/t. Today, the new products, along with the progress made in the field of calculation methods, still enable to increase further the fuel performances with respect to the safety margins. Thus, the conditions are met to implement in the next years new fuel managements on each NPPs series of the EDF fleet with increased enrichment (up to 4.5%) and irradiation limits (up to 62 GWd/t). The recycling of plutonium is part of EDF's reprocessing/recycling strategy. Up to now, 20 PWR 900 MW reactors are managed in MOX hybrid management. The feedback experience of 18 years of PWR operation with MOX is satisfactory, without any specific problem regarding manoeuvrability or plant availability. EDF is now looking to introduce MOX fuels with a higher plutonium content (up to 8.6%) equivalent to natural uranium enriched to 3.7%. It is the goal of the MOX Parity core management which achieve balance of MOX and UOX fuel performance with a significant increase of the MOX average discharge burn-up (BU max: 52 GWd/t for MOX and UOX). The industrial maturity of new FAs designs, with increased performances, allows the implementation in the next years of new fuel managements on each NPPs series of the EDF fleet. The scheduling of the implementation of the new fuel managements on the PWRs fleet is a great challenge for EDF, with important stakes: the nuclear KWh cost decrease with the improvement of the plant operation performance. (author)

  6. 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

  7. 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.

  8. 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.

  9. Experiments of progressive replacement in Cesar at operation temperature. Uranium-plutonium fuels. Study performed within the frame of the CEA-EURATOM - No. 002 64 9 TRUF contract - 'Plutonium recycling'

    International Nuclear Information System (INIS)

    Bosser, Roland; Cuny, Gerard; Hoffmann, Alain; Langlet, Gerard; Laponche, Bernard; Morier, Francis; Penet, Francois; Charbonneau, Serge

    1969-08-01

    Experiments of progressive replacement (or substitution) of uranium-plutonium alloy fuels are part of a general program of experimental studies which are aimed at testing the methods used by the CEA to calculate the evolution of nuclear power reactors (calculation of spectrum in plutonium-containing fuels and validity of data used in these calculations, calculation of cross sections). Such progressive replacements have been performed in Aquilon (with heavy water as moderator) and measurements have been performed by oscillation in Marius and Cesar (graphite moderator). Herein reported experiments have been performed at 20, 100 and 200 C during a first campaign in 1966, and at 300, 400 and 450 C during a second campaign in 1968. Measurements are interpreted by means of the Coregraf 2 code. The report presents experimental conditions in Cesar, the measurement principle and the interpretation method (substitution experiments, enriched uranium calibration, interpretation steps, and temperature coefficient measurement), the obtained results and their discussion [fr

  10. 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)

  11. Fuel performance experience

    International Nuclear Information System (INIS)

    Sofer, G.A.

    1986-01-01

    The history of LWR fuel supply has been characterized by a wide range of design developments and fuel cycle cost improvements. Exxon Nuclear Company, Inc. has pursued an aggressive fuel research and development program aimed at improved fuel performance. Exxon Nuclear has introduced many design innovations which have improved fuel cycle economics and operating flexibility while fuel failures remain at very low levels. The removable upper tie plate feature of Exxon Nuclear assemblies has helped accelerate this development, enabling repeated inspections during successive plant outages. Also, this design feature has made it possible to repair damaged fuel assemblies during refueling outages, thereby minimizing the economic impact of fuel failure from all causes

  12. Effect of compressive force on PEM fuel cell performance

    Science.gov (United States)

    MacDonald, Colin Stephen

    Polymer electrolyte membrane (PEM) fuel cells possess the potential, as a zero-emission power source, to replace the internal combustion engine as the primary option for transportation applications. Though there are a number of obstacles to vast PEM fuel cell commercialization, such as high cost and limited durability, there has been significant progress in the field to achieve this goal. Experimental testing and analysis of fuel cell performance has been an important tool in this advancement. Experimental studies of the PEM fuel cell not only identify unfiltered performance response to manipulation of variables, but also aid in the advancement of fuel cell modelling, by allowing for validation of computational schemes. Compressive force used to contain a fuel cell assembly can play a significant role in how effectively the cell functions, the most obvious example being to ensure proper sealing within the cell. Compression can have a considerable impact on cell performance beyond the sealing aspects. The force can manipulate the ability to deliver reactants and the electrochemical functions of the cell, by altering the layers in the cell susceptible to this force. For these reasons an experimental study was undertaken, presented in this thesis, with specific focus placed on cell compression; in order to study its effect on reactant flow fields and performance response. The goal of the thesis was to develop a consistent and accurate general test procedure for the experimental analysis of a PEM fuel cell in order to analyse the effects of compression on performance. The factors potentially affecting cell performance, which were a function of compression, were identified as: (1) Sealing and surface contact; (2) Pressure drop across the flow channel; (3) Porosity of the GDL. Each factor was analysed independently in order to determine the individual contribution to changes in performance. An optimal degree of compression was identified for the cell configuration in

  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. 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

  16. Improving the service life and performance of CANDU fuel channels

    International Nuclear Information System (INIS)

    Causey, A.R.; Cheadle, B.A.; Coleman, C.E.; Price, E.G.

    1997-01-01

    The development objective for CANDU fuel channels is to produce a design that can operate for 40 years at 90% capacity. Steady progress toward this objective is being made. The factors that determine the life of the channel are reviewed and the processes necessary to achieve the objectives identified. Performance of future fuel channels will be enhanced by reduced operating costs, increased safety margins to postulated accident conditions, and reduced retubing costs compared to current channels. The approaches to these issues are discussed briefly in the paper. (author)

  17. 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

  18. 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.

  19. TACO: fuel pin performance analysis

    International Nuclear Information System (INIS)

    Stoudt, R.H.; Buchanan, D.T.; Buescher, B.J.; Losh, L.L.; Wilson, H.W.; Henningson, P.J.

    1977-08-01

    The thermal performance of fuel in an LWR during its operational lifetime must be described for LOCA analysis as well as for other safety analyses. The determination of stored energy in the LOCA analysis, for example, requires a conservative fuel pin thermal performance model that is capable of calculating fuel and cladding behavior, including the gap conductance between the fuel and cladding, as a function of burnup. The determination of parameters that affect the fuel and cladding performance, such as fuel densification, fission gas release, cladding dimensional changes, fuel relocation, and thermal expansion, should be accounted for in the model. Babcock and Wilcox (B and W) has submitted a topical report, BAW-10087P, December 1975, which describes their thermal performance model TACO. A summary of the elements that comprise the TACO model and an evaluation are presented

  20. 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)

  1. 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)

  2. Handbook of fuel cell performance

    Energy Technology Data Exchange (ETDEWEB)

    Benjamin, T.G.; Camara, E.H.; Marianowski, L.G.

    1980-05-01

    The intent of this document is to provide a description of fuel cells, their performances and operating conditions, and the relationship between fuel processors and fuel cells. This information will enable fuel cell engineers to know which fuel processing schemes are most compatible with which fuel cells and to predict the performance of a fuel cell integrated with any fuel processor. The data and estimates presented are for the phosphoric acid and molten carbonate fuel cells because they are closer to commercialization than other types of fuel cells. Performance of the cells is shown as a function of operating temperature, pressure, fuel conversion (utilization), and oxidant utilization. The effect of oxidant composition (for example, air versus O/sub 2/) as well as fuel composition is examined because fuels provided by some of the more advanced fuel processing schemes such as coal conversion will contain varying amounts of H/sub 2/, CO, CO/sub 2/, CH/sub 4/, H/sub 2/O, and sulfur and nitrogen compounds. A brief description of fuel cells and their application to industrial, commercial, and residential power generation is given. The electrochemical aspects of fuel cells are reviewed. The phosphoric acid fuel cell is discussed, including how it is affected by operating conditions; and the molten carbonate fuel cell is discussed. The equations developed will help systems engineers to evaluate the application of the phosphoric acid and molten carbonate fuel cells to commercial, utility, and industrial power generation and waste heat utilization. A detailed discussion of fuel cell efficiency, and examples of fuel cell systems are given.

  3. Advanced fuel technology and performance

    International Nuclear Information System (INIS)

    1985-10-01

    The purpose of the Advisory Group Meeting on Advanced Fuel Technology and Performance was to review the experience of advanced fuel fabrication technology, its performance, peculiarities of the back-end of the nuclear fuel cycle with regard to all types of reactors and to outline the future trends. As a result of the meeting recommendations were made for the future conduct of work on advanced fuel technology and performance. A separate abstract was prepared for each of the 20 papers in this issue

  4. Nuclear fuel elements design, fabrication and performance

    CERN Document Server

    Frost, Brian R T

    1982-01-01

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

  5. 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)

  6. Spent fuel and materials performance in wet and dry storage

    Energy Technology Data Exchange (ETDEWEB)

    Zuloaga, P [ENRESA (Spain)

    2012-07-01

    According to the 6th General Radioactive Waste Plan, spent fuel in Spain shall have to be gathered in a Centralised Temporary Storage (CTS) during some decades in order to have time for a decision concerning its final fate: direct disposal at a geological repository or partitioning and transmutation if technology opens this possibility when the decision will be taken, expected in 2050. The CTS technology has already been chosen as a vault type building based in spent fuel dry storage. To support the use of this technology, a number of programmes have been completed or are still in progress, mostly concerned about high burnup fuel issues and new cladding materials. These programmes are directly managed by ENRESA alone or in joint venture with other parties, at a national and international level. Apart from that, there are contacts with other countries organisms who share similar interests with Spanish ones. The objectives are: Review of spent fuel data relevant for future storage in Spain; Perform destructive and non-destructive examinations on irradiated and non-irradiated fuel rods relevant to Spanish spent fuel management.

  7. 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

  8. 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.

  9. 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.

  10. Low-enriched fuel particle performance review

    International Nuclear Information System (INIS)

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

    1978-08-01

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

  11. Reference Material Properties and Standard Problems to Verify the Fuel Performance Models Ver 1.0

    International Nuclear Information System (INIS)

    Yang, Yong Sik; Kim, Jae Yong; Koo, Yang Hyun

    2010-12-01

    All fuel performance models must be validated by in-pile and out-pile tests. However, the model validation requires much efforts and times to confirm its exactness. In many fields, new performance models and codes are confirmed by code-to-code benchmarking process under simplified standard problem analysis. At present, the DUOS, which is the steady state fuel performance analysis code for dual cooled annular fuel, development project is progressing and new FEM module is developed to analyze the fuel performance during transient period. In addition, the verification process is planning to examine the new models and module's rightness by comparing with commercial finite element analysis such as a ADINA, ABAQUS and ANSYS. This reports contains the result of unification of material properties and establishment of standard problem to verify the newly developed models with commercial FEM code

  12. 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

  13. 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.

  14. 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

  15. 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.

  16. 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)

  17. In situ metal ion contamination and the effects on proton exchange membrane fuel cell performance

    Science.gov (United States)

    Sulek, Mark; Adams, Jim; Kaberline, Steve; Ricketts, Mark; Waldecker, James R.

    Automotive fuel cell technology has made considerable progress, and hydrogen fuel cell vehicles are regarded as a possible long-term solution to reduce carbon dioxide emissions, reduce fossil fuel dependency and increase energy efficiency. Even though great strides have been made, durability is still an issue. One key challenge is controlling MEA contamination. Metal ion contamination within the membrane and the effects on fuel cell performance were investigated. Given the possible benefits of using stainless steel or aluminum for balance-of-plant components or bipolar plates, cations of Al, Fe, Ni and Cr were studied. Membranes were immersed in metal sulfide solutions of varying concentration and then assembled into fuel cell MEAs tested in situ. The ranking of the four transition metals tested in terms of the greatest reduction in fuel cell performance was: Al 3+ ≫ Fe 2+ > Ni 2+, Cr 3+. For iron-contaminated membranes, no change in cell performance was detected until the membrane conductivity loss was greater than approximately 15%.

  18. Performance and fuel cycle cost analysis of one Janus 30 conceptual design for several fuel element design options

    Energy Technology Data Exchange (ETDEWEB)

    Nurdin, Martias [Research Centre for Nuclear Techniques, National Atomic Energy Agency (Indonesia); Matos, J E; Freese, K E [RERTR Program, Argonne National Laboratory (United States)

    1983-09-01

    The performance and fuel cycle costs for a 25 MW, JANUS 30 reactor conceptual design by INTERATOM, Federal Republic of Germany, for BATAN, Republic of Indonesia have been studied using 19.75% enriched uranium in four fuel element design options. All of these fuel element designs have either been proposed by INTERATOM for various reactors or are currently in use with 93% enriched uranium in reactors in the Federal Republic of Germany. Aluminide, oxide, and silicide fuels were studied for selected designs using the range of uranium densities that are either currently qualified or are being developed and demonstrated internationally. These uranium densities include 1.7-2.3 g/cm{sup 3} in aluminide fuel, 1.7-3.2 g/cm{sup 3} in oxide fuel, and 2.9-6.8 g/cm{sup 3} in silicide fuel. As of November 1982) both the aluminide and the oxide fuels with about 1.7 g U/cm{sup 3} are considered to be fully-proven for licensing purposes. Irradiation screening and proof testing of fuels with uranium densities greater than 1.7 g/cm{sup 3} are currently in progress, and these tests need to be completed in order to obtain licensing authorization for routine reactor use. To assess the long-term fuel adaptation strategy as well as the present fuel acceptance, reactor performance and annual fuel cycle costs were computed for seventeen cases based on a representative end-of-cycle excess reactivity and duty factor. In addition, a study was made to provide data for evaluating the trade-off between the increased safety associated with thicker cladding and the economic penalty due to increased fuel consumption. (author)

  19. LWR fuel performance during anticipated transients with scram

    International Nuclear Information System (INIS)

    Martinson, Z.R.; McCardell, R.K.; MacDonanl, P.E.; Rowland, T.C.; Tokar, M.

    1983-01-01

    Operational transients occur occasionally in light water reactors when minor malfunctions of certain system components affect the reactor core. Potential effects of such malfunctions include a loss of the secondary heat sink, an increase in system pressure, and, in boiling water reactors, void collapse and a brief increase in reactor power. The most severe postulated Boiling Water Reactor (BWR) anticipated transient is characterized by a power peak of up to 495% rated power for about 1 second (according to a recent General Electric Co., generic analysis). The results of a series of fuel behaviour tests in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory are presented in this paper. Four progressively higher and broader power transients at a constant coolant flow rate were performed. The first transient simulated a BWR-5 turbine trip without steam bypass with fuel rods operating at BWR-6 core average rod powers. The second transient simulated a generator load rejection without steam bypass with fuel rods operating at above core average powers. The last two transients were performed at higher powers than safety analysis predicts to be possible in commercial reactors to be defined failure threshold margins. The test rods did not fail and were not damaged during any of the four transients. (author)

  20. 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

  1. High energy-density liquid rocket fuel performance

    Science.gov (United States)

    Rapp, Douglas C.

    1990-01-01

    A fuel performance database of liquid hydrocarbons and aluminum-hydrocarbon fuels was compiled using engine parametrics from the Space Transportation Engine Program as a baseline. Propellant performance parameters are introduced. General hydrocarbon fuel performance trends are discussed with respect to hydrogen-to-carbon ratio and heat of formation. Aluminum-hydrocarbon fuel performance is discussed with respect to aluminum metal loading. Hydrocarbon and aluminum-hydrocarbon fuel performance is presented with respect to fuel density, specific impulse and propellant density specific impulse.

  2. 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.

  3. AREVA's fuel assemblies addressing high performance requirements of the worldwide PWR fleet

    International Nuclear Information System (INIS)

    Anniel, Marc; Bordy, Michel-Aristide

    2009-01-01

    Taking advantage of its presence in the fuel activities since the start of commercial nuclear worldwide operation, AREVA is continuing to support the customers with the priority on reliability, to: >participate in plant operational performance for the in core fuel reliability, the Zero Tolerance for Failure ZTF as a continuous improvement target and the minimisation of manufacturing/quality troubles, >guarantee the supply chain a proven product stability and continuous availability, >support performance improvements with proven design and technology for fuel management updating and cycle cost optimization, >support licensing assessments for fuel assembly and reloads, data/methodologies/services, >meet regulatory challenges regarding new phenomena, addressing emergent performance issues and emerging industry challenges for changing operating regimes. This capacity is based on supplies by AREVA accumulating very large experience both in manufacturing and in plant operation, which is demonstrated by: >manufacturing location in 4 countries including 9 fuel factories in USA, Germany, Belgium and France. Up to now about 120,000 fuel assemblies and 8,000 RCCA have been released to PWR nuclear countries, from AREVA European factories, >irradiation performed or in progress in about half of PWR world wide nuclear plants. Our optimum performances cover rod burn ups of to 82GWD/tU and fuel assemblies successfully operated under various world wide fuel management types. AREVA's experience, which is the largest in the world, has the extensive support of the well known fuel components such as the M5'TM'cladding, the MONOBLOC'TM'guide tube, the HTP'TM' and HMP'TM' structure components and the comprehensive services brought in engineering, irradiation and post irradiation fields. All of AREVA's fuel knowledge is devoted to extend the definition of fuel reliability to cover the whole scope of fuel vendor support. Our Top Reliability and Quality provide customers with continuous

  4. Irradiation performance of metallic fuels

    International Nuclear Information System (INIS)

    Pahl, R.G.; Lahm, C.E.; Porter, D.L.; Batte, G.L.; Hofman, G.L.

    1989-01-01

    Argonne National Laboratory has been working for the past five years to develop and demonstrate the Integral Fast Reactor (IFR) concept. The concept involves a closed system for fast-reactor power generation and on-site fuel reprocessing, both designed specifically around the use of metallic fuel. The Experimental Breeder Reactor-II (EBR-II) has used metallic fuel for all of its 25-year life. In 1985, tests were begun to examine the irradiation performance of advanced-design metallic fuel systems based on U-Zr or U-Pu-Zr fuels. These tests have demonstrated the viable performance of these fuel systems to high burnup. The initial testing program will be described in this paper. 2 figs

  5. Fuel performance analysis code 'FAIR'

    International Nuclear Information System (INIS)

    Swami Prasad, P.; Dutta, B.K.; Kushwaha, H.S.; Mahajan, S.C.; Kakodkar, A.

    1994-01-01

    For modelling nuclear reactor fuel rod behaviour of water cooled reactors under severe power maneuvering and high burnups, a mechanistic fuel performance analysis code FAIR has been developed. The code incorporates finite element based thermomechanical module, physically based fission gas release module and relevant models for modelling fuel related phenomena, such as, pellet cracking, densification and swelling, radial flux redistribution across the pellet due to the build up of plutonium near the pellet surface, pellet clad mechanical interaction/stress corrosion cracking (PCMI/SSC) failure of sheath etc. The code follows the established principles of fuel rod analysis programmes, such as coupling of thermal and mechanical solutions along with the fission gas release calculations, analysing different axial segments of fuel rod simultaneously, providing means for performing local analysis such as clad ridging analysis etc. The modular nature of the code offers flexibility in affecting modifications easily to the code for modelling MOX fuels and thorium based fuels. For performing analysis of fuel rods subjected to very long power histories within a reasonable amount of time, the code has been parallelised and is commissioned on the ANUPAM parallel processing system developed at Bhabha Atomic Research Centre (BARC). (author). 37 refs

  6. CANDU reactor experience: fuel performance

    International Nuclear Information System (INIS)

    Truant, P.T.; Hastings, I.J.

    1985-07-01

    Ontario Hydro has more than 126 reactor-years experience in operating CANDU reactors. Fuel performance has been excellent with 47 000 channel fuelling operations successfully completed and 99.9 percent of the more than 380 000 bundles irradiated operating as designed. Fuel performance limits and fuel defects have had a negligible effect on station safety, reliability, the environment and cost. The actual incapability charged to fuel is less than 0.1 percent over the stations' lifetimes, and more recently has been zero

  7. 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

  8. The Fuel Performance Analysis of LWR Fuel containing High Thermal Conductivity Reinforcements

    International Nuclear Information System (INIS)

    Kim, Seung Su; Ryu, Ho Jin

    2015-01-01

    The thermal conductivity of fuel affects many performance parameters including the fuel centerline temperature, fission gas release and internal pressure. In addition, enhanced safety margin of fuel might be expected when the thermal conductivity of fuel is improved by the addition of high thermal conductivity reinforcements. Therefore, the effects of thermal conductivity enhancement on the fuel performance of reinforced UO2 fuel with high thermal conductivity compounds should be analyzed. In this study, we analyzed the fuel performance of modified UO2 fuel with high thermal conductivity reinforcements by using the FRAPCON-3.5 code. The fissile density and mechanical properties of the modified fuel are considered the same with the standard UO2 fuel. The fuel performance of modified UO2 with high thermal conductivity reinforcements were analyzed by using the FRAPCON-3.5 code. The thermal conductivity enhancement factors of the modified fuels were obtained from the Maxwell model considering the volume fraction of reinforcements

  9. Water Reactor Fuel Performance Meeting 2008

    International Nuclear Information System (INIS)

    2008-10-01

    This meeting contains articles of the Water Reactor Fuel Performance Meeting 2008 of Korean Nuclear Society, Atomic Energy Society of Japan, Chinese Nuclear Society, European Nuclear Society and American Nuclear Society. It was held on Oct. 19-23, 2008 in Seoul, Korea and subject of Meeting is 'New Clear' Fuel - A green energy solution. This proceedings is comprised of 5 tracks. The main topic titles of track are as follows: Advances in water reactor fuel technology, Fuel performance and operational experience, Transient fuel behavior and safety-related issues, Fuel cycle, spent fuel storage and transportations and Fuel modeling and analysis. (Yi, J. H.)

  10. HTGR Fuel performance basis

    International Nuclear Information System (INIS)

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

    1982-05-01

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

  11. Low-enriched fuel particle performance review

    International Nuclear Information System (INIS)

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

    1978-08-01

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

  12. Structure of fuel performance audit code for SFR metal fuel

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Yong Sik; Kim, Hyo Chan [KAERI, Daejeon (Korea, Republic of); Jeong, Hye Dong; Shin, An Dong; Suh, Nam Duk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2012-10-15

    A Sodium Cooled Fast Reactor (SFR) is a promising option to solve the spent fuel problems, but, there are still much technical issues to commercialize a SFR. One of issues is a development of advanced fuel which can solve the safety and the economic issues at the same time. Since a nuclear fuel is the first barrier to protect radioactive isotope release, the fuel's integrity must be secured. In Korea Institute of Nuclear Safety (KINS), the new project has been started to develop the regulatory technology for SFR system including a fuel area. To evaluate the fuel integrity and safety during an irradiation, the fuel performance code must be used for audit calculation. To develop the new code system, the code structure design and its requirements need to be studied. Various performance models and code systems are reviewed and their characteristics are analyzed in this paper. Based on this study, the fundamental performance models are deduced and basic code requirements and structure are established.

  13. 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

  14. Proceedings of the Water Reactor Fuel Performance Meeting - WRFPM / Top Fuel 2009

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2009-06-15

    SFEN, ENS, SNR, ANS, AESJ, CNS KNS, IAEA and NEA are jointly organizing the 2009 International Water Reactor Fuel Performance / TopFuel 2009 Meeting following the 2008 KNS Water Reactor Performance Meeting held during October 19-23, 2008 in Seoul, Korea. This meeting is held annually on a tri-annual rotational basis in Europe, USA and Asia. In 2009, this meeting will be held in Paris, September 6-10, 2009 in coordination with the Global 2009 Conference at the same date and place. That would lead to a common opening session, some common technical presentations, a common exhibition and common social events. The technical scope of the meeting includes all aspects of nuclear fuel from fuel rod to core design as well as manufacturing, performance in commercial and test reactors or on-going and future developments and trends. Emphasis will be placed on fuel reliability in the general context of nuclear 'Renaissance' and recycling perspective. The meeting includes selectively front and/or back end issues that impact fuel designs and performance. In this frame, the conference track devoted to 'Concepts for transportation and interim storage of spent fuels and conditioned waste' will be shared with 'GLOBAL' conference. Technical Tracks: - 1. Fuel Performance, Reliability and Operational Experience: Fuel operating experience and performance; experience with high burn-up fuels; water side corrosion; stress corrosion cracking; MOX fuel performance; post irradiation data on lead fuel assemblies; radiation effects; water chemistry and corrosion counter-measures. - 2. Transient Fuel Behaviour and Safety Related Issues: Transient fuel behavior and criteria (RIA, LOCA, ATWS, Ramp tests..). Fuel safety-related issues such as PCI (pellet cladding interaction), transient fission gas releases and cladding bursting/ballooning during transient events - Advances in fuel performance modeling and core reload methodology, small and large-scale fuel testing

  15. Proceedings of the Water Reactor Fuel Performance Meeting - WRFPM / Top Fuel 2009

    International Nuclear Information System (INIS)

    2009-06-01

    SFEN, ENS, SNR, ANS, AESJ, CNS KNS, IAEA and NEA are jointly organizing the 2009 International Water Reactor Fuel Performance / TopFuel 2009 Meeting following the 2008 KNS Water Reactor Performance Meeting held during October 19-23, 2008 in Seoul, Korea. This meeting is held annually on a tri-annual rotational basis in Europe, USA and Asia. In 2009, this meeting will be held in Paris, September 6-10, 2009 in coordination with the Global 2009 Conference at the same date and place. That would lead to a common opening session, some common technical presentations, a common exhibition and common social events. The technical scope of the meeting includes all aspects of nuclear fuel from fuel rod to core design as well as manufacturing, performance in commercial and test reactors or on-going and future developments and trends. Emphasis will be placed on fuel reliability in the general context of nuclear 'Renaissance' and recycling perspective. The meeting includes selectively front and/or back end issues that impact fuel designs and performance. In this frame, the conference track devoted to 'Concepts for transportation and interim storage of spent fuels and conditioned waste' will be shared with 'GLOBAL' conference. Technical Tracks: - 1. Fuel Performance, Reliability and Operational Experience: Fuel operating experience and performance; experience with high burn-up fuels; water side corrosion; stress corrosion cracking; MOX fuel performance; post irradiation data on lead fuel assemblies; radiation effects; water chemistry and corrosion counter-measures. - 2. Transient Fuel Behaviour and Safety Related Issues: Transient fuel behavior and criteria (RIA, LOCA, ATWS, Ramp tests..). Fuel safety-related issues such as PCI (pellet cladding interaction), transient fission gas releases and cladding bursting/ballooning during transient events - Advances in fuel performance modeling and core reload methodology, small and large-scale fuel testing facilities. - 3. Advances in Water

  16. Preparation, characterisation, engine performance and emission characteristics of coconut oil based hybrid fuels

    Energy Technology Data Exchange (ETDEWEB)

    Singh, Pranil J.; Singh, Anirudh [Division of Physics, School of Engineering and Physics, Faculty of Science, Technology and Environment, University of the South Pacific, 325 Fletcher Road, Suva (Fiji); Khurma, Jagjit [Division of Chemistry, School of Biological, Chemical and Environmental Sciences, Faculty of Science, Technology and Environment, University of the South Pacific, Suva (Fiji)

    2010-09-15

    In this study, hybrid fuels consisting of coconut oil, aqueous ethanol and a surfactant (butan-1-ol) were prepared and tested as a fuel in a direct injection diesel engine. After determining fuel properties such as the density, viscosity and gross calorific values of these fuels, they were used to run a diesel engine. The engine performance and exhaust emissions were investigated and compared with that of diesel. The experimental results show that the efficiency of the hybrid fuels is comparable to that of diesel. As the viscosity of the hybrid fuels decreased and approached that of diesel, the efficiency increased progressively towards that of diesel. The exhaust emissions were lower than those for diesel, except carbon monoxide emissions, which increased. Hence, it is concluded that these hybrid fuels can be used successfully as an alternative fuel in diesel engines without any modifications. Their completely renewable nature ensures that they are environmentally friendly with regard to their emissions characteristics. (author)

  17. Analytic models for fuel pin transient performance

    International Nuclear Information System (INIS)

    Bard, F.E.; Fox, G.L.; Washburn, D.F.; Hanson, J.E.

    1976-09-01

    HEDL's ability to analyze various mechanisms that operate within a fuel pin has progressed substantially through development of codes such as PECTCLAD, which solves cladding response, and DSTRESS, which solves fuel response. The PECTCLAD results show good correlation with a variety of mechanical tests on cladding material and also demonstrate the significance of cladding strength when applying the life fraction rule. The DSTRESS results have shown that fuel deforms sufficiently during overpower transient tests that available volumes are filled, whether in the form of a central cavity or start-up cracks

  18. 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

  19. Fuel performance-REP, Seminars on nuclear fuel performance based on basic underlining phenomena, proceedings

    International Nuclear Information System (INIS)

    2008-01-01

    Description: The need for further improving the understanding of basic phenomena underlying nuclear fuel behaviour has been recognised both by fuel vendors, experts in fuel research in the different laboratories and committees and working groups coordinating international activities. The OECD/NEA Nuclear Science Committee has established an Experts Group addressing this issue. This has led to establishing an International Fuel Performance Experiments Database (IFPE) that should help model evaluation and validation. Many years ago the IAEA established an International Working Group on Fuel Performance and Technology (IWGFPT) that led to the FUMEX-I and FUMEX-II (Fuel Modelling Exercise) which has had an important impact on code improvements. Both international organisations, with the support of national organisations, co-operate in establishing and maintaining the Database and to build confidence in the predictive power of the models through international comparison exercises. But above all the different parties have agreed that seminars focussed on specific phenomena would be beneficial to exchange current knowledge, identify outstanding problems and agree on common action that would lead to improved understanding of the phenomena. A series of three seminars has been initiated by the Commissariat a l'Energie Atomique (CEA), Electricite de France (EdF), Framatome and Cogema under the aegis of the OECD/NEA and the IAEA. 1. Thermal Performance of High Burn-Up LWR Fuel at Cadarache, France, from 3 to 6 of March 1998. Thermal performance occupies the most important aspect of the fuel performance modelling. Not only is it extremely important from a safety point of view, but also many of the material properties of interest and behaviour, such as transport properties like fuel creep and fission gas release are thermally activated processes. Thus, in order to model these processes correctly, it is critical to calculate temperatures and their distribution as accurately as

  20. Fuel performance of DOE fuels in water storage

    International Nuclear Information System (INIS)

    Hoskins, A.P.; Scott, J.G.; Shelton-Davis, C.V.; McDannel, G.E.

    1993-01-01

    Westinghouse Idaho Nuclear Company operates the Idaho Chemical Processing Plant (ICPP) at the Idaho National Engineering Laboratory. In April of 1992, the U.S. Department of Energy (DOE) decided to end the fuel reprocessing mission at ICPP. Fuel performance in storage received increased emphasis as the fuel now needs to be stored until final dispositioning is defined and implemented. Fuels are stored in four main areas: an original underwater storage facility, a modern underwater storage facility, and two dry fuel storage facilities. As a result of the reactor research mission of the DOE and predecessor agencies, the Energy Research and Development Administration and the Atomic Energy Commission, many types of nuclear fuel have been developed, used, and assigned to storage at the ICPP. Fuel clad with stainless steel, zirconium, aluminum, and graphite are represented. Fuel matrices include uranium oxide, hydride, carbide, metal, and alloy fuels, resulting in 55 different fuel types in storage. Also included in the fuel storage inventory is canned scrap material

  1. 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

  2. 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

  3. Nitride fuels irradiation performance data base

    International Nuclear Information System (INIS)

    Brozak, D.E.; Thomas, J.K.; Peddicord, K.L.

    1987-01-01

    An irradiation performance data base for nitride fuels has been developed from an extensive literature search and review that emphasized uranium nitride, but also included performance data for mixed nitrides [(U,Pu)N] and carbonitrides [(U,Pu)C,N] to increase the quantity and depth of pin data available. This work represents a very extensive effort to systematically collect and organize irradiation data for nitride-based fuels. The data base has many potential applications. First, it can facilitate parametric studies of nitride-based fuels to be performed using a wide range of pin designs and operating conditions. This should aid in the identification of important parameters and design requirements for multimegawatt and SP-100 fuel systems. Secondly, the data base can be used to evaluate fuel performance models. For detailed studies, it can serve as a guide to selecting a small group of pin specimens for extensive characterization. Finally, the data base will serve as an easily accessible and expandable source of irradiation performance information for nitride fuels

  4. 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.

  5. Proceedings of the 2007 LWR Fuel Performance Meeting / TopFuel 2007 'Zero by 2010'

    International Nuclear Information System (INIS)

    2007-01-01

    ANS, ENS, AESJ and KNS are jointly organizing the 2007 International LWR Fuel Performance Meeting following the successful ENS TopFuel meeting held during 22-26 October, 2006 in Salamaca, Spain. Merging three premier nuclear fuel design and performance meetings: the ANS LWR Fuel Performance Meeting, the ENS TopFuel and Asian Water Reactor Fuel Performance Meeting (WRFPM) created this international meeting. The meeting will be held annually on a tri-annual rotational basis in USA, Asia, and Europe. The technical scope of the meeting includes all aspects of nuclear fuel from fuel rod to core design as well as performance experience in commercial and test reactors. The meeting excludes front end and back end fuel issues, however, it covers all front and/or back issues that impact fuel designs and performance

  6. 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.

  7. Fuels Performance | Transportation Research | NREL

    Science.gov (United States)

    chemistry, combustion, and emissions to answer two primary questions: How can improved fuel chemistry lead initiative Combustion and engines Fuel chemistry Emissions and fuel economy Find out about related NREL Chemistry. (2014) Saturated Monoglyceride Effects on Low-Temperature Performance of Biodiesel Blends. G.M

  8. Review on Fuel Loading Process and Performance for Advanced Fuel Handling Equipment

    International Nuclear Information System (INIS)

    Chang, Sang-Gyoon; Lee, Dae-Hee; Kim, Young-Baik; Lee, Deuck-Soo

    2007-01-01

    The fuel loading process and the performance of the advanced fuel handling equipment for OPR 1000 (Optimized Power Plant) are analyzed and evaluated. The fuel handling equipment, which acts critical processes in the refueling outage, has been improved to reduce fuel handling time. The analysis of the fuel loading process can be a useful tool to improve the performance of the fuel handling equipment effectively. Some recommendations for further improvement are provided based on this study

  9. 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.

  10. 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.

  11. 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.

  12. Alternate-Fueled Combustor-Sector Performance

    Science.gov (United States)

    Thomas, Anna E.; Saxena, Nikita T.; Shouse, Dale T.; Neuroth, Craig; Hendricks, Robert C.; Lynch, Amy; Frayne, Charles W.; Stutrud, Jeffrey S.; Corporan, Edwin; Hankins, Terry

    2013-01-01

    In order to realize alternative fueling for military and commercial use, the industry has set forth guidelines that must be met by each fuel. These aviation fueling requirements are outlined in MIL-DTL-83133F(2008) or ASTM D 7566 Annex (2011) standards, and are classified as "drop-in" fuel replacements. This report provides combustor performance data for synthetic-paraffinic-kerosene- (SPK-) type (Fischer-Tropsch (FT)) fuel and blends with JP-8+100, relative to JP-8+100 as baseline fueling. Data were taken at various nominal inlet conditions: 75 psia (0.52 MPa) at 500 degF (533 K), 125 psia (0.86 MPa) at 625 degF (603 K), 175 psia (1.21 MPa) at 725 degF (658 K), and 225 psia (1.55 MPa) at 790 degF (694 K). Combustor performance analysis assessments were made for the change in flame temperatures, combustor efficiency, wall temperatures, and exhaust plane temperatures at 3, 4, and 5 percent combustor pressure drop (DP) for fuel:air ratios (F/A) ranging from 0.010 to 0.025. Significant general trends show lower liner temperatures and higher flame and combustor outlet temperatures with increases in FT fueling relative to JP-8+100 fueling. The latter affects both turbine efficiency and blade and vane lives.

  13. Irradiation test and performance evaluation of DUPIC fuel

    International Nuclear Information System (INIS)

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

    2002-05-01

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

  14. Fuel performance annual report for 1981

    International Nuclear Information System (INIS)

    Bailey, W.J.; Tokar, M.

    1982-12-01

    This annual report, the fourth in a series, provides a brief description of fuel performance during 1981 in commercial nuclear power plants. Brief summaries of fuel operating experience, fuel problems, fuel design changes and fuel surveillance programs, and high-burnup fuel experience are provided. References to additional, more detailed information and related NRC evaluations are included

  15. [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.

  16. Surveillance of WWER-440 fuel performance

    International Nuclear Information System (INIS)

    Simko, J.; Urban, P.

    1999-01-01

    In this lecture next problems of surveillance of WWER-440 fuel performance are presented: surveillance of WWER-440 fuel performance at Mochovce NPP; basic data of WWER-440 reactor; in-core reactor measuring system 'SVRK'; basic level of SVRK; information output of basic level of SVRK; surveillance of fuel performance; table of permissible operation conditions of the reactor; limitation of the unit 1 power at the beginning of the operation; cyclic changes of power; future perspectives

  17. BISON Fuel Performance Analysis of FeCrAl cladding with updated properties

    Energy Technology Data Exchange (ETDEWEB)

    Sweet, Ryan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); George, Nathan M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wirth, Brian [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-08-30

    operating conditions used are based off the Peach Bottom BWR and design consideration was given to minimize the neutronic penalty of the FeCrAl cladding by changing fuel enrichment and cladding thickness. As this study progressed, systematic parametric analysis of the fuel and cladding creep responses were also performed.

  18. Fuel performance annual report for 1990

    International Nuclear Information System (INIS)

    Preble, E.A.; Painter, C.L.; Alvis, J.A.; Berting, F.M.; Beyer, C.E.; Payne, G.A.; Wu, S.L.

    1993-11-01

    This annual report, the thirteenth in a series, provides a brief description of fuel performance during 1990 in commercial nuclear power plants. Brief summaries of fuel design changes, fuel surveillance programs, fuel operating experience and trends, fuel problems high-burnup fuel experience, and items of general significance are provided . References to additional, more detailed information, and related NRC evaluations are included where appropriate

  19. MONJU fuel pin performance analysis

    International Nuclear Information System (INIS)

    Kitagawa, H.; Yamanaka, T.; Hayashi, H.

    1979-01-01

    Monju fuel pin has almost the same properties as other LMFBR fuel pins, i.e. Phenix, PFR, CRBR, but would be irradiated under severe conditions: maximum linear heat rate of 381 watt/cm, hot spot cladding temperature of 675 deg C, peak burnup of 131,000 MWd/t, peak fluence (E greater than 0.1 MeV) of 2.3 10 23 n/cm 2 . In order to understand in-core performance of Monju fuel pin, its thermal and mechanical behaviour was predicted using the fast running performance code SIMPLE. The code takes into account pellet-cladding interaction due to thermal expansion and swelling, gap conductance, structural changes of fuel pellets, fission product gas release with burnup and temperature increase, swelling and creep of fuel pellets, corrosion of cladding due to sodium flow and chemical attack by fission products, and cumulative damage of the cladding due to thermal creep

  20. Fuel performance annual report for 1989

    International Nuclear Information System (INIS)

    Bailey, W.J.; Berting, F.M.; Wu, S.

    1992-06-01

    This annual report, the twelfth in a series, provides a brief description of fuel performance during 1989 in commercial nuclear power plants and an indication of trends. Brief summaries of fuel design changes, fuel surveillance programs, fuel operating experience, fuel problems, high-burnup fuel experience, and items of general significance are provided. References to more detailed information and related US Nuclear Regulatory Commission evaluations are included

  1. Reactor fuel performance data file, 1985 edition

    International Nuclear Information System (INIS)

    Harayama, Yasuo; Fujita, Misao; Watanabe, Kohji.

    1986-07-01

    In safety evaluation and integrity studies of reactor fuel, data on fuel performance are the most basic materials. The Fuel Reliability Laboratory No.1 has obtained the fuel performance data by joining in some international programs to study the safety and integrity of fuel. Those data have only used for the studies in the above two fields. However, if the data are rearranged and compiled in a easily usable form, they can be utilized in other field of studies. Then, a 'data file' on fuel performance is beeing compiled by adding data from open literatures to those obtained in international programs. The present report is prepared on the basis of the data file compiled by March in 1986. (author)

  2. U.S. report on fuel performance and technology

    Energy Technology Data Exchange (ETDEWEB)

    Cook, T [Department of Energy, Washington, DC (United States). Office of Engineering and Technology Development

    1997-12-01

    The report reviews the following aspects of fuel performance and technology: increased demand on fuel performance;improved fuel failure rate; operating fuel cycles; capacity factor for US nuclear electric generating plants; potential reduction of SNF due to improved fuel burnup.

  3. High Burnup Fuel Performance and Safety Research

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Je Keun; Lee, Chan Bok; Kim, Dae Ho (and others)

    2007-03-15

    The worldwide trend of nuclear fuel development is to develop a high burnup and high performance nuclear fuel with high economies and safety. Because the fuel performance evaluation code, INFRA, has a patent, and the superiority for prediction of fuel performance was proven through the IAEA CRP FUMEX-II program, the INFRA code can be utilized with commercial purpose in the industry. The INFRA code was provided and utilized usefully in the universities and relevant institutes domesticallly and it has been used as a reference code in the industry for the development of the intrinsic fuel rod design code.

  4. 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

  5. Nuclear fuel performance in boiling water reactors

    International Nuclear Information System (INIS)

    Elkins, R.B.; Baily, W.E.; Proebstle, R.A.; Armijo, J.S.; Klepfer, H.H.

    1981-01-01

    A major development program is described to improve the performance of Boiling Water Reactor fuel. This sustained program is described in four parts: 1) performance monitoring, 2) fuel design changes, 3) plant operating recommendations, and 4) advanced fuel programs

  6. Code structure for U-Mo fuel performance analysis in high performance research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Gwan Yoon; Cho, Tae Won; Lee, Chul Min; Sohn, Dong Seong [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Lee, Kyu Hong; Park, Jong Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    A performance analysis modeling applicable to research reactor fuel is being developed with available models describing fuel performance phenomena observed from in-pile tests. We established the calculation algorithm and scheme to best predict fuel performance using radio-thermo-mechanically coupled system to consider fuel swelling, interaction layer growth, pore formation in the fuel meat, and creep fuel deformation and mass relocation, etc. In this paper, we present a general structure of the performance analysis code for typical research reactor fuel and advanced features such as a model to predict fuel failure induced by combination of breakaway swelling and pore growth in the fuel meat. Thermo-mechanical code dedicated to the modeling of U-Mo dispersion fuel plates is being under development in Korea to satisfy a demand for advanced performance analysis and safe assessment of the plates. The major physical phenomena during irradiation are considered in the code such that interaction layer formation by fuel-matrix interdiffusion, fission induced swelling of fuel particle, mass relocation by fission induced stress, and pore formation at the interface between the reaction product and Al matrix.

  7. 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

  8. Proceedings of the Topical meeting on the reactor fuel performance - TopFuel 2012 Transactions

    International Nuclear Information System (INIS)

    2012-01-01

    TopFuel is an annual topical meeting organised by ENS, the American Nuclear Society and the Atomic Energy Society of Japan. TopFuel's primary objective is to bring together leading specialists in the field from around the world to analyse advances in nuclear fuel management technology and to use the findings of the latest cutting-edge research to help manufacture the high performance nuclear fuels of today and tomorrow. Aim is to discuss the challenges facing the developers and manufacturers of new high-performance nuclear fuels - fuels that will help meet current and future energy demand and reduce man's over dependence upon CO 2 -emitting fossil fuels. The technical scope of Top Fuel 2012 includes all aspects of nuclear fuel from fuel rod to core design as well as manufacturing, performance in commercial and test reactors or on-going and future developments and trends. The meeting includes selectively front and/or back end issues that impact fuel designs and performance. (authors)

  9. 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)

  10. Irradiation performance of HTGR recycle fissile fuel

    International Nuclear Information System (INIS)

    Homan, F.J.; Long, E.L. Jr.

    1976-08-01

    The irradiation performance of candidate HTGR recycle fissile fuel under accelerated testing conditions is reviewed. Failure modes for coated-particle fuels are described, and the performance of candidate recycle fissile fuels is discussed in terms of these failure modes. The bases on which UO 2 and (Th,U)O 2 were rejected as candidate recycle fissile fuels are outlined, along with the bases on which the weak-acid resin (WAR)-derived fissile fuel was selected as the reference recycle kernel. Comparisons are made relative to the irradiation behavior of WAR-derived fuels of varying stoichiometry and conclusions are drawn about the optimum stoichiometry and the range of acceptable values. Plans for future testing in support of specification development, confirmation of the results of accelerated testing by real-time experiments, and improvement in fuel performance and reliability are described

  11. Fuel performance annual report for 1986

    International Nuclear Information System (INIS)

    Bailey, W.J.; Wu, S.

    1988-03-01

    This annual report, the ninth in a series, provides a brief description of fuel performance during 1986 in commercial nuclear power plants and an indication of trends. Brief summaries of fuel design changes, fuel surveillance programs, fuel operating experience, fuel problems, high-burnup fuel experience, and items of general significance are provided. References to more detailed information and related U.S. Nuclear Regulatory Commission evaluations are included. 550 refs., 12 figs., 31 tabs

  12. Fuel performance: Annual report for 1987

    International Nuclear Information System (INIS)

    Bailey, W.J.; Wu, S.

    1989-03-01

    This annual report, the tenth in a series, provides a brief description of fuel performance during 1987 in commercial nuclear power plants and an indication of trends. Brief summaries of fuel design changes, fuel surveillance programs, fuel operating experience, fuel problems, high-burnup fuel experience, and items of general significance are provided. References to more detailed information and related US Nuclear Regulator Commission evaluations are included. 384 refs., 13 figs., 33 tabs

  13. HTGR fuel performance basis

    International Nuclear Information System (INIS)

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

    1982-01-01

    The safety characteristics of the high-temperature gas-cooled reactor (HTGR) during normal and accident conditions are determined in part by HTGR fuel performance. During normal operation, less than 0.1% fuel failure occurs, primarily from defective particles. This low fuel failure fraction limits circulating activity to acceptable levels. During severe accidents, the radiological consequence is influenced by high-temperature fuel particle behavior. An empirical fuel failure model, supported by recent experimental data, is presented. The onset of significant fuel particle failure occurs at temperatures in excess of 1600 0 C, and complete fuel failure occurs at 2660 0 C. This indicates that the fuel is more retentive at higher temperatures than previously assumed. The more retentive nature of the fuel coupled with the high thermal capacitance of the core results in slow release of fission products from the core during severe accidents. The slow release of fission products over hundreds of hours allows for decay of short-lived isotopes. The slow and limited release of fission products under HTGR accident conditions results in very low off-site doses. The slow nature of the accident provides more time for operator action to mitigate the accident and for local and state authorities to respond. These features can be used to take advantage of close-in siting for process applications, flexibility in site selection, and emergency planning

  14. Design and Performance of LPG Fuel Mixer for Dual Fuel Diesel Engine

    Science.gov (United States)

    Desrial; Saputro, W.; Garcia, P. P.

    2018-05-01

    Small horizontal diesel engines are commonly used for agricultural machinery, however, availability of diesel fuel become one of big problems especially in remote area. Conversely, in line with government policy for conversion of kerosene into LPG for cooking, then LPG become more popular and available even in remote area. Therefore, LPG is potential fuel to replace the shortage of diesel fuel for operating diesel engine in remote area. The purpose of this study was to design mixing device for using dual fuel i.e. LPG and diesel fuel and evaluate its performance accordingly. Simulation by using CFD was done in order to analyze mixture characteristics of LPG in air intake manifold. The performance test was done by varying the amount of LPG injected in intake air at 20%, 25%, 30%, 35%, until 40%, respectively. Result of CFD contour simulation showed the best combination when mixing 30% LPG into the intake air. Performance test of this research revealed that mixing LPG in air intake can reduce the diesel fuel consumption about 0.7 l/hour (without load) and 1.14 l/hour (with load). Diesel engine revolution increases almost 300 rpm faster than when using diesel fuel only. Based on economic analysis, using the fuel combination (diesel fuel – LPG) is not recommended in the area near SPBU where the price of diesel fuel is standard. However, using the fuel combination LPG-diesel fuel is highly recommended in the remote areas in Indonesia where price of diesel fuel is comparatively expensive which will provide cheaper total fuel cost for diesel engine operation.

  15. Delivering high performance BWR fuel reliably

    International Nuclear Information System (INIS)

    Schardt, J.F.

    1998-01-01

    Utilities are under intense pressure to reduce their production costs in order to compete in the increasingly deregulated marketplace. They need fuel, which can deliver high performance to meet demanding operating strategies. GE's latest BWR fuel design, GE14, provides that high performance capability. GE's product introduction process assures that this performance will be delivered reliably, with little risk to the utility. (author)

  16. 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.

  17. Application of the BISON Fuel Performance Code of the FUMEX-III Coordinated Research Project

    International Nuclear Information System (INIS)

    Williamson, R.L.; Novascone, S.R.

    2013-01-01

    Since 1981, the International Atomic Energy Agency (IAEA) has sponsored a series of Coordinated Research Projects (CRP) in the area of nuclear fuel modeling. These projects have typically lasted 3-5 years and have had broad international participation. The objectives of the projects have been to assess the maturity and predictive capability of fuel performance codes, support interaction and information exchange between countries with code development and application needs, build a database of well- defined experiments suitable for code validation, transfer a mature fuel modeling code to developing countries, and provide guidelines for code quality assurance and code application to fuel licensing. The fourth and latest of these projects, known as FUMEX-III1 (FUel Modeling at EXtended Burnup- III), began in 2008 and ended in December of 2011. FUMEX-III was the first of this series of fuel modeling CRP's in which the INL participated. Participants met at the beginning of the project to discuss and select a set of experiments ('priority cases') for consideration during the project. These priority cases were of broad interest to the participants and included reasonably well-documented and reliable data. A meeting was held midway through the project for participants to present and discuss progress on modeling the priority cases. A final meeting was held at close of the project to present and discuss final results and provide input for a final report. Also in 2008, the INL initiated development of a new multidimensional (2D and 3D) multiphysics nuclear fuel performance code called BISON, with code development progressing steadily during the three-year FUMEX-III project. Interactions with international fuel modeling researchers via FUMEX-III played a significant role in the BISON evolution, particularly influencing the selection of material and behavioral models which are now included in the code. The FUMEX-III cases are generally integral fuel rod experiments occurring

  18. 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

  19. Fuel Retrieval Sub-Project (FRS) Stuck Fuel Station Performance Test Data Report

    International Nuclear Information System (INIS)

    THIELGES, J.R.

    2000-01-01

    This document provides the test data report for Stuck Fuel Station Performance Testing in support of the Fuel Retrieval Sub-Project. The stuck fuel station was designed to provide a means of cutting open a canister barrel to release fuel elements, etc

  20. Delivering high performance BWR fuel reliably

    Energy Technology Data Exchange (ETDEWEB)

    Schardt, J.F. [GE Nuclear Energy, Wilmington, NC (United States)

    1998-07-01

    Utilities are under intense pressure to reduce their production costs in order to compete in the increasingly deregulated marketplace. They need fuel, which can deliver high performance to meet demanding operating strategies. GE's latest BWR fuel design, GE14, provides that high performance capability. GE's product introduction process assures that this performance will be delivered reliably, with little risk to the utility. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-04-01

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

  2. 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.

  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. Performance and reliability of LWR fuel

    International Nuclear Information System (INIS)

    Bairiot, H.; Deramaix, P.; Vandenberg, C.

    1977-01-01

    boric acid (since 1963), variable D 2 O-H 2 O moderation (1966), the utilization of pyrex (1972) and of gadolinium, first in a BWR (1973) and then in a PWR (1976). The succession and progressive completion of the experience which is detailed in the paper is mainly based on the fact that before loading a fuel in a reactor at least 3 years of preparation are required if it is a first of the kind. The results can only be fully assessed if after the life-time in the power reactor (3 years for a PWR, 5 years for a BWR), a proper postirradiation assessment is performed which requires another 1 to 3 years taking into account the extent of the programme. The data acquisition is therefore a process requiring 9 to 10 years between the decision to go ahead and the full availability of the experience [fr

  5. 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

  6. Transient performance of EBR-II driver fuel

    International Nuclear Information System (INIS)

    Buzzell, J.A.; Hudman, G.D.; Porter, D.L.

    1981-01-01

    The first phases of qualification of the EBR-II driver fuel for repeated transient overpower operation have recently been completed. The accomplishments include prediction of the transient fuel and cladding performance through ex-core testing and fuel-element modeling studies, localized in-core power testing during steady-state operation, and whole-core multiple transient testing. The metallic driver fuel successfully survived 56 transients, spaced over a 45-day period, with power increases of approx. 160% at rates of approx. 1%/s with a 720-second hold at full power. The performance results obtained from both ex-core and n-core tests indicate that the fuel is capable of repeated transient operation

  7. Irradiation and performance evaluation of DUPIC fuel

    International Nuclear Information System (INIS)

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

    2000-05-01

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

  8. Irradiation and performance evaluation of DUPIC fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-05-01

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

  9. 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

  10. Fuel performance evaluation for the CAFE experimental device

    International Nuclear Information System (INIS)

    Giovedi, Claudia; Hirota, Leandro T.

    2011-01-01

    Fuel rod cladding material is the second barrier to prevent the release of radioactive inventories in a PWR reactor. In this sense, an important safety aspect is to assess the fuel behavior under operational conditions. This can be made by means of fuel performance codes and confirmed by experimental measurements. In order to evaluate the fuel behavior of fuel rods in steady-state conditions, it was designed an experimental irradiation device, the Nuclear Fuel Irradiation Circuit (CAFE-Mod1). This device will allow controlling the surface rod temperature, to measure the power associated to the rod and the evolution of fission gas release for a typical PWR fuel pin. However, to support the experimental irradiation program, it is extremely important to simulate the experimental conditions using a fuel performance code. The aim of this paper is to evaluate some parameters and aspects related to the fuel rod behavior during the irradiation program. This evaluation was carried out by means of an adapted fuel performance code. Obtained results have shown that besides of the variation observed for parameters, such as, fuel temperature and fission gas release as a function of fuel enrichment level, the fuel rod integrity was preserved in all studied conditions. (author)

  11. FFTF fuel pin design bases and performance

    International Nuclear Information System (INIS)

    Cox, C.M.; Hanson, J.E.; Roake, W.E.; Slember, R.J.; Weber, C.E.; Millunzi, A.C.

    1975-04-01

    The FFTF fuel pin was conservatively designed to meet thermal and structural performance requirements in the categories normal operation, upset events, emergency events, and hypothetical, faulted events. The fuel pin operating limits consistent with these requirements were developed from a strong fuel pin irradiation testing program scoped to define the performance capability under relevant steady state and transient conditions. Comparison of the results of the irradiation testing program with design requirements indicates that the FFTF fuel pin can exceed its goal burnup of 80,000 MWd/MTM. (U.S.)

  12. Alternate-Fueled Combustor-Sector Performance: Part A: Combustor Performance Part B: Combustor Emissions

    Science.gov (United States)

    Shouse, D. T.; Neuroth, C.; Henricks, R. C.; Lynch, A.; Frayne, C.; Stutrud, J. S.; Corporan, E.; Hankins, T.

    2010-01-01

    Alternate aviation fuels for military or commercial use are required to satisfy MIL-DTL-83133F(2008) or ASTM D 7566 (2010) standards, respectively, and are classified as drop-in fuel replacements. To satisfy legacy issues, blends to 50% alternate fuel with petroleum fuels are certified individually on the basis of feedstock. Adherence to alternate fuels and fuel blends requires smart fueling systems or advanced fuel-flexible systems, including combustors and engines without significant sacrifice in performance or emissions requirements. This paper provides preliminary performance (Part A) and emissions and particulates (Part B) combustor sector data for synthetic-parafinic-kerosene- (SPK-) type fuel and blends with JP-8+100 relative to JP-8+100 as baseline fueling.

  13. Fuel performance annual report, period through December 1978

    International Nuclear Information System (INIS)

    Houston, M.D.

    1979-12-01

    This annual report, intended to be the first in a series, provides a brief description of fuel performance in commercial nuclear power plants. Brief summaries are given for the reporting period of on-site fuel surveillance programs, fuel performance problems, and changes in commercial fuel designs. The report provides many references to more detailed information and to related NRC evaluations. 57 references

  14. Fuel performance annual report for 1991. Volume 9

    International Nuclear Information System (INIS)

    Painter, C.L.; Alvis, J.M.; Beyer, C.E.; Marion, A.L.; Kendrick, E.D.

    1994-08-01

    This report is the fourteenth in a series that provides a compilation of information regarding commercial nuclear fuel performance. The series of annual reports were developed as a result of interest expressed by the public, advising bodies, and the US Nuclear Regulatory Commission (NRC) for public availability of information pertaining to commercial nuclear fuel performance. During 1991, the nuclear industry's focus regarding fuel continued to be on extending burnup while maintaining fuel rod reliability. Utilities realize that high-burnup fuel reduces the amount of generated spent fuel, reduces fuel costs, reduces operational and maintenance costs, and improves plant capacity factors by extending operating cycles. Brief summaries of fuel operating experience, fuel design changes, fuel surveillance programs, high-burnup experience, problem areas, and items of general significance are provided

  15. 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

  16. Performance of candu-6 fuel bundles manufactured in romania nuclear fuel plant

    International Nuclear Information System (INIS)

    Bailescu, A.; Barbu, A.; Din, F.; Dinuta, G.; Dumitru, I.; Musetoiu, A.; Serban, G.; Tomescu, A.

    2013-01-01

    The purpose of this article is to present the performance of nuclear fuel produced by Nuclear Fuel Plant (N.F.P.) - Pitesti during 1995 - 2012 and irradiated in units U1 and U2 from Nuclear Power Plant (N.P.P.) Cernavoda and also present the Nuclear Fuel Plant (N.F.P.) - Pitesti concern for providing technology to prevent the failure causes of fuel bundles in the reactor. This article presents Nuclear Fuel Plant (N.F.P.) - Pitesti experience on tracking performance of nuclear fuel in reactor and strategy investigation of fuel bundles notified as suspicious and / or defectives both as fuel element and fuel bundle, it analyzes the possible defects that can occur at fuel bundle or fuel element and can lead to their failure in the reactor. Implementation of modern technologies has enabled optimization of manufacturing processes and hence better quality stability of achieving components (end caps, chamfered sheath), better verification of end cap - sheath welding. These technologies were qualified by Nuclear Fuel Plant (N.F.P.) - Pitesti on automatic and Computer Numerical Control (C.N.C.) programming machines. A post-irradiation conclusive analysis which will take place later this year (2013) in Institute for Nuclear Research Pitesti (the action was initiated earlier this year by bringing a fuel bundle which has been reported defective by pool visual inspection) will provide additional information concerning potential damage causes of fuel bundles due to manufacturing processes. (authors)

  17. Thorium fuel performance assessment in HTRs

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-01

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

  18. Fuel performance update at Japanese BWRs

    International Nuclear Information System (INIS)

    Otsuka, Yasuyuki; Abe, Moriyasu

    2008-01-01

    Fuel Performance of Japanese BWRs has shown excellency with a very low fuel failure rate over the past couple of decades. In order to eliminate debris-fretting failures, which is considered as a dominant cause, efforts have been made to work out strict FME program and introduce lower tie plate with debris filters. Regarding a measure to detect failures without delay, an online off-gas monitor is installed, and there occurred no significant secondary failures leading to unnecessary plant shutdowns thanks to the monitor. Current standard fuel designs are 9x9 lattice configuration with the design maximum burnup set at 55 GWd/t and the bundle average discharge burnup at 45 GWd/t. Satisfactory performance of many fuels has gained so far to the burnup level. However recent data show an increase in hydrogen pickup of fuel claddings and spacers in the region of high burnup or high irradiation period, and this is one of the issues that should be addressed to maintain and improve fuel reliability for the reasons that hydrogen concentration may affect ductility of zirconium alloy. Furthermore, a recent study showed abnormally high hydrogen concentration in fuel cladding. Comprehensive root cause analyses were made, which include a poolside measurement for a number of high burnup fuels and development of new technique for tenacious crud removal, with the result that it is confirmed that a unique combination of cladding material, irradiation period and environment caused the phenomenon and it is not expected to occur again. New iron enhanced zirconium alloys, which have been investigated in for a long time, are expected to have greater resistance to hydrogen pickup, and introduction of the alloys is considered as the future materials. Regarding extensive inspections of fuel assemblies and channel boxes at Kashiwazaki-Kariwa NPS, no abnormality due to Niigata-ken Chuetsu-oki earthquake was observed for all plants. This paper deals with the current fuel performance at Japanese

  19. Probabilistic assessment of light water reactor fuel performance

    International Nuclear Information System (INIS)

    Misfeldt, I.

    1978-10-01

    A computer system for the statistical evaluation of LWR fuel performance has been developed. The computer code FRP, Fuel Reliability Predictor, calculates the distributions for parameters characterizing the fuel performance and failure probability. The statistical methods employed are either Monte Carlo simulations or low order Taylor approximation. Included in the computer system is a deterministic fuel performance code, which has been verified by comparison with data from irradiation experiments. The distributions for all material data utilized in the fuel simulations are estimations from the best available information in the literature. For the failure prediction, a stress corrosion failure criterion has been derived. The failure criterion is based on data from out-of-reactor stress corrosion experiments performed on unirradiated and irradiated zircaloy with iodine present. By means of an example the typical distributions of the variables characterizing the fuel performance and the accuracy of the methods themselves have been investigated. The application of the computer system is illustrated by a number of examples, these include the evaluation of irradiation experiments, design comparisons, and analyses of minor accidents. (author)

  20. Water reactor fuel element fabrication, with special emphasis on its effects on fuel performance

    International Nuclear Information System (INIS)

    1979-01-01

    Full text: The performance of nuclear fuel has improved over the years and is now a minor cause of outages and of power limitations in nuclear power plants. On the other hand, an increasing number of countries are in the process of developing or implementing their own capability for manufacturing fuel elements. In this context, the IAEA International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) advised that a symposium be organized devoted to the relationship between fuel fabrication and performance The Czechoslovak Atomic Energy Commission agreed to co-operate in the organization of this symposium and to host it in Prague. Those factors which influence fuel fabrication requirements are now well ascertained: as little reactor primary circuit contamination as possible, the tendency to increased burnups, reactor manoeuverability to match power grid demands, the desirability of an autonomous fabrication capability. It is the general experience of fuel element suppliers that fuel quality and performance has increased over the years, the importance of quality assurance and process monitoring has been decisive in this respect The ever increasing mass-production aspect of nuclear fuel leads to some processing steps being revised and alternatives being developed. The relation between fabrication processes and fuel performance characteristics, although generally well perceived, are still the subject of a large amount of experiment and assessment in most countries, both industrial and developing This evidence is most encouraging; it means indeed that nuclear power, which is already amongst the cheapest and safest sources of energy, will continue to be improved. The performance of Zircaloy fuel cladding - presently the material used in most water reactors - is under particular consideration. Better understanding of this quite recent alloy will pave the way for broader fuel utilization limits in the future. The panel discussion, which noted some

  1. The fuel performance code future

    International Nuclear Information System (INIS)

    Ronchi, C.; Van de Laar, J.

    1988-01-01

    The paper describes the LWR version of the fuel performance code FUTURE, which was recently developed to calculate the fuel response (swelling, cladding deformation, release) to reactor transient conditions, starting from a broad-based description of the processes of major concern. The main physical models assumed are presented together with the scheme of the computer program

  2. 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

  3. Performance tests for integral reactor nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Dong-Seong; Yim, Jeong-Sik; Lee, Chong-Tak; Kim, Han-Soo; Koo, Yang-Hyun; Lee, Byung-Ho; Cheon, Jin-Sik; Oh, Je-Yong

    2006-02-15

    An integral type reactor SMART plans to utilize metallic Zr-U fuel which is Zr-based alloy with 34{approx}38 wt% U. In order to verify the technologies for the design and manufacturing of the fuel and get a license, performance tests were carried out. Experimental Fuel Assembly (EFA) manufactured in KAERI is being successfully irradiated in the MIR reactor of RIAR from September 4 2004, and it has achieved burnup of 0.21 g/cc as of January 25 2006. Thermal properties of irradiated Zr-U fuel were measured. Up to the phase transformation temperature, thermal diffusivity increased linearly in proportion to temperature. However its dependence on the burnup was not significant. RIA tests with 4 unirradiated Zr-U fuel rods were performed in Kurchatov Institute to establish a safety criterion. In the case of the un-irradiated Zr-U fuel, the energy deposition during the control rod ejection accident should be less than 172 cal/g to prevent the failure accompanying fuel fragmentation and dispersal. Finally the irradiation tests of fuel rods have been performed at HANARO. The HITE-2 test was successfully completed up to a burnup of 0.31 g/cc. The HITE-3 test began in February 2004 and will be continued up to a target burnup of 0.6 g/cc.

  4. 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.

  5. 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)

  6. Effect of Fuel Additives on Spray Performance of Alternative Jet Fuels

    Science.gov (United States)

    Kannaiyan, Kumaran; Sadr, Reza

    2015-11-01

    Role of alternative fuels on reducing the combustion pollutants is gaining momentum in both land and air transport. Recent studies have shown that addition of nanoscale metal particles as fuel additives to liquid fuels have a positive effect not only on their combustion performance but also in reducing the pollutant formation. However, most of those studies are still in the early stages of investigation with the addition of nanoparticles at low weight percentages. Such an addition can affect the hydrodynamic and thermo-physical properties of the fuel. In this study, the near nozzle spray performance of gas-to-liquid jet fuel with and without the addition of alumina nanoparticles are investigated at macro- and microscopic levels using optical diagnostic techniques. At macroscopic level, the addition of nanoparticles is seen to enhance the sheet breakup process when compared to that of the base fuel. Furthermore, the microscopic spray characteristics such as droplet size and velocity are also found to be affected. Although the addition of nanoscale metal particles at low weight percentages does not affect the bulk fluid properties, the atomization process is found to be affected in the near nozzle region. Funded by Qatar National Research Fund.

  7. Performance optimization of a PEM hydrogen-oxygen fuel cell

    Energy Technology Data Exchange (ETDEWEB)

    Sadiq Al-Baghdadi, Maher A.R. [Fuel Cell Research Center, International Energy and Environment Foundation, Al-Najaf, P.O.Box 39 (Iraq)

    2013-07-01

    The objective was to develop a semi-empirical model that would simulate the performance of proton exchange membrane (PEM) fuel cells without extensive calculations. A fuel cell mathematical module has been designed and constructed to determine the performance of a PEM fuel cell. The influence of some operating parameters on the performance of PEM fuel cell has been investigated using pure hydrogen on the anode side and oxygen on the cathode side. The present model can be used to investigate the influence of process variables for design optimization of fuel cells, stacks, and complete fuel cell power system. The possible mechanisms of the parameter effects and their interrelationships are discussed. In order to assess the validity of the developed model a real PEM fuel cell system has been used to generate experimental data. The comparison shows good agreements between the modelling results and the experimental data. The model is shown a very useful for estimating the performance of PEM fuel cell stacks and optimization of fuel cell system integration and operation.

  8. Probabilistic assessments of fuel performance

    International Nuclear Information System (INIS)

    Kelppe, S.; Ranta-Puska, K.

    1998-01-01

    The probabilistic Monte Carlo Method, coupled with quasi-random sampling, is applied for the fuel performance analyses. By using known distributions of fabrication parameters and real power histories with their randomly selected combinations, and by making a large number of ENIGMA code calculations, one expects to find out the state of the whole reactor fuel. Good statistics requires thousands of runs. A sample case representing VVER-440 reactor fuel indicates relatively low fuel temperatures and mainly athermal fission gas release if any. The rod internal pressure remains typically below 2.5 MPa, which leaves a large margin to the system pressure of 12 MPa Gap conductance, an essential parameter in the accident evaluations, shows no decrease from its start-of-life value. (orig.)

  9. Advances in HTGR fuel performance models

    International Nuclear Information System (INIS)

    Stansfield, O.M.; Goodin, D.T.; Hanson, D.L.; Turner, R.F.

    1985-01-01

    Advances in HTGR fuel performance models have improved the agreement between observed and predicted performance and contributed to an enhanced position of the HTGR with regard to investment risk and passive safety. Heavy metal contamination is the source of about 55% of the circulating activity in the HTGR during normal operation, and the remainder comes primarily from particles which failed because of defective or missing buffer coatings. These failed particles make up about 5 x 10 -4 fraction of the total core inventory. In addition to prediction of fuel performance during normal operation, the models are used to determine fuel failure and fission product release during core heat-up accident conditions. The mechanistic nature of the models, which incorporate all important failure modes, permits the prediction of performance from the relatively modest accident temperatures of a passively safe HTGR to the much more severe accident conditions of the larger 2240-MW/t HTGR. (author)

  10. State of art in FE-based fuel performance codes

    International Nuclear Information System (INIS)

    Kim, Hyo Chan; Yang, Yong Sik; Kim, Dae Ho; Bang, Je Geon; Kim, Sun Ki; Koo, Yang Hyun

    2013-01-01

    Fuel performance codes approximate this complex behavior using an axisymmetric, axially-stacked, one-dimensional radial representation to save computation cost. However, the need for improved modeling of PCMI and, particularly, the importance of multidimensional capability for accurate fuel performance simulation has been identified as safety margin decreases. Finite element (FE) method that is reliable and proven solution in mechanical field has been introduced into fuel performance codes for multidimensional analysis. The present state of the art in numerical simulation of FE-based fuel performance predominantly involves 2-D axisymmetric model and 3-D volumetric model. The FRAPCON and FRAPTRAN own 1.5-D and 2-D FE model to simulate PCMI and cladding ballooning. In 2-D simulation, the FALCON code, developed by EPRI, is a 2-D (R-Z and R-θ) fully thermal-mechanically coupled steady-state and transient FE-based fuel behavior code. The French codes TOUTATIS and ALCYONE which are 3-D, and typically used to investigate localized behavior. In 2008, the Idaho National Laboratory (INL) has been developing multidimensional (2-D and 3-D) nuclear fuel performance code called BISON. In this paper, the current state of FE-based fuel performance code and their models are presented. Based on investigation into the codes, requirements and direction of development for new FE-based fuel performance code can be discussed. Based on comparison of models in FE-based fuel performance code, status of art in the codes can be discussed. A new FE-based fuel performance code should include typical pellet and cladding models which all codes own. In particular, specified pellet and cladding model such as gaseous swelling and high burnup structure (HBS) model should be developed to improve accuracy of code as well as consider AC condition. To reduce computation cost, the approximated gap and the optimized contact model should be also developed

  11. Validation of Fuel Performance Uncertainty for RIA Safety Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Park, Nam-Gyu; Yoo, Jong-Sung; Jung, Yil-Sup [KEPCO Nuclear Fuel Co., Daejeon (Korea, Republic of)

    2016-10-15

    To achieve this the computer code performance has to be validated based on the experimental results. And for the uncertainty quantification, important uncertainty parameters need to be selected, and combined uncertainty has to be evaluated with an acceptable statistical treatment. And important uncertainty parameters to the rod performance such as fuel enthalpy, fission gas release, cladding hoop strain etc. were chosen through the rigorous sensitivity studies. And their validity has been assessed by utilizing the experimental results, which were tested in CABRI and NSRR. Analysis results revealed that several tested rods were not bounded within combined fuel performance uncertainty. Assessment of fuel performance with an extended fuel power uncertainty on tested rods in NSRR and CABRI has been done. Analysis results showed that several tested rods were not bounded within calculated fuel performance uncertainty. This implies that the currently considered uncertainty range of the parameters is not enough to cover the fuel performance sufficiently.

  12. Does rim microstructure formation degrade the fuel rod performance?

    International Nuclear Information System (INIS)

    Baron, D.; Spino, J.

    2002-01-01

    High burnup extension of LWR fuel is progressing to reduce the total process flow and eventually the costs of the nuclear fuel cycle. A particular fuel restructuring at high burnups, commonly observed at the periphery of LWR fuel pellets (rim structure), but also in FBR fuels to some extent and in the Plutonium rich clusters of the MOX Fuels, was considered a priori as a limitation for burnup extension. Since more than ten years this rim effect have been deeply investigated. Its causes and consequences are however not yet totally elucidated. The three steps actually identified of this phenomenon are first a progressive disappearing of the intra-granular Xenon, the outset of numerous 0.5 to 1 m pores and finally a grain subdivision around the pores. Penalty of the porosity increase on the thermal conductivity is obvious. One expect the fission gases to remain trapped in the rim porosity up to a 75 MWd/kgUO 2 local burnup. Above this threshold, 15 to 20 % of the fission gases seem to be quickly released. Microindentation tests conducted at ITU have shown the rim structure to resist fracture extension under punching. It is still open whether this implies certain ductility and viscosity of the material, or if it corresponds to stress relaxation by microcracking. Whatever the case be, it is suggested that the rim material would be able to decrease the interaction stresses and to equalise the cladding strains during a power ramp. Moreover, in the RIA tests, it was concluded so far that the grain de-cohesion caused by gas expansion at the grain boundaries was responsible for the cladding strain and failure. However, not the rim zone was affected by grain de-cohesion but the region adjacent to it. Therefore, in front of the question whether the rim structure degrades the fuel rod behaviour, we continue to argue on its benefit for fuel burnup extension. (author)

  13. Fuel performance annual report for 1983. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Bailey, W.J.; Dunenfeld, M.S.

    1985-03-01

    This annual report, the sixth in a series, provides a brief description of fuel performance during 1983 in commercial nuclear power plants. Brief summaries of fuel design changes, fuel surveillance programs, fuel operating experience, fuel problems, high-burnup fuel experience, and items of general significance are provided. References to additional, more detailed information and related NRC evaluations are included.

  14. Developments of fuel performance analysis codes in KEPCO NF

    International Nuclear Information System (INIS)

    Han, H. T.; Choi, J. M.; Jung, C. D.; Yoo, J. S.

    2012-01-01

    The KEPCO NF has developed fuel performance analysis and design code named as ROPER, and utility codes of XGCOL and XDNB in order to perform fuel rod design evaluation for Korean nuclear power plants. The ROPER code intends to cover full range of fuel performance evaluation. The XGCOL code is for the clad flattening evaluation and the XDNB code is for the extensive DNB propagation evaluation. In addition to these, the KEPCO NF is now in the developing stage for 3-dimensional fuel performance analysis code, named as OPER3D, using 3-dimensional FEM for the nest generation within the joint project CANDU ENERGY in order to analyze PCMI behavior and fuel performance under load following operation. Of these, the ROPER code is now in the stage of licensing activities by Korean regulatory body and the other two are almost in the final developing stage. After finishing the developing, licensing activities are to be performed. These activities are intending to acquire competitiveness, originality, vendor-free ownership of fuel performance codes in the KEPCO NF

  15. Study on severe fuel damage and in-vessel melt progression

    International Nuclear Information System (INIS)

    Kim, Hee Dong; Kim, Sang Baik; Lee, Gyu Jung

    1992-06-01

    In-vessel core melt progression describes the progression of the state of a reactor core from core uncovery up to reactor vessel melt through in uncovered accidents or through temperature stabilization in accidents recovered by core reflooding. Melt progression can be thought as two parts; early melt progression and late melt progression. Early phase of core melt progression includes the progression of core material melting and relocation, which mostly consist of metallic materials. On the other hand, the late phase of core melt progression involves ceramic material melt and relocation to the lower plenum and heat-up the reactor vessel lower head. A large number of information are available for the early melt progression through experiments such as SFD, DF, FLHT test and utilized in the severe accident analysis codes. However, understanding of the late phase melt progression phenomenology is based primary on TMI-2 core examinations and not much experimental information is available. Especilally, the great uncertainties exist in vessel failure mode, melt composition, mass, and temperature. Further research is planned to perform to reduce the uncertainties in understanding of core melt down accidents as parts of long term melt progression research program. A study on the core melt progression at KAERI has been being performed through the Severe Accident Research Program with USNRC. KAERI staff had participated in the PBF SFD experiments at INEL and analyses of experiments were performed using SCDAP code. Experiments of core melt program have not been carried out at KAERI yet. It is planned that further research on core melt down accidents will be performed, which is related to design of future generations of nuclear reactors as parts of long-term project for improvement of nuclear reactor safety. (Author)

  16. 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

  17. Design of JMTR high-performance fuel element

    International Nuclear Information System (INIS)

    Sakurai, Fumio; Shimakawa, Satoshi; Komori, Yoshihiro; Tsuchihashi, Keiichiro; Kaminaga, Fumito

    1999-01-01

    For test and research reactors, the core conversion to low-enriched uranium fuel is required from the viewpoint of non-proliferation of nuclear weapon material. Improvements of core performance are also required in order to respond to recent advanced utilization needs. In order to meet both requirements, a high-performance fuel element of high uranium density with Cd wires as burnable absorbers was adopted for JMTR core conversion to low-enriched uranium fuel. From the result of examination of an adaptability of a few group constants generated by a conventional transport-theory calculation with an isotropic scattering approximation to a few group diffusion-theory core calculation for design of the JMTR high-performance fuel element, it was clear that the depletion of Cd wires was not able to be predicted accurately using group constants generated by the conventional method. Therefore, a new generation method of a few group constants in consideration of an incident neutron spectrum at Cd wire was developed. As the result, the most suitable high-performance fuel element for JMTR was designed successfully, and that allowed extension of operation duration without refueling to almost twice as long and offer of irradiation field with constant neutron flux. (author)

  18. PWR and WWER fuel performance. A comparison of major characteristics

    International Nuclear Information System (INIS)

    Weidinger, H.

    2006-01-01

    PWR and WWER fuel technologies have the same basic performance targets: most effective use of the energy stored in the fuel and highest possible reliability. Both fuel technologies use basically the same strategies to reach these targets: 1) Optimized reload strategies; 2) Maximal use of structural material with low neutron cross sections; 3) Decrease the fuel failure frequency towards a 'zero failure' performance by understanding and eliminating the root causes of those defects. The key driving force of the technology of both, PWR and WWER fuel is high burn-up. Presently a range of 45 - 50 MWD/kgU have been reached commercially for PWR and WWER fuel. The main technical limitations to reach high burn-up are typically different for PWR and WWER fuel: for PWR fuel it is the corrosion and hydrogen uptake of the Zr-based materials; for WWER fuel it is the mechanical and dimensional stability of the FA (and the whole core). Corrosion and hydrogen uptake of Zr-materials is a 'non-problem' for WWER fuel. Other performance criteria that are important for high burn-up are the creep and growth behaviour of the Zr materials and the fission gas release in the fuel rod. There exists a good and broad data base to model and design both fuel types. FA and fuel rod vibration appears to be a generic problem for both fuel types but with more evidence for PWR fuel performance reliability. Grid-to-rod fretting is still a major issue in the fuel failure statistics of PWR fuel. Fuel rod cladding defects by debris fretting is no longer a key problem for PWR fuel, while it still appears to be a significant root cause for WWER fuel failures. 'Zero defect' fuel performance is achievable with a high probability, as statistics for US PWR and WWER-1000 fuel has shown

  19. Review of CAGR fuel performance

    Energy Technology Data Exchange (ETDEWEB)

    Boocock, P.M. (South of Scotland Electricity Board, Glasgow (UK)); Eldred, V.W. (UKAEA Windscale Nuclear Power Development Labs.); Hines, G.F. (UKAEA Atomic Energy Establishment, Winfrith); Seeley, T.A. (British Nuclear Fuels Ltd., Springfields); Shaw, R.A. (UKAEA Springfields Nuclear Power Development Labs.); Waddington, J.S. (Central Electricity Generating Board, Berkeley (UK). Berkeley Nuclear Labs.)

    1983-04-01

    The performance of CAGR fuel in the Electricity Board's reactors is reviewed in the light of operating experience and the post-irradiation examination assessments which have been carried out. The results show that the extrapolation of the design from the Windscale AGR prototype has been very successful. Future trends in fuel design in order to achieve the higher burn-up targets now being specified are discussed.

  20. State of art in FE-based fuel performance codes

    International Nuclear Information System (INIS)

    Kim, Hyo Chan; Yang, Yong Sik; Kim, Dae Ho; Bang, Je Geon; Kim, Sun Ki; Koo, Yang Hyun

    2013-01-01

    Finite element (FE) method that is reliable and proven solution in mechanical field has been introduced into fuel performance codes for multidimensional analysis. The present state of the art in numerical simulation of FE-based fuel performance predominantly involves 2-D axisymmetric model and 3-D volumetric model. The FRAPCON and FRAPTRAN own 1.5-D and 2-D FE model to simulate PCMI and cladding ballooning. In 2-D simulation, the FALCON code, developed by EPRI, is a 2-D (R-Z and R-θ) fully thermal-mechanically coupled steady-state and transient FE-based fuel behavior code. The French codes TOUTATIS and ALCYONE which are 3-D, and typically used to investigate localized behavior. In 2008, the Idaho National Laboratory (INL) has been developing multidimensional (2-D and 3-D) nuclear fuel performance code called BISON. In this paper, the current state of FE-based fuel performance code and their models are presented. Based on investigation into the codes, requirements and direction of development for new FE-based fuel performance code can be discussed. Based on comparison of models in FE-based fuel performance code, status of art in the codes can be discussed. A new FE-based fuel performance code should include typical pellet and cladding models which all codes own. In particular, specified pellet and cladding model such as gaseous swelling and high burnup structure (HBS) model should be developed to improve accuracy of code as well as consider AC condition. To reduce computation cost, the approximated gap and the optimized contact model should be also developed. Nuclear fuel operates in an extreme environment that induces complex multiphysics phenomena, occurring over distances ranging from inter-atomic spacing to meters, and times scales ranging from microseconds to years. This multiphysics behavior is often tightly coupled, a well known example being the thermomechanical behavior. Adding to this complexity, important aspects of fuel behavior are inherently

  1. Fuel performance experience at TVO nuclear power plant

    International Nuclear Information System (INIS)

    Patrakka, E.T.

    1985-01-01

    TVO nuclear power plant consists of two BWR units of ASEA-ATOM design. The fuel performance experience extending through six cycles at TVO I and four cycles at TVO II is reported. The experience obtained so far is mainly based on ASEA-ATOM 8 x 8 fuel and has been satisfactory. Until autumn 1984 one leaking fuel assembly had been identified at TVO I and none at TVO II. Most of the problems encountered have been related to leaf spring screws and channel screws. The experience indicates that satisfactory fuel performance can be achieved when utilizing strict operational rules and proper control of fuel design and manufacture. (author)

  2. Fuels Performance | Transportation Research | NREL

    Science.gov (United States)

    Biodiesel Performance in Modern Engines NREL is working cooperatively with the National Biodiesel Board on research that assesses several aspects of biodiesel compatibility with engines, vehicles, and fuel-handling infrastructure. Several cooperative studies were recently conducted that examined how biodiesel performs with new

  3. Performance optimization of a PEM hydrogen-oxygen fuel cell

    OpenAIRE

    Maher A.R. Sadiq Al-Baghdadi

    2013-01-01

    The objective was to develop a semi-empirical model that would simulate the performance of proton exchange membrane (PEM) fuel cells without extensive calculations. A fuel cell mathematical module has been designed and constructed to determine the performance of a PEM fuel cell. The influence of some operating parameters on the performance of PEM fuel cell has been investigated using pure hydrogen on the anode side and oxygen on the cathode side. The present model can be used to investigate t...

  4. 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.

  5. Degradation of solid oxide fuel cell metallic interconnects in fuels containing sulfur

    Energy Technology Data Exchange (ETDEWEB)

    Ziomek-Moroz, M.; Hawk, Jeffrey A.

    2005-01-01

    Hydrogen is the main fuel for all types of fuel cells except direct methanol fuel cells. Hydrogen can be generated from all manner of fossil fuels, including coal, natural gas, diesel, gasoline, other hydrocarbons, and oxygenates (e.g., methanol, ethanol, butanol, etc.). Impurities in the fuel can cause significant performance problems and sulfur, in particular, can decrease the cell performance of fuel cells, including solid oxide fuel cells (SOFC). In the SOFC, the high (800-1000°C) operating temperature yields advantages (e.g., internal fuel reforming) and disadvantages (e.g., material selection and degradation problems). Significant progress in reducing the operating temperature of the SOFC from ~1000 ºC to ~750 ºC may allow less expensive metallic materials to be used for interconnects and as balance of plant (BOP) materials. This paper provides insight on the material performance of nickel, ferritic steels, and nickel-based alloys in fuels containing sulfur, primarily in the form of H2S, and seeks to quantify the extent of possible degradation due to sulfur in the gas stream.

  6. Probabilistic evaluation of fuel element performance by the combined use of a fast running simplistic and a detailed deterministic fuel performance code

    International Nuclear Information System (INIS)

    Misfeldt, I.

    1980-01-01

    A comprehensive evaluation of fuel element performance requires a probabilistic fuel code supported by a well bench-marked deterministic code. This paper presents an analysis of a SGHWR ramp experiment, where the probabilistic fuel code FRP is utilized in combination with the deterministic fuel models FFRS and SLEUTH/SEER. The statistical methods employed in FRP are Monte Carlo simulation or a low-order Taylor approximation. The fast-running simplistic fuel code FFRS is used for the deterministic simulations, whereas simulations with SLEUTH/SEER are used to verify the predictions of FFRS. The ramp test was performed with a SGHWR fuel element, where 9 of the 36 fuel pins failed. There seemed to be good agreement between the deterministic simulations and the experiment, but the statistical evaluation shows that the uncertainty on the important performance parameters is too large for this ''nice'' result. The analysis does therefore indicate a discrepancy between the experiment and the deterministic code predictions. Possible explanations for this disagreement are discussed. (author)

  7. 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

  8. Technical specifications and performance of CANDU fuel

    International Nuclear Information System (INIS)

    Sejnoha, R.

    1997-01-01

    The relations between Technical Specifications and fuel performance are discussed in terms of design limits and margins. The excellent performance record of CANDU reactor fuel demonstrates that the fuel design defined in the Technical Specifications (and with it other components of the procurement cycle, such as fuel manufacturing), satisfy the requirements. New requirements, changing conditions of fuel application and accumulating experience make periodic updates of the Technical Specifications necessary. Under the CANDU Owners Group (COG) Working Party 9, a Work Package has been conducted to support the review of the Specifications and the documentation of the rationales for their requirements. So far, the review has been completed for 4 Specifications: 1 for Zircaloy tubing, and 3 for uranium dioxide powder. It is planned to complete the review of all 11 currently used specifications by 1999. The paper summarizes the results achieved to mid 1997. (author)

  9. 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)

  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. Progress of the DUPIC fuel compatibility analysis (II) - thermal-hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Choi, Hang Bok

    2005-03-01

    Thermal-hydraulic compatibility of the DUPIC fuel bundle with a 713 MWe Canada deuterium uranium (CANDU-6) reactor was studied by using both the single channel and sub-channel analysis methods. The single channel analysis provides the fuel channel flow rate, pressure drop, critical channel power, and the channel exit quality, which are assessed against the thermal-hydraulic design requirements of the CANDU-6 reactor. The single channel analysis by the NUCIRC code showed that the thermal-hydraulic performance of the DUPIC fuel is not different from that of the standard CANDU fuel. Regarding the local flow characteristics, the sub-channel analysis also showed that the uncertainty of the critical channel power calculation for the DUPIC fuel channel is very small. As a result, both the single and sub-channel analyses showed that the key thermal-hydraulic parameters of the DUPIC fuel channel do not deteriorate compared to the standard CANDU fuel channel.

  12. Home Environmental Factors Influencing Performance and Progress ...

    African Journals Online (AJOL)

    Our findings support other studies which found that parents' educational level and income level have a bearing on school progress and performance. Contrary to most research findings mother tongue instruction did not eme1rge as an important explanatory factor on school progress and performance, however; home ...

  13. 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.

  14. Performance of Combustion Engineering fuel in operating PWRs

    International Nuclear Information System (INIS)

    Andrews, M.G.; Freeburn, H.R.; Wohlsen, W.D.

    1979-01-01

    Performance data on fuel assembly components from seven (7) operating reactors are presented, and discussed in detail where potential problems have occurred and been resolved. Fuel rod performance has continually improved over the last four (4) years with the gradual changeover to the current C-E fuel design. The reliability level is estimated at better than 99.99%, based on activity levels obtained through January 1979 at each plant. Control rod guide tubes have shown various degrees of wear caused by vibration of the control rods in their fully-withdrawn position. The retrofit of wear sleeves within the top portion of the affected guide tubes during routine refueling has permitted the use of these fuel assemblies with no significant loss in performance or safety margins

  15. Irradiation performance of uranium-molybdenum alloy dispersion fuels

    International Nuclear Information System (INIS)

    Almeida, Cirila Tacconi de

    2005-01-01

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

  16. Fuel Combustion and Engine Performance | Transportation Research | NREL

    Science.gov (United States)

    Fuel Combustion and Engine Performance Fuel Combustion and Engine Performance Photo of a gasoline emissions in advanced engine technologies. Photo by Dennis Schroeder, NREL NREL's combustion research and combustion and engine research activities include: Developing experimental and simulation research platforms

  17. 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)

  18. Fuel performance of rod-type research reactor fuel using a centrifugally atomized U-Mo powder

    International Nuclear Information System (INIS)

    Ryu, Ho Jin; Park, Jong Man; Lee, Yoon Sang; Kim, Chang Kyu

    2009-01-01

    A low enriched uranium nuclear fuel for research reactors has been developed in order to replace a highly enriched uranium fuel according to the non-proliferation policy under the reduced enrichment for research and test reactors (RERTR) program. In KAERI, a rod-type U 3 Si dispersion fuel has been developed for a localization of the HANARO fuel and a U 3 Si/Al dispersion fuel of 3.15 gU/cc has been used at HANARO as a driver fuel since 2005. Although uranium silicide dispersion fuels such as U 3 Si 2 /Al and U 3 Si/Al are being used widely, high uranium density dispersion fuels (8-9 g/cm 3 ) are required for some high performance research reactors. U-Mo alloys have been considered as one of the most promising uranium alloys for a dispersion fuel due to their good irradiation performance. An international qualification program on U-Mo fuel to replace a uranium silicide dispersion fuel with a U-Mo dispersion fuel has been carried out

  19. Effects of variations in fuel pellet composition and size on mixed-oxide fuel pin performance

    International Nuclear Information System (INIS)

    Makenas, B.J.; Jensen, B.W.; Baker, R.B.

    1980-10-01

    Experiments have been conducted which assess the effects on fuel pin performance of specific minor variations from nominal in both fuel pellet size and pellet composition. Such pellets are generally referred to in the literature as rogue pellets. The effect of these rogue pellets on fuel pin and reactor performance is shown to be minimal

  20. 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.)

  1. A comparison of thermal algorithms of fuel rod performance code systems

    International Nuclear Information System (INIS)

    Park, C. J.; Park, J. H.; Kang, K. H.; Ryu, H. J.; Moon, J. S.; Jeong, I. H.; Lee, C. Y.; Song, K. C.

    2003-11-01

    The goal of the fuel rod performance is to identify the robustness of a fuel rod with cladding material. Computer simulation of the fuel rod performance becomes one of important parts to designed and evaluate new nuclear fuels and claddings. To construct a computing code system for the fuel rod performance, several algorithms of the existing fuel rod performance code systems are compared and are summarized as a preliminary work. Among several code systems, FRAPCON, and FEMAXI for LWR, ELESTRES for CANDU reactor, and LIFE for fast reactor are reviewed. Thermal algorithms of the above codes are investigated including methodologies and subroutines. This work will be utilized to construct a computing code system for dry process fuel rod performance

  2. A comparison of thermal algorithms of fuel rod performance code systems

    Energy Technology Data Exchange (ETDEWEB)

    Park, C. J.; Park, J. H.; Kang, K. H.; Ryu, H. J.; Moon, J. S.; Jeong, I. H.; Lee, C. Y.; Song, K. C

    2003-11-01

    The goal of the fuel rod performance is to identify the robustness of a fuel rod with cladding material. Computer simulation of the fuel rod performance becomes one of important parts to designed and evaluate new nuclear fuels and claddings. To construct a computing code system for the fuel rod performance, several algorithms of the existing fuel rod performance code systems are compared and are summarized as a preliminary work. Among several code systems, FRAPCON, and FEMAXI for LWR, ELESTRES for CANDU reactor, and LIFE for fast reactor are reviewed. Thermal algorithms of the above codes are investigated including methodologies and subroutines. This work will be utilized to construct a computing code system for dry process fuel rod performance.

  3. Anode-Engineered Protonic Ceramic Fuel Cell with Excellent Performance and Fuel Compatibility

    NARCIS (Netherlands)

    Hua, B.; Yan, N.; Li, M.; Sun, Y.-F.; Zhang, Y.-Q.; Li, J.; Etsell, T.; Sarkar, P.; Luo, J.L.

    2016-01-01

    Directly utilizing hydrocarbon fuels, particularly methane, is advantageous yet challenging in high-performance protonic ceramic fuel cells. In this work, this technological hurdle is well addressed by selective deposition of secondary electrocatalysts within the porous Ni-cermet anode. This novel

  4. Fuel technology and performance of non-water cooled reactors. Proceedings of an advisory group meeting held in Vienna, 5-8 December 1994

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-12-01

    The IAEA Division of Nuclear Fuel Cycle and Waste Management has been closely involved for many years in the collection, analysis and exchange of information relating to the global development of advanced reactor fuel technology and performance. Meetings of experts in this field have been held in 1984 and 1989 and more recently in December 1994 as part of the IAEA`s programme. This publication reviews progress in advanced reactor fuel technology and performance over the past five years, principally related to non-water cooled reactors, namely high temperature gas reactors (HTGRs) and fast reactors (FRs), as well as developments pertaining to thorium fuels and the fuel fabrication technologies. It includes papers from the participants and provides recommendations in key areas where further global co-operation in this field might be usefully initiated or strengthened. The previous two Advisory Group Meetings on Advanced Fuel Technology and Performance, on which separate reports have been published (IAEA-TECDOC-352 (1985) and IAEA-TECDOC-577 (1990)), focused on all types of commercial nuclear reactors. Refs, figs and tabs.

  5. Manufacture of fuel and fuel channels and their performance in Indian PHWRs'

    International Nuclear Information System (INIS)

    Kalidas, R.

    2005-01-01

    Nuclear Fuel Complex (NFC) at Hyderabad is conglomeration of chemical, metallurgical and mechanical plants, processing uranium and zirconium in two separate streams and culminating in the fuel assembly plant. Apart from manufacturing fuel for Pressurised Heavy Water Reactors (PHWRs) and Boiling Water Reactors (BWRs), NFC is also engaged in the manufacture of reactor core structurals for these reactors. NFC has carried our several technological developments over the years and implemented them for the manufacture of fuel, calandria tubes and pressure tubes for PHWRs. Keeping in pace with the Nuclear Power Programme envisaged by the Department of Atomic Energy, NFC had augmented its production capacities in all these areas. The paper highlights several actions initiated in the areas of fuel design, fuel manufacturing, manufacturing of zirconium alloy core structurals, fuel clad tubes and components and their performance in Indian PHWRs. (author)

  6. SGHWR fuel performance, safety and reliability

    International Nuclear Information System (INIS)

    Pickman, D.O.; Inglis, G.H.

    1977-01-01

    The design principles involved in fuel pins and elements need to take account of the sometimes conflicting requirements of performance, safety and reliability. The principal factors involved in this optimisation are discussed and it is shown from fuel irradiation experience in the Winfrith S.G.H.W.R. that the necessary bias toward safety has not resulted in a reliability level lower than that shown by other successful water reactor designs. Reliability has important economic implications and has to be paid for. By a detailed evaluation of S.G.H.W.R. fuel defects it is shown that very few defects can be shown to be related to design, rating or burn-up. This demonstrates that economic aspects have not over-ridden necessary criteria that must be met to achieve the desirable reliability level. It is possible that large-scale experience with S.G.H.W.R. fuel may eventually demonstrate that the balance is too much in favour of reliability and consideration may be given to whether design changes favouring economy could be achieved without compromising safety. The safety criteria applied to S.G.H.W.R. fuel are designed to avoid any possibility of a temperature runaway in any credible accident situation. The philosophy and supporting experimental work programme are outlined and the fuel design features which particularly contribute to maximising safety margins are outlined. Reference is made to new 60 pin fuel element to be used in the commercial S.G.H.W.R.'s and how it compares in design and performance aspects with the 36 pin element that has been used to date in the Winfrith S.G.H.W.R

  7. Effect of different fuel options on performance of high-temperature PEMFC (proton exchange membrane fuel cell) systems

    International Nuclear Information System (INIS)

    Authayanun, Suthida; Saebea, Dang; Patcharavorachot, Yaneeporn; Arpornwichanop, Amornchai

    2014-01-01

    High-temperature proton exchange membrane fuel cells (HT-PEMFCs) have received substantial attention due to their high CO (carbon monoxide) tolerance and simplified water management. The hydrogen and CO fractions affect the HT-PEMFC performance and different fuel sources for hydrogen production result in different product gas compositions. Therefore, the aim of this study is to investigate the theoretical performance of HT-PEMFCs fueled by the reformate gas derived from various fuel options (i.e., methane, methanol, ethanol, and glycerol). Effects of fuel types and CO poisoning on the HT-PEMFC performance are analyzed. Furthermore, the necessity of a water-gas shift (WGS) reactor as a CO removal unit for pretreating the reformate gas is investigated for each fuel type. The methane steam reforming shows the highest possibility of CO formation, whereas the methanol steam reforming produces the lowest quantity of CO in the reformate gas. The methane fuel processing gives the maximum fraction of hydrogen (≈0.79) when the WGS reactor is included. The most suitable fuel is the one with the lowest CO poisoning effect and the maximum fuel cell performance. It is found that the HT-PEMFC system fueled by methanol without the WGS reactor and methane with WGS reactor shows the highest system efficiency (≈50%). - Highlights: • Performance of HT-PEMFC run on different fuel options is theoretically investigated. • Glycerol, methanol, ethanol and methane are hydrogen sources for the HT-PEMFC system. • Effect of CO poisoning on the HT-PEMFC performance is taken into account. • The suitable fuel for HT-PEMFC system is identified regarding the system efficiency

  8. Influence of oxygen-metal ratio on mixed-oxide fuel performance

    International Nuclear Information System (INIS)

    Lawrence, L.A.; Leggett, R.D.

    1979-04-01

    The fuel oxygen-to-metal ratio (O/M) is recognized as an important consideration for performance of uranium--plutonium oxide fuels. An overview of the effects of differing O/M's on the irradiation performance of reference design mixed-oxide fuel in the areas of chemical and mechanical behavior, thermal performance, and fission gas behavior is presented. The pellet fuel has a nominal composition of 75 wt% UO 2 + 25 wt% PuO 2 at a pellet density of approx. 90% TD. for nominal conditions this results in a smeared density of approx. 85%. The cladding in all cases is 20% CW type 316 stainless steel with an outer diameter of 5.84 to 6.35 mm. O/M has been found to significantly influence fuel pin chemistry, mainly FCCI and fission product and fuel migration. It has little effect on thermal performance and overall mechanical behavior or fission gas release. The effects of O/M (ranging from 1.938 to 1.984) in the areas of fuel pin chemistry, to date, have not resulted in any reduction in fuel pin performance capability to goal burnups of approx. 8 atom% or more

  9. Performance of metallic fuels in liquid-metal fast reactors

    International Nuclear Information System (INIS)

    Seidel, B.R.; Walters, L.C.; Kittel, J.H.

    1984-01-01

    Interest in metallic fuels for liquid-metal fast reactors has come full circle. Metallic fuels are once again a viable alternative for fast reactors because reactor outlet temperature of interest to industry are well within the range where metallic fuels have demonstrated high burnup and reliable performance. In addition, metallic fuel is very tolerant of off-normal events of its high thermal conductivity and fuel behavior. Futhermore, metallic fuels lend themselves to compact and simplified reprocessing and refabrication technologies, a key feature in a new concept for deployment of fast reactors called the Integral Fast Reactor (IFR). The IFR concept is a metallic-fueled pool reactor(s) coupled to an integral-remote reprocessing and fabrication facility. The purpose of this paper is to review recent metallic fuel performance, much of which was tested and proven during the twenty years of EBR-II operation

  10. Effect of pilot fuel quantity on the performance of a dual fuel engine

    Energy Technology Data Exchange (ETDEWEB)

    Abd Alla, G.H.; Soliman, H.A.; Badr, O.A.; Abd Rabbo, M.F. [Zagazig University, Cairo (Egypt). Shoubra Faculty of Engineering

    2000-04-01

    It is well known that the operation of dual fuel engines at lower loads suffers from lower thermal efficiency and higher unburned percentages of fuel. To rectify this problem, tests have been conducted on a special single cylinder compression ignition research engine (Ricardo E6) to investigate the effect of pilot fuel quantity on the performance of an indirect injection diesel engine fuelled with gaseous fuel. Diesel fuel was used as the pilot fuel and methane or propane was used as the main fuel which was inducted into the intake manifold to mix with the intake air. Through experimental investigations, it is shown that, the low efficiency and excess emissions at light loads can be improved significantly by increasing the amount of pilot fuel, while increasing the amount of pilot fuel at high loads led to early knocking. (author)

  11. Thermodynamic properties of the DUPIC fuel and its performance

    Energy Technology Data Exchange (ETDEWEB)

    Park, Kwang Heon; Kim, Hee Moon [Kyung Hee Univ., Seoul (Korea, Republic of)

    1997-07-01

    This study describes thermodynamic properties of DUPIC fuel and performance. In initial state, DUPIC fuel which contains fissile materials is different from general nuclear fuel. So this study analyzed oxygen potential, thermal conductivity and specific heat of the DUPIC fuel.

  12. What Happens Inside a Fuel Cell? Developing an Experimental Functional Map of Fuel Cell Performance

    KAUST Repository

    Brett, Daniel J. L.; Kucernak, Anthony R.; Aguiar, Patricia; Atkins, Stephen C.; Brandon, Nigel P.; Clague, Ralph; Cohen, Lesley F.; Hinds, Gareth; Kalyvas, Christos; Offer, Gregory J.; Ladewig, Bradley; Maher, Robert; Marquis, Andrew; Shearing, Paul; Vasileiadis, Nikos; Vesovic, Velisa

    2010-01-01

    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

  13. WWER reactor fuel performance, modelling and experimental support. Proceedings

    International Nuclear Information System (INIS)

    Stefanova, S.; Chantoin, P.; Kolev, I.

    1994-01-01

    This publication is a compilation of 36 papers presented at the International Seminar on WWER Reactor Fuel Performance, Modelling and Experimental Support, organised by the Institute for Nuclear Research and Nuclear Energy (BG), in cooperation with the International Atomic Energy Agency. The Seminar was attended by 76 participants from 16 countries, including representatives of all major Russian plants and institutions responsible for WWER reactor fuel manufacturing, design and research. The reports are grouped in four chapters: 1) WWER Fuel Performance and Economics: Status and Improvement Prospects: 2) WWER Fuel Behaviour Modelling and Experimental Support; 3) Licensing of WWER Fuel and Fuel Analysis Codes; 4) Spent Fuel of WWER Plants. The reports from the corresponding four panel discussion sessions are also included. All individual papers are recorded in INIS as separate items

  14. WWER reactor fuel performance, modelling and experimental support. Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    Stefanova, S; Chantoin, P; Kolev, I [eds.

    1994-12-31

    This publication is a compilation of 36 papers presented at the International Seminar on WWER Reactor Fuel Performance, Modelling and Experimental Support, organised by the Institute for Nuclear Research and Nuclear Energy (BG), in cooperation with the International Atomic Energy Agency. The Seminar was attended by 76 participants from 16 countries, including representatives of all major Russian plants and institutions responsible for WWER reactor fuel manufacturing, design and research. The reports are grouped in four chapters: (1) WWER Fuel Performance and Economics: Status and Improvement Prospects: (2) WWER Fuel Behaviour Modelling and Experimental Support; (3) Licensing of WWER Fuel and Fuel Analysis Codes; (4) Spent Fuel of WWER Plants. The reports from the corresponding four panel discussion sessions are also included. All individual papers are recorded in INIS as separate items.

  15. Make use of EDF orientations in PWR fuel management

    International Nuclear Information System (INIS)

    Gloaguen, A.

    1989-01-01

    The EDF experience acquired permits to allow the PWR fuel performances and to make use of better management. In this domain low progress can be given considerable financial profits. The industrial and commercial structures, the time constant of the fuel cycle, has for consequence that the electric utilities can take advantage only progressively of the expected profits [fr

  16. Advanced fuel technology and performance: Current status and trends

    International Nuclear Information System (INIS)

    1990-11-01

    During the last years the Nuclear Fuel Cycle and Waste Management Division of the IAEA has been giving great attention to the collection, analysis and exchange of information in the field of reactor fuel technology. Most of these activities are being conducted in the framework of the International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT). The purpose of this Advisory Group Meeting on Advanced Fuel Technology and Performance was to update and to continue the previous work, and to review the experience of advanced fuel technology, its performance with regard to all types of reactors and to outline the future trends on the basis of national experience and discussions during the meeting. As a result of the meeting a Summary Report was prepared which reflected the status of the advanced nuclear fuel technology up to 1990. The 10 papers presented by participants of this meeting are also published here. A separate abstract was prepared for each of these papers. Refs, figs and tabs

  17. High-quality thorium TRISO fuel performance in HTGRs

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-11-01

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

  18. High-quality thorium TRISO fuel performance in HTGRs

    International Nuclear Information System (INIS)

    Verfondern, Karl; Allelein, Hans-Josef; Nabielek, Heinz; Kania, Michael J.

    2013-01-01

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

  19. 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)

  20. CANDU type fuel activities in Argentina

    International Nuclear Information System (INIS)

    Lavarez, L.; Casario, J.A.; Moreno, C.

    2003-01-01

    Domestic fuel performance in Embalse NPP during the last two years has been excellent without a significant occurrence of fuel failures. The defect rate level was reasonably low with a lowest value of 0.02 % in 2002. The implementation of fuel design optimizations to increase uranium content was fully completed by the end of year 2000. The in-reactor performance was not affected and shows the high degree of maturity reached for both the design and the manufacturing procedures and capabilities. A feasibility study for the utilization of SEU in Embalse NPP mainly conducted by NA-SA and AECL is almost completed. Some fuel related activities are still in progress. As part of them fuel behavior simulations using simplified power histories were performed to assess the influence of SEU fuel burnup extension. (author)

  1. SP-100 Fuel Pin Performance: Results from Irradiation Testing

    Science.gov (United States)

    Makenas, Bruce J.; Paxton, Dean M.; Vaidyanathan, Swaminathan; Marietta, Martin; Hoth, Carl W.

    1994-07-01

    A total of 86 experimental fuel pins with various fuel, liner, and cladding candidate materials have been irradiated in the Experimental Breeder Reactor-II (EBR-II) and the Fast Flux Test Facility (FFTF) reactor as part of the SP-100 fuel pin irradiation testing program. Postirradiation examination results from these fuel pins are key in establishing performance correlations and demonstrating the lifetime and safety of the reactor fuel system. This paper provides a brief description of the in-reactor fuel pin tests and presents the most recent irradiation data on the performance of wrought rhenium (Re) liner material and high density UN fuel at goal burnup of 6 atom percent (at. %). It also provides an overview of the significant variety of other fuel/liner/cladding combinations which were irradiated as part of this program and which may be of interest to more advanced efforts.

  2. Performance of PARR-1 with LEU Fuel

    International Nuclear Information System (INIS)

    Pervez, S.; Latif, M.; Bokhari, I.H.; Bakhtyar, S.

    2005-01-01

    Pakistan Research Reactor (PARR-1) went critical in 1965 with HEU fuel. The reactor core was converted to LEU fuel with power upgradation from 5 MW to 10 MW in 1992. The reactor has been operated with LEU fuel for about 10,000 hours and has produced about 66,000 MWh energy up to now. Average burn up of the irradiated fuel is about 42 %. The fuel performance during the last 12 years has been excellent. Post irradiation visual inspection of the fuel has revealed no abnormality. During operation there have been no signs of releases in the pool water establishing the full integrity of this fuel. The reactor has been mainly utilized for radioisotope production, beam tube experiments including neutron diffraction studies, neutron radiography etc. Studies have been completed to operate the reactor with a mixed core (HEU + LEU) to utilize the less burned HEU fuel elements. A major project of production of fission Moly using PARR-1 is in the final stages. (author)

  3. Fuel performance-experience to date and future potential

    International Nuclear Information System (INIS)

    Proebstle, R.A.; Klepfer, H.H.

    1987-01-01

    The experience in the USA to date, as reported in the Federal Energy Regulatory Commission data, conforms a very favorable cost trend for nuclear fuel costs relative to fossil fuel costs. The nuclear fuel cost promose relative to other fuels looks even better in future. Uranium supply surplus and advances in enrichment technology suggest that this trend should continue. Threats to the economic potential for nuclear fuel costs include unexpected problems in actural versus projected core and fuel technical performance. The New designs for BWR's nuclear fuel are extended to 38,000 MWd/MTU and the fuel rod reliabilities of 0.999994 are achievable. This reliability is equivalent to less than 3 fuel rod failures over the 40 year life of a reactor. (Liu)

  4. Isotopic modelling using the ENIGMA-B fuel performance code

    International Nuclear Information System (INIS)

    Rossiter, G.D.; Cook, P.M.A.; Weston, R.

    2001-01-01

    A number of experimental programmes by BNFL and other MOX fabricators have now shown that the in-pile performance of MOX fuel is generally similar to that of conventional UO 2 fuel. Models based on UO 2 fuel experience form a good basis for a description of MOX fuel behaviour. However, an area where the performance of MOX fuel is sufficiently different from that of UO 2 to warrant model changes is in the radial power and burnup profile. The differences in radial power and burnup profile arise from the presence of significant concentrations of plutonium in MOX fuel, at beginning of life, and their subsequent evolution with burnup. Amongst other effects, plutonium has a greater neutron absorption cross-section than uranium. This paper focuses on the development of a new model for the radial power and burnup profile within a UO 2 or MOX fuel rod, in which the underlying fissile isotope concentration distributions are tracked during irradiation. The new model has been incorporated into the ENIGMA-B fuel performance code and has been extended to track the isotopic concentrations of the fission gases, xenon and krypton. The calculated distributions have been validated against results from rod puncture measurements and electron probe micro-analysis (EPMA) linescans, performed during the M501 post irradiation examination (PIE) programme. The predicted gas inventory of the fuel/clad gap is compared with the isotopic composition measured during rod puncture and the measured radial distributions of burnup (from neodymium measurements) and plutonium in the fuel are compared with the calculated distributions. It is shown that there is good agreement between the code predictions and the measurements. (author)

  5. Use of advanced simulations in fuel performance codes

    International Nuclear Information System (INIS)

    Van Uffelen, P.

    2015-01-01

    The simulation of the cylindrical fuel rod behaviour in a reactor or a storage pool for spent fuel requires a fuel performance code. Such tool solves the equations for the heat transfer, the stresses and strains in fuel and cladding, the evolution of several isotopes and the behaviour of various fission products in the fuel rod. The main equations along with their limitations are briefly described. The current approaches adopted for overcoming these limitations and the perspectives are also outlined. (author)

  6. French programme for HTR fuel

    International Nuclear Information System (INIS)

    Gillet, R.M.

    1991-01-01

    It is reported that in the frameworks of the French HTR research program, stopped in 1979 the HTR coated particle fuel, fuel rod and prismatic fuel element design have been successfully developed and irradiation tested in France and specific examination methods for irradiated fuel particles, rods and graphite blocks have been developed. Currently CEA is involved in fission product transport experiments sponsored by the US Department of Energy and performed in the COMEDIE loop. Finally the CEA follows progress and developments in HTR fuel research and development throughout the world. 1 tab

  7. 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.

  8. In-core fuel management and perspectives

    International Nuclear Information System (INIS)

    Waeckel, N.

    2009-01-01

    The management of nuclear fuel inside the core has to take into account the necessity to stop the reactor periodically to renew the fuel partially and to perform maintenance operations. The fuel management strategy determines the cost of the fuel (through the number of assemblies that have been changed and their enrichment rate) and the duration of the campaign till next stop. Fuel management strategies have to conciliate different objectives: -) the safety of the reactor, -) the reliability of the fuel assemblies, -) the optimization of the fuel cost by increasing the discharge burnup. The necessity of spent fuel processing implies a maximal discharge burnup. During the 1990-2000 period, the discharge burnups have been progressively increased through the following fuel management strategies: Garance, Cyclades and Gemmes. During the years 2000-2009, the progressive absorption of the nuclear over-equipment, the opening of the European electricity markets favored power production through the MOX-parity, Alcade and Galice fuel management strategies. The perspective for next decade is to favor production to the prejudice of higher burnups. (A.C.)

  9. A statistical approach to nuclear fuel design and performance

    Science.gov (United States)

    Cunning, Travis Andrew

    As CANDU fuel failures can have significant economic and operational consequences on the Canadian nuclear power industry, it is essential that factors impacting fuel performance are adequately understood. Current industrial practice relies on deterministic safety analysis and the highly conservative "limit of operating envelope" approach, where all parameters are assumed to be at their limits simultaneously. This results in a conservative prediction of event consequences with little consideration given to the high quality and precision of current manufacturing processes. This study employs a novel approach to the prediction of CANDU fuel reliability. Probability distributions are fitted to actual fuel manufacturing datasets provided by Cameco Fuel Manufacturing, Inc. They are used to form input for two industry-standard fuel performance codes: ELESTRES for the steady-state case and ELOCA for the transient case---a hypothesized 80% reactor outlet header break loss of coolant accident. Using a Monte Carlo technique for input generation, 105 independent trials are conducted and probability distributions are fitted to key model output quantities. Comparing model output against recognized industrial acceptance criteria, no fuel failures are predicted for either case. Output distributions are well removed from failure limit values, implying that margin exists in current fuel manufacturing and design. To validate the results and attempt to reduce the simulation burden of the methodology, two dimensional reduction methods are assessed. Using just 36 trials, both methods are able to produce output distributions that agree strongly with those obtained via the brute-force Monte Carlo method, often to a relative discrepancy of less than 0.3% when predicting the first statistical moment, and a relative discrepancy of less than 5% when predicting the second statistical moment. In terms of global sensitivity, pellet density proves to have the greatest impact on fuel performance

  10. Spent fuel metal storage cask performance testing and future spent fuel concrete module performance testing

    International Nuclear Information System (INIS)

    McKinnon, M.A.; Creer, J.M.

    1988-10-01

    REA-2023 Gesellshaft fur Nuklear Service (GNS) CASTOR-V/21, Transnuclear TN-24P, and Westinghouse MC-10 metal storage casks, have been performance tested under the guidance of the Pacific Northwest Laboratory to determine their thermal and shielding performance. The REA-2023 cask was tested under Department of Energy (DOE) sponsorship at General Electric's facilities in Morris, Illinois, using BWR spent fuel from the Cooper Reactor. The other three casks were tested under a cooperative agreement between Virginia Power Company and DOE at the Idaho National Engineering Laboratory (INEL) by EGandG Idaho, Inc., using intact spent PWR fuel from the Surry reactors. The Electric Power Research Institute (EPRI) made contributions to both programs. A summary of the various cask designs and the results of the performance tests is presented. The cask designs include: solid and liquid neutron shields; lead, steel, and nodular cast iron gamma shields; stainless steel, aluminum, and copper baskets; and borated materials for criticality control. 4 refs., 8 figs., 6 tabs

  11. Modeling of high-density U-MO dispersion fuel plate performance

    International Nuclear Information System (INIS)

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

    2002-01-01

    Results from postirradiation examinations (PIE) of highly loaded U-Mo/Al dispersion fuel plates over the past several years have shown that the interaction between the metallic fuel particles and the matrix aluminum can be extensive, reducing the volume of the high-conductivity matrix phase and producing a significant volume of low-conductivity reaction-product phase. This phenomenon results in a significant decrease in fuel meat thermal conductivity during irradiation. PIE has further shown that the fuel-matrix interaction rate is a sensitive function of irradiation temperature. The interplay between fuel temperature and fuel-matrix interaction makes the development of a simple empirical correlation between the two difficult. For this reason a comprehensive thermal model has been developed to calculate temperatures throughout the fuel plate over its lifetime, taking into account the changing volume fractions of fuel, matrix and reaction-product phases within the fuel meat owing to fuel-matrix interaction; this thermal model has been incorporated into the dispersion fuel performance code designated PLATE. Other phenomena important to fuel thermal performance that are also treated in PLATE include: gas generation and swelling in the fuel and reaction-product phases, incorporation of matrix aluminum into solid solution with the unreacted metallic fuel particles, matrix extrusion resulting from fuel swelling, and cladding corrosion. The phenomena modeled also make possible a prediction of fuel plate swelling. This paper presents a description of the models and empirical correlations employed within PLATE as well as validation of code predictions against fuel performance data for U-Mo experimental fuel plates from the RERTR-3 irradiation test. (author)

  12. Transmutation Fuel Performance Code Thermal Model Verification

    Energy Technology Data Exchange (ETDEWEB)

    Gregory K. Miller; Pavel G. Medvedev

    2007-09-01

    FRAPCON fuel performance code is being modified to be able to model performance of the nuclear fuels of interest to the Global Nuclear Energy Partnership (GNEP). The present report documents the effort for verification of the FRAPCON thermal model. It was found that, with minor modifications, FRAPCON thermal model temperature calculation agrees with that of the commercial software ABAQUS (Version 6.4-4). This report outlines the methodology of the verification, code input, and calculation results.

  13. 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

  14. Investigation of microalgae HTL fuel effects on diesel engine performance and exhaust emissions using surrogate fuels

    International Nuclear Information System (INIS)

    Hossain, Farhad M.; Nabi, Md. Nurun; Rainey, Thomas J.; Bodisco, Timothy; Rahman, Md. Mostafizur; Suara, Kabir; Rahman, S.M.A.; Van, Thuy Chu; Ristovski, Zoran; Brown, Richard J.

    2017-01-01

    Highlights: • Development of a microalgae HTL surrogate of biocrude fuel using chemical compounds. • Physiochemical properties of surrogate blends were analysed. • Experimentally investigated diesel engine performance and emissions using surrogate fuels. • No significant changes in engine performance were observed with HTL surrogate blends. • Major emissions including PM, PN and CO were reduced significantly with increasing of NOx emission. - Abstract: This paper builds on previous work using surrogate fuel to investigate advanced internal combustion engine fuels. To date, a surrogate fuel of this nature has not been used for microalgae hydrothermal liquefaction (HTL) biocrude. This research used five different chemical groups found in microalgae HTL biocrude to design a surrogate fuel. Those five chemical groups constitute around 65% (by weight) of a microalgae biocrude produced by HTL. Weight percentage of the microalgae HTL biocrude chemical compounds were used to design the surrogate fuel, which was miscible with diesel at all percentages. The engine experiments were conducted on a EURO IIIA turbocharged common-rail direct-injection six-cylinder diesel engine to test engine performance and emissions. Exhaust emissions, including particulate matter and other gaseous emissions, were measured with the surrogate fuel and a reference diesel fuel. Experimental results showed that without significantly deteriorating engine performance, lower particulate mass, particulate number and CO emissions were observed with a penalty in NOx emissions for all surrogate blends compared to those of the reference diesel.

  15. 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

  16. Spent Fuel Transportation Package Performance Study - Experimental Design Challenges

    International Nuclear Information System (INIS)

    Snyder, A. M.; Murphy, A. J.; Sprung, J. L.; Ammerman, D. J.; Lopez, C.

    2003-01-01

    Numerous studies of spent nuclear fuel transportation accident risks have been performed since the late seventies that considered shipping container design and performance. Based in part on these studies, NRC has concluded that the level of protection provided by spent nuclear fuel transportation package designs under accident conditions is adequate. [1] Furthermore, actual spent nuclear fuel transport experience showcase a safety record that is exceptional and unparalleled when compared to other hazardous materials transportation shipments. There has never been a known or suspected release of the radioactive contents from an NRC-certified spent nuclear fuel cask as a result of a transportation accident. In 1999 the United States Nuclear Regulatory Commission (NRC) initiated a study, the Package Performance Study, to demonstrate the performance of spent fuel and spent fuel packages during severe transportation accidents. NRC is not studying or testing its current regulations, a s the rigorous regulatory accident conditions specified in 10 CFR Part 71 are adequate to ensure safe packaging and use. As part of this study, NRC currently plans on using detailed modeling followed by experimental testing to increase public confidence in the safety of spent nuclear fuel shipments. One of the aspects of this confirmatory research study is the commitment to solicit and consider public comment during the scoping phase and experimental design planning phase of this research

  17. Performance Evaluation of Metallic Dispersion Fuel for Advanced Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Ho Jin; Park, Jong Man; Kim, Chang Kyu; Chae, Hee Taek; Song, Kee Chan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Yeon Soo [Argonne National Laboratory, New York (United States)

    2007-07-01

    Uranium alloys with a high uranium density has been developed for high power research reactor fuel using low-enriched uranium (LEU). U-Mo alloys have been developed as candidate fuel material because of excellent irradiation behavior. Irradiation behavior of U-Mo/Al dispersion fuel has been investigated to develop high performance research reactor fuel as RERTR international research program. While plate-type and rod-type dispersion fuel elements are used for research reactors, HANARO uses rod-type dispersion fuel elements. PLATE code is developed by Argonne National Laboratory for the performance evaluation of plate-type dispersion fuel, but there is no counterpart for rod-type dispersion fuel. Especially, thermal conductivity of fuel meat decreases during the irradiation mainly because of interaction layer formation at the interface between the U-Mo fuel particle and Al matrix. The thermal conductivity of the interaction layer is not as high as the Al matrix. The growth of interaction layer is interactively affected by the temperature of fuel because it is associated with a diffusion reaction which is a thermally activated process. It is difficult to estimate the temperature profile during irradiation test due to the interdependency of fuel temperature and thermal conductivity changed by interaction layer growth. In this study, fuel performance of rod-type U-Mo/Al dispersion fuels during irradiation tests were estimated by considering the effect of interaction layer growth on the thermal conductivity of fuel meat.

  18. Performance Evaluation of Metallic Dispersion Fuel for Advanced Research Reactors

    International Nuclear Information System (INIS)

    Ryu, Ho Jin; Park, Jong Man; Kim, Chang Kyu; Chae, Hee Taek; Song, Kee Chan; Kim, Yeon Soo

    2007-01-01

    Uranium alloys with a high uranium density has been developed for high power research reactor fuel using low-enriched uranium (LEU). U-Mo alloys have been developed as candidate fuel material because of excellent irradiation behavior. Irradiation behavior of U-Mo/Al dispersion fuel has been investigated to develop high performance research reactor fuel as RERTR international research program. While plate-type and rod-type dispersion fuel elements are used for research reactors, HANARO uses rod-type dispersion fuel elements. PLATE code is developed by Argonne National Laboratory for the performance evaluation of plate-type dispersion fuel, but there is no counterpart for rod-type dispersion fuel. Especially, thermal conductivity of fuel meat decreases during the irradiation mainly because of interaction layer formation at the interface between the U-Mo fuel particle and Al matrix. The thermal conductivity of the interaction layer is not as high as the Al matrix. The growth of interaction layer is interactively affected by the temperature of fuel because it is associated with a diffusion reaction which is a thermally activated process. It is difficult to estimate the temperature profile during irradiation test due to the interdependency of fuel temperature and thermal conductivity changed by interaction layer growth. In this study, fuel performance of rod-type U-Mo/Al dispersion fuels during irradiation tests were estimated by considering the effect of interaction layer growth on the thermal conductivity of fuel meat

  19. Metal fuel manufacturing and irradiation performance

    International Nuclear Information System (INIS)

    Pedersen, D.R.; Walters, L.C.

    1992-01-01

    The advances in metal fuel by the Integral Fast Reactor Program at Argonne National Laboratory are the subject of this paper. The Integral Fast Reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The advances stressed in the paper include fuel irradiation performance, and improved passive safety. The goals and the safety philosophy of the Integral Fast Reactor Program are stressed

  20. Effect of Biodiesel Fuel Injection Timing and Venture for Gaseous Fuel Induction on the Performance, Emissions and Combustion Characteristics of Dual Fuel Engine

    Directory of Open Access Journals (Sweden)

    Mallikarjun Bhovi

    2018-02-01

    Full Text Available Advancing or retarding pilot fuel injection timing in a diesel engine provided with either conventional mechanical fuel injection (CMFIS or high pressure injection as in common rail fuel injection (CRDI systems can significantly affect its performance and tail pipe emissions. Performance of diesel engine when fueled with various biofuels as well as gaseous fuels tends to vary with subsequent changes in pilot fuel injection timings. Biodiesel derived from rubber seed oil called Rubber Seed Oil Methyl Ester (RuOME and hydrogen (H2 and hydrogen enriched compressed natural gas called (HCNG both being renewable fuels when used in diesel engines modified to operate in dual fuel mode can provide complete replacement for fossil diesel. In the present study, effect of injection timings and venture design for gas mixing on the performance, combustion and emission characteristics of dual fuel engine fitted with both CMFIS and CRDI injection systems and operated on RuOME and HCNG/hydrogen has been investigated. Results showed that high pressure CRDI assisted injection of RuOME with optimized mixing chamber (carburetor for hydrogen induction in dual fuel engine performed improved compared to that with CMFIS. In addition, for the same fuel combinations, CRDI resulted in lower biodiesel consumption, lower carbon monoxide (BSCO and hydrocarbon (BSHC emissions and increased NOx emissions than CMFIS operation.

  1. Alternate-Fueled Combustor-Sector Performance. Parts A and B; (A) Combustor Performance; (B) Combustor Emissions

    Science.gov (United States)

    Shouse, D. T.; Hendricks, R. C.; Lynch, A.; Frayne, C. W.; Stutrud, J. S.; Corporan, E.; Hankins, T.

    2012-01-01

    Alternate aviation fuels for military or commercial use are required to satisfy MIL-DTL-83133F(2008) or ASTM D 7566 (2010) standards, respectively, and are classified as "drop-in" fuel replacements. To satisfy legacy issues, blends to 50% alternate fuel with petroleum fuels are certified individually on the basis of processing and assumed to be feedstock agnostic. Adherence to alternate fuels and fuel blends requires "smart fueling systems" or advanced fuel-flexible systems, including combustors and engines, without significant sacrifice in performance or emissions requirements. This paper provides preliminary performance (Part A) and emissions and particulates (Part B) combustor sector data. The data are for nominal inlet conditions at 225 psia and 800 F (1.551 MPa and 700 K), for synthetic-paraffinic-kerosene- (SPK-) type (Fisher-Tropsch (FT)) fuel and blends with JP-8+100 relative to JP-8+100 as baseline fueling. Assessments are made of the change in combustor efficiency, wall temperatures, emissions, and luminosity with SPK of 0%, 50%, and 100% fueling composition at 3% combustor pressure drop. The performance results (Part A) indicate no quantifiable differences in combustor efficiency, a general trend to lower liner and higher core flow temperatures with increased FT fuel blends. In general, emissions data (Part B) show little differences, but with percent increase in FT-SPK-type fueling, particulate emissions and wall temperatures are less than with baseline JP-8. High-speed photography illustrates both luminosity and combustor dynamic flame characteristics.

  2. MCNP Progress & Performance Improvements

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Bull, Jeffrey S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-04-14

    Twenty-eight slides give information about the work of the US DOE/NNSA Nuclear Criticality Safety Program on MCNP6 under the following headings: MCNP6.1.1 Release, with ENDF/B-VII.1; Verification/Validation; User Support & Training; Performance Improvements; and Work in Progress. Whisper methodology will be incorporated into the code, and run speed should be increased.

  3. IRRADIATION PERFORMANCE OF U-Mo MONOLITHIC FUEL

    Directory of Open Access Journals (Sweden)

    M.K. MEYER

    2014-04-01

    Full Text Available High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. UMo alloys represent the best known tradeoff in these properties. Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.

  4. Irradiation performance of U-Mo monolithic fuel

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, M. K.; Gan, J.; Jue, J. F.; Keiser, D. D.; Perez, E.; Robinson, A.; Wachs, D. M.; Woolstenhulme, N. [Idaho National Laboratory, Idaho (Korea, Republic of); Kim, Y.S.; Hofman, G. L. [Argonne National Laboratory, Lemont (United States)

    2014-04-15

    High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. U-Mo alloys represent the best known tradeoff in these properties. Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.

  5. 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

  6. Performance of the Westinghouse WWER-1000 fuel design

    International Nuclear Information System (INIS)

    Hoglund, J.; Riznychenko, O.; Latorre, R.; Lashevych, P.

    2011-01-01

    In 2005 six (6) Westinghouse WWER-1000 Lead Test Assemblies (LTAs) were loaded in the South Ukraine Unit 3. This design has demonstrated full compatibility with resident fuel designs and all associated fuel handling and reactor components. Operations have further demonstrated adequacy of performance margins and the reliability requirements for multiple cycles of operation. The LTA's have now been discharged after completing the planned four cycles of operation and having reached an average assembly burnup in excess of 43 MWd/kgU. Post Irradiation Examinations were performed after completion of each cycle. The final LTA inspection program at end of Cycle 20 in 2010 yielded satisfactory results on all counts, and it was concluded that the 6 Westinghouse LTA's performed as expected during their operational regimes. Very good performance was demonstrated in the WWER-1000 reactor environment for the Zr-1%Nb as grid material, and ZIRLO fuel cladding and structural components. Control Rod Assemblies drop times and drag forces were all within the accepted values. The LTA program demonstrated that this fuel design is suitable for full core applications. However, the topic of fuel assembly distortion resistance was re-visited and Westinghouse therefore considered operational experience and design features from multiple development programs to enhance the basic Westinghouse WWER-1000 fuel design for Ukrainian reactors. The design now includes features that further mitigate assembly bow while at the same time improving the fuel cycle economy. This paper describes briefly the development of the Westinghouse WWER-1000 fuel design and how test results and operational experiences from multiple sources have been utilized to produce a most suitable fuel design. Early in 2011 a full region of the Westinghouse WWER-1000 design completed another full cycle of operation at South Ukraine Unit 3, all with excellent results. All 42 fuel assemblies were examined for visible damage or non

  7. Pushing back the boundaries of PWR fuel performance

    International Nuclear Information System (INIS)

    Sofer, G.A.; Skogen, F.B.; Brown, C.A.; Fresk, Y.U.

    1985-01-01

    In today's fiercely competitive PWR reload market utilities are benefiting from a variety of design innovations which are helping to cut fuel cycle costs and to improve fuel performance. An advanced PWR fuel design from Exxon, for example, currently under evaluation at the Ginna plant in the United States, offers higher burn-up and greater power cycling. (author)

  8. 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.

  9. 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.

  10. 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.

  11. Dissolution performance of plutonium nitride based fuel materials

    Energy Technology Data Exchange (ETDEWEB)

    Aneheim, E.; Hedberg, M. [Nuclear Chemistry, Chemistry and Chemical Engineering, Chalmers University of Technology, Kemivaegen 4, Gothenburg, SE41296 (Sweden)

    2016-07-01

    Nitride fuels have been regarded as one viable fuel option for Generation IV reactors due to their positive features compared to oxides. To be able to close the fuel cycle and follow the Generation IV concept, nitrides must, however, demonstrate their ability to be reprocessed. This means that the dissolution performance of actinide based nitrides has to be thoroughly investigated and assessed. As the zirconium stabilized nitrides show even better potential as fuel material than does the pure actinide containing nitrides, investigations on the dissolution behavior of both PuN and (Pu,Zr)N has been undertaken. If possible it is desirable to perform the fuel dissolutions using nitric acid. This, as most reprocessing strategies using solvent-solvent extraction are based on a nitride containing aqueous matrix. (Pu,Zr)N/C microspheres were produced using internal gelation. The spheres dissolution performance was investigated using nitric acid with and without additions of HF and Ag(II). In addition PuN fuel pellets were produced from powder and their dissolution performance were also assessed in a nitric acid based setting. It appears that both PuN and (Pu,Zr)N/C fuel material can be completely dissolved in nitric acid of high concentration with the use of catalytic amounts of HF. The amount of HF added strongly affects dissolution kinetics of (Pu, Zr)N and the presence of HF affects the 2 solutes differently, possibly due to inhomogeneity o the initial material. Large additions of Ag(II) can also be used to facilitate the dissolution of (Pu,Zr)N in nitric acid. PuN can be dissolved by pure nitric acid of high concentration at room temperature while (Pu, Zr)N is unaffected under similar conditions. At elevated temperature (reflux), (Pu,Zr)N can, however, also be dissolved by concentrated pure nitric acid.

  12. Proceedings of DUPIC fuel workshop 97

    International Nuclear Information System (INIS)

    1997-07-01

    The researchers discuss the technical aspects of DUPIC fuel fabrication in the workshop as follows; 1) The DUPIC fuel development program in KAERI 2) AECL's progress in developing the DUPIC fuel fabrication process 3) Mechanical decladding 4) Nonproliferation and safeguards aspects of the DUPIC fuel cycle concept 5) Assessment of DUPIC fuel compatibility with CANDU-6 6) The development of combination software for spent PWR fuel to fabricate the homogeneous DUPIC fuel 7) Thermodynamic properties of the DUPIC fuel and its performance 8) Captural properties of cesium and ruthenium 9) A secondary fuel removal process : Plasma processing 10) Technology development for DUPIC process safeguards

  13. Proceedings of DUPIC fuel workshop 97

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-07-01

    The researchers discuss the technical aspects of DUPIC fuel fabrication in the workshop as follows; (1) The DUPIC fuel development program in KAERI (2) AECL`s progress in developing the DUPIC fuel fabrication process (3) Mechanical decladding (4) Nonproliferation and safeguards aspects of the DUPIC fuel cycle concept (5) Assessment of DUPIC fuel compatibility with CANDU-6 (6) The development of combination software for spent PWR fuel to fabricate the homogeneous DUPIC fuel (7) Thermodynamic properties of the DUPIC fuel and its performance (8) Captural properties of cesium and ruthenium (9) A secondary fuel removal process : Plasma processing (10) Technology development for DUPIC process safeguards.

  14. Performance of commercially produced mixed-oxide fuels in EBR-II

    International Nuclear Information System (INIS)

    Hales, J.W.; Lawrence, L.A.

    1980-11-01

    Commercially produced fuels for the Fast Flux Test Facility (FFTF) were irradiated in EBR-II under conditions of high cladding temperature (approx. 700 0 C) and low power (approx. 200 W/cm) to verify that manufacturing processes did not introduce variables which significantly affect general fuel performance. Four interim examinations and a terminal examination were completed to a peak burnup of 5.2 at. % to provide irradiation data pertaining to fuel restructuring and dimensional stability at low fuel temperature, fuel-cladding reactions at high cladding temperature and general fuel behavior. The examinations indicate completely satisfactory irradiation performance for low heat rates and high cladding temperatures to 5.2 at. % burnup

  15. Experience of safety and performance improvement for fuel handling equipment

    International Nuclear Information System (INIS)

    Gyoon Chang, Sang; Hee Lee, Dae

    2014-01-01

    The purpose of this study is to provide experience of safety and performance improvement of fuel handling equipment for nuclear power plants in Korea. The fuel handling equipment, which is used as an important part of critical processes during the refueling outage, has been improved to enhance safety and to optimize fuel handling procedures. Results of data measured during the fuel reloading are incorporated into design changes. The safety and performance improvement for fuel handling equipment could be achieved by simply modifying the components and improving the interlock system. The experience provided in this study can be useful lessons for further improvement of the fuel handling equipment. (authors)

  16. Fuel performance analysis for the HAMP-1 mini plate test

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Byoung Jin; Tahka, Y. W.; Yim, J. S.; Lee, B. H. [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    U-7wt%Mo/Al- 5wt%Si dispersion fuel with 8gU/cm{sup 3} is chosen to achieve more efficiency and higher performance than the conventional U{sub 3}Si{sub 2} fuel. As part of the fuel qualification program for the KiJang research reactor (KJRR), three irradiation tests with mini-plates are on the way at the High-flux Advanced Neutron Application Reactor (HANARO). The first test among three HANARO Mini-Plate Irradiation tests (HAMP-1, 2, 3) has completed. PLATE code has been initially developed to analyze the thermal performance of high density U-Mo/Al dispersion fuel plates during irradiation [1]. We upgraded the PLATE code with the latest irradiation results which were implemented by corrosion, thermal conductivity and swelling model. Fuel performance analysis for HAMP-1 was conducted with updated PLATE. This paper presents results of performance evaluation of the HAMP-1. Maximum fuel temperature was obtained 136 .deg., which is far below the preset limit of 200 .deg. for the irradiation test. The meat swelling and corrosion thickness was also confirmed that the developed fuel would behave as anticipated.

  17. Regulatory activities in the area of fuel safety and performance

    International Nuclear Information System (INIS)

    Viktorov, A.; Couture, M.

    2005-01-01

    Generic Action Item 94G02 'Impact of Fuel Bundle Condition on Reactor Safety' in many ways determined the present priorities in regulatory activities related to fuel performance. As one of the closure criteria it required that all licensees establish 'an effective formal and systematic process for integrating fuel design, fuel and channel inspection, laboratory examination, research, operating limits and safety analysis'. To date, such a process has been, to a large extent, put in place by all licensees. To assure that such processes remain operational and effective after the GAI closure, CNSC required, through S-99, to report annually on fuel performance and major activities in the fuel safety area. The scope of reported information has been defined to allow CNSC staff evaluation of key events and trends in fuel performance. To compliment reporting by the industry, CNSC staff has conducted targeted inspections of fuel compliance programs at all sites. Combined together, these activities provide the regulator with the confidence that CANDU fuel is robust and operates with safety margins. The scrutiny, to which fuel performance has been subjected lately, has allowed identification of certain programmatic weaknesses and gaps in the knowledge concerning the fuel behaviour under various conditions. It has become apparent that top-level strategies for assessment of fuel performance may have been inadequate and far from systematic; fuel inspection practices and capabilities have varied significantly from site to site; certain issues were identified but remained unaddressed for significant time; priorities in experimental or design support activities were not assigned consistently. The presentation gives examples of areas where, in the opinion of the CNSC staff, further work is required to support fuel design and safety envelopes. The implementation of new CANFLEX fuel designs is currently being considered by the industry and CNSC staff has been engaged in the review

  18. Review of cell performance in anion exchange membrane fuel cells

    Science.gov (United States)

    Dekel, Dario R.

    2018-01-01

    Anion exchange membrane fuel cells (AEMFCs) have recently received increasing attention since in principle they allow for the use of non-precious metal catalysts, which dramatically reduces the cost per kilowatt of power in fuel cell devices. Until not long ago, the main barrier in the development of AEMFCs was the availability of highly conductive anion exchange membranes (AEMs); however, improvements on this front in the past decade show that newly developed AEMs have already reached high levels of conductivity, leading to satisfactory cell performance. In recent years, a growing number of research studies have reported AEMFC performance results. In the last three years, new records in performance were achieved. Most of the literature reporting cell performance is based on hydrogen-AEMFCs, although an increasing number of studies have also reported the use of fuels others than hydrogen - such as alcohols, non-alcohol C-based fuels, as well as N-based fuels. This article reviews the cell performance and performance stability achieved in AEMFCs through the years since the first reports in the early 2000s.

  19. The irradiation performance of austenitic stainless steel clade PWR fuel rods

    International Nuclear Information System (INIS)

    Teixeira e Silva, A.; Esteves, A.M.

    1988-01-01

    The steady state irradiation performance of austenitic stainless steel clad pressurized water reactor fuel rods is modeled with fuel performance codes of the FRAP series. These codes, originally developed to model the thermal-mechanical behavior of zircaloy clad fuel rods, are modified to model stainless steel clad fuel rods. The irradiation thermal-mechanical behavior of type 348 stainless steel and zircaloy fuel rods is compared. (author) [pt

  20. Key differences in the fabrication of US and German TRISO-coated particle fuel, and their implications on fuel performance

    International Nuclear Information System (INIS)

    Petti, D.A.; Buongiorno, J.; Maki, J.T.; Miller, G.K.; Hobbins, R.R.

    2002-01-01

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

  1. Fuel Performance Calculations for FeCrAl Cladding in BWRs

    Energy Technology Data Exchange (ETDEWEB)

    George, Nathan [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Sweet, Ryan [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Maldonado, G. Ivan [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Wirth, Brian D. [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Worrall, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    This study expands upon previous neutronics analyses of the reactivity impact of alternate cladding concepts in boiling water reactor (BWR) cores and directs focus toward contrasting fuel performance characteristics of FeCrAl cladding against those of traditional Zircaloy. Using neutronics results from a modern version of the 3D nodal simulator NESTLE, linear power histories were generated and supplied to the BISON-CASL code for fuel performance evaluations. BISON-CASL (formerly Peregrine) expands on material libraries implemented in the BISON fuel performance code and the MOOSE framework by providing proprietary material data. By creating material libraries for Zircaloy and FeCrAl cladding, the thermomechanical behavior of the fuel rod (e.g., strains, centerline fuel temperature, and time to gap closure) were investigated and contrasted.

  2. Performance and Exhaust Emissions in a Natural-Gas Fueled Dual-Fuel Engine

    Science.gov (United States)

    Shioji, Masahiro; Ishiyama, Takuji; Ikegami, Makoto; Mitani, Shinichi; Shibata, Hiroaki

    In order to establish the optimum fueling in a natural gas fueled dual fuel engine, experiments were done for some operational parameters on the engine performances and the exhaust emissions. The results show that the pilot fuel quantity should be increased and its injection timing should be advanced to suppress unburned hydrocarbon emission in the middle and low output range, while the quantity should be reduced and the timing retarded to avoid onset of knock at high loads. Unburned hydrocarbon emission and thermal efficiency are improved by avoiding too lean natural gas mixture by restricting intake charge air. However, the improvement is limited because the ignition of pilot fuel deteriorates with excessive throttling. It is concluded that an adequate combination of throttle control and equivalence ratio ensures low hydrocarbon emission and the thermal efficiency comparable to diesel operation.

  3. Fuel Performance Modeling of U-Mo Dispersion Fuel: The thermal conductivity of the interaction layers of the irradiated U-Mo dispersion fuel

    Energy Technology Data Exchange (ETDEWEB)

    Mistarhi, Qusai M.; Ryu, Ho Jin [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    U-Mo/Al dispersion fuel performed well at a low burn-up. However, higher burn-up and higher fission rate irradiation testing showed enhanced fuel meat swelling which was caused by high interaction layer growth and pore formation. The performance of the dispersion type fuel in the irradiation and un-irradiation environment is very important. During the fabrication of the dispersion type fuel an Interaction Layer (IL) is formed due to the inter-diffusion between the U-Mo fuel particles and the Al matrix which is an intermetallic compound (U,Mo)Alx. During irradiation, the IL becomes amorphous causing a further decrease in the thermal conductivity and an increase in the centerline temperature of the fuel meat. Several analytical models and numerical methods were developed to study the performance of the unirradiated U-Mo/Al dispersion fuel. Two analytical models were developed to study the performance of the irradiated U-Mo/Al dispersion fuel. In these models, the thermal conductivity of the IL was assumed to be constant. The properties of the irradiated U-Mo dispersion fuel have been investigated recently by Huber et al. The objective of this study is to develop a correlation for IL thermal conductivity during irradiation as a function of the temperature and fission density from the experimentally measured thermal conductivity of the irradiated U-Mo/Al dispersion fuel. The thermal conductivity of IL during irradiation was calculated from the experimentally measured data and a correlation was developed from the thermal conductivity of IL as a function of T and fission density.

  4. Heterogeneous fuels for minor actinides transmutation: Fuel performance codes predictions in the EFIT case study

    Energy Technology Data Exchange (ETDEWEB)

    Calabrese, R., E-mail: rolando.calabrese@enea.i [ENEA, Innovative Nuclear Reactors and Fuel Cycle Closure Division, via Martiri di Monte Sole 4, 40129 Bologna (Italy); Vettraino, F.; Artioli, C. [ENEA, Innovative Nuclear Reactors and Fuel Cycle Closure Division, via Martiri di Monte Sole 4, 40129 Bologna (Italy); Sobolev, V. [SCK.CEN, Belgian Nuclear Research Centre, Boeretang 200, B-2400 Mol (Belgium); Thetford, R. [Serco Technical and Assurance Services, 150 Harwell Business Centre, Didcot OX11 0QB (United Kingdom)

    2010-06-15

    Plutonium recycling in new-generation fast reactors coupled with minor actinides (MA) transmutation in dedicated nuclear systems could achieve a decrease of nuclear waste long-term radiotoxicity by two orders of magnitude in comparison with current once-through strategy. In a double-strata scenario, purpose-built accelerator-driven systems (ADS) could transmute minor actinides. The innovative nuclear fuel conceived for such systems demands significant R and D efforts in order to meet the safety and technical performance of current fuel systems. The Integrated Project EUROTRANS (EUROpean research programme for the TRANSmutation of high level nuclear waste in ADS), part of the EURATOM Framework Programme 6 (FP6), undertook some of this research. EUROTRANS developed from the FP5 research programmes on ADS (PDS-XADS) and on fuels dedicated to MA transmutation (FUTURE, CONFIRM). One of its main objectives is the conceptual design of a small sub-critical nuclear system loaded with uranium-free fuel to provide high MA transmutation efficiency. These principles guided the design of EFIT (European Facility for Industrial Transmutation) in the domain DESIGN of IP EUROTRANS. The domain AFTRA (Advanced Fuels for TRAnsmutation system) identified two composite fuel systems: a ceramic-ceramic (CERCER) where fuel particles are dispersed in a magnesia matrix, and a ceramic-metallic (CERMET) with a molybdenum matrix in the place of MgO matrix to host a ceramic fissile phase. The EFIT fuel is composed of plutonium and MA oxides in solid solution with isotopic vectors typical of LWR spent fuel with 45 MWd/kg{sub HM} discharge burnup and 30 years interim storage before reprocessing. This paper is focused on the thermomechanical state of the hottest fuel pins of two EFIT cores of 400 MW{sub (th)} loaded with either CERCER or CERMET fuels. For calculations three fuel performance codes were used: FEMALE, TRAFIC and TRANSURANUS. The analysis was performed at the beginning of fuel life

  5. Neutronics Benchmarks for the Utilization of Mixed-Oxide Fuel: Joint U.S./Russian Progress Report for Fiscal Year 1997 Volume 2-Calculations Performed in the United States

    Energy Technology Data Exchange (ETDEWEB)

    Primm III, RT

    2002-05-29

    This volume of the progress report provides documentation of reactor physics and criticality safety studies conducted in the US during fiscal year 1997 and sponsored by the Fissile Materials Disposition Program of the US Department of Energy. Descriptions of computational and experimental benchmarks for the verification and validation of computer programs for neutron physics analyses are included. All benchmarks include either plutonium, uranium, or mixed uranium and plutonium fuels. Calculated physics parameters are reported for all of the computational benchmarks and for those experimental benchmarks that the US and Russia mutually agreed in November 1996 were applicable to mixed-oxide fuel cycles for light-water reactors.

  6. Neutronics Benchmarks for the Utilization of Mixed-Oxide Fuel: Joint U.S./Russian Progress Report for Fiscal Year 1997 Volume 2-Calculations Performed in the United States

    International Nuclear Information System (INIS)

    Primm III, RT

    2002-01-01

    This volume of the progress report provides documentation of reactor physics and criticality safety studies conducted in the US during fiscal year 1997 and sponsored by the Fissile Materials Disposition Program of the US Department of Energy. Descriptions of computational and experimental benchmarks for the verification and validation of computer programs for neutron physics analyses are included. All benchmarks include either plutonium, uranium, or mixed uranium and plutonium fuels. Calculated physics parameters are reported for all of the computational benchmarks and for those experimental benchmarks that the US and Russia mutually agreed in November 1996 were applicable to mixed-oxide fuel cycles for light-water reactors

  7. High performance direct methanol fuel cell with thin electrolyte membrane

    Science.gov (United States)

    Wan, Nianfang

    2017-06-01

    A high performance direct methanol fuel cell is achieved with thin electrolyte membrane. 320 mW cm-2 of peak power density and over 260 mW cm-2 at 0.4 V are obtained when working at 90 °C with normal pressure air supply. It is revealed that the increased anode half-cell performance with temperature contributes primarily to the enhanced performance at elevated temperature. From the comparison of iR-compensated cathode potential of methanol/air with that of H2/air fuel cell, the impact of methanol crossover on cathode performance decreases with current density and becomes negligible at high current density. Current density is found to influence fuel efficiency and methanol crossover significantly from the measurement of fuel efficiency at different current density. At high current density, high fuel efficiency can be achieved even at high temperature, indicating decreased methanol crossover.

  8. Comparison of Core Performance with Various Oxide fuels on Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jin Ha; Kim, Myung Hyun [Kyung Hee University, Yongin (Korea, Republic of)

    2016-05-15

    The system is called Prototype GenIV Sodium-cooled Fast Reactor (PGSFR). Ultimate goal of PGSFR is test for capability of TRU transmutation. Purpose of this study is test for evaluation of in-core performance and TRU transmutation performance by applying various oxide fuel loaded TRU. Fuel type of reference core is changed to uranium-based oxide fuel. Oxide fuel has a lot of experience through fuel fabrication and reactor operation. This study performed by compared and analyzed a core performance of various oxide fuels. (U,Pu)O{sub 2} and (U,TRU)O{sub 2} which various oxide fuel types are selected as extreme case for comparison with core performance and transmutation capability of TRU isotopes. Thorium-based fuel is known that it has good performance for burner reactor due to low proliferation characteristic. To check the performance of TRU incineration for comparison with uranium-based fuel on prototype SFR, Thorium-based fuel, (Th,U)O{sub 2}, (Th,Pu)O{sub 2} and (Th,TRU)O{sub 2}, is selected. Calculations of core performance for various oxide fuel are performed using the fast calculation tool, TRANSX / DANTSTS / REBUS-3. In this study, comparison of core performance and transmutation performance is conducted with various fuel types in a sodium-cooled fast reactor. Mixed oxide fuel with TRU can produce the energy with small amount of fissile material. However, the TRU fuel is confirmed to bring a potential decline of the safety parameters. In case of (Th,U)O2 fuel, the flux level in thermal neutron region becomes lower because of higher capture cross-section of Th-232 than U-238. However, Th-232 has difficulty in converting to TRU isotopes. Therefore, the TRU consumption mass is relatively high in mixed oxide fuel with thorium and TRU.

  9. Comparison of Core Performance with Various Oxide fuels on Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Choi, Jin Ha; Kim, Myung Hyun

    2016-01-01

    The system is called Prototype GenIV Sodium-cooled Fast Reactor (PGSFR). Ultimate goal of PGSFR is test for capability of TRU transmutation. Purpose of this study is test for evaluation of in-core performance and TRU transmutation performance by applying various oxide fuel loaded TRU. Fuel type of reference core is changed to uranium-based oxide fuel. Oxide fuel has a lot of experience through fuel fabrication and reactor operation. This study performed by compared and analyzed a core performance of various oxide fuels. (U,Pu)O_2 and (U,TRU)O_2 which various oxide fuel types are selected as extreme case for comparison with core performance and transmutation capability of TRU isotopes. Thorium-based fuel is known that it has good performance for burner reactor due to low proliferation characteristic. To check the performance of TRU incineration for comparison with uranium-based fuel on prototype SFR, Thorium-based fuel, (Th,U)O_2, (Th,Pu)O_2 and (Th,TRU)O_2, is selected. Calculations of core performance for various oxide fuel are performed using the fast calculation tool, TRANSX / DANTSTS / REBUS-3. In this study, comparison of core performance and transmutation performance is conducted with various fuel types in a sodium-cooled fast reactor. Mixed oxide fuel with TRU can produce the energy with small amount of fissile material. However, the TRU fuel is confirmed to bring a potential decline of the safety parameters. In case of (Th,U)O2 fuel, the flux level in thermal neutron region becomes lower because of higher capture cross-section of Th-232 than U-238. However, Th-232 has difficulty in converting to TRU isotopes. Therefore, the TRU consumption mass is relatively high in mixed oxide fuel with thorium and TRU.

  10. 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

  11. Impact on burnup performance of coated particle fuel design in pebble bed reactor with ROX fuel

    International Nuclear Information System (INIS)

    Ho, Hai Quan; Obara, Toru

    2015-01-01

    The pebble bed reactor (PBR), a kind of high-temperature gas-cooled reactor (HTGR), is expected to be among the next generation of nuclear reactors as it has excellent passive safety features, as well as online refueling and high thermal efficiency. Rock-like oxide (ROX) fuel has been studied at the Japan Atomic Energy Agency (JAEA) as a new once-through type fuel concept. Rock-like oxide used as fuel in a PBR can be expected to achieve high burnup and improve chemical stabilities. In the once-through fuel concept, the main challenge is to achieve as high a burnup as possible without failure of the spent fuel. The purpose of this study was to investigate the impact on burnup performance of different coated fuel particle (CFP) designs in a PBR with ROX fuel. In the study, the AGR-1 Coated Particle design and Deep-Burn Coated Particle design were used to make the burnup performance comparison. Criticality and core burnup calculations were performed by MCPBR code using the JENDL-4.0 library. Results at equilibrium showed that the two reactors utilizing AGR-1 Coated Particle and Deep-Burn Coated Particle designs could be critical with almost the same multiplication factor k eff . However, the power peaking factor and maximum power per fuel ball in the AGR-1 coated particle design was lower than that of Deep-Burn coated particle design. The AGR-1 design also showed an advantage in fissions per initial fissile atoms (FIFA); the AGR-1 coated particle design produced a higher FIFA than the Deep-Burn coated particle design. These results suggest that the difference in coated particle fuel design can have an effect on the burnup performance in ROX fuel. (author)

  12. Gas-Cooled Reactor Programs annual progress report for period ending December 31, 1973

    International Nuclear Information System (INIS)

    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. 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

  14. Effects of ambient conditions on fuel cell vehicle performance

    Science.gov (United States)

    Haraldsson, K.; Alvfors, P.

    Ambient conditions have considerable impact on the performance of fuel cell hybrid vehicles. Here, the vehicle fuel consumption, the air compressor power demand, the water management system and the heat loads of a fuel cell hybrid sport utility vehicle (SUV) were studied. The simulation results show that the vehicle fuel consumption increases with 10% when the altitude increases from 0 m up to 3000 m to 4.1 L gasoline equivalents/100 km over the New European Drive Cycle (NEDC). The increase is 19% on the more power demanding highway US06 cycle. The air compressor is the major contributor to this fuel consumption increase. Its load-following strategy makes its power demand increase with increasing altitude. Almost 40% of the net power output of the fuel cell system is consumed by the air compressor at the altitude of 3000 m with this load-following strategy and is thus more apparent in the high-power US06 cycle. Changes in ambient air temperature and relative humidity effect on the fuel cell system performance in terms of the water management rather in vehicle fuel consumption. Ambient air temperature and relative humidity have some impact on the vehicle performance mostly seen in the heat and water management of the fuel cell system. While the heat loads of the fuel cell system components vary significantly with increasing ambient temperature, the relative humidity did not have a great impact on the water balance. Overall, dimensioning the compressor and other system components to meet the fuel cell system requirements at the minimum and maximum expected ambient temperatures, in this case 5 and 40 °C, and high altitude, while simultaneously choosing a correct control strategy are important parameters for efficient vehicle power train management.

  15. 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

  16. Irradiation Performance of HTGR Fuel in WWR-K Research Reactor

    International Nuclear Information System (INIS)

    Ueta, Shohei; Sakaba, Nariaki; Shaimerdenov, Asset; Gizatulin, Shamil; Chekushina, Lyudmila; Chakrov, Petr; Honda, Masaki; Takahashi, Masashi; Kitagawa, Kenichi

    2014-01-01

    A capsule irradiation test with the high temperature gas-cooled reactor (HTGR) fuel is being carried out using WWR-K research reactor in the Institute of Nuclear Physics of the Republic of Kazakhstan (INP) to attain 100 GWd/t-U of burnup under normal operating condition of a practical small-sized HTGR. This is the first HTGR fuel irradiation test for INP in Kazakhstan collaborated with Japan Atomic Energy Agency (JAEA) in frame of International Science and Technology Center (ISTC) project. In the test, TRISO coated fuel particle with low-enriched UO_2 (less than 10 % of "2"3"5U) is used, which was newly designed by JAEA to extend burnup up to 100 GWd/t-U comparing with that of the HTTR (33 GWd/t-U). Both TRISO and fuel compact as the irradiation test specimen were fabricated in basis of the HTTR fuel technology by Nuclear Fuel Industries, Ltd. in Japan. A helium-gas-swept capsule and a swept-gas sampling device installed in WWR-K were designed and constructed by INP. The irradiation test has been started in October 2012 and will be completed up to the end of February 2015. The irradiation test is in the progress up to 69 GWd/t of burnup, and integrity of new TRISO fuel has been confirmed. In addition, as predicted by the fuel design, fission gas release was observed due to additional failure of as-fabricated SiC-defective fuel. (author)

  17. 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

  18. Proceedings of the 2006 International Meeting on LWR fuel performance 'Nuclear Fuel: Addressing the future' - TopFuel 2006 Transactions

    International Nuclear Information System (INIS)

    2006-01-01

    From 22-26 October, 340 researchers, nuclear engineers and scientists from across Europe and beyond congregated in the ancient university city of Salamanca, Spain, to discuss the challenges facing the developers and manufacturers of new high-performance nuclear fuels-fuels that will help meet current and future energy demand and reduce man's over dependence upon CO 2 -emitting fossil fuels. TopFuel is an annual topical meeting organised by ENS, the American Nuclear Society and the Atomic Energy Society of Japan. This year it was co-sponsored by the IAEA, the OECD/NEA and the Spanish Nuclear Society (SNE). TopFuel's primary objective was to bring together leading specialists in the field from around the world to analyse advances in nuclear fuel management technology and to use the findings of the latest cutting-edge research to help manufacture the high performance nuclear fuels of today and tomorrow. The TopFuel 2006 agenda revolved around ten technical sessions dedicated to priority issues such as security of supply, new fuel and reactor core designs, fuel cycle strategies and spent fuel management. Among the many topics under discussion were new developments in fuel performance modelling, advanced fuel assembly design and the improved conditioning and processing of spent fuel. During the week, a poster exhibition also gave delegates the opportunity to display and discuss the results of their latest work and to network with fellow professionals. One important statement to emerge from TopFuel 2006 was that the world has enough reserves of uranium to support the large-scale and long-term production of nuclear energy. The OECD/NEA and the IAEA recently published a report entitled Uranium 2005: Resources, Production and Demand (the Red Book). The report, which makes a comprehensive assessment of uranium supplies and projected demand up until the year 2025, concludes by saying 'the uranium resource base is adequate to meet projected future requirements'. With the

  19. MARMOT update for oxide fuel modeling

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Yongfeng [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schwen, Daniel [Idaho National Lab. (INL), Idaho Falls, ID (United States); Chakraborty, Pritam [Idaho National Lab. (INL), Idaho Falls, ID (United States); Jiang, Chao [Idaho National Lab. (INL), Idaho Falls, ID (United States); Aagesen, Larry [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ahmed, Karim [Idaho National Lab. (INL), Idaho Falls, ID (United States); Jiang, Wen [Idaho National Lab. (INL), Idaho Falls, ID (United States); Biner, Bulent [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bai, Xianming [Virginia Polytechnic Inst. and State Univ. (Virginia Tech), Blacksburg, VA (United States); Tonks, Michael [Pennsylvania State Univ., University Park, PA (United States); Millett, Paul [Univ. of Arkansas, Fayetteville, AR (United States)

    2016-09-01

    This report summarizes the lower-length-scale research and development progresses in FY16 at Idaho National Laboratory in developing mechanistic materials models for oxide fuels, in parallel to the development of the MARMOT code which will be summarized in a separate report. This effort is a critical component of the microstructure based fuel performance modeling approach, supported by the Fuels Product Line in the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. The progresses can be classified into three categories: 1) development of materials models to be used in engineering scale fuel performance modeling regarding the effect of lattice defects on thermal conductivity, 2) development of modeling capabilities for mesoscale fuel behaviors including stage-3 gas release, grain growth, high burn-up structure, fracture and creep, and 3) improved understanding in material science by calculating the anisotropic grain boundary energies in UO$_2$ and obtaining thermodynamic data for solid fission products. Many of these topics are still under active development. They are updated in the report with proper amount of details. For some topics, separate reports are generated in parallel and so stated in the text. The accomplishments have led to better understanding of fuel behaviors and enhance capability of the MOOSE-BISON-MARMOT toolkit.

  20. Development of MHI PWR fuel assembly with high thermal performance

    International Nuclear Information System (INIS)

    Yasushi Makino; Masaya Hoshi; Masaji Mori; Hidetoshi Kido; Kazuo Ikeda

    2005-01-01

    Mitsubishi Heavy Industries, Ltd. (MHI) has been developing a PWR fuel assembly to meet the needs of Japanese fuel market with mainly improving its reliability such as a mechanical strength, a seismic strength and endurance. For burn-up extension of the fuel to 55 GWd/t, MHI has introduced a Zircaloy spacer grid with better neutron economics with retaining the reliability in an operating core. However, for a future power up-rating and a longer cycle operation, a higher thermal performance is required for PWR fuel assembly. To meet the needs of fuel market, MHI has developed an advanced type of Zircaloy spacer grid with a greater DNB performance while retaining the reliability of a fuel and a relatively low pressure drop. For the greater DNB performance, MHI optimized geometrical shape of mixing vane to promote a fluid mixing performance. In this report, higher DNB performance provided by the advanced Zircaloy spacer grid is presented. The results of 3D simulation for the flow behavior in 5 x 5 partial assembly, a mixing test and a water DNB test were compared between the current and the advanced spacer grids. Consequently, it was confirmed that a crossover vane enhanced a fluid mixing and the advanced spacer grid could significantly improve DNB performance compared with the current design of spacer grids. (authors)

  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. The neutronic and fuel cycle performance of interchangeable 3500 MWth metal and oxide fueled LMRs

    International Nuclear Information System (INIS)

    Fujita, E.K.; Wade, D.C.

    1990-01-01

    This study summarizes the neutronic and fuel cycle analysis performed at Argonne National Laboratory for an oxide and a metal fueled 3500 MWth LMR. These reactor designs formed the basis for a joint US/European study of LMR ATWS events. The oxide and metal core designs were developed to meet reactor performance specifications that are constrained by requirements for core loading interchangeability and for a small burnup reactivity swing. Differences in the computed performance parameters of the oxide and metal cores, arising from basic differences in their neutronic characteristics, are identified and discussed. It is shown that metal and oxide cores designed to the same ground rules exhibit many similar performance characteristics; however, they differ substantially in reactivity coefficients, control strategies, and fuel cycle options. 12 refs., 2 figs., 12 tabs

  3. Features of fuel performance at high fuel burnups

    International Nuclear Information System (INIS)

    Proselkov, V.N.; Scheglov, A.S.; Smirnov, A.V.; Smirnov, V.P.

    2001-01-01

    Some features of fuel behavior at high fuel burnups, in particular, initiation and development of rim-layer, increase in the rate of fission gas release from the fuel and increase in the inner gas pressure in the fuel rod are briefly described. Basing on the analysis of the data of post-irradiation examinations of fuel rods of WWER-440 working FA and CR fuel followers, that have been operated for five fuel cycles and got the average fuel burnup or varies as 50MW-day/kgU, a conclusion is made that the WWER-440 fuel burnup can be increased at least to average burnups of 55-58 MW-day/kgU per fuel assembly (Authors)

  4. Westinghouse fuel manufacturing systems: a step change in performance improvements

    International Nuclear Information System (INIS)

    Mutyala, Meena

    2009-01-01

    Today's competitive electrical generation industry demands that nuclear power plant operators minimize total operating costs, including fuel cycle cost while maintaining flawless fuel performance. The mission of Westinghouse Nuclear Fuel is to be the industry's most responsive supplier of flawless, value added fuel products and services, as judged by our customers. As nuclear is fast becoming the choice of many countries, existing manufacturing plants and facilities are once again running at full capacity. In this context Westinghouse Nuclear Fuel is committed to deliver a step change in performance improvement worldwide through its manufacturing operations by the introduction of a set of fundamentals collectively named the 'Westinghouse Fuel Manufacturing System' (WFMS), whose key principles are discussed in this paper. (author)

  5. Burnup performance of rock-like oxide (ROX) fuel in small pebble bed reactor with accumulative fuel loading scheme

    International Nuclear Information System (INIS)

    Simanullang, Irwan Liapto; Obara, Toru

    2017-01-01

    Highlights: • Burnup performance using ROX fuel in PBR with accumulative fuel loading scheme was analyzed. • Initial excess reactivity was suppressed by reducing 235 U enrichment in the startup condition. • Negative temperature coefficient was achieved in all condition of PBR with accumulative fuel loading scheme using ROX fuel. • Core lifetime of PBR with accumulative fuel loading scheme using ROX fuel was shorter than with UO 2 fuel. • In PBR with accumulative fuel loading scheme using ROX fuel, achieved discharged burnup can be as high as that for UO 2 fuel. - Abstract: The Japan Atomic Energy Agency (JAEA) has proposed rock-like oxide (ROX) fuel as a new, once-through type fuel concept. Here, burnup performance using ROX fuel was simulated in a pebble bed reactor with an accumulative fuel loading scheme. The MVP-BURN code was used to simulate the burnup calculation. Fuel of 5 g-HM/pebble with 20% 235 U enrichment was selected as the optimum composition. Discharged burnup could reach up to 218 GWd/t, with a core lifetime of about 8.4 years. However, high excess reactivity occurred in the initial condition. Initial fuel enrichment was therefore reduced from 20% to 4.65% to counter the initial excess reactivity. The operation period was reduced by the decrease of initial fuel enrichment, but the maximum discharged burnup was 198 GWd/t. Burnup performance of ROX fuel in this reactor concept was compared with that of UO 2 fuel obtained previously. Discharged burnup for ROX fuel in the PBR with an accumulative fuel loading scheme was as high as UO 2 fuel. Maximum power density could be lowered by introducing ROX fuel compared to UO 2 fuel. However, PBR core lifetime was shorter with ROX fuel than with UO 2 fuel. A negative temperature coefficient was achieved for both UO 2 and ROX fuels throughout the operation period.

  6. 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)

  7. SGHWR fuel performance, safety and reliability

    International Nuclear Information System (INIS)

    Pickman, D.O.; Inglis, G.H.

    1977-05-01

    The design principles involved in fuel pins and elements need to take account of the sometimes conflicting requirements of safety and reliability. The principal factors involved in this optimisation are discussed and it is shown from fuel irradiation experience in the Winfrith SGHWR that the necessary bias towards safety has not resulted in a reliability level lower than that shown by other successful water reactor designs. Reliability has important economic implications. By a detailed evaluation of SGHWR fuel defects it is shown that very few defects can be shown to be related to design, rating, or burn-up. This demonstrates that economic aspects have not over-ridden necessary criteria that most be met to achieve the desirable reliability level. It is possible that large scale experience on SGHWR fuel may eventually demonstrate that the balance is too much in favour of reliability and consideration may be given to whether design changes favouring economy could be achieved without compromising safety. The safety criteria applied to SGHWR fuel are designed to avoid any possibility of a temperature runaway in any credible accident situation. the philosophy and supporting experimental work programme are outlines and the fuel design features which particularly contribute to maximising safety margins are outlined. Reference is made to the new 60-pin fuel element to be used in the commercial SGHWRs and to its comparison in design and performance aspects with the 36-pin element that has been used to date in the Winfrith SGHWR. (author)

  8. TRISO Fuel Performance: Modeling, Integration into Mainstream Design Studies, and Application to a Thorium-fueled Fusion-Fission Hybrid Blanket

    Energy Technology Data Exchange (ETDEWEB)

    Powers, Jeffrey James [Univ. of California, Berkeley, CA (United States)

    2011-11-30

    This study focused on creating a new tristructural isotropic (TRISO) coated particle fuel performance model and demonstrating the integration of this model into an existing system of neutronics and heat transfer codes, creating a user-friendly option for including fuel performance analysis within system design optimization and system-level trade-off studies. The end product enables both a deeper understanding and better overall system performance of nuclear energy systems limited or greatly impacted by TRISO fuel performance. A thorium-fueled hybrid fusion-fission Laser Inertial Fusion Energy (LIFE) blanket design was used for illustrating the application of this new capability and demonstrated both the importance of integrating fuel performance calculations into mainstream design studies and the impact that this new integrated analysis had on system-level design decisions. A new TRISO fuel performance model named TRIUNE was developed and verified and validated during this work with a novel methodology established for simulating the actual lifetime of a TRISO particle during repeated passes through a pebble bed. In addition, integrated self-consistent calculations were performed for neutronics depletion analysis, heat transfer calculations, and then fuel performance modeling for a full parametric study that encompassed over 80 different design options that went through all three phases of analysis. Lastly, side studies were performed that included a comparison of thorium and depleted uranium (DU) LIFE blankets as well as some uncertainty quantification work to help guide future experimental work by assessing what material properties in TRISO fuel performance modeling are most in need of improvement. A recommended thorium-fueled hybrid LIFE engine design was identified with an initial fuel load of 20MT of thorium, 15% TRISO packing within the graphite fuel pebbles, and a 20cm neutron multiplier layer with beryllium pebbles in flibe molten salt coolant. It operated

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

    International Nuclear Information System (INIS)

    Ryu, Ho Jin; Park, Jong Man; Oh, Seok Jin; Jang, Se Jung; Yu, Byung Ok; Lee, Choong Seong; Seo, Chul Gyo; Chae, Hee Taek; Kim, Chang Kyu

    2008-01-01

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

  10. BISON Fuel Performance Analysis of IFA-796 Rod 3 & 4 and Investigation of the Impact of Fuel Creep

    Energy Technology Data Exchange (ETDEWEB)

    Wirth, Brian [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sweet, Ryan T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-08-01

    In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace the currently used zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromiumaluminum (FeCrAl) alloys because they exhibit slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and slow cladding consumption in the presence of high temperature steam. These alloys should also exhibit increased “coping time” in the event of an accident scenario by improving the mechanical performance at high temperatures, allowing greater flexibility to achieve core cooling. As a continuation of the development of these alloys, in-reactor irradiation testing of FeCrAl cladded fuel rods has started. In order to provide insight on the possible behavior of these fuel rods as they undergo irradiation in the Halden Boiling Water Reactor, engineering analysis has been performed using FeCrAl material models implemented into the BISON fuel performance code. This milestone report provides an update on the ongoing development of modeling capability to predict FeCrAl cladding fuel performance and to provide an early look at the possible behavior of planned in-reactor FeCrAl cladding experiments. In particular, this report consists of two separate analyses. The first analysis consists of fuel performance simulations of IFA-796 rod 4 and two segments of rod 3. These simulations utilize previously implemented material models for the C35M FeCrAl alloy and UO2 to provide a bounding behavior analysis corresponding to variation of the initial fuel cladding gap thickness within the fuel rod. The second analysis is an assessment of the fuel and cladding stress states after modification of the fuel creep model that is currently implemented in the BISON fuel performance code. Effects from modifying the fuel creep model were identified for the BISON simulations

  11. The U.S. RERTR program status and progress

    International Nuclear Information System (INIS)

    Travelli, A.

    1998-01-01

    The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program since its inception in 1978 is described. A brief summary of the results which the RERTR Program had achieved by the end of 1996 in collaboration with its many international partners is followed by a detailed review of the major events, findings, and activities of 1997. Significant progress has been made during the past year. In the area of U.S. acceptance of spent fuel from foreign research reactors, several shipments have taken place and additional are being planned. Intense fuel development activities are in progress, including procurement of equipment, screening of candidate materials, and production of microplates. Irradiation of the first series of microplates began in August 1997 in the Advanced Test Reactor, in Idaho. Progress has been made in the Russian RERTR program, which aims to develop and demonstrate within five years the technical means needed to convert Russian-supplied research reactors to LEU fuels. The study of an alternative LEU core for the FRM-II design has been extended to address, with favorable results, controversial performance issues which were raised at last year's meeting. Progress was also made on several aspects of producing molybdenum-99 from fission targets utilizing LEU instead of HEU. Various types of targets and processes are being pursued, with FDA approval of an LEU process projected to occur within two years. The feasibility of LEU Fuel conversion for three important DOE research reactors (BMRR, HFBR, and HFIR) has been evaluated by the RERTR program. In spite of the many momentous events which have occurred during the intervening years, and the excellent progress achieved, the most important challenges that the RERTR program faces today are not very different in type from those that were faced during the first RERTR meeting. Now, as then, the most important task is to develop new LEU fuels satisfying requirements which cannot be satisfied by any

  12. The U.S. RERTR Program. Status and progress

    International Nuclear Information System (INIS)

    Travelli, A.

    1997-01-01

    The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program since its inception in 1978 is described. A brief summary of the results which the RERTR Program had achieved by the end of 1996 in collaboration with its many international partners is followed by a detailed review of the major events, findings, and activities of 1997. Significant progress has been made during the past year. In the area of U.S. acceptance of spent fuel from foreign research reactors, several shipments have taken place and additional are being planned. Intense fuel development activities are in progress, including procurement of equipment, screening of candidate materials, and production of microplates. Irradiation of the first series of microplates began in August 1997 in the Advanced Test Reactor, in Idaho. Progress has been made in the Russian RERTR program, which aims to develop and demonstrate within five years the technical means needed to convert Russian-supplied research reactors to LEU fuels. The study of an alternative LEU core for the FRM-II design has been extended to address, with favorable results, controversial performance issues which were raised at last year's meeting. Progress was also made on several aspects of producing molybdenum-99 from fission targets utilizing LEU instead of HEU. Various types of targets and processes are being pursued, with FDA approval of an LEU process projected to occur within two years. The feasibility of LEU fuel conversion for three important DOE research reactors (BMRR, HFBR, and HFIR) has been evaluated by the RERTR program. In spite of the many momentous events which have occurred during the intervening years, and the excellent progress achieved, the most important challenges that the RERTR program faces today are not very different in type from those that were faced during the first RERTR meeting. Now, as then, the most important task is to develop new LEU fuels satisfying requirements which cannot be satisfied by any

  13. Manufacture of fuel and fuel channels and their performance in Indian PHWRS - an overview

    International Nuclear Information System (INIS)

    Kalidas, R.

    2005-01-01

    Nuclear Fuel Complex (NFC) at Hyderabad is a conglomeration of chemical, metallurgical and mechanical plants, processing uranium and zirconium in two separate streams and culminating in the fuel assembly plant. Apart from manufacturing fuel for Pressurised Heavy Water Reactors (PHWRs) and Boiling Water Reactors (BWRs), NFC is also engaged in the manufacture of reactor core structurals for these reactors. NFC has carried out several technological developments over the years and implemented them for the manufacture of fuel, calandria tubes and pressure tubes for PHWRs. Keeping in pace with the Nuclear Power Programme envisaged by the Department of Atomic Energy, NFC had augmented its production capacities in all these areas. The paper highlights several actions initiated in the areas of fuel design, fuel manufacturing, manufacturing of zirconium alloy core structurals, fuel clad tubes and components and their performance in Indian PHWRs. (author)

  14. Annual progress report on the NSRR experiments, (19)

    International Nuclear Information System (INIS)

    1991-08-01

    Fuel behavior studies in simulated reactivity initiated accident (RIA) conditions have been performed by utilizing Nuclear Safety Research Reactor (NSRR) since October 1975. This report describes the results obtained from experiments performed from January through December, 1987. A total of 38 tests was carried out during this period : those are 5 NSRR standard fuel test, 6 fuel design parameter tests (2 fuel tests with iodine, 3 tests with SUS claddings, and 1 for destructive force measurement test), 7 cooling parameter tests (3 test with flow tube and 4 bundle tests), 8 advanced type fuel tests (1 NRC tests, 5 PCI-resistant fuel tests, 2 pressurized hollow fuel tests), 1 high pressure / high temperature tests, 2 high temperature / high pressure loop tests, 3 high temperature flooding tests, 3 fuel motion observation tests, and 3 miscellaneous atmospheric pressure capsule tests (1 fuel damage test, 2 acceleration tests). The progress of computer code development is also described. (author)

  15. Used Nuclear Fuel Loading and Structural Performance Under Normal Conditions of Transport- Demonstration of Approach and Results on Used Fuel Performance Characterization

    Energy Technology Data Exchange (ETDEWEB)

    Adkins, Harold [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Geelhood, Ken [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Koeppel, Brian [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Coleman, Justin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bignell, John [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Flores, Gregg [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wang, Jy-An [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sanborn, Scott [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Spears, Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States); Klymyshyn, Nick [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-09-30

    This document addresses Oak Ridge National Laboratory milestone M2FT-13OR0822015 Demonstration of Approach and Results on Used Nuclear Fuel Performance Characterization. This report provides results of the initial demonstration of the modeling capability developed to perform preliminary deterministic evaluations of moderate-to-high burnup used nuclear fuel (UNF) mechanical performance under normal conditions of storage (NCS) and normal conditions of transport (NCT) conditions. This report also provides results from the sensitivity studies that have been performed. Finally, discussion on the long-term goals and objectives of this initiative are provided.

  16. Spent nuclear fuel integrity during dry storage - performance tests and demonstrations

    International Nuclear Information System (INIS)

    McKinnon, M.A.; Doherty, A.L.

    1997-06-01

    This report summarizes the results of fuel integrity surveillance determined from gas sampling during and after performance tests and demonstrations conducted from 1983 through 1996 by or in cooperation with the US DOE Office of Commercial Radioactive Waste Management (OCRWM). The cask performance tests were conducted at Idaho National Engineering Laboratory (INEL) between 1984 and 1991 and included visual observation and ultrasonic examination of the condition of the cladding, fuel rods, and fuel assembly hardware before dry storage and consolidation of fuel, and a qualitative determination of the effects of dry storage and fuel consolidation on fission gas release from the spent fuel rods. The performance tests consisted of 6 to 14 runs involving one or two loading, usually three backfill environments (helium, nitrogen, and vacuum backfills), and one or two storage system orientations. The nitrogen and helium backfills were sampled and analyzed to detect leaking spent fuel rods. At the end of each performance test, periodic gas sampling was conducted on each cask. A spent fuel behavior project (i.e., enhanced surveillance, monitoring, and gas sampling activities) was initiated by DOE in 1994 for intact fuel in a CASTOR V/21 cask and for consolidated fuel in a VSC-17 cask. The results of the gas sampling activities are included in this report. Information on spent fuel integrity is of interest in evaluating the impact of long-term dry storage on the behavior of spent fuel rods. Spent fuel used during cask performance tests at INEL offers significant opportunities for confirmation of the benign nature of long-term dry storage. Supporting cask demonstration included licensing and operation of an independent spent fuel storage installation (ISFSI) at the Virginia Power (VP) Surry reactor site. A CASTOR V/21, an MC-10, and a Nuclear Assurance NAC-I28 have been loaded and placed at the VP ISFSI as part of the demonstration program. 13 refs., 14 figs., 9 tabs

  17. Water channel reactor fuels and fuel channels: Design, performance, research and development. Proceedings of a technical committee meeting

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-01-01

    The International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) recommended holding a Technical Committee Meeting on Water Channel Reactor Fuel including into this category fuels and pressure tubes/fuel channels for Atucha-I and II, BWR, CANDU, FUGEN and RBMK reactors. The IWGFPT considered that even if the characteristics of Atucha, CANDUs, BWRs, FUGEN and RBMKs differ considerably, there are also common features. These features include materials aspects, as well as core, fuel assembly and fuel rod design, and some safety issues. There is also some similarity in fuel power history and operating conditions (Atucha-I and II, FUGEN and RBMK). Experts from 11 countries participated at the meeting and presented papers on technology, performance, safety and design, and materials aspects of fuels and pressure tubes/fuel channels for the above types of water channel reactors. Refs, figs, tabs.

  18. Water channel reactor fuels and fuel channels: Design, performance, research and development. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    1998-01-01

    The International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) recommended holding a Technical Committee Meeting on Water Channel Reactor Fuel including into this category fuels and pressure tubes/fuel channels for Atucha-I and II, BWR, CANDU, FUGEN and RBMK reactors. The IWGFPT considered that even if the characteristics of Atucha, CANDUs, BWRs, FUGEN and RBMKs differ considerably, there are also common features. These features include materials aspects, as well as core, fuel assembly and fuel rod design, and some safety issues. There is also some similarity in fuel power history and operating conditions (Atucha-I and II, FUGEN and RBMK). Experts from 11 countries participated at the meeting and presented papers on technology, performance, safety and design, and materials aspects of fuels and pressure tubes/fuel channels for the above types of water channel reactors

  19. DUPIC fuel performance from reactor physics viewpoint

    International Nuclear Information System (INIS)

    Choi, H.; Rhee, B.W.; Park, H.

    1995-01-01

    A preliminary study was performed for the evaluation of Stress Corrosion Cracking (SCC) parameters of nominal DUPIC fuel in CANDU reactor. For the reference 2-bundle shift refueling scheme, the predicted ramped power and power increase of the 43-element DUPIC fuel in the equilibrium core are below the SCC thresholds of CANDU natural uranium fuel. For 4-bundle shift refueling scheme, the envelope of element ramped power and power increase upon refueling are 8% and 44% higher than those of 2-bundle shift refueling scheme on the average, respectively, and both schemes are not expected to cause SCC failures. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-08-01

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

  1. Influence of performance characteristic of a gaseous fuel supply system on hydrocarbon emissions of a dual-fuel engine

    Energy Technology Data Exchange (ETDEWEB)

    Ren, J.; Wang, Z.Y.; Zhong, H.; Hao, S.H. [Xi' an Jiaotong Univ., Dept. of Automobile Engineering, Xi' an (China)

    2000-11-01

    The performance of the gaseous fuel supply and its influence on hydrocarbon (HC) emissions of dual-fuel engines have been investigated. A new design of manifold respirators with mixers is also presented in the paper. The design of the gaseous fuel supply system has a great influence on HC emissions in the dual-fuel engine at light load. The problem of scavenging is discussed and solved by using the manifold respirators in the dual-fuel engine. It performs the function of retarding the gaseous fuel entry timing from the moment of intake valve opening, and its delaying effects have been measured and tested. Experimental results show that the manifold respirator gives the best performance in reducing HC emissions compared with a common pipe mixer and a respirator with bo miser. In addition, the mixing effects are sensitive to the mixer configuration. (Author)

  2. Performance and fuel conversion efficiency of a spark ignition engine fueled with iso-butanol

    International Nuclear Information System (INIS)

    Irimescu, Adrian

    2012-01-01

    Highlights: ► Iso-butanol use in a port injection spark ignition engine. ► Fuel conversion efficiency calculated based on chassis dynamometer measurements. ► Combined study of engine efficiency and air–fuel mixture temperature. ► Excellent running characteristics with minor fuel system modifications. ► Up to 11% relative drop in part load efficiency due to incomplete fuel vaporization. -- Abstract: Alcohols are increasingly used as fuels for spark ignition engines. While ethanol is most commonly used, long chain alcohols such as butanol feature several advantages like increased heating value and reduced corrosive action. This study investigated the effect of fueling a port injection engine with iso-butanol, as compared to gasoline operation. Performance levels were maintained within the same limits as with the fossil fuel without modifications to any engine component. An additional electronic module was used for increasing fuel flow by extending the injection time. Fuel conversion efficiency decreased when the engine was fueled with iso-butanol by up to 9% at full load and by up to 11% at part load, calculated as relative values. Incomplete fuel evaporation was identified as the factor most likely to cause the drop in engine efficiency.

  3. Extending the application range of a fuel performance code from normal operating to design basis accident conditions

    International Nuclear Information System (INIS)

    Van Uffelen, P.; Gyori, C.; Schubert, A.; Laar, J. van de; Hozer, Z.; Spykman, G.

    2008-01-01

    Two types of fuel performance codes are generally being applied, corresponding to the normal operating conditions and the design basis accident conditions, respectively. In order to simplify the code management and the interface between the codes, and to take advantage of the hardware progress it is favourable to generate a code that can cope with both conditions. In the first part of the present paper, we discuss the needs for creating such a code. The second part of the paper describes an example of model developments carried out by various members of the TRANSURANUS user group for coping with a loss of coolant accident (LOCA). In the third part, the validation of the extended fuel performance code is presented for LOCA conditions, whereas the last section summarises the present status and indicates needs for further developments to enable the code to deal with reactivity initiated accident (RIA) events

  4. performance calculations of gadolinium oxide and boron nitride coated fuel

    International Nuclear Information System (INIS)

    Tanker, E.; Uslu, I.; Disbudak, H.; Guenduez, G.

    1997-01-01

    A comparative study was performed on the behaviour of natural uranium dioxide-gadolinium oxide mixture fuel and boron nitride coated low enriched fuel in a pressurized water reactor. A fuel element containing one burnable poison fuel pins was modeled with the computer code WIMS, and burn-up dependent critically, fissile isotope inventory and two dimensional power distribution were obtained. Calculations were performed for burnable poison fuels containing 5% and 10% gadolinium oxide and for those coated with 1μ,5μ and 10μ of boron nitride. Boron nitride coating was found superior to gadolinium oxide on account of its smoother criticality curve, lower power peaks and insignificant change in fissile isotope content

  5. 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.

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

    International Nuclear Information System (INIS)

    Mujid S. Kazimi; Pavel Hejzlar

    2006-01-01

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

  7. Solid electrolyte fuel cells

    Science.gov (United States)

    Isaacs, H. S.

    Progress in the development of functioning solid electrolyte fuel cells is summarized. The solid electrolyte cells perform at 1000 C, a temperature elevated enough to indicate high efficiencies are available, especially if the cell is combined with a steam generator/turbine system. The system is noted to be sulfur tolerant, so coal containing significant amounts of sulfur is expected to yield satisfactory performances with low parasitic losses for gasification and purification. Solid oxide systems are electrically reversible, and are usable in both fuel cell and electrolysis modes. Employing zirconium and yttrium in the electrolyte provides component stability with time, a feature not present with other fuel cells. The chemical reactions producing the cell current are reviewed, along with materials choices for the cathodes, anodes, and interconnections.

  8. Performance of bio fuels in diesel engines

    International Nuclear Information System (INIS)

    Nunez I, Manuel L; Prada V, Laura P

    2007-01-01

    This paper shows the preliminary results of pilot plant tests developed in oil catalytic hydrotreating process, where the crude palm oil or a mixture of crude palm oil and mineral diesel is treated with an injection of 99% pure hydrogen flux, in a fixed bed reactor at high pressures and temperatures, in a presence of Nickel Molybdenum catalyst supported on alumina bed. The main product of this process is a fuel (bio diesel) which has the same or better properties than the diesel obtained by petroleum refining. It has been made some performance fuel tests in diesel engine? with good results in terms of power, torque and fuel consumption, without any changes in engine configuration. Considering the characteristics of the Catalytic hydrotreated bio diesel compare to conventional diesel, both fuels have similar distillation range? however, bio diesel has better flash point, cetane index and thermal stability. Gas fuels (methane, ethane, and propane) CO 2 and water are the secondary products of the process.

  9. 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

  10. Improving performance with accident tolerant-fuels

    International Nuclear Information System (INIS)

    Gomes, Daniel S.; Muniz, Rafael O.R.; Giovedi, Claudia; Universidade de São Paulo

    2017-01-01

    After the Fukushima reactor accident, efforts to improve risk management in nuclear operations have included the intensification of research on accident-tolerant fuels (ATFs). In this investigation, the physical properties of recently developed ATFs were compared with those of the current standard fuel, UO 2 - Zr. The goals for innovative fuel design include a rigorous characterization of the thermal, mechanical, and chemical considerations. The intentions are to lengthen the burnup cycle, raise the power density, and improve safety. Fuels must have a high uranium density - above that supported by UO 2 - and possess a coating that exhibits better oxidation resistance than Zircaloys. ATFs such as U 3 Si 2 , UN, and UC contain a higher uranium density and thermal conductivity than UO 2 , providing significant benefits. The ideal combination of fuel and cladding must increase performance in a loss-of-coolant accident. However, U 3 Si 2 , UN, and UC have a disadvantage; their respective swelling rates are higher than that of UO 2 . These ATFs also have thermal conductivities approximately four times higher than that of UO 2 . A study was conducted investigating the hydrogen generated by the oxidation of zirconium alloys in contact with steam using cladding options such as Fe-Cr-Al and silicon carbide. It was confirmed that ferritic alloys offer a better response under severe conditions, because of their mechanical properties as creep rate. The findings of this study indicate that advanced fuels should replace UO 2 - Zr as the fuel system of choice. (author)

  11. Improving performance with accident tolerant-fuels

    Energy Technology Data Exchange (ETDEWEB)

    Gomes, Daniel S.; Muniz, Rafael O.R.; Giovedi, Claudia, E-mail: dsgomes@ipen.br, E-mail: rafael.orm@gmail.com, E-mail: claudia.giovedi@ctmsp.mar.mil.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Universidade de São Paulo (USP), SP (Brazil). Departamento de Engenharia Naval e Oceânica

    2017-07-01

    After the Fukushima reactor accident, efforts to improve risk management in nuclear operations have included the intensification of research on accident-tolerant fuels (ATFs). In this investigation, the physical properties of recently developed ATFs were compared with those of the current standard fuel, UO{sub 2} - Zr. The goals for innovative fuel design include a rigorous characterization of the thermal, mechanical, and chemical considerations. The intentions are to lengthen the burnup cycle, raise the power density, and improve safety. Fuels must have a high uranium density - above that supported by UO{sub 2} - and possess a coating that exhibits better oxidation resistance than Zircaloys. ATFs such as U{sub 3}Si{sub 2}, UN, and UC contain a higher uranium density and thermal conductivity than UO{sub 2}, providing significant benefits. The ideal combination of fuel and cladding must increase performance in a loss-of-coolant accident. However, U{sub 3}Si{sub 2}, UN, and UC have a disadvantage; their respective swelling rates are higher than that of UO{sub 2}. These ATFs also have thermal conductivities approximately four times higher than that of UO{sub 2}. A study was conducted investigating the hydrogen generated by the oxidation of zirconium alloys in contact with steam using cladding options such as Fe-Cr-Al and silicon carbide. It was confirmed that ferritic alloys offer a better response under severe conditions, because of their mechanical properties as creep rate. The findings of this study indicate that advanced fuels should replace UO{sub 2} - Zr as the fuel system of choice. (author)

  12. Effects of UO2 fuel microstructure and density on fuel in-reactor performance

    International Nuclear Information System (INIS)

    Hansson, L.

    1988-02-01

    The volume changes of UO 2 fuel pellets, produced by neutron irradiation, can be characterized by two processes: fission spike induced densification through pore skrinkage and later fission produced induced swelling of UO 2 matrix. In-pile densification is controlled by the initial density and microstructure of the fuel, particularly by the pore size distribution. The extent of swelling depends mainly on the amount of fission products produced, but the fission gas release as well as the swelling may be reduced by increasing the grain size of UO 2 . Fabrication of fuel pellets having certain in-reactor properties requires detailed knowledge of the effects of individual fabrication parameters. The irradiation experience of fuels fabricated by using different conversion and pelletizing methods is extensive. Based on this experience, some general characteristics of stable/well-performing fuel microstructures have been summarized

  13. Modelling fuel cell performance using artificial intelligence

    Science.gov (United States)

    Ogaji, S. O. T.; Singh, R.; Pilidis, P.; Diacakis, M.

    Over the last few years, fuel cell technology has been increasing promisingly its share in the generation of stationary power. Numerous pilot projects are operating worldwide, continuously increasing the amount of operating hours either as stand-alone devices or as part of gas turbine combined cycles. An essential tool for the adequate and dynamic analysis of such systems is a software model that enables the user to assess a large number of alternative options in the least possible time. On the other hand, the sphere of application of artificial neural networks has widened covering such endeavours of life such as medicine, finance and unsurprisingly engineering (diagnostics of faults in machines). Artificial neural networks have been described as diagrammatic representation of a mathematical equation that receives values (inputs) and gives out results (outputs). Artificial neural networks systems have the capacity to recognise and associate patterns and because of their inherent design features, they can be applied to linear and non-linear problem domains. In this paper, the performance of the fuel cell is modelled using artificial neural networks. The inputs to the network are variables that are critical to the performance of the fuel cell while the outputs are the result of changes in any one or all of the fuel cell design variables, on its performance. Critical parameters for the cell include the geometrical configuration as well as the operating conditions. For the neural network, various network design parameters such as the network size, training algorithm, activation functions and their causes on the effectiveness of the performance modelling are discussed. Results from the analysis as well as the limitations of the approach are presented and discussed.

  14. Modelling fuel cell performance using artificial intelligence

    Energy Technology Data Exchange (ETDEWEB)

    Ogaji, S.O.T.; Singh, R.; Pilidis, P.; Diacakis, M. [Power Propulsion and Aerospace Engineering Department, Centre for Diagnostics and Life Cycle Costs, Cranfield University (United Kingdom)

    2006-03-09

    Over the last few years, fuel cell technology has been increasing promisingly its share in the generation of stationary power. Numerous pilot projects are operating worldwide, continuously increasing the amount of operating hours either as stand-alone devices or as part of gas turbine combined cycles. An essential tool for the adequate and dynamic analysis of such systems is a software model that enables the user to assess a large number of alternative options in the least possible time. On the other hand, the sphere of application of artificial neural networks has widened covering such endeavours of life such as medicine, finance and unsurprisingly engineering (diagnostics of faults in machines). Artificial neural networks have been described as diagrammatic representation of a mathematical equation that receives values (inputs) and gives out results (outputs). Artificial neural networks systems have the capacity to recognise and associate patterns and because of their inherent design features, they can be applied to linear and non-linear problem domains. In this paper, the performance of the fuel cell is modelled using artificial neural networks. The inputs to the network are variables that are critical to the performance of the fuel cell while the outputs are the result of changes in any one or all of the fuel cell design variables, on its performance. Critical parameters for the cell include the geometrical configuration as well as the operating conditions. For the neural network, various network design parameters such as the network size, training algorithm, activation functions and their causes on the effectiveness of the performance modelling are discussed. Results from the analysis as well as the limitations of the approach are presented and discussed. (author)

  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. Performance of a Fuel-Cell-Powered, Small Electric Airplane Assessed

    Science.gov (United States)

    Berton, Jeffrey J.

    2004-01-01

    Rapidly emerging fuel-cell-power technologies may be used to launch a new revolution of electric propulsion systems for light aircraft. Future small electric airplanes using fuel cell technologies hold the promise of high reliability, low maintenance, low noise, and - with the exception of water vapor - zero emissions. An analytical feasibility and performance assessment was conducted by NASA Glenn Research Center's Airbreathing Systems Analysis Office of a fuel-cell-powered, propeller-driven, small electric airplane based on a model of the MCR-01 two-place kitplane (Dyn'Aero, Darois, France). This assessment was conducted in parallel with an ongoing effort by the Advanced Technology Products Corporation and the Foundation for Advancing Science and Technology Education. Their project - partially funded by a NASA grant - is to design, build, and fly the first manned, continuously propelled, nongliding electric airplane. In our study, an analytical performance model of a proton exchange membrane (PEM) fuel cell propulsion system was developed and applied to a notional, two-place light airplane modeled after the MCR-01 kitplane. The PEM fuel cell stack was fed pure hydrogen fuel and humidified ambient air via a small automotive centrifugal supercharger. The fuel cell performance models were based on chemical reaction analyses calibrated with published data from the fledgling U.S. automotive fuel cell industry. Electric propeller motors, rated at two shaft power levels in separate assessments, were used to directly drive a two-bladed, variable-pitch propeller. Fuel sources considered were compressed hydrogen gas and cryogenic liquid hydrogen. Both of these fuel sources provided pure, contaminant-free hydrogen for the PEM cells.

  17. Operational limitations of light water reactors relating to fuel performance

    International Nuclear Information System (INIS)

    Cheng, H.S.

    1976-07-01

    General aspects of fuel performance for typical Boiling and Pressurized Water Reactors are presented. Emphasis is placed on fuel failures in order to make clear important operational limitations. A discussion of fuel element designs is first given to provide the background information for the subsequent discussion of several fuel failure modes that have been identified. Fuel failure experiences through December 31, 1974, are summarized. The operational limitations that are required to mitigate the effects of fuel failures are discussed

  18. Sensitivity assessment of fuel performance codes for LOCA accident scenario

    Energy Technology Data Exchange (ETDEWEB)

    Abe, Alfredo; Gomes, Daniel; Silva, Antonio Teixeira e; Muniz, Rafael O.R. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Giovedi, Claudia; Martins, Marcelo, E-mail: ayabe@ipen.br, E-mail: claudia.giovedi@labrisco.usp.br [Universidade de Sao Paulo (LABRISCO/USP), Sao Paulo, SP (Brazil). Lab. de Análise, Avaliação e Gerenciamento de Risco

    2017-07-01

    FRAPCON code predicts fuel rod performance in LWR (Light Water Reactor) by modeling fuel responses under normal operating conditions and anticipated operational occurrences; FRAPTRAN code is applied for fuel transient under fast transient and accident conditions. The codes are well known and applied for different purposes and one of the use is to address sensitivity analysis considering fuel design parameters associated to fabrication, moreover can address the effect of physical models bias. The objective of this work was to perform an assessment of fuel manufacturing parameters tolerances and fuel models bias using FRAPCON and FRAPTRAN codes for Loss of Coolant Accident (LOCA) scenario. The preliminary analysis considered direct approach taken into account most relevant manufacturing tolerances (lower and upper bounds) related to design parameters and physical models bias without considering their statistical distribution. The simulations were carried out using the data available in the open literature related to the series of LOCA experiment performed at the Halden reactor (specifically IFA-650.5). The manufacturing tolerances associated to design parameters considered in this paper were: enrichment, cladding thickness, pellet diameter, pellet density, and filling gas pressure. The physical models considered were: fuel thermal expansion, fission gas release, fuel swelling, irradiation creep, cladding thermal expansion, cladding corrosion, and cladding hydrogen pickup. The results obtained from sensitivity analysis addressed the impact of manufacturing tolerances and physical models in the fuel cladding burst time observed for the IFA-650.5 experiment. (author)

  19. Sensitivity assessment of fuel performance codes for LOCA accident scenario

    International Nuclear Information System (INIS)

    Abe, Alfredo; Gomes, Daniel; Silva, Antonio Teixeira e; Muniz, Rafael O.R.; Giovedi, Claudia; Martins, Marcelo

    2017-01-01

    FRAPCON code predicts fuel rod performance in LWR (Light Water Reactor) by modeling fuel responses under normal operating conditions and anticipated operational occurrences; FRAPTRAN code is applied for fuel transient under fast transient and accident conditions. The codes are well known and applied for different purposes and one of the use is to address sensitivity analysis considering fuel design parameters associated to fabrication, moreover can address the effect of physical models bias. The objective of this work was to perform an assessment of fuel manufacturing parameters tolerances and fuel models bias using FRAPCON and FRAPTRAN codes for Loss of Coolant Accident (LOCA) scenario. The preliminary analysis considered direct approach taken into account most relevant manufacturing tolerances (lower and upper bounds) related to design parameters and physical models bias without considering their statistical distribution. The simulations were carried out using the data available in the open literature related to the series of LOCA experiment performed at the Halden reactor (specifically IFA-650.5). The manufacturing tolerances associated to design parameters considered in this paper were: enrichment, cladding thickness, pellet diameter, pellet density, and filling gas pressure. The physical models considered were: fuel thermal expansion, fission gas release, fuel swelling, irradiation creep, cladding thermal expansion, cladding corrosion, and cladding hydrogen pickup. The results obtained from sensitivity analysis addressed the impact of manufacturing tolerances and physical models in the fuel cladding burst time observed for the IFA-650.5 experiment. (author)

  20. Integral Fast Reactor Program annual progress report, FY 1994

    International Nuclear Information System (INIS)

    Chang, Y.I.; Walters, L.C.; Laidler, J.J.; Pedersen, D.R.; Wade, D.C.; Lineberry, J.J.

    1994-12-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1994. Technical accomplishments are presented in the following areas of the IFR technology development activities: metal fuel performance; pyroprocess development; safety experiments and analyses; core design development; fuel cycle demonstration; and LMR technology R ampersand D

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

    Energy Technology Data Exchange (ETDEWEB)

    Mujid S. Kazimi; Pavel Hejzlar

    2006-01-31

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

  2. Performance of fuel system at different diesel temperature

    Science.gov (United States)

    Xu, Xiaoyong; Li, Xiaolu; Sun, Zai

    2010-08-01

    This paper presents the findings about performance of the fuel system of a diesel engine at different diesel temperature obtained through simulation and experiment. It can be seen from these findings that at the same rotational speed of fuel pump, the initial pressure in the fuel pipe remain unchanged as the fuel temperature increases, the peak pressure at the side of fuel pipe near the injector delays, and its largest value of pressure decreases. Meanwhile, at the same temperature, as the rotational speed increases, the initial pressure of fuel pipe is also essentially the same, the arrival of its peaks delays, and its largest value of pressure increases. The maximum fuel pressure at the side of fuel pipe near the injector has an increase of 28.9 %, 22.3%, and 13.9% respectively than the previous ones according to its conditions. At the same rotational speed, as the temperature increases, the injection quantity through the nozzle orifice decreases. At the same temperature, as the rotational speed increases, the injection quantity through the nozzle orifice increases. These experimental results are consistent with simulation results.

  3. Progress Report 1994

    International Nuclear Information System (INIS)

    1995-01-01

    This document is the 1994 annual progress report of the CEA-Direction of Waste Management (DGD). It comprises four chapters. The first chapter is a general presentation of radioactive wastes, of the management of liquid effluents, solid wastes, sealed sources, of the relations with the ANDRA (The French Agency for the Management of Radioactive Wastes), and of the research and development studies in progress for the improvement of waste management. The second chapter concerns the spent fuels and their reprocessing, in particular AGR and PWR type reactor fuels, the ''Caramel'' fuel from Osiris reactor and the cover elements from the Rapsodie reactor core. The long time storage of ancient fuels is also discussed. The third chapter concerns the dismantling of decommissioned installations, the actions in progress and the planning of dismantling actions up to the year 2000. Chapter four is devoted to the management of wastes from the Direction of Military Applications (DAM), the actions in progress in the different DAM centers and the cleansing projects at Marcoule plant. (J.S.). 5 figs., 28 tabs., 21 photos., 3 appendix

  4. An innovative fuel design concept for improved Light Water Reactor performance and safety

    International Nuclear Information System (INIS)

    Tulenko, J.S.; Connell, R.G.

    1993-01-01

    The primary goal of this research is to develop a new fuel design which will have improved thermal/mechanical performance characteristics greatly superior to current thermal and mechanical design performance. The mechanical/thermal constraints define the lifetime of the fuel, the maximum power at which the fuel can be operated, the probability of fuel failure over core lifetime, and the integrity of a core during a transient excursion. The thermal/mechanical limits act to degrade fuel integrity when they are violated. The purpose of this project is to investigate a novel design for light water reactor fuel which will extend fuel performance limits and improve reactor safety even further than is currently achieved. This project is investigating liquid metal bonding of LWR fuel in order to radically decrease fuel centerline temperatures which has major performance and safety benefits. The project will verify the compatibility of the liquid metal bond with both the fuel pellets and cladding material, verify the performance enhancement features of the new design over the fuel lifetime, and verify the economic fabricability of the concept and will show how this concept will benefit the LWR nuclear industry

  5. Fuel R and D international programmes, a way to demonstrate future fuel performances

    International Nuclear Information System (INIS)

    Vanderborck, Y.; Mertens, L.; Dekeyser, J.; Sannen, L.

    1997-01-01

    As a MOX fuel manufacturer, BELGONUCLEAIRE have spent more than 15 years promoting and managing International R and D Programmes, many of them in close cooperation with SCK''centrdot'' CEN. Such programmes dedicated to MOX versus UO 2 fuel behaviour are most of the time based on irradiation in research reactors in which the investigated fuel is submitted to power variations and to ramp testing or are performed in commercial reactors. This paper is focused on recent programmes concerned by high and medium burn-up in BWR and PWR conditions for MOX fuel. It will present also the new opportunities for new programmes. The goals, the programmes descriptions and the expected data being part of these R and D programmes is presented. (author)

  6. Lessons from the SPA European performance assessment exercise for spent fuel disposal in granite

    International Nuclear Information System (INIS)

    Baudoin, P.; Besnus, F.; Gay, D.; Serres, C.

    2000-01-01

    The SPA project, addressing performance assessment of spent fuel disposal, is a direct continuation of the efforts made by the European Community since 1982 to build a common understanding of the methods applicable to performance assessment of a deep geological disposal. It enabled to draw preliminary conclusions on the respective importance of the radionuclides present in the different types of waste, on the influence of some of the main assumptions used in the modelling and helped to specify the expectable role of disposal system components. From 1996 to 1999, six national research institutions in six member countries of EU (Spain, Germany, France, The Netherlands, Belgium and Finland) were involved in this project devoted to the case of spent fuel. This paper only addresses the assessments made by IPSN (France) for the granite site in the case of the normal evolution scenario. According to the results obtained, the maximum effective dose is in a first time mainly caused by two fission products ( 129 I and 79 Se). Then, radionuclides from radioactive decay chains ( 226 Ra , 230 Th and 229 Th) progressively become the main contributors. (author)

  7. Severe fuel-damage scoping test performance

    International Nuclear Information System (INIS)

    Gruen, G.E.; Buescher, B.J.

    1983-01-01

    As a result of the Three Mile Island Unit-2 (TMI-2) accident, the Nuclear Regulatory Commission has initiated a severe fuel damage test program to evaluate fuel rod and core response during severe accidents similar to TMI-2. The first test of Phase I of this series has been successfully completed in the Power Burst Facility at the Idaho National Engineering Laboratory. Following the first test, calculations were performed using the TRAC-BD1 computer code with actual experimental boundary conditions. This paper discusses the test conduct and performance and presents the calculated and measured test bundle results. The test resulted in a slow heatup to 2000 K over about 4 h, with an accelerated reaction of the zirconium cladding at temperatures above 1600 K in the lower part or the bundle and 2000 K in the upper portion of the bundle

  8. Performance of a pellet boiler fired with agricultural fuels

    International Nuclear Information System (INIS)

    Carvalho, Lara; Wopienka, Elisabeth; Pointner, Christian; Lundgren, Joakim; Verma, Vijay Kumar; Haslinger, Walter; Schmidl, Christoph

    2013-01-01

    Highlights: ► Performance evaluation of a pellet boiler operated with different agricultural fuels. ► Agricultural fuels could be burn in the tested boiler for a certain period of time. ► All the fuels (except straw and Sorghum) satisfied the European legal requirements. ► Boilers for burning agricultural fuels should have a flexible control system. - Abstract: The increasing demand for woody biomass increases the price of this limited resource, motivating the growing interest in using woody materials of lower quality as well as non-woody biomass fuels for heat production in Europe. The challenges in using non-woody biomass as fuels are related to the variability of the chemical composition and in certain fuel properties that may induce problems during combustion. The objective of this work has been to evaluate the technical and environmental performance of a 15 kW pellet boiler when operated with different pelletized biomass fuels, namely straw (Triticum aestivum), Miscanthus (Miscanthus × giganteus), maize (Zea mays), wheat bran, vineyard pruning (from Vitis vinifera), hay, Sorghum (Sorghum bicolor) and wood (from Picea abies) with 5% rye flour. The gaseous and dust emissions as well as the boiler efficiency were investigated and compared with the legal requirements defined in the FprEN 303-5 (final draft of the European standard 303-5). It was found that the boiler control should be improved to better adapt the combustion conditions to the different properties of the agricultural fuels. Additionally, there is a need for a frequent cleaning of the heat exchangers in boilers operated with agricultural fuels to avoid efficiency drops after short term operation. All the agricultural fuels satisfied the legal requirements defined in the FprEN 303-5, with the exception of dust emissions during combustion of straw and Sorghum. Miscanthus and vineyard pruning were the best fuels tested showing comparable emission values to wood combustion

  9. Integral Fast Reactor Program. Annual progress report, FY 1992

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.; Walters, L.C.; Laidler, J.J.; Pedersen, D.R.; Wade, D.C.; Lineberry, M.J.

    1993-06-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1992. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R&D.

  10. Report of the collaboration project for research and development of sphere-pac fuel among JNC-PSI-NRG (1). Planning, fuel design, pin fabrication

    International Nuclear Information System (INIS)

    Morihira, Masayuki; Ozawa, Takayuki; Tomita, Yutaka; Suzuki, Masahiro; Kihara, Yoshiyuki; Shigetome, Yoshiaki; Kohno, Shusaku

    2004-07-01

    The collaboration project concerning sphere-pac fuel among JNC, Swiss PSI (Paul Scherrer Institut) and Dutch NRG (Nuclear Research and Consultancy Group) is in progress. Final target of the project is comparative irradiation tests of sphere-pac fuel in the HFR (High Flux Reactor) in Petten in the Netherlands with pellet type fuel and vipack fuel. Total 16 fuel segments (8 pins) of these three types of fuel are planned to be irradiated. Two sphere-pac fuel segments contain 5%Np in addition to 20%Pu-MOX. Other segments contain no Np. The objective of the irradiation tests is to obtain the restructuring data in the early beginning of life for SPF as well as power-to-melt test data for the potential study of SPF. At the same time introduction of modeling technique for irradiation performance analysis, fuel design, fuel fabrication is also important objective for JNC. Fabrication of irradiation test pins was completed till May 2003 in PSI. After transportation of the fuel pins to Petten, two times of irradiation were performed in January to March in 2004 and now post irradiation tests are in progress. Later two irradiations will be done till the autumn in 2004. This report summarized the basic plan, fuel design, and fabrication of irradiation test pins concerning this collaboration project. (author)

  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. The development of the fuel rod transient performance analysis code FTPAC

    International Nuclear Information System (INIS)

    Han Zhijie; Ji Songtao

    2014-01-01

    Fuel rod behavior, especially the integrity of cladding, played an important role in fuel safety research during reactor transient and hypothetical accidents conditions. In order to study fuel rod performance under transient accidents, FTPAC (Fuel Transient Performance Analysis Code) has been developed for simulating light water reactor fuel rod transient behavior when power or coolant boundary conditions are rapidly changing. It is composed of temperature, mechanical deformation, cladding oxidation and gas pressure model. The assessment was performed by comparing FTPAC code analysis result to experiments data and FRAPTRAN code calculations. Comparison shows that, the FTPAC gives reasonable agreement in temperature, deformation and gas pressure prediction. And the application of slip coefficient is more suitable for simulating the sliding between pellet and cladding when the gap is closed. (authors)

  13. 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)

  14. Performance concerns for high duty fuel cycle

    International Nuclear Information System (INIS)

    Esposito, V.J.; Gutierrez, J.E.

    1999-01-01

    One of the goals of the nuclear industry is to achieve economic performance such that nuclear power plants are competitive in a de-regulated market. The manner in which nuclear fuel is designed and operated lies at the heart of economic viability. In this sense reliability, operating flexibility and low costs are the three major requirements of the NPP today. The translation of these three requirements to the design is part of our work. The challenge today is to produce a fuel design which will operate with long operating cycles, high discharge burnup, power up-rating and while still maintaining all design and safety margins. European Fuel Group (EFG) understands that to achieve the required performance high duty/energy fuel designs are needed. The concerns for high duty design includes, among other items, core design methods, advanced Safety Analysis methodologies, performance models, advanced material and operational strategies. The operational aspects require the trade-off and evaluation of various parameters including coolant chemistry control, material corrosion, boiling duty, boron level impacts, etc. In this environment MAEF is the design that EFG is now offering based on ZIRLO alloy and a robust skeleton. This new design is able to achieve 70 GWd/tU and Lead Test Programs are being executed to demonstrate this capability. A number of performance issues which have been a concern with current designs have been resolved such as cladding corrosion and incomplete RCCA insertion (IRI). As the core duty becomes more aggressive other new issues need to be addressed such as Axial Offset Anomaly. These new issues are being addressed by combination of the new design in concert with advanced methodologies to meet the demanding needs of NPP. The ability and strategy to meet high duty core requirements, flexibility of operation and maintain acceptable balance of all technical issues is the discussion in this paper. (authors)

  15. Castor-V/21 PWR spent fuel storage cask performance test

    International Nuclear Information System (INIS)

    Creer, J.M.; Schoonen, D.H.

    1986-01-01

    Performance testing of a CASTOR-V/21 PWR spent fuel storage cask manufactured by Gesellschaft fur Nuklear Service (GNS) was performed as part of a cooperative program between Virginia Power and the US Department of Energy. The performance test consisted of obtaining cask handling experience and heat transfer, shielding, and limited fuel integrity data. Five heat transfer test runs were performed with 21 Surry reactor spent fuel assemblies generating approximately 28 kW. Test runs were performed vacuum, nitrogen, and helium backfill environments with the cask in both vertical and horizontal orientations. Cask exterior surface gamma and neutron dose rates were measured with the cask fully loaded. Gas samples were obtained at the beginning and end of each run with nitrogen or helium environments to verify fuel integrity. The heat transfer performance of the CASTOR-V/21 cask was exceptionally good. Peak clad temperatures with helium and nitrogen environments with the cask in a vertical orientation and with helium with the cask in a horizontal orientation were less than 380 0 C. Vertical vacuum and horizontal nitrogen test runs resulted in peak clad temperatures over 380 0 , but the temperatures were not excessively high ( 0 C). The shielding performance of the cask met the design goal of less than 200 mrem/hr. Cask surface dose rates of <75 mrem/hr can easily be established with minor gamma shielding design refinements if desired. Gas samples obtained during testing indicated no leaking fuel rods were present in the cask. It was concluded that the cask performed satisfactorily from heat transfer and shielding perspectives

  16. CASTOR-V/21 PWR spent fuel storage cask performance test

    International Nuclear Information System (INIS)

    Creer, J.M.; Schoonen, D.H.

    1986-01-01

    Performance testing of a CASTOR-V/21 PWR spent fuel storage cask manufactured by Gesellschaft fur Nuklear Service (GNS) was performed as part of a cooperative program between Virginia Power and the US Department of Energy. The performance test consisted of obtaining cask handling experience and heat transfer, shielding, and limited fuel integrity data. Five heat transfer test runs were performed with 21 Surry reactor spent fuel assemblies generating approximately 28 kW. Test runs were performed with vacuum, nitrogen, and helium backfills in both vertical and horizontal orientations. Cask exterior surface gamma and neutron dose rates were measured with the cask fully loaded. Gas samples were obtained at the beginning and end of each run with nitrogen or helium backfills to verify fuel integrity. The heat transfer performance of the CASTOR-V/21 cask was exceptionally good. Peak clad temperatures with helium and nitrogen backfills in a vertical orientation and with helium in a horizontal orientation were less than 380 0 C. Vertical vacuum and horizontal nitrogen runs resulted in peak clad temperatures over 380 0 , but the temperatures were not excessively high ( 0 C). The shielding performance of the cask met the design expectation of less than 200 mrem/h. Cask surface dose rates of <75 mrem/h can easily be established with minor gamma shielding design refinements if desired. Gas samples obtained during testing indicated no leaking fuel rods were present in the cask. It was concluded that the cask performed satisfactorily from heat transfer and shielding perspectives

  17. High performance fuel technology development : Development of high performance cladding materials

    International Nuclear Information System (INIS)

    Park, Jeongyong; Jeong, Y. H.; Park, S. Y.

    2012-04-01

    The superior in-pile performance of the HANA claddings have been verified by the successful irradiation test and in the Halden research reactor up to the high burn-up of 67GWD/MTU. The in-pile corrosion and creep resistances of HANA claddings were improved by 40% and 50%, respectively, over Zircaloy-4. HANA claddings have been also irradiated in the commercial reactor up to 2 reactor cycles, showing the corrosion resistance 40% better than that of ZIRLO in the same fuel assembly. Long-term out-of-pile performance tests for the candidates of the next generation cladding materials have produced the highly reliable test results. The final candidate alloys were selected and they showed the corrosion resistance 50% better than the foreign advanced claddings, which is beyond the original target. The LOCA-related properties were also improved by 20% over the foreign advanced claddings. In order to establish the optimal manufacturing process for the inner and outer claddings of the dual-cooled fuel, 18 different kinds of specimens were fabricated with various cold working and annealing conditions. Based on the performance tests and various out-of-pile test results obtained from the specimens, the optimal manufacturing process was established for the inner and outer cladding tubes of the dual-cooled fuel

  18. Oxide thickness measurement for monitoring fuel performance at high burnup

    International Nuclear Information System (INIS)

    Jaeger, M.A.; Van Swam, L.F.P.; Brueck-Neufeld, K.

    1991-01-01

    For on-site monitoring of the fuel performance at high burnup, Advanced Nuclear Fuels uses the linear scan eddy current method to determine the oxide thickness of irradiated Zircaloy fuel cans. Direct digital data acquisition methods are employed to collect the data on magnetic storage media. This field-proven methodology allows oxide thickness measurements and rapid interpretation of the data during the reactor outages and makes it possible to immediately reinsert the assemblies for the next operating cycle. The accuracy of the poolside measurements and data acquisition/interpretation techniques have been verified through hot cell metallographic measurements of rods previously measured in the fuel pool. The accumulated data provide a valuable database against which oxide growth models have been benchmarked and allow for effective monitoring of fuel performance. (orig.) [de

  19. The history, genotoxicity and carcinogenicity of carbon-based fuels and their emissions: part 4 - alternative fuels.

    Science.gov (United States)

    Claxton, Larry D

    2015-01-01

    Much progress has been made in reducing the pollutants emitted from various combustors (including diesel engines and power plants) by the use of alternative fuels; however, much more progress is needed. Not only must researchers improve fuels and combustors, but also there is a need to improve the toxicology testing and analytical chemistry methods associated with these complex mixtures. Emissions from many alternative carbonaceous fuels are mutagenic and carcinogenic. Depending on their source and derivation, alternative carbonaceous fuels before combustion may or may not be genotoxic; however, in order to know their genotoxicity, appropriate chemical analysis and/or bioassay must be performed. Newly developed fuels and combustors must be tested to determine if they provide a public health advantage over existing technologies - including what tradeoffs can be expected (e.g., decreasing levels of PAHs versus increasing levels of NOx and possibly nitroarenes in ambient air). Another need is to improve exposure estimations which presently are a weak link in doing risk analyses. Copyright © 2014 Elsevier B.V. All rights reserved.

  20. Theoretical and experimental analysis of fast reactor fuel performance

    International Nuclear Information System (INIS)

    Kummerer, K.R.; Freund, D.; Steiner, H.

    1982-09-01

    In order to predict behavior, performance, and capability of prototypic fuel pins a standard operational scheme for the SNR-300 fast breeder reactor is established considering besides normal operation unscheduled power changes and shutdowns. The behavior during the whole lifetime is calculated using the updated SATURN codes and - for special conditions as power transients and skewed fuel rod power - the new TRANSIENT and TEXDIF codes. The results of these calculations are compared to experimental findings. It is demonstrated that the level of modeling and the knowledge of material properties under irradiation are sufficient for a quantitative description of the fuel pin performance under the above mentioned conditions. (orig.) [de

  1. Theoretical performance of hydrogen-bromine rechargeable SPE fuel cell

    Science.gov (United States)

    Savinell, Robert F.; Fritts, S. D.

    1987-01-01

    A mathematical model was formulated to describe the performance of a hydrogen-bromine fuel cell. Porous electrode theory was applied to the carbon felt flow-by electrode and was coupled to theory describing the solid polymer electrolyte (SPE) system. Parametric studies using the numerical solution to this model were performed to determine the effect of kinetic, mass transfer, and design parameters on the performance of the fuel cell. The results indicate that the cell performance is most sensitive to the transport properties of the SPE membrane. The model was also shown to be a useful tool for scale-up studies.

  2. Spent fuel performance in geologic repository environments

    International Nuclear Information System (INIS)

    Bradley, D.J.

    1985-10-01

    The performance assessment of the waste package is a current area of study in the United States program to develop a geologic repository for nuclear waste isolation. The waste package is presently envisioned as the waste form and its surrounding containers and possibly a packing material composed of crushed host rock or mixtures of that rock with clays. This waste package is tied to performance criteria set forth in recent legislation. It is the goal of the Civilian Radioactive Waste Management Program to obtain the necessary information on the waste package, in several geologic environments, to show that the waste package provides reasonable assurance of meeting established performance criteria. This paper discusses the United States program directed toward managing high-level radioactive waste, with emphasis on the current effort to define the behavior of irradiated spent fuel in repository groundwaters. Current studies are directed toward understanding the rate and nature (such as valence state, colloid form if any, solid phase controlling solubility) of radionuclide release from the spent fuel. Due to the strong interactive effect of radiation, thermal fields, and waste package components on this release, current spent fuel studies are being conducted primarily in the presence of waste package components over a wide range of potential environments

  3. Final Report - Effects of Impurities on Fuel Cell Performance and Durability

    Energy Technology Data Exchange (ETDEWEB)

    Trent Molter

    2012-08-18

    This program is focused on the experimental determination of the effects of key hydrogen side impurities on the performance of PEM fuel cells. Experimental data has been leveraged to create mathematical models that predict the performance of PEM fuel cells that are exposed to specific impurity streams. These models are validated through laboratory experimentation and utilized to develop novel technologies for mitigating the effects of contamination on fuel cell performance. Results are publicly disseminated through papers, conference presentations, and other means.

  4. WWER fuel performance, development, QA and future prospects at Loviisa NPS

    Energy Technology Data Exchange (ETDEWEB)

    Loesoenen, P [Imatran Voima Oy, Vantaa (Finland)

    1994-12-31

    The essential characteristics of Loviisa fuel service are presented. A brief description is given of the steps in developing the performance values including: burnup dependent linear heat rate for fuel rods, comparison of measured and calculated cladding creep down of some rods, development of average discharge burnup for unit 1 and unit 2, results from experimental irradiation of test rods in research reactors. Some noticeable design changes performed and in-reactor behaviour of the lead fuel assemblies are also discussed. The quality assurance of the fuel procurement and operation is explained. Future prospects are connected with power raise of Loviisa reactors, discontinuing of spent fuel transportation to Russia for final disposal and new WWER fuel deliverers. 4 figs., 1 tab., 8 refs.

  5. 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

  6. 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

  7. The Fuel Accident Condition Simulator (FACS) furnace system for high temperature performance testing of VHTR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Demkowicz, Paul A., E-mail: paul.demkowicz@inl.gov [Idaho National Laboratory, 2525 Fremont Avenue, MS 3860, Idaho Falls, ID 83415-3860 (United States); Laug, David V.; Scates, Dawn M.; Reber, Edward L.; Roybal, Lyle G.; Walter, John B.; Harp, Jason M. [Idaho National Laboratory, 2525 Fremont Avenue, MS 3860, Idaho Falls, ID 83415-3860 (United States); Morris, Robert N. [Oak Ridge National Laboratory, 1 Bethel Valley Road, Oak Ridge, TN 37831 (United States)

    2012-10-15

    Highlights: Black-Right-Pointing-Pointer A system has been developed for safety testing of irradiated coated particle fuel. Black-Right-Pointing-Pointer FACS system is designed to facilitate remote operation in a shielded hot cell. Black-Right-Pointing-Pointer System will measure release of fission gases and condensable fission products. Black-Right-Pointing-Pointer Fuel performance can be evaluated at temperatures as high as 2000 Degree-Sign C in flowing helium. - Abstract: The AGR-1 irradiation of TRISO-coated particle fuel specimens was recently completed and represents the most successful such irradiation in US history, reaching peak burnups of greater than 19% FIMA with zero failures out of 300,000 particles. An extensive post-irradiation examination (PIE) campaign will be conducted on the AGR-1 fuel in order to characterize the irradiated fuel properties, assess the in-pile fuel performance in terms of coating integrity and fission metals release, and determine the fission product retention behavior during high temperature safety testing. A new furnace system has been designed, built, and tested to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000 Degree-Sign C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, and Eu), iodine, and fission gases (Kr, Xe). Fission gases released from the fuel to the sweep gas are monitored in real time using dual cryogenic traps fitted with high purity germanium detectors. Condensable fission products are collected on a plate attached to a water-cooled cold finger that can be exchanged periodically without interrupting the test. Analysis of fission products on the condensation plates involves dry gamma counting followed by chemical analysis of selected isotopes. This paper will describe design and operational details of the Fuel Accident Condition Simulator furnace system and the associated

  8. Solid Polymer Fuel Cells. Electrode and membrane performance studies

    Energy Technology Data Exchange (ETDEWEB)

    Moeller-Holst, S.

    1996-12-31

    This doctoral thesis studies aspects of fuel cell preparation and performance. The emphasis is placed on preparation and analysis of low platinum-loading solid polymer fuel cell (SPEC) electrodes. A test station was built and used to test cells within a wide range of real operating conditions, 40-150{sup o}C and 1-10 bar. Preparation and assembling equipment for single SPFCs was designed and built, and a new technique of spraying the catalyst layer directly onto the membrane was successfully demonstrated. Low Pt-loading electrodes (0.1 mg Pt/cm{sup 2}) prepared by the new technique exhibited high degree of catalyst utilization. The performance of single cells holding these electrodes is comparable to state-of-the-art SPFCs. Potential losses in single cell performance are ascribed to irreversibilities by analysing the efficiency of the Solid Oxide Fuel Cell by means of the second law of thermodynamics. The water management in membranes is discussed for a model system and the results are relevant to fuel cell preparation and performance. The new spray deposition technique should be commercially interesting as it involves few steps as well as techniques that are adequate for larger scale production. 115 refs., 43 figs., 18 tabs.

  9. Solid Polymer Fuel Cells. Electrode and membrane performance studies

    Energy Technology Data Exchange (ETDEWEB)

    Moeller-Holst, S

    1997-12-31

    This doctoral thesis studies aspects of fuel cell preparation and performance. The emphasis is placed on preparation and analysis of low platinum-loading solid polymer fuel cell (SPEC) electrodes. A test station was built and used to test cells within a wide range of real operating conditions, 40-150{sup o}C and 1-10 bar. Preparation and assembling equipment for single SPFCs was designed and built, and a new technique of spraying the catalyst layer directly onto the membrane was successfully demonstrated. Low Pt-loading electrodes (0.1 mg Pt/cm{sup 2}) prepared by the new technique exhibited high degree of catalyst utilization. The performance of single cells holding these electrodes is comparable to state-of-the-art SPFCs. Potential losses in single cell performance are ascribed to irreversibilities by analysing the efficiency of the Solid Oxide Fuel Cell by means of the second law of thermodynamics. The water management in membranes is discussed for a model system and the results are relevant to fuel cell preparation and performance. The new spray deposition technique should be commercially interesting as it involves few steps as well as techniques that are adequate for larger scale production. 115 refs., 43 figs., 18 tabs.

  10. Yucca Mountain Project - Argonne National Laboratory annual progress report, FY 1994

    Energy Technology Data Exchange (ETDEWEB)

    Bates, J.K.; Fortner, J.A.; Finn, P.A.; Wronkiewicz, D.J.; Hoh, J.C.; Emery, J.W.; Buck, E.C.; Wolf, S.F.

    1995-02-01

    This document reports on the work done by the Nuclear Waste Management Section of the Chemical Technology Division (CMT), Argonne National Laboratory, in the period October 1993-September 1994. Studies have been performed to evaluate the performance of nuclear waste glass and spent fuel samples under unsaturated conditions (low volume water contact) that are likely to exist in the Yucca Mountain environment being considered as a potential site for a high-level waste repository. Tests with simulated waste glasses have been in progress for over eight years and demonstrate that actinides from initially fresh glass surfaces will be released as a result of the spallation of reacted glass layers from the surface, as the small volume of water passes over the waste form. Studies are also underway to evaluate the performance of spent fuel samples and unirradiated UO{sub 2} in projected repository conditions. Tests with UO{sub 2} have been ongoing for nine years and show that the oxidation of UO{sub 2} occurs rapidly, and the resulting paragenetic sequence of secondary phases that form on the sample surface is similar to that observed in natural analogues. The reaction of spent fuel samples under conditions similar to those used with UO{sub 2} have been in progress for nearly two years, and the results suggest that spent fuel follows the same reaction progress as UO{sub 2}. The release of individual fission products and transuranic elements was not congruent, with the release being controlled by the formation of small particles or colloids that are suspended in solution and transported away from the waste form. The reaction progress depends on the composition of the spent fuel samples used and, likely, on the composition of the groundwater that contacts the waste form.

  11. Perspectives of the LPG-fuel market

    International Nuclear Information System (INIS)

    Anon.

    1996-01-01

    In today's context of environment protection cares and in agreement with the actions carried out by the local authorities and organizations, the LPG-fuel appears in the media with a reviving good and credible reputation. However, one can ask about its real perspectives of development. This short paper analyses the conditions for this development at the international and at the national scale. LPG-fuel is the most used and most performing alternative fuel in the world and can contribute to the energy independence of the countries with national LPG resources. Its technological mastery leads progressively the car manufacturers to propose vehicles already equipped with dual-fuel engines, but to reach its expected success, LPG-fuel needs a financial aid from the national authorities in the form of taxes reduction. Its privileged markets are the taxicabs, the public transportation systems and the fleets of intervention vehicles. In France, the network of LPG distribution stations is progressively reinforcing but a revision of the legal texts concerning the training of personnel and the approval of repair shops must be updated. (J.S.)

  12. Research and development into power reactor fuel performance

    International Nuclear Information System (INIS)

    Notley, M.J.F.

    1983-07-01

    The nuclear fuel in a power reactor must perform reliably during normal operation, and the consequences of abnormal events must be researched and assessed. The present highly reliable operation of the natural UO 2 in the CANDU power reactors has reduced the need for further work in this area; however a core of expertise must be retained for purposes such as training of new staff, retaining the capability of reacting to unforeseen circumstances, and participating in the commercial development of new ideas. The assessment of fuel performance during accidents requires research into many aspects of materials, fuel and fission product behaviour, and the consolidation of that knowledge into computer codes used to evaluate the consequences of any particular accident. This work is growing in scope, much is known from out-reactor work at temperatures up to about 1500 degreesC, but the need for in-reactor verification and investigation of higher-temperature accidents has necessitated the construction of a major new in-reactor test loop and the initiation of the associated out-reactor support programs. Since many of the programs on normal and accident-related performance are generic in nature, they will be applicable to advanced fuel cycles. Work will therefore be gradually transferred from the present, committed power reactor system to support the next generation of thorium-based reactor cycles

  13. Study on thermal performance and margins of BWR fuel elements

    International Nuclear Information System (INIS)

    Stosic, Zoran

    1999-01-01

    This paper contributes to developing a methodology of predicting and analyzing thermal performance and margins of Boiling Water Reactor (BWR) fuel assemblies under conditions of reaching high quality Boiling Crisis and subsequent post-dryout thermal hydraulics causing temperature excursion of fuel cladding. Operational margins against dryout and potential for increasing fuel performance with appropriate benefits are discussed. The philosophy of modeling with its special topics are demonstrated on the HECHAN (HEated CHannel ANalyzer) model as the state-of-art for thermal-hydraulics analysis of BWR fuel assemblies in pre- and post-dryout two-phase flow regimes. The scope of further work either being or has to be performed concerning implementation of new physical aspects, including domain extension of HECHAN model applications to the Pressurized Water Reactors (PWRs), is discussed. Finally, a comprehensive overview of the literature dealing with development of the model is given. (author)

  14. Integral Fast Reactor Program annual progress report, FY 1991

    International Nuclear Information System (INIS)

    1992-06-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1991. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R ampersand D

  15. Integral Fast Reactor Program. Annual progress report, FY 1993

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.; Walters, L.C.; Laidler, J.J.; Pedersen, D.R.; Wade, D.C.; Lineberry, M.J.

    1994-10-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1993. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R and D.

  16. Integral Fast Reactor Program. Annual progress report, FY 1993

    International Nuclear Information System (INIS)

    Chang, Y.I.; Walters, L.C.; Laidler, J.J.; Pedersen, D.R.; Wade, D.C.; Lineberry, M.J.

    1994-10-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1993. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R and D

  17. Performance study of sugar-yeast-ethanol bio-hybrid fuel cells

    Science.gov (United States)

    Jahnke, Justin P.; Mackie, David M.; Benyamin, Marcus; Ganguli, Rahul; Sumner, James J.

    2015-05-01

    Renewable alternatives to fossil hydrocarbons for energy generation are of general interest for a variety of political, economic, environmental, and practical reasons. In particular, energy from biomass has many advantages, including safety, sustainability, and the ability to be scavenged from native ecosystems or from waste streams. Microbial fuel cells (MFCs) can take advantage of microorganism metabolism to efficiently use sugar and other biomolecules as fuel, but are limited by low power densities. In contrast, direct alcohol fuel cells (DAFCs) take advantage of proton exchange membranes (PEMs) to generate electricity from alcohols at much higher power densities. Here, we investigate a novel bio-hybrid fuel cell design prepared using commercial off-the-shelf DAFCs. In the bio-hybrid fuel cells, biomass such as sugar is fermented by yeast to ethanol, which can be used to fuel a DAFC. A separation membrane between the fermentation and the DAFC is used to purify the fermentate while avoiding any parasitic power losses. However, shifting the DAFCs from pure alcohol-water solutions to filtered fermented media introduces complications related to how the starting materials, fermentation byproducts, and DAFC waste products affect both the fermentation and the long-term DAFC performance. This study examines the impact of separation membrane pore size, fermentation/fuel cell byproducts, alcohol and salt concentrations, and load resistance on fuel cell performance. Under optimized conditions, the performance obtained is comparable to that of a similar DAFC run with a pure alcohol-water mixture. Additionally, the modified DAFC can provide useable amounts of power for weeks.

  18. An introduction to current modeling techniques in nuclear fuel performance analysis

    International Nuclear Information System (INIS)

    Faya, S.C.S.

    1981-10-01

    An overview of several important phenomena considered in fuel performance analysis, such as fuel restructuring, fuel swelling, fission gas release, pellet cracking, etc... are provided. All these phenomena are independent. For the purpose of illustration of integral fuel rod analysis the LIFE-1 computer code is discussed. (Author) [pt

  19. Influence of the fuel and dosage on the performance of double-compartment microbial fuel cells.

    Science.gov (United States)

    Asensio, Y; Fernandez-Marchante, C M; Lobato, J; Cañizares, P; Rodrigo, M A

    2016-08-01

    This manuscript focuses on the evaluation of the use of different types and dosages of fuels in the performance of double-compartment microbial fuel cell equipped with carbon felt electrodes and cationic membrane. Five types of fuels (ethanol, glycerol, acetate, propionate and fructose) have been tested for the same organic load (5,000 mg L(-1) measured as COD) and for one of them (acetate), the range of dosages between 500 and 20,000 mg L(-1) of COD was also studied. Results demonstrate that production of electricity depends strongly on the fuel used. Carboxylic acids are much more efficient than alcohols or fructose for the same organic load and within the range 500-5,000 mg L(-1) of acetate the production of electricity increases linearly with the amount of acetate fed but over these concentrations a change in the population composition may explain a worse performance. Copyright © 2016 Elsevier Ltd. All rights reserved.

  20. In-reactor performance of methods to control fuel-cladding chemical interaction

    International Nuclear Information System (INIS)

    Weber, E.T.; Gibby, R.L.; Wilson, C.N.; Lawrence, L.A.; Adamson, M.G.

    1979-01-01

    Inner surface corrosion of austenitic stainless steel cladding by oxygen and reactive fission product elements requires a 50 μm wastage allowance in current FBR reference oxide fuel pin design. Elimination or reduction of this wastage allowance could result in better reactor efficiency and economics through improvements in fuel pin performance and reliability. Reduction in cladding thickness and replacement of equivalent volume with fuel result in improved breeding capability. Of the factors affecting fuel-cladding chemical interaction (FCCI), oxygen activity within the fuel pin can be most readily controlled and/or manipulated without degrading fuel pin performance or significantly increasing fuel fabrication costs. There are two major approaches to control oxygen activity within an oxide fuel pin: (1) control of total oxygen inventory and chemical activity (Δ anti GO 2 ) by use of low oxygen-to-metal ratio (O/M) fuel; and (2) incorporation of a material within the fuel pin to provide in-situ control of oxygen activity (Δ anti GO 2 ) and fixation of excess oxygen prior to, or in preference to reaction with the cladding. The paper describes irradiation tests which were conducted in EBR-II and GETR incorporating oxygen buffer/getter materials and very low O/M fuel to control oxygen activity in sealed fuel pins

  1. Development of Out-pile Test Technology for Fuel Assembly Performance Verification

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Tae Hyun; In, W. K.; Oh, D. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)] (and others)

    2007-03-15

    Out-pile tests with full scale fuel assembly are to verify the design and to evaluate the performance of the final products. HTL for the hydraulic tests and FAMeCT for mechanical/structural tests were constructed in this project. The maximum operating conditions of HTL are 30 bar, 320 .deg. C, and 500 m3/hr. This facility can perform the pressure drop test, fuel assembly uplift test, and flow induced vibration test. FAMeCT can perform the bending and vibration tests. The verification of the developed facilities were carried out by comparing the reference data of the fuel assembly which was obtained at the Westinghouse Co. The compared data showed a good coincidence within uncertainties. FRETONUS was developed for high temperature and high pressure fretting wear simulator and performance test. A performance test was conducted for 500 hours to check the integrity, endurance, data acquisition capability of the simulator. The technology of turbulent flow analysis and finite element analysis by computation was developed. From the establishments of out-pile test facilities for full scale fuel assembly, the domestic infrastructure for PWR fuel development has been greatly upgraded.

  2. Performance of PEM Liquid-Feed Direct Methanol-Air Fuel Cells

    Science.gov (United States)

    Narayanan, S. R.

    1995-01-01

    A direct methanol-air fuel cell operating at near atmospheric pressure, low-flow rate air, and at temperatures close to 60oC would tremendously enlarge the scope of potential applications. While earlier studies have reported performance with oxygen, the present study focuses on characterizing the performance of a PEM liquid feed direct methanol-air cell consisting of components developed in house. These cells employ Pt-Ru catalyst in the anode, Pt at the cathode and Nafion 117 as the PEM. The effect of pressure, flow rate of air and temperature on cell performance has been studied. With air, the performance level is as high as 0.437 V at 300 mA/cm2 (90oC, 20 psig, and excess air flow) has been attained. Even more significant is the performance level at 60oC, 1 atm and low flow rates of air (3-5 times stoichiometric), which is 0.4 V at 150 mA/cm2. Individual electrode potentials for the methanol and air electrode have been separated and analyzed. Fuel crossover rates and the impact of fuel crossover on the performance of the air electrode have also been measured. The study identifies issues specific to the methanol-air fuel cell and provides a basis for improvement strategies.

  3. Calculation study of the WWER-440 fuel performance for extended burnup

    International Nuclear Information System (INIS)

    Kujal, J.; Pazdera, F.; Barta, O.

    1984-01-01

    The results of preliminary calculational study of extended burnup cycling schemes impact on WWER-440 fuel performance are presented. Two high burnup schemes were proposed with three and four cycles, resp. Comparison was made with three cycle reference case. The thermal mechanical analysis was performed with PIN and RELA codes. The values of rod internal pressure, fuel centerline temperatures and fuel-cladding gap are expressed as function of power history. (author)

  4. PWR fuel performance and future trend in Japan

    International Nuclear Information System (INIS)

    Kondo, Y.

    1987-01-01

    Since the first PWR power plant Mihama Unit 1 initiated its commercial operation in 1970, Japanese utilities and manufacturers have expended much of their resources and efforts to improve PWR technology. The results are already seen in significantly improved performance of 16 PWR plants now in operation. Mitsubishi Heavy Industries Ltd. (MHI) has been supplying them with nuclear fuel assemblies, which are over 5700. As the reliability of the current design fuel has been achieved, the direction of R and D on nuclear fuel has changed to make nuclear power more competitive to the other power generation methods. The most important R and D targets are the burnup extension, Gd contained fuel, Pu utilizatoin and the load follow capacility. (author)

  5. Solid oxide fuel cell performance comparison fueled by methane, MeOH, EtOH and gasoline surrogate C_8H_1_8

    International Nuclear Information System (INIS)

    Liso, Vincenzo; Cinti, Giovanni; Nielsen, Mads P.; Desideri, Umberto

    2016-01-01

    Carbon deposition is a major cause of degradation in solid oxide fuel cell systems. The ability to predict carbon formation in reforming processes is thus absolutely necessary for stable operation of solid oxide fuel cell systems. In the open literature it is found that the steam input is always considered in large excess compared to what required by the reforming process with the purpose of reducing carbon formation and avoiding rapid degradation of the cell performance. This makes it difficult to consistently compare system performance with different fuels. In this work, the molar compositions at equilibrium are calculated for a minimum steam to carbon ratio for each fuel type. We carry out a thermodynamic analysis of fuel/steam system using Gibbs Free Energy minimization method. A mathematical relationship between Lagrange's multipliers and carbon activity in the gas phase was deduced. Minimum steam required for the reforming process for each fuel was related to the heat required for the reforming process and fuel cell open circuit voltage. Furthermore, in an experimental test, steam reforming product compositions were used to evaluate and compare SOFC performance with different hydrocarbons. Comparing the model to the experimental activity, it is revealed that at temperatures exceeding 800 °C the gas composition is dominated by hydrogen and carbon monoxide for any of the fuels considered leading to similar cell polarization curves performance for different fuels. The main effect on the performance is related to OCV values which are dependent on different steam content for each fuel. It was concluded that the magnitude of the heat requested for the fuel reforming process is the major difference in system performance when comparing different fuels. However, reforming kinetic effects can become predominant rather than thermodynamics, especially at lower temperatures.

  6. Transient performance simulation of aircraft engine integrated with fuel and control systems

    International Nuclear Information System (INIS)

    Wang, C.; Li, Y.G.; Yang, B.Y.

    2017-01-01

    Highlights: • A new performance simulation method for engine hydraulic fuel systems is introduced. • Time delay of engine performance due to fuel system model is noticeable but small. • The method provides details of fuel system behavior in engine transient processes. • The method could be used to support engine and fuel system designs. - Abstract: A new method for the simulation of gas turbine fuel systems based on an inter-component volume method has been developed. It is able to simulate the performance of each of the hydraulic components of a fuel system using physics-based models, which potentially offers more accurate results compared with those using transfer functions. A transient performance simulation system has been set up for gas turbine engines based on an inter-component volume (ICV) method. A proportional-integral (PI) control strategy is used for the simulation of engine controller. An integrated engine and its control and hydraulic fuel systems has been set up to investigate their coupling effect during engine transient processes. The developed simulation system has been applied to a model aero engine. The results show that the delay of the engine transient response due to the inclusion of the fuel system model is noticeable although relatively small. The developed method is generic and can be applied to any other gas turbines and their control and fuel systems.

  7. Tradeoffs in fuel cycle performance for most promising options - 15346

    International Nuclear Information System (INIS)

    Taiwo, T.; Kim, T.K.; Feng, B.; Stauff, N.; Hoffman, E.; Ganda, F.; Todosow, M.; Brown, N.; Raitses, G.; Gehin, J.; Powers, J.; Youinou, G.; Hiruta, H.; Wigeland, R.

    2015-01-01

    A recent Evaluation and Screening (E/S) study of nuclear fuel cycle options was conducted by grouping all potential options into 40 Evaluation Groups (EGs) based on similarities in fundamental physics characteristics and fuel cycle performance. Through a rigorous evaluation process considering benefit and challenge metrics, 4 of these EGs were identified by the E/S study as 'most promising'. All 4 involve continuous recycle of U/Pu or U/TRU with natural uranium feed in fast critical reactors. However, these most promising EGs also include fuel cycle groups with variations on feed materials, neutron spectra, and reactor criticality. Therefore, the impacts of the addition of natural thorium fuel feed to a system that originally only used natural uranium fuel feed, using an intermediate spectrum instead of a fast spectrum, and using externally-driven systems versus critical reactors were evaluated. It was found that adding thorium to the natural uranium feed mixture leads to lower burnup, higher mass flows, and degrades fuel cycle benefit metrics (waste management, resource utilization, etc.) for fuel cycles that continuously recycle U/Pu or U/TRU. Adding thorium results in fissions of 233 U instead of just 239 Pu and in turn results in a lower average number of neutrons produced per absorption (η) for the fast reactor system. For continuous recycling systems, the lower η results in lower excess reactivity and subsequently lower achievable fuel burnup. This in turn leads to higher mass flows (fabrication, reprocessing, disposal, etc.) to produce a given amount of energy and subsequent lower metrics performance. The investigated fuel cycle options with intermediate spectrum reactors also exhibited degraded performance in the benefit metrics compared to fast spectrum reactors. Similarly, this is due to lower η values as the spectrum softens. The best externally-driven systems exhibited similar performance as fast critical reactors in terms of mass flows

  8. Performance enhancement of a spark ignition engine fed by different fuel types

    International Nuclear Information System (INIS)

    Hedfi, Hachem; Jbara, Abdessalem; Jedli, Hedi; Slimi, Khalifa; Stoppato, Anna

    2016-01-01

    Highlights: • Biogas mixed with hydrogen is checked for a spark ignition engine. • An engine fed by biogas, hydrogen, natural gas or liquid petroleum gas is studied. • Efficiency is optimized with respect to consumption and exhaust gas recirculation. • Combustion reaction progress is characterized in real time. - Abstract: A numerical model based on thermodynamic and kinetic analyses has been established in order to evaluate biogas, hydrogen, natural gas or liquid petroleum gas as fuels in a spark ignition engine. For each fuel type, consumption as well as efficiency have been compared to gasoline in order to generate the same engine work (in the range of 0.28–0.43 W h/cycle). It was found that the spark ignition engine can be fed by an equimolar mixture of biogas and hydrogen. Moreover, thermal efficiency has been enhanced with respect to fuel consumption and exhaust gas recirculation (EGR). It was shown that an equimolar mixture between biogas and hydrogen increases the ITE by around 2.2% and decreases the mass consumption by less than 0.01 g/cycle. In addition, the combustion reaction progresses as well as CO and CO_2 emissions have been characterized in real time.

  9. Relations between Air-Fuel Ratio and Dynamic Performance of Small Race Cars

    OpenAIRE

    位田, 晴良; Ida, Haruyoshi; 漁, 佑一郎; Sunadori, Yuichiro; 牧田, 俊太郎; Makita, Syuntaro; 宮﨑, 真央; Miyazaki, Manaka; 磯松, 弥司; Isomatsu, Yatsuka

    2017-01-01

    'It goes without saying that engine output power characteristics greatly affect the dynamic performance of the race car. One of the methods of changing the output power of the engine is to adjust the set amount of fuel supply. This method changes the air-fuel ratio of the air fuel mixture supplied to the engine. In this study, a slalom test run of a small race car was used to examine dynamic performance with attention to the air-fuel ratio changed by adjusting the set amount of fuel supply. T...

  10. Performance Evaluation of a High Bandwidth Liquid Fuel Modulation Valve for Active Combustion Control

    Science.gov (United States)

    Saus, Joseph R.; DeLaat, John C.; Chang, Clarence T.; Vrnak, Daniel R.

    2012-01-01

    At the NASA Glenn Research Center, a characterization rig was designed and constructed for the purpose of evaluating high bandwidth liquid fuel modulation devices to determine their suitability for active combustion control research. Incorporated into the rig s design are features that approximate conditions similar to those that would be encountered by a candidate device if it were installed on an actual combustion research rig. The characterized dynamic performance measures obtained through testing in the rig are planned to be accurate indicators of expected performance in an actual combustion testing environment. To evaluate how well the characterization rig predicts fuel modulator dynamic performance, characterization rig data was compared with performance data for a fuel modulator candidate when the candidate was in operation during combustion testing. Specifically, the nominal and off-nominal performance data for a magnetostrictive-actuated proportional fuel modulation valve is described. Valve performance data were collected with the characterization rig configured to emulate two different combustion rig fuel feed systems. Fuel mass flows and pressures, fuel feed line lengths, and fuel injector orifice size was approximated in the characterization rig. Valve performance data were also collected with the valve modulating the fuel into the two combustor rigs. Comparison of the predicted and actual valve performance data show that when the valve is operated near its design condition the characterization rig can appropriately predict the installed performance of the valve. Improvements to the characterization rig and accompanying modeling activities are underway to more accurately predict performance, especially for the devices under development to modulate fuel into the much smaller fuel injectors anticipated in future lean-burning low-emissions aircraft engine combustors.

  11. Performance evaluation of UO2-Zr fuel in power ramp tests

    International Nuclear Information System (INIS)

    Knudsen, P.; Bagger, C.

    1977-01-01

    In power reactors using UO 2 -Zr fuel, rapid power increases may lead to failures in fuel pins that have been irradiated at steady or decreasing heat loads. This paper presents results which extend the experience with power ramp performance of high burn-up fuel pins. A test fuel element containing both pellet and vipac UO 2 -Zr fuel pins was irradiated in the HBWR at Halden for effectively 2 1/2 years to an average burn-up of 21,000 MWD/te UO 2 at gradually decreasing power levels. The subsequent non-destructive characterization revealed formation of transverse cracks in the vipac fuel columns. After the HBWR irradiation, five of the fuel pins were power ramp tested individually in the DR 3 Reactor at Riso. The ramp rates in this test series were in the range 3-60 W/cm min. The maximum local heat loads seen in the ramp tests were 20-120% above the highest levels experienced at the same axial positions during the HBWR irradiation. Three pellets and one vipac fuel pin failed, whereas another vipac pin gave no indication of clad penetration. Profilometry after the ramp testing indicated the formation of small ridges for both types of fuel pins. For vipac fuel, the ridges were less regularly distributed along the pin length than for pellet fuel. Neutron radiography revealed the formation of additional transverse and longitudinal fuel cracks during the power ramps for both types of fuel pins. The observed failures seemed to be marginal since little or no indication as to the locations of the clad penetrations could be derived from the non-destructive post-irradiation examinations. The cases have been analyzed by means of the Danish fuel performance codes. The calculations, which are in general agreement with the observations, are discussed. The results of the investigations indicate qualitative similarities in over power performance of the two fuel types

  12. Joyo progress report, vol. 8

    International Nuclear Information System (INIS)

    1983-01-01

    Following Joyo Reactor Technology Progress Reports (Vol. 1 to Vol. 7), the name was changed to Joyo Progress Report from this volume, and the activities concerning the fast breeder experimental reactor Joyo as a whole are to be reported as quarterly report. In the fast breeder experimental reactor Joyo, the change to the core for irradiation (MK-2) from the core for breeding (MK-1) was carried out since January, 1982, in order to utilize the reactor as an irradiation facility for the development of fuel and materials. The main work was the construction of the core for irradiation by exchanging 290 fuel elements, and the exchange of upper and lower guide pipes for control rods, the reconstruction of the driving mechanism, the installation of standby neutron detector system, the acceptance and inspection of new fuel, and the transfer of spent fuel between pools were carried out. As scheduled, the core for irradiation attained the initial criticality on November 22, and the works of constructing the core were completed on December 23, 1982. Thereafter, the 100 MW performance test was begun. Various experience and valuable data were obtained in the regular inspection and the maintenance and repair works carried out at the same time, regarding the operation and maintenance of the Joyo facilities. (Kako, I.)

  13. Operating performance and reliability of CANDU PHWR fuel channels in Canada

    International Nuclear Information System (INIS)

    Strachan, B.; Brown, D.R.

    1983-03-01

    CANDU nuclear plants use many small-diameter high-pressure fuel channels. Good operating performance from the CANDU fuel channels has made a major contribution to the world-leading operating record of the CANDU nuclear power plants. As of 1982 December 31, there were 7,480 fuel channels installed in 18 CANDU reactors over 500 MW(e) in size. Eight of these reactors have been declared in-service and have accumulated 24,000 fuel channel-years of operation. The only significant operating problems with fuel channels have been the occurrence of leaking cracks in 70 fuel channels and a larger amount of axial creep on the early reactors than was originally provided for in the design. Both of these problems have been corrected on all CANDU reactors built since the Bruce GS 'A' station and the newer reactors should exhibit even better performance

  14. Design characteristics of metallic fuel rod on its in-LMR performance

    International Nuclear Information System (INIS)

    Hwang, Woan; Kang Hee Young; Nam, Cheol; Kim, Jong Oh

    1997-01-01

    Fuel design is a key feature to assure LMR safety goals. To date, a large effort had been devoted to develop metallic fuels at ANL's experimental breeder reactor (EBR-II). The major design and performance parameters investigated include; thermal conductivity and temperature profile; smear density; axial plenum; FCMI and cladding deformation including creep, and fission gas release. In order to evaluate the sensitivity of each parameter, in-LMR performances of metallic fuels are not only reviewed by the experiment results in literatures, but also key design characteristics according to the variation of metallic fuel rod design parameters are analyzed by using the MACSIS code which simulates in-reactor behaviors of metal fuel rod. In this study, key design characteristics and the criteria which must be considered to design fuel rod in LMR, are proposed and discussed. (author). 14 refs., 4 figs

  15. Evaluation of fuel performance with different enrichment degrees for an experimental device

    International Nuclear Information System (INIS)

    Giovedi, Claudia; Pino, Eddy S.; Gomes, Daniel S.; Abe, Alfredo Y.; Silva, Antonio Teixeira e

    2013-01-01

    Evaluation of fuel performance is conventionally carried out using specific codes developed to this aim. The obtained data are confirmed by experimental measurements performed using devices, which are located inside research reactors, projected to simulate reactor conditions under normal operation. Due to the limitations of the available reactor core length for irradiation in research reactors core, fuel rods used to obtain experimental data must present the same characteristics of the real fuel rod, but with a shorter length. Then, in order to compare the obtained results to the expected behavior of the real fuel rod, the experimental fuel rod should be designed with a free volume to fuel volume ratio very closed to the one of the full scale fuel rod. The aim of this paper is to evaluate some parameters and aspects related to the fuel rod behavior in a rod applied to the experimental irradiation device called Nuclear Fuel Irradiation Circuit (CAFE-Mod1) considering two fuel enrichment degrees: a typical commercial PWR enrichment and a value about 4 times higher. This evaluation is carried out by means of an adapted fuel performance code. Some of the parameter evaluated were fuel temperature and fission gas release as function of the fuel enrichment level. The results obtained in this paper were very similar to the ones previously obtained without consider similar free volume between the experimental and the full length fuel rod, regardless of low increases observed for the internal rod pressure and the amount of fission gas released. (author)

  16. Romanian nuclear fuel fabrication and in-reactor fuel operational experience

    International Nuclear Information System (INIS)

    Budan, O.

    2003-01-01

    A review of the Romanian nuclear program since mid 60's is made. After 1990, the new Romanian nuclear power authority, RENEL-GEN, elaborated a realistic Nuclear Fuel Program. This program went through the Romanian nuclear fuel plant qualification with the Canadian (AECL and ZPI) support, restarting in January 1995 of the industrial nuclear fuel production, quality evaluation of the fuel produced before 1990 and the recovery of this fuel. This new policy produced good results. FCN is since 1995 the only CANDU fuel supplier from outside Canada recognised by AECL as an authorised CANDU fuel manufacturer. The in-reactor performances and behaviour of the fuel manufactured by FCN after its qualification have been excellent. Very low - more then five times lesser than the design value - fuel defect rate has been recorded up to now and the average discharge of this fuel was with about 9% greater than the design value. Since mid 1998 when SNN took charge of the production of nuclear generated electricity, FCN made significant progresses in development and procurement of new and more efficient equipment and is now very close to double its fuel production capacity. After the completion of the recovery of the fuel produced before June 1990, FCN is already prepared to shift its fuel production to the so-called 'heavy' bundle containing about 19.3 kg of Uranium per bundle

  17. Performance of the Westinghouse WWER-1000 fuel design

    International Nuclear Information System (INIS)

    Höglund, J.; Jansson, A.; Latorre, R.; Davis, D.

    2015-01-01

    In 2005, six (6) Westinghouse WWER-1000 Lead Test Assemblies (LTAs) were loaded in South Ukraine Unit 3 (SU3). The LTAs completed the planned four cycles of operation and reached an average assembly burnup in excess of 43 MWd/ kgU. Post Irradiation Examination (PIE) inspections were performed after completion of each cycle and it was concluded that the 6 Westinghouse LTAs performed as expected during their operational regimes. In 2010, a full region of 42 assemblies of an enhanced WWER-1000 fuel design for Ukrainian reactors, designated WFA, was loaded in SU3. The WFA includes features that further mitigate assembly bow while at the same time improving the fuel cycle economy. In 2015, 26 WFAs completed their planned four cycles of operation reaching an average assembly burnup in excess of 42 MWd/ kgU. Currently 36 WFAs continue operating their fourth cycle in SU3. In addition, South Ukraine Unit 2 (SU2) has been loaded with WFAs and 27 assemblies have completed two cycles of operation reaching an average assembly burnup above 24 MWd/kgU. PIE for the WFAs has been completed after each cycle of operation. All assemblies have been examined for visible damage or non-standard position of fuel assembly components during unloading and reloading. All WFAs have also been subject to the standard leak testing process, with all fuel rods found to be hermetically sealed and non-leaking. Each outage, six WFAs have been subject to a more extensive inspection program. In 2012, 2013, and 2015, the Westinghouse Fuel Inspection and Repair Equipment (FIRE) workstation were used for the SU3 inspections. Excellent irradiation fuel performance has been observed and measured on all WFAs. The fuel assembly growth, rod cluster control assembly (RCCA) drag forces, oxide thickness, total fuel rod-to-nozzle gap channel closure, and fuel assembly bow data were within the bounds of the Westinghouse experience database. Results and concluding remarks from the PIEs are provided in this paper. In

  18. Metal fuel safety performance

    International Nuclear Information System (INIS)

    Miles, K.J. Jr.; Tentner, A.M.

    1988-01-01

    The current development of breeder reactor systems has lead to the renewed interest in metal fuels as the driver material. Modeling efforts were begun to provide a mechanistic description of the metal fuel during anticipated and hypothetical transients within the context of the SAS4A accident analysis code system. Through validation exercises using experimental results of metal fuel TREAT tests, confidence is being developed on the nature and accuracy of the modeling and implementation. Prefailure characterization, transient pin response, margins to failure, axial in-pin fuel relocation prior to cladding breach, and molten fuel relocation after cladding breach are considered. Transient time scales ranging from milliseconds to many hours can be studied with all the reactivity feedbacks evaluated

  19. Hydrogen-enriched fuels

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-08-01

    NRG Technologies, Inc. is attempting to develop hardware and infrastructure that will allow mixtures of hydrogen and conventional fuels to become viable alternatives to conventional fuels alone. This commercialization can be successful if the authors are able to achieve exhaust emission levels of less than 0.03 g/kw-hr NOx and CO; and 0.15 g/kw-hr NMHC at full engine power without the use of exhaust catalysts. The major barriers to achieving these goals are that the lean burn regimes required to meet exhaust emissions goals reduce engine output substantially and tend to exhibit higher-than-normal total hydrocarbon emissions. Also, hydrogen addition to conventional fuels increases fuel cost, and reduces both vehicle range and engine output power. Maintaining low emissions during transient driving cycles has not been demonstrated. A three year test plan has been developed to perform the investigations into the issues described above. During this initial year of funding research has progressed in the following areas: (a) a cost effective single-cylinder research platform was constructed; (b) exhaust gas speciation was performed to characterize the nature of hydrocarbon emissions from hydrogen-enriched natural gas fuels; (c) three H{sub 2}/CH{sub 4} fuel compositions were analyzed using spark timing and equivalence ratio sweeping procedures and finally; (d) a full size pick-up truck platform was converted to run on HCNG fuels. The testing performed in year one of the three year plan represents a baseline from which to assess options for overcoming the stated barriers to success.

  20. Hydrocarbon Fuel Thermal Performance Modeling based on Systematic Measurement and Comprehensive Chromatographic Analysis

    Science.gov (United States)

    2016-07-31

    distribution unlimited Hydrocarbon Fuel Thermal Performance Modeling based on Systematic Measurement and Comprehensive Chromatographic Analysis Matthew...vital importance for hydrocarbon -fueled propulsion systems: fuel thermal performance as indicated by physical and chemical effects of cooling passage... analysis . The selection and acquisition of a set of chemically diverse fuels is pivotal for a successful outcome since test method validation and

  1. Highlights from the IAEA coordinated research programme on fuel performance and fission product data

    International Nuclear Information System (INIS)

    Nabielek, H.; Schenk, W.; Verfondern, K.

    1996-01-01

    Seven countries are cooperating with the objectives (i) to document the status of the experimental data base and of the predictive methods for Gas-Cooled Reactor fuel performance and fission product behaviour; (ii) to verify and validate methods in fuel performance and fission product retention prediction. These countries are China, France, Germany, Japan, Russia, USA and the UK. Duration of the programme is 1993-96. The technology areas addressed in this IAEA Coordinated Research Programme are: Fuel design and manufacture, Normal operation fuel performance and fission product behaviour, Accident condition fuel performance and fission product behaviour, -core heatup, -fast transients, -oxidising conditions (water and air ingress), Plateout, re-entrainment of plateout, fission product behaviour in the reactor building, and Performance of advanced fuels. Work performed so far has generated a 300-page draft document with important information for normal operations (Germany, Japan, China, Russia) and accident conditions (USA, Japan, Germany, Russia) and, additionally, a special chapter on advanced fuels (Japan). (author)

  2. 11. International conference on WWER fuel performance, modelling and experimental support. Proceedings

    International Nuclear Information System (INIS)

    Manolova, M.; Boneva, S.; Mitev, M.

    2015-01-01

    This publication is a compilation of the papers presented in 11th International Conference on WWER Fuel Performance, Modeling and Experimental Support, organized by the Institute for Nuclear Research and Nuclear Energy (INRNE) of the Bulgarian Academy of Sciences in co-operation with the International Atomic Energy Agency (IAEA), Vienna, Austria, supported by the Kozloduy Nuclear Power Plant (KNPP), the Bulgarian Nuclear Regulatory Agency, and TVEL Fuel Company, Russia. The Conference took place in hotel Bolero, Golden Sands Resort, Bulgaria, from 26 September 2015 to 3 October 2015. It was attended by 117 participants, among them more than 100 experts and specialists from 22 countries, including representatives of 3 international organizations, 16 Russian organizations and other 36 foreign institutes, nuclear fuel plants, nuclear power plants and organizations responsible for WWER and PWR fuel design, manufacturing and research, and 3 Bulgarian organizations, working for the Bulgarian nuclear industry. 70 papers have been presented in the Conference in 6 oral and 1 poster session, covering: (1) general overview lectures; (2) fuel performance and operational experience; (3) fuel modeling and experimental support; (4) fuel safety and QA; (5) spent fuel performance and management; (6) specific issues of WWER-1000 fuel reliability. The proceedings provide Summary, Conclusions and Recommendations of the Conference, together with the full text of the presentations. IAEA Technical Meeting (TM) “Achieving zero fuel failure rates: challenges and perspectives”, 1 – 2 October 2015 was organized in conjunction with the 11th International Conference on WWER Fuel Performance, Modelling and Experimental Support. The reports presented on TM sessions are included in the Conference Proceedings too

  3. Study on the micro direct ethanol fuel cell (Micro-DEFC) performance

    Science.gov (United States)

    Saisirirat, Penyarat; Joommanee, Bordindech

    2018-01-01

    The direct ethanol fuel cell (DEFC) is selected for this research. DEFC uses ethanol in the fuel cell instead of the more toxic methanol. Ethanol is more attractive than methanol by many reasons. Ethanol is a hydrogen-rich liquid and it has a higher specific energy (8.0 kWh/kg) compared to that of methanol (6.1 kWh/kg). Ethanol can be obtained in great quantity from biomass through a fermentation process from renewable resources such as sugar cane, wheat, corn, and even straw. The use of ethanol would also overcome both the storage and infrastructure challenge of hydrogen for fuel cell applications. The experimental apparatus on the micro direct ethanol fuel cell for measuring the cell performance has been set for this research. The objective is to study the micro direct ethanol fuel cell performance for applying with the portable electronic devices. The cell performance is specified in the terms of cell voltage, cell current and power of the cell at room operating temperature and 1 atm for the pressure and also includes the ethanol fuel consumption. The effect of operating temperature change on the electrical production performance is also studied. The steady-state time for collecting each data value is about 5-10 minutes. The results show that with the increase of concentrations of ethanol by volume, the reactant concentration at the reaction sites increases so the electrochemical rate also increases but when it reaches the saturated point the performance gradually drops.

  4. Hybrid Direct Carbon Fuel Cell Performance with Anode Current Collector Material

    DEFF Research Database (Denmark)

    Deleebeeck, Lisa; Kammer Hansen, Kent

    2015-01-01

    collectors were studied: Au, Ni, Ag, and Pt. It was shown that the performance of the direct carbon fuel cell (DCFC) is dependent on the current collector materials, Ni and Pt giving the best performance, due to their catalytic activity. Gold is suggested to be the best material as an inert current collector......The influence of the current collector on the performance of a hybrid direct carbon fuel cell (HDCFC), consisting of solid oxide fuel cell (SOFC) with a molten carbonate-carbon slurry in contact with the anode, has been investigated using current-voltage curves. Four different anode current...

  5. Proceedings of American Nuclear Society topical meeting on light water reactor fuel performance

    International Nuclear Information System (INIS)

    1979-01-01

    Separate abstracts are included for 32 of the 39 papers presented concerning fuel performance; fuel performance under transient and off-normal operations; fuel-cladding interactions; and fission product release. Seven papers have previously been input into the data base

  6. Shielding Performance Measurements of Spent Fuel Transportation Container

    Directory of Open Access Journals (Sweden)

    SUN Hong-chao

    2015-11-01

    Full Text Available The safety supervision of radioactive material transportation package has been further stressed and implemented. The shielding performance measurements of spent fuel transport container is the important content of supervision. However, some of the problems and difficulties reflected in practice need to be solved, such as the neutron dose rate on the surface of package is too difficult to measure exactly, the monitoring results are not always reliable, etc. The monitoring results using different spectrometers were compared and the simulation results of MCNP runs were considered. An improvement was provided to the shielding performance measurements technique and management of spent fuel transport.

  7. 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....

  8. Fuel performance in water storage

    International Nuclear Information System (INIS)

    Hoskins, A.P.; Scott, J.G.; Shelton-Davis, C.V.; McDannel, G.E.

    1993-11-01

    Westinghouse Idaho Nuclear Company operates the Idaho Chemical Processing Plant (ICPP) at the Idaho National Engineering Laboratory (INEL) for the Department of Energy (DOE). A variety of different types of fuels have been stored there since the 1950's prior to reprocessing for uranium recovery. In April of 1992, the DOE decided to end fuel reprocessing, changing the mission at ICPP. Fuel integrity in storage is now viewed as long term until final disposition is defined and implemented. Thus, the condition of fuel and storage equipment is being closely monitored and evaluated to ensure continued safe storage. There are four main areas of fuel storage at ICPP: an original underwater storage facility (CPP-603), a modern underwater storage facility (CPP-666), and two dry fuel storage facilities. The fuels in storage are from the US Navy, DOE (and its predecessors the Energy Research and Development Administration and the Atomic Energy Commission), and other research programs. Fuel matrices include uranium oxide, hydride, carbide, metal, and alloy fuels. In the underwater storage basins, fuels are clad with stainless steel, zirconium, and aluminum. Also included in the basin inventory is canned scrap material. The dry fuel storage contains primarily graphite and aluminum type fuels. A total of 55 different fuel types are currently stored at the Idaho Chemical Processing Plant. The corrosion resistance of the barrier material is of primary concern in evaluating the integrity of the fuel in long term water storage. The barrier material is either the fuel cladding (if not canned) or the can material

  9. Performance and emissions of a supercharged dual-fuel engine fueled by hydrogen-rich coke oven gas

    Energy Technology Data Exchange (ETDEWEB)

    Roy, M.M.; Tomita, E.; Kawahara, N.; Harada, Y.; Sakane, A. [Okayama University, Okayama (Japan). Dept. of Mechanical Engineering

    2009-12-15

    This study investigated the engine performance and emissions of a supercharged dual-fuel engine fueled by hydrogen-rich coke oven gas and ignited by a pilot amount of diesel fuel. The engine was tested for use as a cogeneration engine, so power output while maintaining a reasonable thermal efficiency was important. Experiments were carried out at a constant pilot injection pressure and pilot quantity for different fuel-air equivalence ratios and at various injection timings without and with exhaust gas recirculation (EGR). The experimental strategy was to optimize the injection timing to maximize engine power at different fuel-air equivalence ratios without knocking and within the limit of the maximum cylinder pressure. The engine was tested first without EGR condition up to the maximum possible fuel-air equivalence ratio of 0.65. A maximum indicated mean effective pressure (IMEP) of 1425 kPa and a thermal efficiency of 39% were obtained. However, the nitrogen oxides (NOx) emissions were high. A simulated EGR up to 50% was then performed to obtain lower NOx emissions. The maximum reduction of NOx was 60% or more maintaining the similar levels of IMEP and thermal efficiency. Two-stage combustion was obtained; this is an indicator of maximum power output conditions and a precursor of knocking combustion.

  10. A fuel performance code TRUST VIc and its validation

    Energy Technology Data Exchange (ETDEWEB)

    Ishida, M; Kogai, T [Nippon Nuclear Fuel Development Co. Ltd., Oarai, Ibaraki (Japan)

    1997-08-01

    This paper describes a fuel performance code TRUST V1c developed to analyze thermal and mechanical behavior of LWR fuel rod. Submodels in the code include FP gas models depicting gaseous swelling, gas release from pellet and axial gas mixing. The code has FEM-based structure to handle interaction between thermal and mechanical submodels brought by the gas models. The code is validated against irradiation data of fuel centerline temperature, FGR, pellet porosity and cladding deformation. (author). 9 refs, 8 figs.

  11. A fuel performance code TRUST VIc and its validation

    International Nuclear Information System (INIS)

    Ishida, M.; Kogai, T.

    1997-01-01

    This paper describes a fuel performance code TRUST V1c developed to analyze thermal and mechanical behavior of LWR fuel rod. Submodels in the code include FP gas models depicting gaseous swelling, gas release from pellet and axial gas mixing. The code has FEM-based structure to handle interaction between thermal and mechanical submodels brought by the gas models. The code is validated against irradiation data of fuel centerline temperature, FGR, pellet porosity and cladding deformation. (author). 9 refs, 8 figs

  12. Performance of U-Pu-Zr fuel cast into zirconium molds

    International Nuclear Information System (INIS)

    Crawford, D.C.; Lahm, C.E.; Tsai, H.

    1992-01-01

    Current fabrication techniques for the integral fast reactor (IFR) fuel utilize injection casting into quartz molds after reprocessing in the IFR fuel cycle facility. The quartz molds are destroyed during the fuel demolding process, and the quartz residue must therefore be treated as contaminated waste. Alternatively, if the fuel can be cast into molds that remain as part of the fuel slugs (i.e., if the fuel can be left inside the molds for irradiation), then the quartz mold contribution to the waste stream can be eliminated. This possibility is being addresssed in an ongoing effort to evaluate the irradiation performance of fuel cast into zirconium sheaths rather than quartz molds. Zirconium was chosen as the sheath material because it is the component of the U-Pu-Zr fuel alloy that raises the alloy solidus temperatures and provides resistance to fuel-cladding chemical interaction (FCCI)

  13. Fuel cycle related parametric study considering long lived actinide production, decay heat and fuel cycle performances

    International Nuclear Information System (INIS)

    Raepsaet, X.; Damian, F.; Lenain, R.; Lecomte, M.

    2001-01-01

    One of the very attractive HTGR reactor characteristics is its highly versatile and flexible core that can fulfil a wide range of diverse fuel cycles. Based on a GTMHR-600 MWth reactor, analyses of several fuel cycles were carried out without taking into account common fuel particle performance limits (burnup, fast fluence, temperature). These values are, however, indicated in each case. Fuel derived from uranium, thorium and a wide variety of plutonium grades has been considered. Long-lived actinide production and total residual decay heat were evaluated for the various types of fuel. The results presented in this papers provide a comparison of the potential and limits of each fuel cycle and allow to define specific cycles offering lowest actinide production and residual heat associated with a long life cycle. (author)

  14. An innovative fuel design concept for improved light water reactor performance and safety. Final technical report

    International Nuclear Information System (INIS)

    Tulenko, J.S.; Connell, R.G.

    1995-07-01

    Light water reactor (LWR) fuel performance is limited by thermal and mechanical constraints associated with the design, fabrication, and operation of fuel in a nuclear reactor. The purpose of this research was to explore a technique for extending fuel performance by thermally bonding LWR fuel with a non-alkaline liquid metal alloy. Current LWR fuel rod designs consist of enriched uranium oxide (UO 2 ) fuel pellets enclosed in a zirconium alloy cylindrical clad. The space between the pellets and the clad is filled by an inert gas. Due to the thermal conductivity of the gas, the gas space thermally insulates the fuel pellets from the reactor coolant outside the fuel rod, elevating the fuel temperatures. Filling the gap between the fuel and clad with a high conductivity liquid metal thermally bonds the fuel to the cladding, and eliminates the large temperature change across the gap, while preserving the expansion and pellet loading capabilities. The resultant lower fuel temperature directly impacts fuel performance limit margins and also core transient performance. The application of liquid bonding techniques to LWR fuel was explored for the purposes of increasing LWR fuel performance and safety. A modified version of the ESCORE fuel performance code (ESBOND) has been developed under the program to analyze the in-reactor performance of the liquid metal bonded fuel. An assessment of the technical feasibility of this concept for LWR fuel is presented, including the results of research into materials compatibility testing and the predicted lifetime performance of Liquid Metal Bonded LWR fuel

  15. Performance analysis of hybrid solid oxide fuel cell and gas turbine cycle: Application of alternative fuels

    International Nuclear Information System (INIS)

    Zabihian, Farshid; Fung, Alan S.

    2013-01-01

    Highlights: • Variation of the stream properties in the syngas-fueled hybrid SOFC–GT cycle. • Detailed analysis of the operation of the methane-fueled SOFC–GT cycle. • Investigate effects of inlet fuel type and composition on performance of cycle. • Comparison of system operation when operated with and without anode recirculation. - Abstract: In this paper, the hybrid solid oxide fuel cell (SOFC) and gas turbine (GT) model was applied to investigate the effects of the inlet fuel type and composition on the performance of the cycle. This type of analysis is vital for the real world utilization of manufactured fuels in the hybrid SOFC–GT system due to the fact that these fuel compositions depends on the type of material that is processed, the fuel production process, and process control parameters. In the first part of this paper, it is shown that the results of a limited number of studies on the utilization of non-conventional fuels have been published in the open literature. However, further studies are required in this area to investigate all aspects of the issue for different configurations and assumptions. Then, the results of the simulation of the syngas-fueled hybrid SOFC–GT cycle are employed to explain the variation of the stream properties throughout the cycle. This analysis can be very helpful in understanding cycle internal working and can provide some interesting insights to the system operation. Then, the detailed information of the operation of the methane-fueled SOFC–GT cycle is presented. For both syngas- and methane-fueled cycles, the operating conditions of the equipment are presented and compared. Moreover, the comparison of the characteristics of the system when it is operated with two different schemes to provide the required steam for the cycle, with anode recirculation and with an external source of water, provides some interesting insights to the system operation. For instance, it was shown that although the physical

  16. Integrated Radiation Transport and Nuclear Fuel Performance for Assembly-Level Simulations

    Energy Technology Data Exchange (ETDEWEB)

    Clarno, Kevin T [ORNL; Hamilton, Steven P [ORNL; Philip, Bobby [ORNL; Berrill, Mark A [ORNL; Sampath, Rahul S [ORNL; Allu, Srikanth [ORNL; Pugmire, Dave [ORNL; Dilts, Gary [Los Alamos National Laboratory (LANL); Banfield, James E [ORNL

    2012-02-01

    The Advanced Multi-Physics (AMP) Nuclear Fuel Performance code (AMPFuel) is focused on predicting the temperature and strain within a nuclear fuel assembly to evaluate the performance and safety of existing and advanced nuclear fuel bundles within existing and advanced nuclear reactors. AMPFuel was extended to include an integrated nuclear fuel assembly capability for (one-way) coupled radiation transport and nuclear fuel assembly thermo-mechanics. This capability is the initial step toward incorporating an improved predictive nuclear fuel assembly modeling capability to accurately account for source-terms and boundary conditions of traditional (single-pin) nuclear fuel performance simulation, such as the neutron flux distribution, coolant conditions, and assembly mechanical stresses. A novel scheme is introduced for transferring the power distribution from the Scale/Denovo (Denovo) radiation transport code (structured, Cartesian mesh with smeared materials within each cell) to AMPFuel (unstructured, hexagonal mesh with a single material within each cell), allowing the use of a relatively coarse spatial mesh (10 million elements) for the radiation transport and a fine spatial mesh (3.3 billion elements) for thermo-mechanics with very little loss of accuracy. In addition, a new nuclear fuel-specific preconditioner was developed to account for the high aspect ratio of each fuel pin (12 feet axially, but 1 4 inches in diameter) with many individual fuel regions (pellets). With this novel capability, AMPFuel was used to model an entire 17 17 pressurized water reactor fuel assembly with many of the features resolved in three dimensions (for thermo-mechanics and/or neutronics), including the fuel, gap, and cladding of each of the 264 fuel pins; the 25 guide tubes; the top and bottom structural regions; and the upper and lower (neutron) reflector regions. The final, full assembly calculation was executed on Jaguar using 40,000 cores in under 10 hours to model over 162

  17. Analysis of fuel pin behavior under slow-ramp type transient overpower condition by using the fuel performance evaluation code 'FEMAXI-FBR'

    International Nuclear Information System (INIS)

    Tsuboi, Yasushi; Ninokata, Hisashi; Endo, Hiroshi; Ishizu, Tomoko; Tatewaki, Isao; Saito, Hiroaki

    2012-01-01

    FEMAXI-FBR has been developed as the one module of the core disruptive accident analysis code 'ASTERIA-FBR' in order to evaluate the mixed oxide (MOX) fuel performance under steady, transient and accident conditions of fast reactors consistently. On the basis of light water reactor (LWR) fuel performance evaluation code 'FEMAXI-6', FEMAXI-FBR develops specific models for the fast reactor fuel performance, such as restructuring, material migration during steady state and transient, melting cavity formation and pressure during accident, so that it can evaluate the fuel failure during accident. The analysis of test pin with slow transient over power test of CABRI-2 program was conducted from steady to transient. The test pin was pre-irradiated and tested under transient overpower with several % P 0 /s (P 0 : steady state power) of the power rate. Analysis results of the gas release ratio, pin failure time, and fuel melt radius were compared to measured values. The analysis results of the steady and transient performances were also compared with the measured values. The compared performances are gas release ratio, fuel restructuring for steady state and linear power and melt radius at failure during transient. This analysis result reproduces the measured value. It was concluded that FEMAXI-FBR is effective to evaluate fast reactor fuel performances from steady state to accident conditions. (author)

  18. Development of high performance hybrid rocket fuels

    Science.gov (United States)

    Zaseck, Christopher R.

    . In order to examine paraffin/additive combustion in a motor environment, I conducted experiments on well characterized aluminum based additives. In particular, I investigate the influence of aluminum, unpassivated aluminum, milled aluminum/polytetrafluoroethylene (PTFE), and aluminum hydride on the performance of paraffin fuels for hybrid rocket propulsion. I use an optically accessible combustor to examine the performance of the fuel mixtures in terms of characteristic velocity efficiency and regression rate. Each combustor test consumes a 12.7 cm long, 1.9 cm diameter fuel strand under 160 kg/m 2s of oxygen at up to 1.4 MPa. The experimental results indicate that the addition of 5 wt.% 30 mum or 80 nm aluminum to paraffin increases the regression rate by approximately 15% compared to neat paraffin grains. At higher aluminum concentrations and nano-scale particles sizes, the increased melt layer viscosity causes slower regression. Alane and Al/PTFE at 12.5 wt.% increase the regression of paraffin by 21% and 32% respectively. Finally, an aging study indicates that paraffin can protect air and moisture sensitive particles from oxidation. The opposed burner and aluminum/paraffin hybrid rocket experiments show that additives can alter bulk fuel properties, such as viscosity, that regulate entrainment. The general effect of melt layer properties on the entrainment and regression rate of paraffin is not well understood. Improved understanding of how solid additives affect the properties and regression of paraffin is essential to maximize performance. In this document I investigate the effect of melt layer properties on paraffin regression using inert additives. Tests are performed in the optical cylindrical combustor at ˜1 MPa under a gaseous oxygen mass flux of ˜160 kg/m2s. The experiments indicate that the regression rate is proportional to mu0.08rho 0.38kappa0.82. In addition, I explore how to predict fuel viscosity, thermal conductivity, and density prior to testing

  19. 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.)

  20. Effect of fuel fabrication parameters on performance- designer's point of view

    International Nuclear Information System (INIS)

    Prasad, P.N.; Ravi, M.; Soni, R.; Bajaj, S.S.; Bhardwaj, S.A.

    2004-01-01

    The fuel bundle performance in reactor depends upon the material properties, dimensions of the different components and their inter-compatibility. This paper brings out the fuel parameters required to be optimised to achieve better fuel reliability, operational flexibility, safety and economics from the designer point of view

  1. Effect analysis of air introduced by pressurization on fuel rod performances

    International Nuclear Information System (INIS)

    Ren Qisen; Liu Tong; Sheng Guofu

    2012-01-01

    In the process of pressurization and seal welding, it is common practice to vacuumize before gas filling for the sake of preventing introducing air and other impurities, which would affect the gas composition inside of the fuel rod. However, vacuumization during pressurization is likely not being required sometimes in order to simplify the fabrication procedure. In the present work, based on the AFA3G fuel rod design with 2 MPa of filling gas, analyses on fuel rod performances were carried out under the condition of pressurization with and without vacuumization, respectively. Furthermore, the effect on hydrogen content in fuel rod was preliminarily discussed. Results indicate that the impacts of air composition introduced by pressurization on fuel rod thermal-mechanical performances, such as internal pressure and fuel center temperature, were extremely slight. The gap conductance varies to some extent as a result of the change of gas composition due to air introduced in fuel rod. The impact of humidity on water content in fuel rod is negligible at a low temperature of around 25℃. However, at higher temperature, it is essential to pay attention on the control of fabrication process, and prevent much moisture entering into the fuel rod and increasing the probability of hydriding failure. (authors)

  2. 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

  3. Performance Evaluation and Suggestion of Upgraded Fuel Handling Equipment for Operating OPR1000

    International Nuclear Information System (INIS)

    Chang, Sang Gyoon; Hwang, Jeung Ki; Choi, Taek Sang; Na, Eun Seok; Lee, Myung Lyul; Baek, Seung Jin; Kim, Man Su; Kunik, Jack

    2011-01-01

    The purpose of this study is to evaluate the performance of upgraded FHE (Fuel Handling Equipment) for operating OPR 1000 (Optimized Power Reactor) by using data measured during the fuel reloading, and make some suggestions on enhancing the performance of the FHE. The fuel handling equipment, which serves critical processes in the refueling outage, has been upgraded to increase and improve the operational availability of the plant. The evaluation and suggestion of this study can be a beneficial tool related to the performance of the fuel handling equipment

  4. 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

  5. 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

  6. Effects of coal-derived trace species on performance of molten carbonate fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    1992-05-01

    The Carbonate Fuel Cell is a very promising option for highly efficient generation of electricity from many fuels. If coal-gas is to be used, the interactions of coal-derived impurities on various fuel cell components need to be understood. Thus the effects on Carbonate Fuel Cell performance due to ten different coal-derived contaminants viz., NH{sub 3}, H{sub 2}S, HC{ell}, H{sub 2}Se, AsH{sub 3}, Zn, Pb, Cd, Sn, and Hg, have been studied at Energy Research Corporation. Both experimental and theoretical evaluations were performed, which have led to mechanistic insights and initial estimation of qualitative tolerance levels for each species individually and in combination with other species. The focus of this study was to investigate possible coal-gas contaminant effects on the anode side of the Carbonate Fuel Cell, using both out-of-cell thermogravimetric analysis by isothermal TGA, and fuel cell testing in bench-scale cells. Separate experiments detailing performance decay in these cells with high levels of ammonia contamination (1 vol %) and with trace levels of Cd, Hg, and Sn, have indicated that, on the whole, these elements do not affect carbonate fuel cell performance. However, some performance decay may result when a number of the other six species are present, singly or simultaneously, as contaminants in fuel gas. In all cases, tolerance levels have been estimated for each of the 10 species and preliminary models have been developed for six of them. At this stage the models are limited to isothermal, benchscale (300 cm{sup 2} size) single cells. The information obtained is expected to assist in the development of coal-gas cleanup systems, while the contaminant performance effects data will provide useful basic information for modeling fuel cell endurance in conjunction with integrated gasifier/fuel-cell systems (IGFC).

  7. Pressurized water reactor fuel performance problems connected with fuel cladding corrosion processes

    International Nuclear Information System (INIS)

    Dobrevski, I.; Zaharieva, N.

    2008-01-01

    Generally, Pressurized Water Reactor (WWER, PWR) Fuel Element Performance is connected with fuel cladding corrosion and crud deposition processes. By transient to extended fuel cycles in nuclear power reactors, aiming to achieve higher burnup and better fuel utilization, the role of these processes increases significantly. This evolution modifies the chemical and electrochemical conditions in the reactor primary system, including change of fuel claddings' environment. The higher duty cores are always attended with increased boiling (sub-cooled nucleate boiling) mainly on the feed fuel assemblies. This boiling process on fuel cladding surfaces can cause different consequences on fuel element cladding's environment characteristics. In the case of boiling at the cladding surfaces without or with some cover of corrosion product deposition, the behavior of gases dissolved in water phase is strongly influenced by the vapor generation. The increase of vapor partial pressure will reduce the partial pressures of dissolved gases and will cause their stripping out. By these circumstances the concentrations of dissolved gases in cladding wall water layer can dramatically decrease, including also the case by which all dissolved gases to be stripped out. On the other hand it is known that the hydrogen is added to primary coolant in order to avoid the production of oxidants by radiolysis of water. It is clear that if boiling strips out dissolved hydrogen, the creation of oxidizing conditions at the cladding surfaces will be favored. In this case the local production of oxidants will be a result from local processes of water radiolysis, by which not only both oxygen (O 2 ) and hydrogen (H 2 ) but also hydrogen peroxide (H 2 O 2 ) will be produced. While these hydrogen and oxygen will be stripped out preferentially by boiling, the bigger part of hydrogen peroxide will remain in wall water phase and will act as the most important factor for creation of oxidizing conditions in fuel

  8. 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

  9. Transient fuel and target performance testing for the HWR-NPR

    International Nuclear Information System (INIS)

    Jicha, J.J. Jr.

    1990-01-01

    This paper describes a five year program of fuel target transient performance testing and model development required for the safety assessment of the HWR new production reactor. Technical issues are described, focusing on fuel and target behavior during extremely low probability transients which can lead to fuel melting. Early work on these issues is reviewed. The program to meet remaining needs is described. Three major transient-testing activities are included: in-cell experiments on small samples of irradiated fuel and target, small-scale phenomenological experiments in the ACRR reactor, and limited-integral experiments in the TREAT reactor. A coordinated development of detailed fuel and target behavior models is also described

  10. Dual-fuel natural gas/diesel engines: Technology, performance, and emissions

    Science.gov (United States)

    Turner, S. H.; Weaver, C. S.

    1994-11-01

    An investigation of current dual-fuel natural gas/diesel engine design, performance, and emissions was conducted. The most pressing technological problems associated with dual-fuel engine use were identified along with potential solutions. It was concluded that dual-fuel engines can achieve low NO(sub x) and particulate emissions while retaining fuel-efficiency and BMEP levels comparable to those of diesel engines. The investigation also examined the potential economic impact of dual-fuel engines in diesel-electric locomotives, marine vessels, farm equipment, construction, mining, and industrial equipment, and stand-alone electricity generation systems. Recommendations for further additional funding to support research, development, and demonstration in these applications were then presented.

  11. Fuel performance and operation experience of WWER-440 fuel in improved fuel cycle

    International Nuclear Information System (INIS)

    Gagarinski, A.; Proselkov, V.; Semchenkov, Yu.

    2007-01-01

    The paper summarizes WWER-440 second-generation fuel operation experience in improved fuel cycles using the example of Kola NPP units 3 and 4. Basic parameters of fuel assemblies, fuel rods and uranium-gadolinium fuel rods, as well as the principal neutronic parameters and burn-up achieved in fuel assemblies are presented. The paper also contains some data concerning the activity of coolant during operation (Authors)

  12. Performance and specific emissions contours throughout the operating range of hydrogen-fueled compression ignition engine with diesel and RME pilot fuels

    Directory of Open Access Journals (Sweden)

    Shahid Imran

    2015-09-01

    Full Text Available This paper presents the performance and emissions contours of a hydrogen dual fueled compression ignition (CI engine with two pilot fuels (diesel and rapeseed methyl ester, and compares the performance and emissions iso-contours of diesel and rapeseed methyl ester (RME single fueling with diesel and RME piloted hydrogen dual fueling throughout the engines operating speed and power range. The collected data have been used to produce iso-contours of thermal efficiency, volumetric efficiency, specific oxides of nitrogen (NOX, specific hydrocarbons (HC and specific carbon dioxide (CO2 on a power-speed plane. The performance and emission maps are experimentally investigated, compared, and critically discussed. Apart from medium loads at lower and medium speeds with diesel piloted hydrogen combustion, dual fueling produced lower thermal efficiency everywhere across the map. For diesel and RME single fueling the maximum specific NOX emissions are centered at the mid speed, mid power region. Hydrogen dual fueling produced higher specific NOX with both pilot fuels as compared to their respective single fueling operations. The range, location and trends of specific NOX varied significantly when compared to single fueling cases. The volumetric efficiency is discussed in detail with the implications of manifold injection of hydrogen analyzed with the conclusions drawn.

  13. Performance and management of IPR-R1 fuel elements

    International Nuclear Information System (INIS)

    Stasiulevicius, R.; Maretti Junior, F.

    1983-01-01

    The performance of fuel elements during the 23 years of the reactor operation, is presented aiming to introduce improvements in the fuel load distribution and consequent increase of the reactivity. A computer code CORE was developed aiming to calculate the individual burnup of the fuel elements and the value of the reactivity for several core configurations, establishing a routine to control the nuclear material in the IPR-R1. The values calculated were compared with the experimental results. Some alternatives to augment the reactivity of the present core are presented foreseeing the fuel load availability for operation with 100Km and, for angmenting the power reaction in a next stage. (E.G.) [pt

  14. Evaluation of corrosion on the fuel performance of stainless steel cladding

    Directory of Open Access Journals (Sweden)

    de Souza Gomes Daniel

    2016-01-01

    Full Text Available In nuclear reactors, the use of stainless steel (SS as the cladding material offers some advantages such as good mechanical and corrosion resistance. However, its main advantage is the reduction in the amount of the hydrogen released during loss-of-coolant accident, as observed in the Fukushima Daiichi accident. Hence, research aimed at developing accident tolerant fuels should consider SS as an important alternative to existing materials. However, the available computational tools used to analyze fuel rod performance under irradiation are not capable of assessing the effectiveness of SS as the cladding material. This paper addresses the SS corrosion behavior in a modified fuel performance code in order to evaluate its effect on the global fuel performance. Then, data from the literature concerning to SS corrosion are implemented in the specific code subroutines, and the results obtained are compared to those for Zircaloy-4 (Zy-4 under the same power history. The results show that the effects of corrosion on SS are considerably different from those on Zy-4. The thickness of the oxide layer formed on the SS surface is considerably lower than that formed on Zy-4. As a consequence of this, the global fuel performance of SS under irradiation should be less affected by the corrosion.

  15. Evaluation of fuel performance for fresh and aged CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Jong Yeob; Bae, Jun Ho; Park, Joo Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    Like all other industrial plants, nuclear power plants also undergo degradations, so called ageing, with their operation time. Accordingly, in the recent safety analysis for a refurbished Wolsong 1 NPP, various ageing effects were incorporated into the hydraulic models of a number of the components in the primary heat transport system for conservatism. The ageing data of thermal-hydraulic components for 11 EFPY of Wolsong 1 were derived by using NUCIRC code based on the site operation data and they were modified to the appropriate input data for CATHENA code which is a thermal hydraulic code for a postulated accident analysis. This paper deals with the ageing effect of the PHTS (primary heat transport system) of CANDU reactor on the fuel performance during the normal operation. Initial conditions for fuel performance analysis were derived from the thermal-hydraulic analysis for both fresh and aged core models. Here, fresh core means a core state just right after the refurbishment and the aged core is 11 EFPY state after the refurbishment of Wolsong 1. The fuel performance was analyzed by using ELESTRES code for both fresh and aged core state and the results were compared in order to verify the ageing effect of CANDU HTS on the fuel performance.

  16. Modeling constituent redistribution in U–Pu–Zr metallic fuel using the advanced fuel performance code BISON

    International Nuclear Information System (INIS)

    Galloway, J.; Unal, C.; Carlson, N.; Porter, D.; Hayes, S.

    2015-01-01

    Highlights: • An improved constituent distribution formulation in metallic nuclear fuels. • The new algorithm is implemented into the advanced fuel performance framework BISON. • Experimental Breeder Reactor-II data, T179, DP16, T459 are reanalyzed. • Phase dependent diffusion coefficients are improved. • Most influential phase is gamma, followed by alpha and thirdly the beta phase. - Abstract: An improved robust formulation for constituent distribution in metallic nuclear fuels is developed and implemented into the advanced fuel performance framework BISON. The coupled thermal diffusion equations are solved simultaneously to reanalyze the constituent redistribution in post irradiation data from fuel tests performed in Experimental Breeder Reactor-II (EBR-II). Deficiencies observed in previously published formulation and numerical implementations are also improved. The present model corrects an inconsistency between the enthalpies of solution and the solubility limit curves of the phase diagram while also adding an artificial diffusion term when in the 2-phase regime that stabilizes the standard Galerkin finite element (FE) method used by BISON. An additional improvement is in the formulation of zirconium flux as it relates to the Soret term. With these new modifications, phase dependent diffusion coefficients are revaluated and compared with the previously recommended values. The model validation included testing against experimental data from fuel pins T179, DP16 and T459, irradiated in EBR-II. A series of viable material properties for U–Pu–Zr based materials was determined through a sensitivity study, which resulted in three cases with differing parameters that showed strong agreement with one set of experimental data, rod T179. Subsequently a full-scale simulation of T179 was performed to reduce uncertainties, particularly relating to the temperature boundary condition for the fuel. In addition a new thermal conductivity model combining all

  17. Performance and emissions of a dual-fuel pilot diesel ignition engine operating on various premixed fuels

    International Nuclear Information System (INIS)

    Yousefi, Amin; Birouk, Madjid; Lawler, Benjamin; Gharehghani, Ayatallah

    2015-01-01

    Highlights: • Natural gas/diesel, methanol/diesel, and hydrogen/diesel cases were investigated. • For leaner mixtures, the hydrogen/diesel case has the highest IMEP and ITE. • The methanol/diesel case has the maximum IMEP and ITE for richer mixtures. • Hydrogen/diesel case experiences soot and CO free combustion at rich regions. - Abstract: A multi-dimensional computational fluid dynamics (CFD) model coupled with chemical kinetics mechanisms was applied to investigate the effect of various premixed fuels and equivalence ratios on the combustion, performance, and emissions characteristics of a dual-fuel indirect injection (IDI) pilot diesel ignition engine. The diesel fuel is supplied via indirect injection into the cylinder prior to the end of the compression stroke. Various premixed fuels were inducted into the engine through the intake manifold. The results showed that the dual-fuel case using hydrogen/diesel has a steeper pressure rise rate, higher peak heat release rate (PHRR), more advanced ignition timing, and shorter ignition delay compared to the natural gas/diesel and methanol/diesel dual-fuel cases. For leaner mixtures (Φ_P 0.32). For instance, with an equivalence ratio of 0.35, the ITE is 56.24% and 60.85% for hydrogen/diesel and methanol/diesel dual-fuel cases, respectively. For an equivalence ratio of 0.15, the natural gas/diesel simulation exhibits partial burn combustion and thus results in a negative IMEP. At equivalence ratios of 0.15, 0.2, and 0.25, the methanol/diesel case experiences misfiring phenomenon which consequently deteriorates the engine performance considerably. As for the engine-out emissions, the hydrogen/diesel results display carbon monoxide (CO) free combustion relative to natural gas/diesel and methanol/diesel engines; however, considerable amount of nitrogen oxides (NO_x) emissions are produced at an equivalence ratio of 0.35 which exceeds the Euro 6 NO_x limit. Due to the larger area exposed to high temperature regions

  18. Fuel performance of licensed nuclear power plants through 1974

    International Nuclear Information System (INIS)

    Bobe, P.E.

    1976-01-01

    General aspects of fuel element design and specific design data for typical Pressurized and Boiling Water Reactors are presented. Based on a literature search, failure modes and specific failures incurred through December 31, 1974 are described, together with a discussion of those problems which have had a significant impact upon plant operation. The relationship between fuel element failures and the resultant coolant activity/radioactive gaseous effluents upon radiation exposure, plant availability and capacity factors, economic impact, and waste management, are discussed. An assessment was made regarding the generation, availability, and use of fuel performance data

  19. Fuel leak testing performance at NPP Jaslovske Bohunice

    International Nuclear Information System (INIS)

    Slugen, V.; Krnac, S.; Smiesko, I.

    1995-01-01

    The NPP Bohunice VVER-440 fuel leak testing experience are relatively extensive in comparison with other VVER-440 users. As the first Europe NPP was adapted Siemens (KWU) in core-sipping equipment to VVER-440 units and since this time were have done these tests also for NPP Paks (Hungary) and NPP Dukovany (Czech Republic). The occurrence of leaking fuel assemblies in NPP is in the last 5 years relatively stabilised and low. A significant difference can be observed between type V-230 (31 leaks) and type V-213 (1 leak). None of of the indicated leaking fuel assemblies has been investigated in the hot cell. Therefore cannot be confirm the effective causes of leak occurrence. Nevertheless, the fuel failure rate and the performance of leak testing in NPP Bohunice are comparable to the world standard at PWR's. 1 tab., 2 figs., 3 refs

  20. Fuel leak testing performance at NPP Jaslovske Bohunice

    Energy Technology Data Exchange (ETDEWEB)

    Slugen, V; Krnac, S [Slovak Technical Univ., Bratislava (Slovakia); Smiesko, I [Nuclear Powr Plant EBO, Jaslovske Bohuce (Slovakia)

    1996-12-31

    The NPP Bohunice VVER-440 fuel leak testing experience are relatively extensive in comparison with other VVER-440 users. As the first Europe NPP was adapted Siemens (KWU) in core-sipping equipment to VVER-440 units and since this time were have done these tests also for NPP Paks (Hungary) and NPP Dukovany (Czech Republic). The occurrence of leaking fuel assemblies in NPP is in the last 5 years relatively stabilised and low. A significant difference can be observed between type V-230 (31 leaks) and type V-213 (1 leak). None of of the indicated leaking fuel assemblies has been investigated in the hot cell. Therefore cannot be confirm the effective causes of leak occurrence. Nevertheless, the fuel failure rate and the performance of leak testing in NPP Bohunice are comparable to the world standard at PWR`s. 1 tab., 2 figs., 3 refs.

  1. Performance assessment analyses unique to Department of Energy spent nuclear fuel

    International Nuclear Information System (INIS)

    Loo, H.H.; Duguid, J.J.

    2000-01-01

    This paper describes the iterative process of grouping and performance assessment that has led to the current grouping of the U.S. Department of Energy (DOE) spent nuclear fuel (SNF). The unique sensitivity analyses that form the basis for incorporating DOE fuel into the total system performance assessment (TSPA) base case model are described. In addition, the chemistry that results from dissolution of DOE fuel and high level waste (HLW) glass in a failed co-disposal package, and the effects of disposal of selected DOE SNF in high integrity cans are presented

  2. Improving the performance of dual fuel engines running on natural gas/LPG by using pilot fuel derived from jojoba seeds

    Energy Technology Data Exchange (ETDEWEB)

    Selim, Mohamed Y.E. [Mechanical Engineering Department, College of Engineering, UAE University, Jimmi, Al-Ain, P.O. Box 17555, Abu Dhabi (United Arab Emirates); Radwan, M.S.; Saleh, H.E. [Mechanical Power Engineering Department, Faculty of Engineering at Mattaria, Helwan University, Cairo (Egypt)

    2008-06-15

    The use of jojoba methyl ester as a pilot fuel was investigated for almost the first time as a way to improve the performance of dual fuel engine running on natural gas or liquefied petroleum gas (LPG) at part load. The dual fuel engine used was Ricardo E6 variable compression diesel engine and it used either compressed natural gas (CNG) or LPG as the main fuel and jojoba methyl ester as a pilot fuel. Diesel fuel was used as a reference fuel for the dual fuel engine results. During the experimental tests, the following have been measured: engine efficiency in terms of specific fuel consumption, brake power output, combustion noise in terms of maximum pressure rise rate and maximum pressure, exhaust emissions in terms of carbon monoxide and hydrocarbons, knocking limits in terms of maximum torque at onset of knocking, and cyclic variability data of 100 engine cycles in terms of maximum pressure and its pressure rise rate average and standard deviation. The tests examined the following engine parameters: gaseous fuel type, engine speed and load, pilot fuel injection timing, pilot fuel mass and compression ratio. Results showed that using the jojoba fuel with its improved properties has improved the dual fuel engine performance, reduced the combustion noise, extended knocking limits and reduced the cyclic variability of the combustion. (author)

  3. 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)

  4. Performance effects of coal-derived contaminants on the carbonate fuel cell

    Energy Technology Data Exchange (ETDEWEB)

    Pigeaud, A. [Energy Research Corp., Danbury, CT (United States); Wilemski, G. [Physical Sciences, Inc., Andover, MA (United States)

    1993-05-01

    Coal-derived contaminant studies have been pursued at ERC since the early 1980`s when the pace of carbonate fuel cell development began to markedly increase. Initial work was concerned with performance effects on laboratory and bench-scale carbonate fuel cells primarily due to sulfur compounds. Results have now also been obtained with respect to nine additional coal-gas contaminants, including volatile trace metal species. Thermochemical calculations, out-of-cell experiments, and cell performance as well as endurance testshave recently been conducted which have involved the following species: NH{sub 3}, H{sub 2}S [COS], HCl, AsH{sub 3}[As{sub 2}(v)], Zn(v), Pb(v), Cd(v), H{sub 2} Se, Hg(v), Sn(v). Employing thermochemically calculated results, thermogravimetric (TGA) and pre-, and post-test analytical data as well as fuel cell performance observations, it has been shown that there are four main mechanisms of contaminant interaction with the carbonate fuel cell. These have been formulated into performance models for six significant contaminant species, thus providing long-term endurance estimations.

  5. Performance effects of coal-derived contaminants on the carbonate fuel cell

    Energy Technology Data Exchange (ETDEWEB)

    Pigeaud, A. (Energy Research Corp., Danbury, CT (United States)); Wilemski, G. (Physical Sciences, Inc., Andover, MA (United States))

    1993-01-01

    Coal-derived contaminant studies have been pursued at ERC since the early 1980's when the pace of carbonate fuel cell development began to markedly increase. Initial work was concerned with performance effects on laboratory and bench-scale carbonate fuel cells primarily due to sulfur compounds. Results have now also been obtained with respect to nine additional coal-gas contaminants, including volatile trace metal species. Thermochemical calculations, out-of-cell experiments, and cell performance as well as endurance testshave recently been conducted which have involved the following species: NH[sub 3], H[sub 2]S [COS], HCl, AsH[sub 3][As[sub 2](v)], Zn(v), Pb(v), Cd(v), H[sub 2] Se, Hg(v), Sn(v). Employing thermochemically calculated results, thermogravimetric (TGA) and pre-, and post-test analytical data as well as fuel cell performance observations, it has been shown that there are four main mechanisms of contaminant interaction with the carbonate fuel cell. These have been formulated into performance models for six significant contaminant species, thus providing long-term endurance estimations.

  6. Core performance of equilibrium fast reactors for different coolant materials and fuel types

    International Nuclear Information System (INIS)

    Mizutani, Akihiko; Sekimoto, Hiroshi

    1998-01-01

    Parametric studies with several coolant and fuel materials in the equilibrium state are performed for fast reactors in which natural uranium is fed and all of the actinides are confined. Sodium, sodium-potassium, lead, lead-bismuth and helium coolant materials, and oxide, nitride and metal fuels are employed to compare the neutronic characteristics in the equilibrium state. As to the criticality performance, sodium-potassium shows the best performance among the liquid metal coolants and the metallic fuel indicates the best performance

  7. Oxide fuel pin transient performance analysis and design with the TEMECH code

    International Nuclear Information System (INIS)

    Bard, F.E.; Dutt, S.P.; Hinman, C.A.; Hunter, C.W.; Pitner, A.L.

    1986-01-01

    The TEMECH code is a fast-running, thermal-mechanical-hydraulic, analytical program used to evaluate the transient performance of LMR oxide fuel pins. The code calculates pin deformation and failure probability due to fuel-cladding differential thermal expansion, expansion of fuel upon melting, and fission gas pressurization. The mechanistic fuel model in the code accounts for fuel cracking, crack closure, porosity decrease, and the temperature dependence of fuel creep through the course of the transient. Modeling emphasis has been placed on results obtained from Fuel Cladding Transient Test (FCTT) testing, Transient Fuel Deformation (TFD) tests and TREAT integral fuel pin experiments

  8. Fuel performance and fission product behaviour in gas cooled reactors

    International Nuclear Information System (INIS)

    1997-11-01

    The Co-ordinated Research Programme (CRP) on Validation of Predictive Methods for Fuel and Fission Product Behaviour was organized within the frame of the International Working Group on Gas Cooled Reactors. This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs), and supports the conduct of these activities. The objectives of this CRP were to review and document the status of the experimental data base and of the predictive methods for GCR fuel performance and fission product behaviour; and to verify and validate methodologies for the prediction of fuel performance and fission product transport

  9. Optimization of the fuel assembly for the Canadian Supercritical Water-cooled Reactor (SCWR)

    Energy Technology Data Exchange (ETDEWEB)

    French, C.; Bonin, H.; Chan, P., E-mail: Corey.French@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada)

    2013-07-01

    A parametric optimization of the Canadian Supercritical Water-cooled Reactor (SCWR) lattice geometry and fresh fuel content is performed in this work. With the potential to improve core physics and performance, significant gains to operating and safety margins could be achieved through slight progressions. The fuel performance codes WIMS-AECL and SERPENT are used to calculate performance factors, and use them as inputs to an optimization algorithm. (author)

  10. Application of advanced validation concepts to oxide fuel performance codes: LIFE-4 fast-reactor and FRAPCON thermal-reactor fuel performance codes

    Energy Technology Data Exchange (ETDEWEB)

    Unal, C., E-mail: cu@lanl.gov [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Williams, B.J. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Yacout, A. [Argonne National Laboratory, 9700 S. Cass Avenue, Lemont, IL 60439 (United States); Higdon, D.M. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States)

    2013-10-15

    Highlights: ► The application of advanced validation techniques (sensitivity, calibration and prediction) to nuclear performance codes FRAPCON and LIFE-4 is the focus of the paper. ► A sensitivity ranking methodology narrows down the number of selected modeling parameters from 61 to 24 for FRAPCON and from 69 to 35 for LIFE-4. ► Fuel creep, fuel thermal conductivity, fission gas transport/release, crack/boundary, and fuel gap conductivity models of LIFE-4 are identified for improvements. ► FRAPCON sensitivity results indicated the importance of the fuel thermal conduction and the fission gas release models. -- Abstract: Evolving nuclear energy programs expect to use enhanced modeling and simulation (M and S) capabilities, using multiscale, multiphysics modeling approaches, to reduce both cost and time from the design through the licensing phases. Interest in the development of the multiscale, multiphysics approach has increased in the last decade because of the need for predictive tools for complex interacting processes as a means of eliminating the limited use of empirically based model development. Complex interacting processes cannot be predicted by analyzing each individual component in isolation. In most cases, the mathematical models of complex processes and their boundary conditions are nonlinear. As a result, the solutions of these mathematical models often require high-performance computing capabilities and resources. The use of multiscale, multiphysics (MS/MP) models in conjunction with high-performance computational software and hardware introduces challenges in validating these predictive tools—traditional methodologies will have to be modified to address these challenges. The advanced MS/MP codes for nuclear fuels and reactors are being developed within the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program of the US Department of Energy (DOE) – Nuclear Energy (NE). This paper does not directly address challenges in calibration

  11. Performance of a diesel engine operating on raw coal-diesel fuel and solvent refined coal-diesel fuel slurries. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, H.P.

    1980-03-01

    Performance tests using an 11 kW single cylinder diesel engine were made to determine the effects of three different micronized coal-fuel oil slurries being considered as alternative fuels. Slurries containing 20, 32, and 40%-wt micronized raw coal in No. 2 fuel oil were used. Results are presented indicating the changes in the concentrations of SO/sub X/ and NO/sub X/ in the exhaust, exhaust opacity, power and efficiency, and in wear rates relative to operation on fuel oil No. 2. The engine was operated for 10 h at full load and 1400 rpm on al fuels except the 40%-wt slurry. This test was discontinued because of extremely poor performance.

  12. A comparative study of fission gas behaviour in UO2 and MOX fuels using the meteor fuel performance code

    International Nuclear Information System (INIS)

    Struzik, C.; Garcia, Ph.; Noirot, L.

    2002-01-01

    The paper reviews some of the fission-gas-related differences observed between MOX MIMAS AUC fuels and homogeneous UO 2 fuels. Under steady-state conditions, the apparently higher fractional release in MOX fuels is interpreted with the METEOR fuel performance code as a consequence of their lower thermal conductivity and the higher linear heat rates to which MOX fuel rods are subjected. Although more fundamental diffusion properties are needed, the apparently greater swelling of MOX fuel rods at higher linear heat rates can be ascribed to enhanced diffusion properties. (authors)

  13. Assessment of causes for degrading fuel performance at Darlington NGS

    International Nuclear Information System (INIS)

    Judah, J.; Goodchild, S.

    2013-01-01

    Fuel performance at the Darlington nuclear generating station has historically been excellent. Until recently, the majority of these few fuel defects have been attributed to fretting by heat transport system debris. The minority have been linked to manufacturing issues. Recently, Darlington has experienced an increase in the number of fuel defects. Although the defect rate remains low with respect to industry standards, this defect experience is considered to be unacceptable given current industry expectations and the OPG zero defect policy. Nine fuel defects have been discharged since 2007 from the four Darlington reactors. This represents a fuel defect rate of just 0.35 defects per year per reactor. At the time of this writing three additional defects are suspected to be in core. Although a definitive defect cause has yet to be identified, these fuel performance issues appear to be due to the coincidental degradation of manufacturing and operational factors, thereby decreasing the margins to fuel failure due to fuelling power ramps. All of the confirmed defected bundles have been long bundles and all experienced a relatively high power ramp when shifted from Position 2 to Position 6. High bundle uranium masses and low internal clearances are thought to be significant contributing factors. Bundle burnups at the time of the power ramps were low and these bundles were not identified by existing power ramp defect predictive tools. Our assessment has resulted in a number of recommendations which are designed to mitigate these adverse conditions by restoring the margins to power ramp failures. These recommendations impact broadly across a number of organizations including reactor physics, fuel design, fuel manufacturing, reactor design, inspections and PIE. (author)

  14. BWR spent fuel storage cask performance test. Volume 2. Pre- and post-test decay heat, heat transfer, and shielding analyses

    International Nuclear Information System (INIS)

    Wiles, L.E.; Lombardo, N.J.; Heeb, C.M.; Jenquin, U.P.; Michener, T.E.; Wheeler, C.L.; Creer, J.M.; McCann, R.A.

    1986-06-01

    This report describes the decay heat, heat transfer, and shielding analyses conducted in support of performance testing of a Ridhihalgh, Eggers and Associates REA 2033 boiling water reactor (BWR) spent fuel storage cask. The cask testing program was conducted for the US Department of Energy (DOE) Commercial Spent Fuel Management Program by the Pacific Northwest Laboratory (PNL) and by General Electric at the latters' Morris Operation (GE-MO) as reported in Volume I. The analyses effort consisted of performing pretest calculations to (1) select spent fuel for the test; (2) symmetrically load the spent fuel assemblies in the cask to ensure lateral symmetry of decay heat generation rates; (3) optimally locate temperature and dose rate instrumentation in the cask and spent fuel assemblies; and (4) evaluate the ORIGEN2 (decay heat), HYDRA and COBRA-SFS (heat transfer), and QAD and DOT (shielding) computer codes. The emphasis of this second volume is on the comparison of code predictions to experimental test data in support of the code evaluation process. Code evaluations were accomplished by comparing pretest (actually pre-look, since some predictions were not completed until testing was in progress) predictions with experimental cask testing data reported in Volume I. No attempt was made in this study to compare the two heat transfer codes because results of other evaluations have not been completed, and a comparison based on one data set may lead to erroneous conclusions

  15. 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.

  16. Modelling of LOCA Tests with the BISON Fuel Performance Code

    Energy Technology Data Exchange (ETDEWEB)

    Williamson, Richard L [Idaho National Laboratory; Pastore, Giovanni [Idaho National Laboratory; Novascone, Stephen Rhead [Idaho National Laboratory; Spencer, Benjamin Whiting [Idaho National Laboratory; Hales, Jason Dean [Idaho National Laboratory

    2016-05-01

    BISON is a modern finite-element based, multidimensional nuclear fuel performance code that is under development at Idaho National Laboratory (USA). Recent advances of BISON include the extension of the code to the analysis of LWR fuel rod behaviour during loss-of-coolant accidents (LOCAs). In this work, BISON models for the phenomena relevant to LWR cladding behaviour during LOCAs are described, followed by presentation of code results for the simulation of LOCA tests. Analysed experiments include separate effects tests of cladding ballooning and burst, as well as the Halden IFA-650.2 fuel rod test. Two-dimensional modelling of the experiments is performed, and calculations are compared to available experimental data. Comparisons include cladding burst pressure and temperature in separate effects tests, as well as the evolution of fuel rod inner pressure during ballooning and time to cladding burst. Furthermore, BISON three-dimensional simulations of separate effects tests are performed, which demonstrate the capability to reproduce the effect of azimuthal temperature variations in the cladding. The work has been carried out in the frame of the collaboration between Idaho National Laboratory and Halden Reactor Project, and the IAEA Coordinated Research Project FUMAC.

  17. Experience with advanced driver fuels in EBR-II

    International Nuclear Information System (INIS)

    Lahm, C.E.; Koenig, J.F.; Pahl, R.G.; Porter, D.L.; Crawford, D.C.

    1992-01-01

    The Experimental Breeder Reactor II (EBR-II) is a complete nuclear power plant, incorporating a pool-type liquid-metal reactor (LMR) with a fuel-power thermal output of 62.5 MW and an electrical output of 20 MW. Initial criticality was in 1961, utilizing a metallic driver fuel design called the Mark-I. The fuel design has evolved over the last 30 yr, and significant progress has been made on improving performance. The first major innovations were incorporated into the Mark-II design, and burnup then increased dramatically. This design performed successfully, and fuel element lifetime was limited by subassembly hardware performance rather than the fuel element itself. Transient performance of the fuel was also acceptable and demonstrated the ability of EBR-II to survive severe upsets such as a loss of flow without scram. In the mid 1980s, with renewed interest in metallic fuels and Argonne's integral fast reactor (IFR) concept, the Mark-II design was used as the basis for new designs, the Mark-III and Mark-IV. In 1987, the Mark-III design began qualification testing to become a driver fuel for EBR-II. This was followed in 1989 by the Mark-IIIA and Mark-IV designs. The next fuel design, the Mark-V, is being planned to demonstrate the utilization of recycled fuel. The fuel cycle facility attached to EBR-II is being refurbished to produce pyroprocessed recycled fuel as part of the demonstration of the IFR

  18. Performance limits of coated particle fuel. Part I. The significance of empirical performance diagrams and mathematical models in fuel development and power reactor studies

    Energy Technology Data Exchange (ETDEWEB)

    Graham, L. W.; Hick, H.

    1973-06-15

    This report introduces a general survey of our present knowledge and understanding of coated particle fuel performance. It defines first the reference power reactor conditions and the reference coated particle design on which the survey is centred. It describes then the typical strategy which has been followed in coated particle fuel development by the Dragon Project R & D Branch. Finally it shows the priorities which have governed the time scale and scope of fuel development and of the present review.

  19. High temperature compression tests performed on doped fuels

    International Nuclear Information System (INIS)

    Duguay, C.; Mocellin, A.; Dehaudt, P.; Fantozzi, G.

    1997-01-01

    The use of additives of corundum structure M 2 O 3 (M=Cr, Al) is an effective way of promoting grain growth of uranium dioxide. The high-temperature compressive deformation of large-grained UO 2 doped with these oxides has been investigated and compared with that of pure UO 2 with a standard microstructure. Such doped fuels are expected to exhibit enhanced plasticity. Their use would therefore reduce the pellet-cladding mechanical interaction and thus improve the performances of the nuclear fuel. (orig.)

  20. 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)

  1. 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.

  2. 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.

  3. 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.

  4. Qualification status of LEU [low enriched uranium] fuels

    International Nuclear Information System (INIS)

    Snelgrove, J.L.

    1987-01-01

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

  5. Irradiation performance of experimental fast reactor 'JOYO' MK-1 driver fuel assemblies

    International Nuclear Information System (INIS)

    Itaki, Toshiyuki; Kono, Keiichi; Tachi, Hirokatsu; Yamanouchi, Sadamu; Yuhara, Shunichi; Shibahara, Itaru

    1985-01-01

    The experimental fast reactor ''JOYO'' completed it's breeder core (MK-I) operation in January 1982. The MK-I driver fuel assemblies were removed from the core sequencially in order of burnup increase and have been under postirradiation examination (PIE). The PIE has almost been completed for 30 assemblies including the highest burnup assemblies of 48,000 MWD/MTM. It has been confirmed that all fuel assemblies have exhibited satisfactory performance without detrimental assembly deformation or without any indications of fuel pin breach. The irradiation conditions of the MK-I core were somewhat more moderate than those conditions envisioned for prototypic reactor. However the results of the examination revealed the typical irradiation behavior of LMFBR fuels, although such characteristics were benign as compared with those anticipated in high burnup fuels. Systematic performance data have been accumulated through the fuel fabrication, irradiation and postirradiation examination processes. Based on these data, the MK-I fuel designing and fabrication techniques were totally confirmed. This technical experience and the associated insight into irradiation behavior have established a milestone to the next step of fast reactor fuel development. (author)

  6. Development and validation of a fuel performance analysis code

    International Nuclear Information System (INIS)

    Majalee, Aaditya V.; Chaturvedi, S.

    2015-01-01

    CAD has been developing a computer code 'FRAVIZ' for calculation of steady-state thermomechanical behaviour of nuclear reactor fuel rods. It contains four major modules viz., Thermal module, Fission Gas Release module, Material Properties module and Mechanical module. All these four modules are coupled to each other and feedback from each module is fed back to others to get a self-consistent evolution in time. The computer code has been checked against two FUMEX benchmarks. Modelling fuel performance in Advance Heavy Water Reactor would require additional inputs related to the fuel and some modification in the code.(author)

  7. Achieving high performance in intermediate temperature direct carbon fuel cells with renewable carbon as a fuel source

    International Nuclear Information System (INIS)

    Hao, Wenbin; He, Xiaojin; Mi, Yongli

    2014-01-01

    Highlights: • Bamboo fiber and waste paper were pyrolyzed to generate bamboo carbon and waste paper carbon as anode fuels of IT-DCFC. • Superior cell performance was achieved with the waste paper carbon. • The results suggested the high performance was due to the highest thermal reactivity and the catalytic inherent impurities. • Calcite and kaolinite as inherent impurities favored the thermal decomposition and the electrooxidation of carbon. - Abstract: Three kinds of carbon sources obtained from carbon black, bamboo fiber and waste paper were investigated as anode fuels in an intermediate temperature direct carbon fuel cell. The carbon sources were characterized with X-ray photoelectron spectroscopy, thermal gravimetric analysis, etc. The results indicated that the waste paper carbon was more abundant in calcite and kaolinite, and showed higher thermal reactivity in the intermediate temperature range compared with the other two carbon sources. The cell performance was tested at 650 °C in a hybrid single cell, using Sm 0.20 Ce 0.80 O 2−x as the electrolyte. As a result, the cell fed with waste paper carbon showed the highest performance among the three carbon sources, with a peak power density of 225 mW cm −2 . The results indicated that its inherent impurities, such as calcite and kaolinite, might favor the thermal gasification of renewable carbon sources, which resulted in the enhanced performance of the intermediate temperature direct carbon fuel cell

  8. Effect of engine parameters and type of gaseous fuel on the performance of dual-fuel gas diesel engines. A critical review

    Energy Technology Data Exchange (ETDEWEB)

    Sahoo, B.B. [Centre for Energy, Indian Institute of Technology, Guwahati 781039 (India); Sahoo, N.; Saha, U.K. [Department of Mechanical Engineering, Indian Institute of Technology, Guwahati 781039 (India)

    2009-08-15

    Petroleum resources are finite and, therefore, search for their alternative non-petroleum fuels for internal combustion engines is continuing all over the world. Moreover gases emitted by petroleum fuel driven vehicles have an adverse effect on the environment and human health. There is universal acceptance of the need to reduce such emissions. Towards this, scientists have proposed various solutions for diesel engines, one of which is the use of gaseous fuels as a supplement for liquid diesel fuel. These engines, which use conventional diesel fuel and gaseous fuel, are referred to as 'dual-fuel engines'. Natural gas and bio-derived gas appear more attractive alternative fuels for dual-fuel engines in view of their friendly environmental nature. In the gas-fumigated dual-fuel engine, the primary fuel is mixed outside the cylinder before it is inducted into the cylinder. A pilot quantity of liquid fuel is injected towards the end of the compression stroke to initiate combustion. When considering a gaseous fuel for use in existing diesel engines, a number of issues which include, the effects of engine operating and design parameters, and type of gaseous fuel, on the performance of the dual-fuel engines, are important. This paper reviews the research on above issues carried out by various scientists in different diesel engines. This paper touches upon performance, combustion and emission characteristics of dual-fuel engines which use natural gas, biogas, producer gas, methane, liquefied petroleum gas, propane, etc. as gaseous fuel. It reveals that 'dual-fuel concept' is a promising technique for controlling both NO{sub x} and soot emissions even on existing diesel engine. But, HC, CO emissions and 'bsfc' are higher for part load gas diesel engine operations. Thermal efficiency of dual-fuel engines improve either with increased engine speed, or with advanced injection timings, or with increased amount of pilot fuel. The ignition

  9. Support vector machine to predict diesel engine performance and emission parameters fueled with nano-particles additive to diesel fuel

    Science.gov (United States)

    Ghanbari, M.; Najafi, G.; Ghobadian, B.; Mamat, R.; Noor, M. M.; Moosavian, A.

    2015-12-01

    This paper studies the use of adaptive Support Vector Machine (SVM) to predict the performance parameters and exhaust emissions of a diesel engine operating on nanodiesel blended fuels. In order to predict the engine parameters, the whole experimental data were randomly divided into training and testing data. For SVM modelling, different values for radial basis function (RBF) kernel width and penalty parameters (C) were considered and the optimum values were then found. The results demonstrate that SVM is capable of predicting the diesel engine performance and emissions. In the experimental step, Carbon nano tubes (CNT) (40, 80 and 120 ppm) and nano silver particles (40, 80 and 120 ppm) with nanostructure were prepared and added as additive to the diesel fuel. Six cylinders, four-stroke diesel engine was fuelled with these new blended fuels and operated at different engine speeds. Experimental test results indicated the fact that adding nano particles to diesel fuel, increased diesel engine power and torque output. For nano-diesel it was found that the brake specific fuel consumption (bsfc) was decreased compared to the net diesel fuel. The results proved that with increase of nano particles concentrations (from 40 ppm to 120 ppm) in diesel fuel, CO2 emission increased. CO emission in diesel fuel with nano-particles was lower significantly compared to pure diesel fuel. UHC emission with silver nano-diesel blended fuel decreased while with fuels that contains CNT nano particles increased. The trend of NOx emission was inverse compared to the UHC emission. With adding nano particles to the blended fuels, NOx increased compared to the net diesel fuel. The tests revealed that silver & CNT nano particles can be used as additive in diesel fuel to improve complete combustion of the fuel and reduce the exhaust emissions significantly.

  10. Fuel performance and fission product behaviour in gas cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-11-01

    The Co-ordinated Research Programme (CRP) on Validation of Predictive Methods for Fuel and Fission Product Behaviour was organized within the frame of the International Working Group on Gas Cooled Reactors. This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs), and supports the conduct of these activities. The objectives of this CRP were to review and document the status of the experimental data base and of the predictive methods for GCR fuel performance and fission product behaviour; and to verify and validate methodologies for the prediction of fuel performance and fission product transport. Refs, figs, tabs.

  11. PAPIRUS - a computer code for FBR fuel performance analysis

    International Nuclear Information System (INIS)

    Kobayashi, Y.; Tsuboi, Y.; Sogame, M.

    1991-01-01

    The FBR fuel performance analysis code PAPIRUS has been developed to design fuels for demonstration and future commercial reactors. A pellet structural model was developed to describe the generation, depletion and transport of vacancies and atomic elements in unified fashion. PAPIRUS results in comparison with the power - to - melt test data from HEDL showed validity of the code at the initial reactor startup. (author)

  12. A study on properties-performances of coated particle fuel and on-line DB establishment

    International Nuclear Information System (INIS)

    Kim, Yong Soo; Lee, Hyo Cheol; Jang, Jeong Nam; Kwon, Seok Hwan

    2007-03-01

    Recently national project for HTGR for hydrogen production has been kicked off. However, For the successful development of the high temperature gas cooled reactor high temperature and burn-up dependent properties of the reactor materials are essentially and crucially required. Therefore, it was proposed to build up the materials properties and fuel performance data base. In this study, a phase - 1 properties and performance DB for coated particle fuel was developed. This database report consists two sections: materials properties and fuel performance. The materials properties has three parts: kernel materials, carbide coating materials, and fuel elements and graphite matrix. UO2 and UCO belong to kernel materials while PyC, SiC, and ZrC comprises the coating materials section. Thermal, mechanical and physical properties data of these materials were collected, reviewed, and summarized. Additionally, the property change induced by manufacture process and irradiation were reviewed. Fuel performance data were also collected, reviewed, and analyzed based on the key phenomena and failure mechanism, These performance data are divided into two: normal and accident. All of these data will be accessible in the pc based stand-alone system. These results will be directly used for HTGR fuel design and fabrication and preliminary fuel performance analysis under irradiation

  13. 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)

  14. Effects of dimethyl ether on the performance characteristics of a direct methanol fuel cell

    International Nuclear Information System (INIS)

    Seo, Sang Hern; Lee, Chang Sik

    2013-01-01

    Highlights: • Activation loss is significantly reduced in fuel cell with DME-methanol. • DME crossover through the membrane reduces. • The open circuit voltage of DME-methanol the fuel cell increases. • The overall efficiency of the mixed fuel cell is higher than that of DMFC. - Abstract: The objective of this study was to determine the effects of dimethyl ether (DME) on the performance characteristics of a direct methanol fuel cell. Impedance and crossover experiments were performed in order to investigate the performance losses such as ohmic loss, activation loss and crossover loss accurately. The DME was pressurized to 5 bar to supply with liquid phase was and blended with an aqueous methanol solution. In this experiment, the membrane electrode assembly (MEA) was composed of Nafion 115, anode catalyst loaded Pt–Ru and cathode catalyst loaded Pt-Black. Experimental results showed that fuel cells with DME-methanol enhanced performance when compared to fuel cells with methanol only. Such performance enhancement was due to a decrease in activation losses by DME oxidation reactions. As the DME crossover through the membrane was reduced, the open circuit voltage (OCV) of the fuel cell increased. Other output characteristics are also discussed

  15. Irradiation performance of U-Pu-Zr metal fuels for liquid-metal-cooled reactors

    International Nuclear Information System (INIS)

    Tsai, H.; Cohen, A.B.; Billone, M.C.; Neimark, L.A.

    1994-10-01

    This report discusses a fuel system utilizing metallic U-Pu-Zr alloys which has been developed for advanced liquid metal-cooled reactors (LMRs). Result's from extensive irradiation testing conducted in EBR-II show a design having the following key features can achieve both high reliability and high burnup capability: a cast nominally U-20wt %Pu-10wt %Zr slug with the diameter sized to yield a fuel smear density of ∼75% theoretical density, low-swelling tempered martensitic stainless steel cladding, sodium bond filling the initial fuel/cladding gap, and an as-built plenum/fuel volume ratio of ∼1.5. The robust performance capability of this design stems primarily from the negligible loading on the cladding from either fuel/cladding mechanical interaction or fission-gas pressure during the irradiation. The effects of these individual design parameters, e.g., fuel smear density, zirconium content in fuel, plenum volume, and cladding types, on fuel element performance were investigated in a systematic irradiation experiment in EBR-II. The results show that, at the discharge burnup of ∼11 at. %, variations on zirconium content or plenum volume in the ranges tested have no substantial effects on performance. Fuel smear density, on the other hand, has pronounced but countervailing effects: increased density results in greater cladding strain, but lesser cladding wastage from fuel/cladding chemical interaction

  16. Engine performance testing using variable RON95 fuel brands available in Malaysia

    Directory of Open Access Journals (Sweden)

    Mohd Riduan Aizuddin Fahmi

    2017-01-01

    Full Text Available There are various gasoline fuel producers available in Malaysia. The effects of fuel variations from different manufacturers on vehicle performance have always been a debate among users and currently the facts still remains inconclusive. Hence, this study focuses on analyzing various RON95 fuel brands available in the Malaysian market and finding the differences towards engine performance. In terms of engine output, the important data of power (hp and torque (Nm will be gathered by using an engine dynamometer. Another data that would also be taken into account is the knocking where the relative knock index can be measured in percentage using the knock sensor accelerometer. Results have shown that the performance of different fuel brands tested are indeed different albeit by only a small margin even though all fuels are categorized with the same octane rating. The power and torque results also imply that both are influenced by the amount of vibration generated due to engine knocking. Based from the overall outcome, consumers would not need to only focus on a certain type of gasoline brand as all differentiates the engine performance marginally.

  17. Effects of chemical equilibrium on turbine engine performance for various fuels and combustor temperatures

    Science.gov (United States)

    Tran, Donald H.; Snyder, Christopher A.

    1992-01-01

    A study was performed to quantify the differences in turbine engine performance with and without the chemical dissociation effects for various fuel types over a range of combustor temperatures. Both turbojet and turbofan engines were studied with hydrocarbon fuels and cryogenic, nonhydrocarbon fuels. Results of the study indicate that accuracy of engine performance decreases when nonhydrocarbon fuels are used, especially at high temperatures where chemical dissociation becomes more significant. For instance, the deviation in net thrust for liquid hydrogen fuel can become as high as 20 percent at 4160 R. This study reveals that computer central processing unit (CPU) time increases significantly when dissociation effects are included in the cycle analysis.

  18. Influence of pellet-clad-gap-size on LWR fuel rod performance

    International Nuclear Information System (INIS)

    Brzoska, B.; Fuchs, H.P.; Garzarolli, F.; Manzel, R.

    1979-01-01

    The as-fabricated pellet-clad-gap size varies due to fabricational tolerances of the cladding inner diameter and the pellet outer diameter. The consequences of these variations on the fuel rod behaviour are analyzed using the KWU fuel rod code CARO. The code predictions are compared with experimental results of special pathfinder test fuel rods irradiated in the OBRIGHEIM nuclear power plant. These test fuel rods include gap sizer in the range of 140 μm to 270 μm, prepressurization between 13 bar to 36 bar and Helium and Argon fill gases irradiated up to a local burnup of 35 MWd/kg(U). Post irradiation examination were performed at different burnups. CARC calculations have been performed with special emphasis in cladding creep down, fission gas release and pellet clad gap closure. (orig.)

  19. Performance comparison of low-temperature direct alcohol fuel cells with different anode catalysts

    Science.gov (United States)

    Zhou, W. J.; Zhou, B.; Li, W. Z.; Zhou, Z. H.; Song, S. Q.; Sun, G. Q.; Xin, Q.; Douvartzides, S.; Goula, M.; Tsiakaras, P.

    Low-temperature polymer electrolyte membrane fuel cells directly fed by methanol and ethanol were investigated employing carbon supported Pt, PtSn and PtRu as anode catalysts, respectively. Employing Pt/C as anode catalyst, both direct methanol fuel cell (DMFC) and direct ethanol fuel cell (DEFC) showed poor performances even in presence of high Pt loading on anode. It was found that the addition of Ru or Sn to the Pt dramatically enhances the electro-oxidation of both methanol and ethanol. It was also found that the single cell adopting PtRu/C as anode shows better DMFC performance, while PtSn/C catalyst shows better DEFC performance. The single fuel cell using PtSn/C as anode catalyst at 90 °C shows similar power densities whenever fueled by methanol or ethanol. The cyclic voltammetry (CV) and single fuel cell tests indicated that PtRu is more suitable for DMFC while PtSn is more suitable for DEFC.

  20. Evaluation and demonstration of methods for improved fuel utilization. Second semi-annual progress report, April 1, 1980-September 30, 1980

    International Nuclear Information System (INIS)

    1981-01-01

    Demonstrations are being performed in the Fort Calhoun reactor. The current program consists of two parts, one to demonstrate low leakage fuel management (SAVFUEL - Shimmed And Very Flexible Uranium Element Loading) and the other to demonstrate high burnup. The first part will demonstrate that the power duty cycle which is characteristic of SAVFUEL does not have a deleterious effect on fuel performance, while the second part will demonstrate that the peak rod average burnup of the current 14 x 14 fuel design can be increased to 45 GWD/T. A visual examination conducted at poolside was completed on four fuel assemblies which are scheduled to demonstrate the SAVFUEL power cycle and seventeen fuel assemblies which are scheduled to provide high burnup fuel performance data. Results of visual examinations, shoulder gap closure, fuel assembly growth, and fuel rod channel width measurements are reported which show excellent fuel performance for the high burnup; demonstration assemblies after four exposure cycles. These results support an additional exposure cycle for the high burnup demonstration assemblies which currently have an assembly average burnup up to 37 GWD/T

  1. Tennessee Valley Authority nuclear fuel performance

    International Nuclear Information System (INIS)

    Rahn, B.; Gardner, D.

    1985-01-01

    The Browns Ferry BWR units started operation in the early to mid 1970s, and the Sequoyah PWR units started in the very early 1980s. Operating dates, outage length, capacity factor, and cycle exposure are given for these units. This paper deals with fuel performance at the Browns Ferry and Sequoyah plants. Leaks, off-gas, fraction failed, iodine-131, and corrosion are covered. 14 figs

  2. Performance monitoring of zircaloy-4 square fuel channels at TAPS-1 and 2

    International Nuclear Information System (INIS)

    Akhtar, J.; Ramu, A.; Anilkumar, K.R.; Sharma, B.L.; Bhattacharjee, S.; Ramamurty, U.; Srivastava, S.P.; Prasad, P.N.; Anantharaman, K.

    2006-01-01

    Tarapur Atomic Power Station is a twin unit Boiling Water Reactors. The initial rated capacity of each unit was 210 MWe. Subsequently due to Secondary Steam Generator tube leak problem, the units were de-rated to 160 MWe in the year 1984-85. The station has completed 36 years of successful commercial operation. TAPS reactor fuel channels are made of Zircaloy-4, material. These are used along with 6x6 array nuclear fuel assemblies. The fuel channels need to be discharged once it reaches an optimum exposure limit and based on the surveillance programme, which monitors the channels performance. NFC has indigenously developed fuel channels for TAPS and these are at various stages of exposure in both the reactor cores. The performance review of these channels was carried out by the experts from TAPS-Site, NPCIL-ED and RED, BARC. The two major factors, which affect fuel channels performance, are (a) Bulge and (b) Bow. The phenomenon of longitudinal bow occurs due to the neutron flux gradient across the channels faces. Studies made on this subject by General Electric (GE) indicated that this channel deflection occurs at a slow rate. Therefore, fuel channels surveillance programme is essential to check the irradiated fuel channels performance in order to replace the fuel channels once it reaches the optimum exposure limit. To estimate the useful life of irradiated fuel channels, channel deflection/bulge measurement inspection system and methodology was developed jointly by TAPS and Centre for Design and manufacture (CDM), BARC. This system was successfully deployed at TAPS. This paper briefly describes the developmental efforts made by Nuclear Fuel Complex (NFC), Hyderabad, NPCIL-Fuel Group, Engg.Directorate, RED/BARC, CDM/BARC. (author)

  3. AFA 2G and AFA 3G fuel rod performance analysis

    International Nuclear Information System (INIS)

    Lu Huaquan; Liu Tong; Jiao Yongjun; Pang Hua

    2002-01-01

    For 18-months fuel cycle strategy in GNPJVC DAYA BAY unit 1/2, by means of COCCINEL, the fuel rod performance for AFA 3G and AFA 2G in transition cycle is analyzed. The design criteria which should be respected in fuel rod design are included and the design methodology is introduced. All the criteria mentioned are verified and met

  4. Spent Fuel Performance Assessment and Research. Final Report of a Coordinated Research Project on Spent Fuel Performance Assessment and Research (SPAR-III) 2009–2014

    International Nuclear Information System (INIS)

    2015-10-01

    At the beginning of 2014, there were 437 nuclear power reactors in operation and 72 reactors under construction. To date, around 370 500 t (HM) (tonnes of heavy metal) of spent fuel have been discharged from reactors, and approximately 253 700 t (HM) are stored at various storage facilities. Although wet storage at reactor sites still dominates, the amount of spent fuel being transferred to dry storage technologies has increased significantly since 2005. For example, around 28% of the total fuel inventory in the United States of America is now in dry storage. Although the licensing for the construction of geological disposal facilities is under way in Finland, France and Sweden, the first facility is not expected to be available until 2025 and for most States with major nuclear programmes not for several decades afterwards. Spent fuel is currently accumulating at around 7000 t (HM) per year worldwide. The net result is that the duration of spent fuel storage has increased beyond what was originally foreseen. In order to demonstrate the safety of both spent fuel and the storage system, a good understanding of the processes that might cause deterioration is required. To address this, the IAEA continued the Coordinated Research Project (CRP) on Spent Fuel Performance Assessment and Research (SPAR-III) in 2009 to evaluate fuel and materials performance under wet and dry storage and to assess the impact of interim storage on associated spent fuel management activities (such as handling and transport). This has been achieved through: evaluating surveillance and monitoring programmes of spent fuel and storage facilities; collecting and exchanging relevant experience of spent fuel storage and the impact on associated spent fuel management activities; facilitating the transfer of knowledge by documenting the technical basis for spent fuel storage; creating synergy among research projects of the participating Member States; and developing the capability to assess the impact

  5. Microstructured Electrolyte Membranes to Improve Fuel Cell Performance

    Science.gov (United States)

    Wei, Xue

    Fuel cells, with the advantages of high efficiency, low greenhouse gas emission, and long lifetime are a promising technology for both portable power and stationary power sources. The development of efficient electrolyte membranes with high ionic conductivity, good mechanical durability and dense structure at low cost remains a challenge to the commercialization of fuel cells. This thesis focuses on exploring novel composite polymer membranes and ceramic electrolytes with the microstructure engineered to improve performance in direct methanol fuel cells (DMFCs) and solid oxide fuel cells (SOFCs), respectively. Polymer/particle composite membranes hold promise to meet the demands of DMFCs at lower cost. The structure of composite membranes was controlled by aligning proton conducting particles across the membrane thickness under an applied electric field. The field-induced structural changes caused the membranes to display an enhanced water uptake, proton conductivity, and methanol permeability in comparison to membranes prepared without an applied field. Although both methanol permeability and proton conductivity are enhanced by the applied field, the permeability increase is relatively lower than the proton conductivity improvement, which results in enhanced proton/methanol selectivity and improved DMFC performance. Apatite ceramics are a new class of fast ion conductors being studied as alternative SOFC electrolytes in the intermediate temperature range. An electrochemical/hydrothermal deposition method was developed to grow fully dense apatite membranes containing well-developed crystals with c-axis alignment to promote ion conductivity. Hydroxyapatite seed crystals were first deposited onto a metal substrate electrochemically. Subsequent ion substitution during the hydrothermal growth process promoted the formation of dense, fully crystalline films with microstructure optimal for ion transport. The deposition parameters were systematically investigated, such as

  6. Fuel Cell Handbook, Fifth Edition

    Energy Technology Data Exchange (ETDEWEB)

    Energy and Environmental Solutions

    2000-10-31

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

  7. Fuel properties effect on the performance of a small high temperature rise combustor

    Science.gov (United States)

    Acosta, Waldo A.; Beckel, Stephen A.

    1989-01-01

    The performance of an advanced small high temperature rise combustor was experimentally determined at NASA-Lewis. The combustor was designed to meet the requirements of advanced high temperature, high pressure ratio turboshaft engines. The combustor featured an advanced fuel injector and an advanced segmented liner design. The full size combustor was evaluated at power conditions ranging from idle to maximum power. The effect of broad fuel properties was studied by evaluating the combustor with three different fuels. The fuels used were JP-5, a blend of Diesel Fuel Marine/Home Heating Oil, and a blend of Suntec C/Home Heating Oil. The fuel properties effect on the performance of the combustion in terms of pattern factor, liner temperatures, and exhaust emissions are documented.

  8. Fuel performance computer code simulation of steady-state and transient regimes of the stainless steel fuel rods

    International Nuclear Information System (INIS)

    Gomes, Daniel de Souza

    2014-01-01

    The immediate cause of the accident at the Fukushima Daiichi nuclear plant in March 2011 was the meltdown of the reactor core. During this process, the zirconium cladding of the fuel reacts with water, producing a large amount of hydrogen. This hydrogen, combined with volatile radioactive materials leaked from the containment vessel and entered the building of the reactor, resulting in explosions. In the past, stainless steel was used as the coating in many pressurized water reactors (PWR) under irradiation and their performance was excellent, however, the stainless steel was replaced by a zirconium-based alloy as a coating material mainly due to its lower section shock-absorbing neutrons. Today, the stainless steel finish appears again as a possible solution for security issues related to the explosion and hydrogen production. The objective of this thesis is to discuss the performance under irradiation of fuel rods using stainless steel as a coating material. The results showed that stainless steel rods exhibit lower temperatures and higher fuel pellet width of the gap - coating the coated rods Zircaloy and this gap does not close during the irradiation. The thermal performance of the two fuel rods is very similar, and the penalty of increased absorption of neutrons due to the use of stainless steel can be offset by the combination of a small increase in the enrichment of U- 235 and changes in the size of the spacing between the fuel rods. (author)

  9. Validating the BISON fuel performance code to integral LWR experiments

    Energy Technology Data Exchange (ETDEWEB)

    Williamson, R.L., E-mail: Richard.Williamson@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Gamble, K.A., E-mail: Kyle.Gamble@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Perez, D.M., E-mail: Danielle.Perez@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Novascone, S.R., E-mail: Stephen.Novascone@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Pastore, G., E-mail: Giovanni.Pastore@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Gardner, R.J., E-mail: Russell.Gardner@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Hales, J.D., E-mail: Jason.Hales@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Liu, W., E-mail: Wenfeng.Liu@anatech.com [ANATECH Corporation, 5435 Oberlin Dr., San Diego, CA 92121 (United States); Mai, A., E-mail: Anh.Mai@anatech.com [ANATECH Corporation, 5435 Oberlin Dr., San Diego, CA 92121 (United States)

    2016-05-15

    Highlights: • The BISON multidimensional fuel performance code is being validated to integral LWR experiments. • Code and solution verification are necessary prerequisites to validation. • Fuel centerline temperature comparisons through all phases of fuel life are very reasonable. • Accuracy in predicting fission gas release is consistent with state-of-the-art modeling and the involved uncertainties. • Rod diameter comparisons are not satisfactory and further investigation is underway. - Abstract: BISON is a modern finite element-based nuclear fuel performance code that has been under development at Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and has been used to analyze a variety of fuel forms in 1D spherical, 2D axisymmetric, or 3D geometries. Code validation is underway and is the subject of this study. A brief overview of BISON's computational framework, governing equations, and general material and behavioral models is provided. BISON code and solution verification procedures are described, followed by a summary of the experimental data used to date for validation of Light Water Reactor (LWR) fuel. Validation comparisons focus on fuel centerline temperature, fission gas release, and rod diameter both before and following fuel-clad mechanical contact. Comparisons for 35 LWR rods are consolidated to provide an overall view of how the code is predicting physical behavior, with a few select validation cases discussed in greater detail. Results demonstrate that (1) fuel centerline temperature comparisons through all phases of fuel life are very reasonable with deviations between predictions and experimental data within ±10% for early life through high burnup fuel and only slightly out of these bounds for power ramp experiments, (2) accuracy in predicting fission gas release appears to be consistent with state-of-the-art modeling and with the involved uncertainties and (3) comparison

  10. Multi-Dimensional Simulation of LWR Fuel Behavior in the BISON Fuel Performance Code

    Science.gov (United States)

    Williamson, R. L.; Capps, N. A.; Liu, W.; Rashid, Y. R.; Wirth, B. D.

    2016-11-01

    Nuclear fuel operates in an extreme environment that induces complex multiphysics phenomena occurring over distances ranging from inter-atomic spacing to meters, and times scales ranging from microseconds to years. To simulate this behavior requires a wide variety of material models that are often complex and nonlinear. The recently developed BISON code represents a powerful fuel performance simulation tool based on its material and physical behavior capabilities, finite-element versatility of spatial representation, and use of parallel computing. The code can operate in full three dimensional (3D) mode, as well as in reduced two dimensional (2D) modes, e.g., axisymmetric radial-axial ( R- Z) or plane radial-circumferential ( R- θ), to suit the application and to allow treatment of global and local effects. A BISON case study was used to illustrate analysis of Pellet Clad Mechanical Interaction failures from manufacturing defects using combined 2D and 3D analyses. The analysis involved commercial fuel rods and demonstrated successful computation of metrics of interest to fuel failures, including cladding peak hoop stress and strain energy density. In comparison with a failure threshold derived from power ramp tests, results corroborate industry analyses of the root cause of the pellet-clad interaction failures and illustrate the importance of modeling 3D local effects around fuel pellet defects, which can produce complex effects including cold spots in the cladding, stress concentrations, and hot spots in the fuel that can lead to enhanced cladding degradation such as hydriding, oxidation, CRUD formation, and stress corrosion cracking.

  11. Demonstration of uncertainty quantification and sensitivity analysis for PWR fuel performance with BISON

    International Nuclear Information System (INIS)

    Zhang, Hongbin; Zhao, Haihua; Zou, Ling; Burns, Douglas; Ladd, Jacob

    2017-01-01

    BISON is an advanced fuels performance code being developed at Idaho National Laboratory and is the code of choice for fuels performance by the U.S. Department of Energy (DOE)’s Consortium for Advanced Simulation of Light Water Reactors (CASL) Program. An approach to uncertainty quantification and sensitivity analysis with BISON was developed and a new toolkit was created. A PWR fuel rod model was developed and simulated by BISON, and uncertainty quantification and sensitivity analysis were performed with eighteen uncertain input parameters. The maximum fuel temperature and gap conductance were selected as the figures of merit (FOM). Pearson, Spearman, and partial correlation coefficients were considered for all of the figures of merit in sensitivity analysis. (author)

  12. Demonstration of Uncertainty Quantification and Sensitivity Analysis for PWR Fuel Performance with BISON

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Hongbin; Ladd, Jacob; Zhao, Haihua; Zou, Ling; Burns, Douglas

    2015-11-01

    BISON is an advanced fuels performance code being developed at Idaho National Laboratory and is the code of choice for fuels performance by the U.S. Department of Energy (DOE)’s Consortium for Advanced Simulation of Light Water Reactors (CASL) Program. An approach to uncertainty quantification and sensitivity analysis with BISON was developed and a new toolkit was created. A PWR fuel rod model was developed and simulated by BISON, and uncertainty quantification and sensitivity analysis were performed with eighteen uncertain input parameters. The maximum fuel temperature and gap conductance were selected as the figures of merit (FOM). Pearson, Spearman, and partial correlation coefficients were considered for all of the figures of merit in sensitivity analysis.

  13. Siemens capabilities to perform detailed fuel inspections during short outages

    International Nuclear Information System (INIS)

    Knecht, K.; Reparaz, A.

    1999-01-01

    Fuel inspection data are used to support development activities such as corrosion resistant cladding and advanced fuel assembly designs that will reach higher burnups. Increased inspection efforts are necessary to optimize fuel management and performance strategies. Additionally, there is an increasing trend to reduce outage time in Germany and abroad. Siemens has recently developed several timesaving systems for rapid inspection of fuel assemblies and core components. Siemens' focus in developing these systems has been to obtain data in reduced reactor outage time while increasing both the volume and the quality of the measured data. Mast sipping for PWRs is used for identifying leaking fuel assemblies and allows early detection of leaks during downloading of the fuel assemblies from the reactor. An In-Core sipping system for BWRs based on a hood technique to allow testing a full core within 16 hours is under development. (authors)

  14. High temperature compression tests performed on doped fuels

    Energy Technology Data Exchange (ETDEWEB)

    Duguay, C.; Mocellin, A.; Dehaudt, P. [Commissariat a l`Energie Atomique, CEA Grenoble (France); Fantozzi, G. [INSA Lyon - GEMPPM, Villeurbanne (France)

    1997-12-31

    The use of additives of corundum structure M{sub 2}O{sub 3} (M=Cr, Al) is an effective way of promoting grain growth of uranium dioxide. The high-temperature compressive deformation of large-grained UO{sub 2} doped with these oxides has been investigated and compared with that of pure UO{sub 2} with a standard microstructure. Such doped fuels are expected to exhibit enhanced plasticity. Their use would therefore reduce the pellet-cladding mechanical interaction and thus improve the performances of the nuclear fuel. (orig.) 5 refs.

  15. Poolside fuel assembly inspection campaigns performed at Kernkraftwerk Leibstadt during summer 1997

    International Nuclear Information System (INIS)

    Zwicky, H.U.; Wiktor, C.G.; Schrire, D.

    1998-01-01

    In order to minimise fuel cycle costs, fuel assembly discharge burnup and average U-235 enrichment were increasing over past years in the Kernkraftwerk Leibstadt (KKL) plant. In parallel, high burnup verification programs were defined in collaboration with fuel suppliers. The aim of these programs is to demonstrate safe and reliable fuel performance up to the designed burnup limit and to identify any problems in due time. This is not only achieved by detailed poolside inspections of lead test assemblies, but also by hot cell post-irradiation examination of selected rods. In the frame of a hot cell examination campaign, enhanced localised corrosion in the vicinity of spacers on SVEA-96 fuel rods was identified in May 1997 as a potential problem. The average rod burnup of the investigated rods was around 50 MWd/kgU after 5 one year cycles of operation. As fuel operation up to six cycles is foreseen in KKLs fuel management plants, the risk of fuel failures caused by enhanced localised corrosion could not be excluded. An action plan was therefore developed in order to identify the root cause. Part of the action plan were two poolside inspection campaigns: 1. Visual inspection of 38 assemblies unloaded during refuelling outage 1996 after 5 cycles in operation. This campaign was performed in June 1997. It gave a broader data base to develop a concept for fuel management for the upcoming refuelling outage scheduled in August 1997. 2. Visual inspection, oxide layer thickness measurements, crud sampling and rod diameter measurements on 29 assemblies with different operation histories. This campaign was performed during the outage. A large portion of the inspected bundles was re-inserted for continued operation. The collected data confirmed that assumptions made for reload licensing and safety analyses were conservative. The inspection campaigns performed at KKL during summer 1997 by ABB Atom demonstrated that it is possible to address unexpected problems in a short time

  16. Long Term Performance Study of a Direct Methanol Fuel Cell Fed with Alcohol Blends

    Directory of Open Access Journals (Sweden)

    Eleuterio Mora

    2013-01-01

    Full Text Available The use of alcohol blends in direct alcohol fuel cells may be a more environmentally friendly and less toxic alternative to the use of methanol alone in direct methanol fuel cells. This paper assesses the behaviour of a direct methanol fuel cell fed with aqueous methanol, aqueous ethanol and aqueous methanol/ethanol blends in a long term experimental study followed by modelling of polarization curves. Fuel cell performance is seen to decrease as the ethanol content rises, and subsequent operation with aqueous methanol only partly reverts this loss of performance. It seems that the difference in the oxidation rate of these alcohols may not be the only factor affecting fuel cell performance.

  17. 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.

  18. Fuel Cell Buses in U.S. Transit Fleets: Current Status 2012

    Energy Technology Data Exchange (ETDEWEB)

    Eudy, Leslie [National Renewable Energy Lab. (NREL), Golden, CO (United States); Chandler, Kevin [Battelle, Columbus, OH (United States); Gikakis, Christina [Federal Transit Administration, Washington, DC (United States)

    2012-11-01

    This report is the sixth in an annual series of reports that summarize the progress of fuel cell electric bus (FCEB) development in the United States and discuss the achievements and challenges of introducing fuel cell propulsion in transit. The report also provides a snapshot of current FCEB performance results over the last year.

  19. Analyses of expected rod performance during the dry storage of spent fuel

    International Nuclear Information System (INIS)

    Einziger, R.E.

    1982-08-01

    Within the next ten years, a number of utilities will be forced to increase their interim spent-fuel-storage capability or face the loss of full-core reserve. Dry storage is being considered to fill this need. This paper analyzes the fuel-rod-performance data supporting dry storage and discusses areas where there are still outstanding questions. Three storage temperature ranges (T 0 C, 250 0 C 0 C and T > 400 0 C), two atmospheres (inert, unlimited air) and two initial fuel-rod conditions (intact, breached) are considered. It is concluded that a fuel-performance data base exists that indicates that storage below 250 0 C can be accomplished with long-term fuel pellet and cladding stability. At higher temperatures, analytic studies and laboratory experiments are needed especially to extrapolate and interpret the result of demonstration tests. 2 figures, 2 tables

  20. Impact of nuclear fuel performance and reliability in a deregulated electricity market

    International Nuclear Information System (INIS)

    Barnoski, M.; Kaiser, B.

    1997-01-01

    US nuclear plants are now placing a great deal of importance on fuel performance and reliability because of their direct effect on a plant's capacity factor, availability, operating flexibility, outage duration, and worker exposure. This is reflected by utilities making fuel reliability a principal vendor evaluation criterion and making fuel performance an important bonus/penalty provision. Defect-free fuel, combined with ever improving design features, makes it possible for plants to operate at higher outputs while reducing operating and maintenance costs. Higher electrical output at lower cost is essential for US nuclear plants to meet the expected competition in a deregulated market. Through continued investment in manufacturing improvements and a vigilant, systematic design process, ABB's nuclear fuel has contributed to helping its nuclear utility customers reduce their cost of doing business. (author) 1 fig., 1 tab

  1. Methodological aspects of fuel performance system analysis at raw hydrocarbon processing plants

    Science.gov (United States)

    Kulbjakina, A. V.; Dolotovskij, I. V.

    2018-01-01

    The article discusses the methodological aspects of fuel performance system analysis at raw hydrocarbon (RH) processing plants. Modern RH processing facilities are the major consumers of energy resources (ER) for their own needs. To reduce ER, including fuel consumption, and to develop rational fuel system structure are complex and relevant scientific tasks that can only be done using system analysis and complex system synthesis. In accordance with the principles of system analysis, the hierarchical structure of the fuel system, the block scheme for the synthesis of the most efficient alternative of the fuel system using mathematical models and the set of performance criteria have been developed on the main stages of the study. The results from the introduction of specific engineering solutions to develop their own energy supply sources for RH processing facilities have been provided.

  2. Assessment of dry storage performance of spent LWR fuel assemblies with increasing burnup

    International Nuclear Information System (INIS)

    Peehs, M.; Garzarolli, F.; Goll, W.

    1999-01-01

    Although the safety of a dry long-term spent fuel store is scarcely influenced if a few fuel rods start to leak during extended storage - since all confinement systems are designed to retain gaseous activity safely - it is a very conservative safety goal to avoid the occurrence of systematic rod defects. To assess the extended storage performance of a spent fuel assembly (FA), the experience can be collated into 3 storage modes: I - fast rate of temperature decrease δ max ≥ δ ≥ 300 deg. C, II - medium rate of decrease for the fuel rod dry storage temperature 300 deg. C > δ ≥ 200 deg. C, III - slow to negligible rate of temperature decrease for δ 2 -fuel are practically immobile during storage. Consequently all fission-product-driven defect mechanisms will not take place. The leading defect mechanism - also for fuel rods with increased burnup - remains creep due to the hoop strain resulting from the fuel rod internal fission gas pressure. Limiting the creep to its primary and secondary stages prevents fuel rod degradation. The allowable uniform strain of the cladding is 1 - 2%. Calculations were performed to predict the dry storage performance of fuel assemblies with a burnup ≤ 55 GW · d/tHM based on the fuel assemblies end of life (EOL)-data and on a representative curve T = f(t). The maximum allowable hot spot temperature of a fuel rod in the CASTOR V cask was between 348 deg. C (U FA) and 358 deg. C (MOX FA). The highest hoop strain predicted after 40 years of storage is 0.77% proving that spent LWR fuel dry storage is safe. (author)

  3. Sensitivity Analysis of FEAST-Metal Fuel Performance Code: Initial Results

    International Nuclear Information System (INIS)

    Edelmann, Paul Guy; Williams, Brian J.; Unal, Cetin; Yacout, Abdellatif

    2012-01-01

    This memo documents the completion of the LANL milestone, M3FT-12LA0202041, describing methodologies and initial results using FEAST-Metal. The FEAST-Metal code calculations for this work are being conducted at LANL in support of on-going activities related to sensitivity analysis of fuel performance codes. The objective is to identify important macroscopic parameters of interest to modeling and simulation of metallic fuel performance. This report summarizes our preliminary results for the sensitivity analysis using 6 calibration datasets for metallic fuel developed at ANL for EBR-II experiments. Sensitivity ranking methodology was deployed to narrow down the selected parameters for the current study. There are approximately 84 calibration parameters in the FEAST-Metal code, of which 32 were ultimately used in Phase II of this study. Preliminary results of this sensitivity analysis led to the following ranking of FEAST models for future calibration and improvements: fuel conductivity, fission gas transport/release, fuel creep, and precipitation kinetics. More validation data is needed to validate calibrated parameter distributions for future uncertainty quantification studies with FEAST-Metal. Results of this study also served to point out some code deficiencies and possible errors, and these are being investigated in order to determine root causes and to improve upon the existing code models.

  4. Consequences of Fuel Failure on Criticality Safety of Used Nuclear Fuel

    International Nuclear Information System (INIS)

    Marshall, William J.; Wagner, John C.

    2012-09-01

    This report documents work performed for the Department of Energy's Office of Nuclear Energy (DOENE) Fuel Cycle Technologies Used Fuel Disposition Campaign to assess the impact of fuel reconfiguration due to fuel failure on the criticality safety of used nuclear fuel (UNF) in storage and transportation casks. This work was motivated by concerns related to the potential for fuel degradation during extended storage (ES) periods and transportation following ES, but has relevance to other potential causes of fuel reconfiguration. Commercial UNF in the United States is expected to remain in storage for longer periods than originally intended. Extended storage time and irradiation of nuclear fuel to high-burnup values (>45 GWd/t) may increase the potential for fuel failure during normal and accident conditions involving storage and transportation. Fuel failure, depending on the severity, can result in changes to the geometric configuration of the fuel, which has safety and regulatory implications for virtually all aspects of a UNF storage and transport system's performance. The potential impact of fuel reconfiguration on the safety of UNF in storage and transportation is dependent on the likelihood and extent of the fuel reconfiguration, which is not well understood and is currently an active area of research. The objective of this work is to assess and quantify the impact of postulated failed fuel configurations on the criticality safety of UNF in storage and transportation casks. Although this work is motivated by the potential for fuel degradation during ES periods and transportation following ES, it has relevance to fuel reconfiguration due to the effects of high burnup. Regardless of the ultimate disposition path, UNF will need to be transported at some point in the future. To investigate and quantify the impact of fuel reconfiguration on criticality safety limits, which are given in terms of the effective neutron multiplication factor, a set of failed fuel

  5. Hydrogen as automotive fuel

    International Nuclear Information System (INIS)

    Ambrosini, G.; Ciancia, A.; Pede, G.; Brighigna, M.

    1993-01-01

    Hydrogen fueled vehicles may just be the answer to the air pollution problem in highly polluted urban environments where the innovative vehicle's air pollution abatement characteristics would justify its high operating costs as compared with those of conventional automotive alternatives. This paper examines the feasibility of hydrogen as an automotive fuel by analyzing the following aspects: the chemical-physical properties of hydrogen in relation to its use in internal combustion engines; the modifications necessary to adapt internal combustion engines to hydrogen use; hydrogen fuel injection systems; current production technologies and commercialization status of hydrogen automotive fuels; energy efficiency ratings; environmental impacts; in-vehicle storage systems - involving the use of hydrides, high pressure systems and liquid hydrogen storage systems; performance in terms of pay-load ratio; autonomous operation; and operating costs. With reference to recent trial results being obtained in the USA, an assessment is also made of the feasibility of the use of methane-hydrogen mixtures as automotive fuels. The paper concludes with a review of progress being made by ENEA (the Italian Agency for New Technology, Energy and the Environment) in the development of fuel storage and electronic fuel injection systems for hydrogen powered vehicles

  6. The RERTR [Reduced Enrichment Research and Test Reactor] Program: Progress and plans

    International Nuclear Information System (INIS)

    Travelli, A.

    1987-01-01

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

  7. 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)

  8. 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

  9. Experimental study of dual fuel engine performance using variable LPG composition and engine parameters

    International Nuclear Information System (INIS)

    Elnajjar, Emad; Selim, Mohamed Y.E.; Hamdan, Mohammad O.

    2013-01-01

    Highlights: • The effect of using variable LPG is studied. • Five fuels with propane to butane % volume ratio are: 100-70-55-25-0. • 100% Propane composition shows the highest noise levels with similar performance. • At 45° BTDC injection timing 55% Propane LPG the only fuel experience knocking. • LPG fuels gave similar engine performance, with differences in levels of noise. - Abstract: The present work investigates experimentally the effect of LPG fuel with different composition and engine parameters on the performance of a dual compression engine. Five different blends of LPG fuels are used with Propane to Butane volume ratio of 100:0, 70:30, 55:45, 25:75, and 0:100. A single cylinder, naturally aspirated, four strokes, indirectly injected, water cooled modified Ricardo E6 engine, is used in this study. The study is carried out by measuring the cylinder pressure, engine load, engine speed, crank angle, and the fuel’s flow rate. The engine performance under variable LPG fuel composition, engine load, pilot fuel injection timing, compression ratio, pilot fuel mass and engine speed, are estimated by comparing the following engine parameters: the cylinder maximum pressure, the indicated mean effective pressure, the maximum rate of pressure rise, and the thermal efficiency. The experimental data indicates that the engine parameters are playing a major role on the engine’s performance. Different LPG fuel composition did not show a major effect on the engine efficiency but directly impacted the levels of generated combustion noise

  10. Greek research reactor performance characteristics after addition of beryllium reflector and LEU fuel

    International Nuclear Information System (INIS)

    Deen, J.R.; Snelgrove, J.L.; Papastergiou, C.

    1992-01-01

    The GRR-1 is a 5-MW pool-type, light-water-moderated and-cooled reactor fueled with MTR-type fuel elements. Recently received Be reflector blocks will soon be added to the core to add additional reactivity until fresh LEU fuel arrives. REBUS-3 xy fuel cycle analyses, using burnup dependent cross sections, were performed to assist in fuel management decisions for the water- and Be-reflected HEU nonequilibrium cores. Cross sections generated by EPRI-CELL have been benchmarked to identical VIM Monte Carlo models. The size of the Be-reflected LEU core has been reduced to 30 elements compared to 35 for the HEU water-reflected core, and an equilibrium cycle calculation has been performed

  11. Recent IAEA activities on CANDU-PHWR fuels and fuel cycles

    International Nuclear Information System (INIS)

    Inozemtsev, V.; Ganguly, C.

    2005-01-01

    Pressurized Heavy Water Reactors (PHWR), widely known as CANDU, are in operation in Argentina, Canada, China, India, Pakistan, Republic of Korea and Romania and account for about 6% of the world's nuclear electricity production. The CANDU reactor and its fuel have several unique features, like horizontal calandria and coolant tubes, on-power fuel loading, thin-walled collapsible clad coated with graphite on the inner surface, very high density (>96%TD) natural uranium oxide fuel and amenability to slightly enriched uranium oxide, mixed uranium plutonium oxide (MOX), mixed thorium plutonium oxide, mixed thorium uranium (U-233) oxide and inert matrix fuels. Several Technical Working Groups (TWG) of IAEA periodically discuss and review CANDU reactors, its fuel and fuel cycle options. These include TWGs on water-cooled nuclear power reactor Fuel Performance and Technology (TWGFPT), on Nuclear Fuel Cycle Options and spent fuel management (TWGNFCO) and on Heavy Water Reactors (TWGHWR). In addition, IAEA-INPRO project also covers Advanced CANDU Reactors (ACR) and DUPIC fuel cycles. The present paper summarises the Agency's activities in CANDU fuel and fuel cycle, highlighting the progress during the last two years. In the past we saw HWR and LWR technologies and fuel cycles separate, but nowadays their interaction is obviously growing, and their mutual influence may have a synergetic character if we look at the world nuclear fuel cycle as at an integrated system where the both are important elements in line with fast neutron, gas cooled and other advanced reactors. As an international organization the IAEA considers this challenge and makes concrete steps to tackle it for the benefit of all Member States. (author)

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

    International Nuclear Information System (INIS)

    Talamo, Alberto; Gohar, Yousry

    2011-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  14. FAST: a combined NOC and transient fuel performance model using a commercial FEM environment

    Energy Technology Data Exchange (ETDEWEB)

    Prudil, A.; Bell, J.; Oussoren, A.; Chan, P. [Royal Military College of Canada, Kingston, ON (Canada); Lewis, B. [Univ. of Ontario Inst. of Tech., Oshawa, ON (Canada)

    2014-07-01

    The Fuel And Sheath modelling Tool (FAST) is a combined normal operating conditions (NOC) and transient fuel performance code developed on the COMSOL Multiphysics finite-element platform. The FAST code has demonstrated excellent performance in proof of concept validation tests against experimental data and comparison to the ELESIM, ELESTRES and ELOCA fuel performance codes. In this paper we discuss ongoing efforts to expand the capabilities of the code to include multiple pellet geometries, model stress-corrosion cracking phenomena and modelling of advanced fuels composed of mixed oxides of thorium, uranium, and plutonium for the Canadian Supercritical Water Reactor (SCWR). (author)

  15. Performance of LMFBR fuel pins with (Pu,Th)O/sub 2-x/ and UO2

    International Nuclear Information System (INIS)

    Lawrence, L.A.

    1983-09-01

    The irradiation performance of (Pu,Th)O/sub 2-x/ and UO 2 fueled pins for breeder reactor application were compared to the extensive performance data base for the (U,Pu)O/sub 2-x/ fuel system. Th-Pu and 238 U- 233 U based fuel systems were candidate fuel fertile/fissile isotopic combinations for development of alternatives to the current LMFBR fuel cycle. Initial screening tests were conducted in the EBR-II to obtain comparative performance data because of the limited experience with these fuel systems. In some cases, 235 U was used as a substitute for 233 U because of the difficulties in fabrication of available 233 U due to its high gamma ray emission rate

  16. Atomic scale simulations for improved CRUD and fuel performance modeling

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, Anders David Ragnar [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Cooper, Michael William Donald [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-01-06

    A more mechanistic description of fuel performance codes can be achieved by deriving models and parameters from atomistic scale simulations rather than fitting models empirically to experimental data. The same argument applies to modeling deposition of corrosion products on fuel rods (CRUD). Here are some results from publications in 2016 carried out using the CASL allocation at LANL.

  17. The analysis of the annular fuel performance in steady state condition by using AFPAC code

    International Nuclear Information System (INIS)

    He Xiaojun; Ji Songtao; Zhang Yingchao

    2012-01-01

    The fuel performance code AFPAC v1.0 is used to analyze annular fuel's behavior under steady state conditions, including neutronics, thermal hydraulic, rod deformation, fission gas release and rod internal pressure. The calculation results show that: 1) Annular fuel has a good steady irradiation performance at 150% power level as current LWRs' with burnup up to 50 GWd/t, and all parameters, such as temperature, rod internal pressure and rod deformation, are meet the rod design criteria for current fuel of PWRs: 2) Compared to the solid fuel under the same irradiation condition. annular fuel has lower temperature, smaller deformation, lower fission gas release and lower pressure at EOL. From the point of view of steady irradiation performance, the safety of reactors can significantly improved by u sing the annular fuel. (authors)

  18. Performance analysis of a mixed nitride fuel system for an advanced liquid metal reactor

    International Nuclear Information System (INIS)

    Lyon, W.F.; Baker, R.B.; Leggett, R.D.

    1991-01-01

    In this paper, the conceptual development and analysis of a proposed mixed nitride driver and blanket fuel system for a prototypic advanced liquid metal reactor design is performed. As a first step, an intensive literature survey is completed on the development and testing of nitride fuel systems. Based on the results of this survey, prototypic mixed nitride fuel and blanket pins is designed and analyzed using the SIEX computer code. The analysis predicts that the nitride fuel consistently operated at peak temperatures and cladding strain levels that compared quite favorably with competing fuel designs. These results, along with data available in the literature on nitride fuel performance, indicate that a nitride fuel system should offer enhanced capabilities for advanced liquid metal reactors

  19. Performance analysis of a mixed nitride fuel system for an advanced liquid metal reactor

    International Nuclear Information System (INIS)

    Lyon, W.F.; Baker, R.B.; Leggett, R.D.

    1990-11-01

    The conceptual development and analysis of a proposed mixed nitride driver and blanket fuel system for a prototypic advanced liquid metal reactor design has been performed. As a first step, an intensive literature survey was completed on the development and testing of nitride fuel systems. Based on the results of this survey, prototypic mixed nitride fuel and blanket pins were designed and analyzed using the SIEX computer code. The analysis predicted that the nitride fuel consistently operated at peak temperatures and cladding strain levels that compared quite favorably with competing fuel designs. These results, along with data available in the literature on nitride fuel performance, indicate that a nitride fuel system should offer enhanced capabilities for advanced liquid metal reactors. 13 refs., 10 figs., 2 tabs

  20. Evaluation and demonstration of methods for improved fuel utilization. First semi-annual progress report, September 1979-March 1980

    International Nuclear Information System (INIS)

    Decher, U.

    1980-01-01

    Demonstrations of improved fuel management and burnup are being performed in the Fort Calhoun reactor. More efficient fuel management will be achieved through the implementation of a low leakage concept called SAVFUEL (Shimmed And Very Flexible Uranium Element Loading), which is expected to reduce uranium requirements by 2 to 4%. The burnup will be increased sufficiently to reduce uranium requirements by 5 to 15%. Four fuel assemblies scheduled to demonstrate the SAVFUEL duty cycle and loaded into the core in December 1978 were inspected visually prior to their second exposure cycle. In addition, seventeen fuel assemblies were inspected after their fourth exposure cycle having achieved assembly average burnup up to 36 GWD/T. One assembly has been reinserted into Cycle 6 for a fifth exposure cycle. The preliminary results of all visual fuel inspections which appear to show excellent fuel rod performance are presented in this report. This report also contains the results of a licensing activity which was performed to allow insertion of a highly burned assembly into the reactor for a fifth irradiation cycle