WorldWideScience

Sample records for fuel performance annual

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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.

  7. 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

  8. 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

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

    International Nuclear Information System (INIS)

    Crouthamel, C.E.

    1978-10-01

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

  10. 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

  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. FY2015 ceramic fuels development annual highlights

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-22

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

  13. FY2016 Ceramic Fuels Development Annual Highlights

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-01-24

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

  14. 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

  15. 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

  16. Annual Performance Report - FY 2011

    Science.gov (United States)

    This report summarizes OIG activity, performance, results, and challenges, and provides a financial accounting of resources for fiscal year (FY) 2011 compared to our FY 2011 annual performance targets.

  17. 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

  18. 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.

  19. Fifth annual report to congress. Federal alternative motor fuels programs

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-09-01

    This report presents the status of the US Department of Energy`s alternative fuel vehicle demonstration and performance tracking programs being conducted in accordance with the Energy Policy and Conservation Act. These programs comprise the most comprehensive data collection effort ever undertaken on alternative transportation fuels and alternative fuel vehicles. The report summarizes tests and results from the fifth year. Electric vehicles are not included in these programs, and the annual report does not include information on them. Since the inception of the programs, great strides have been made in developing commercially viable alternative fuel vehicle technologies. However, as is the case in the commercialization of all new technologies, some performance problems have been experienced on vehicles involved in early demonstration efforts. Substantial improvements have been recorded in vehicle practicality, safety, and performance in real-world demonstrations. An aspect of particular interest is emissions output. Results from light duty alternative fuel vehicles have demonstrated superior inservice emissions performance. Heavy duty alternative fuel vehicles have demonstrated dramatic reductions in particulate emissions. However, emissions results from vehicles converted to run on alternative fuel have not been as promising. Although the technologies available today are commercially viable in some markets, further improvements in infrastructure and economics will result in greater market expansion. Information is included in this report on light and heavy duty vehicles, transit buses, vehicle conversions, safety, infrastructure support, vehicle availability, and information dissemination.

  20. 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)

  1. 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

  2. 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)

  3. 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)

  4. 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.

  5. 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.

  6. 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.

  7. 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

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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.

  13. 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.

  14. Canadian Fusion Fuels Technology Project annual report 93/94

    International Nuclear Information System (INIS)

    1994-01-01

    The Canadian Fusion Fuels Technology Project exists to develop fusion technologies and apply them worldwide in today's advanced fusion projects and to apply these technologies in fusion and tritium research facilities. CFFTP concentrates on developing capability in fusion fuel cycle systems, in tritium handling technologies and in remote handling. This is an annual report for CFFTP and as such also includes a financial report

  15. 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

  16. IEA-Advanced Motor Fuels Annual Report 2010

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-12-02

    The annual report from the IEA implementing agreement on Advanced Motor Fuels (AMF) describes the agreement, activities, and projects for the year. A section on the global situation for Advanced Motor Fuels includes country reports from each participating AMF member. A status report on each active annex for the agreement is also included, as is a message from the AMF Chairman. Final sections include an Outlook for Advanced Motor Fuels, further information, and a glossary of terms.

  17. 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

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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

  3. World nuclear fuel market. Seventeenth annual meeting

    International Nuclear Information System (INIS)

    Anon.

    1990-01-01

    The papers presented at the seventeenth World Nuclear Fuels Market meeting are cataloged individually. This volume includes information on the following areas of interest: historical and current aspects of the uranium and plutonium market with respect to supply and demand, pricing, spot market purchasing, and other market phenomena; impact of reprocessing and recycling uranium, plutonium, and mixed oxide fuels; role of individual countries in the market: Hungary, Germany, the Soviet Union, Czechoslovakia, France, and the US; the impact of public opinion and radioactive waste management on the nuclear industry, and a debate regarding long term versus short term contracting by electric utilities for uranium and enrichment services

  4. IEA-Advanced Motor Fuels Annual Report 2011

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-07-01

    The annual report from the IEA implementing agreement on Advanced Motor Fuels (AMF) describes what the agreement is about, how to join, various activities of the agreement, a message from the Chairman, and projects/annexes active for the year. An annual section covers the global situation for the topic of advanced motor fuels. Another section includes highlights coming from each country participating in AMF, and major sections relaying activities on each of the ongoing annexes. Information regarding participating delegations, contact information, publications resulting from AMF, and upcoming meetings rounds out the report.

  5. IEA-Advanced Motor Fuels Annual Report 2012

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-06-15

    The annual report from the IEA implementing agreement on Advanced Motor Fuels (AMF) describes what the agreement is about, how to join, various activities of the agreement, a message from the Chairman, and projects/annexes active for the year. An annual section covers the global situation for the topic of advanced motor fuels. Another section includes highlights coming from each country participating in AMF, and major sections relaying activities on each of the ongoing annexes. Information regarding participating delegations, contact information, publications resulting from AMF, and upcoming meetings rounds out the report.

  6. Liquid Fuels from Lignins: Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    Chum, H. L.; Johnson, D. K.

    1986-01-01

    This task was initiated to assess the conversion of lignins into liquid fuels, primarily of lignins relevant to biomass-to-ethanol conversion processes. The task was composed of a literature review of this area and an experimental part to obtain pertinent data on the conversion of lignins germane to biomass-to-ethanol conversion processes.

  7. Canadian Fusion Fuels Technology Project annual report 93/94

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-12-31

    The Canadian Fusion Fuels Technology Project exists to develop fusion technologies and apply them worldwide in today`s advanced fusion projects and to apply these technologies in fusion and tritium research facilities. CFFTP concentrates on developing capability in fusion fuel cycle systems, in tritium handling technologies and in remote handling. This is an annual report for CFFTP and as such also includes a financial report.

  8. 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.

  9. World nuclear fuel market. Eighteenth annual meeting

    International Nuclear Information System (INIS)

    Anon.

    1991-01-01

    The papers presented at the eighteenth World Nuclear Fuels Market meeting are cataloged separately. This volume includes information on the following areas of interest: world uranium enrichment capacity and enriched uranium inventories; the impact of new enrichment technologies; predictions of future market trends; non-proliferation aspects of nuclear trade; and a debate as to whether uranium can be successfully traded on a commodities exchange

  10. 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

  11. 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

  12. 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.)

  13. 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.

  14. 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.

  15. 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.

  16. 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)

  17. Thermal Performance Benchmarking: Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    Feng, Xuhui [National Renewable Energy Laboratory (NREL), Golden, CO (United States). Transportation and Hydrogen Systems Center

    2017-10-19

    In FY16, the thermal performance of the 2014 Honda Accord Hybrid power electronics thermal management systems were benchmarked. Both experiments and numerical simulation were utilized to thoroughly study the thermal resistances and temperature distribution in the power module. Experimental results obtained from the water-ethylene glycol tests provided the junction-to-liquid thermal resistance. The finite element analysis (FEA) and computational fluid dynamics (CFD) models were found to yield a good match with experimental results. Both experimental and modeling results demonstrate that the passive stack is the dominant thermal resistance for both the motor and power electronics systems. The 2014 Accord power electronics systems yield steady-state thermal resistance values around 42- 50 mm to the 2nd power K/W, depending on the flow rates. At a typical flow rate of 10 liters per minute, the thermal resistance of the Accord system was found to be about 44 percent lower than that of the 2012 Nissan LEAF system that was benchmarked in FY15. The main reason for the difference is that the Accord power module used a metalized-ceramic substrate and eliminated the thermal interface material layers. FEA models were developed to study the transient performance of 2012 Nissan LEAF, 2014 Accord, and two other systems that feature conventional power module designs. The simulation results indicate that the 2012 LEAF power module has lowest thermal impedance at a time scale less than one second. This is probably due to moving low thermally conductive materials further away from the heat source and enhancing the heat spreading effect from the copper-molybdenum plate close to the insulated gate bipolar transistors. When approaching steady state, the Honda system shows lower thermal impedance. Measurement results of the thermal resistance of the 2015 BMW i3 power electronic system indicate that the i3 insulated gate bipolar transistor module has significantly lower junction

  18. 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

  19. 2011 Annual Criticality Safety Program Performance Summary

    Energy Technology Data Exchange (ETDEWEB)

    Andrea Hoffman

    2011-12-01

    The 2011 review of the INL Criticality Safety Program has determined that the program is robust and effective. The review was prepared for, and fulfills Contract Data Requirements List (CDRL) item H.20, 'Annual Criticality Safety Program performance summary that includes the status of assessments, issues, corrective actions, infractions, requirements management, training, and programmatic support.' This performance summary addresses the status of these important elements of the INL Criticality Safety Program. Assessments - Assessments in 2011 were planned and scheduled. The scheduled assessments included a Criticality Safety Program Effectiveness Review, Criticality Control Area Inspections, a Protection of Controlled Unclassified Information Inspection, an Assessment of Criticality Safety SQA, and this management assessment of the Criticality Safety Program. All of the assessments were completed with the exception of the 'Effectiveness Review' for SSPSF, which was delayed due to emerging work. Although minor issues were identified in the assessments, no issues or combination of issues indicated that the INL Criticality Safety Program was ineffective. The identification of issues demonstrates the importance of an assessment program to the overall health and effectiveness of the INL Criticality Safety Program. Issues and Corrective Actions - There are relatively few criticality safety related issues in the Laboratory ICAMS system. Most were identified by Criticality Safety Program assessments. No issues indicate ineffectiveness in the INL Criticality Safety Program. All of the issues are being worked and there are no imminent criticality concerns. Infractions - There was one criticality safety related violation in 2011. On January 18, 2011, it was discovered that a fuel plate bundle in the Nuclear Materials Inspection and Storage (NMIS) facility exceeded the fissionable mass limit, resulting in a technical safety requirement (TSR) violation. The

  20. 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

  1. 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.)

  2. 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)

  3. 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)

  4. 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)

  5. 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

  6. 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.

  7. Commercial waste and spent fuel packaging program. Annual report

    International Nuclear Information System (INIS)

    Hakl, A.R.

    1981-10-01

    This document is a report of activities performed by Westinghouse Advanced Energy Systems Division - Nevada Operations in meeting subtask objectives described in the Nevada Nuclear Waste Storage Investigations (NNWSI) Project Plan and revised planning documentation for Fiscal Year (FY) 1981. Major activities included: completion of the first fuel exchange in the Spent Fuel Test - Climax program; plasma arc welder development; modification and qualification of a canister cutter; installation, and activation of a remote area monitor, constant air monitor and an alpha/beta/gamma counting system; qualification of grapples required to handle pressurized water reactor or boiling water reactor fuel and high level waste (HLW) logs; data acquisition from the 3 kilowatt soil temperature test, 2 kw fuel temperature test, and 2 kw drywell test; calorimetry of the fuel assembly used in the fuel temperature test; evaluation of moisture accumulation in the drywells and recommendations for proposed changes; revision of safety assessment document to include HLW log operations; preparation of quality assurance plan and procedures; development and qualification of all equipment and procedures to receive, handle and encapsulate both the HLW log and spent fuel for the basalt waste isolation program/near surface test facility program; preliminary studies of both the requirements to perform waste packaging for the test and evaluation facility and a cask storage program for the DOE Interim Spent Fuel Management program; and remote handling operations on radioactive source calibration in support of other contractors

  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. H2 and Fuel cell annual activity report - France 2011

    International Nuclear Information System (INIS)

    Lucchese, Paul; Julien, Marianne

    2012-01-01

    This report aims making better known the hydrogen and fuel cell technologies in France and their main actors (large groups, small and medium-sized enterprises, start-ups, research centres). After a presentation of the French energy context, it presents the national programme and strategic actions, and local programmes and initiatives. The next chapter presents the main results and events for the different fields of application: leading edge markets, transports, decentralized or residential stationary applications, hydrogen and renewable energies, portable applications, transverse domains. The annual activity and main results of the different actors are then presented: research and development, small or medium sized enterprises and start-ups, large groups

  10. 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.

  11. 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

  12. An annual performance appraisal for CERN

    CERN Multimedia

    2012-01-01

    The March Council session is the occasion for CERN’s annual performance appraisal, but instead of the MARS form familiar to CERN staff, the lab’s working document is an Annual Progress Report, linked to the Medium-Term Plan, matching achievements to objectives. This year, I think it’s fair to say, we were firmly able to tick the ‘achieved’ box.   Top of the list of objectives was the LHC, which exceeded expectations in 2011 by delivering over five times the anticipated luminosity for protons while improving the lead-ion integrated luminosity by an order of magnitude compared to 2010.  As a result, the experiments published over 190 papers and made a staggering 1900 conference presentations. Underpinning this was the Worldwide LHC Computing Grid, which routinely performs so well that we hardly notice it's there. Non-LHC physics also had its day in the sun in 2011, with experiments at the PS, SPS, AD, n_TOF and ISOLDE, as well ...

  13. 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

  14. 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

  15. 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

  16. 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)

  17. High performance nuclear fuel element

    International Nuclear Information System (INIS)

    Mordarski, W.J.; Zegler, S.T.

    1980-01-01

    A fuel-pellet composition is disclosed for use in fast breeder reactors. Uranium carbide particles are mixed with a powder of uraniumplutonium carbides having a stable microstructure. The resulting mixture is formed into fuel pellets. The pellets thus produced exhibit a relatively low propensity to swell while maintaining a high density

  18. Hydrogen engine performance analysis. First annual report

    Energy Technology Data Exchange (ETDEWEB)

    Adt, Jr., R. R.; Swain, M. R.; Pappas, J. M.

    1978-08-01

    Many problems associated with the design and development of hydrogen-air breathing internal combustion engines for automotive applications have been identified by various domestic and foreign researchers. This project addresses the problems identified in the literature, seeks to evaluate potential solutions to these problems, and will obtain and document a design data-base convering the performance, operational and emissions characteristics essential for making rational decisions regarding the selection and design of prototype hydrogen-fueled, airbreathing engines suitable for manufacture for general automotive use. Information is included on the operation, safety, emission, and cost characteristics of hydrogen engines, the selection of a test engine and testing facilities, and experimental results. Baseline data for throttled and unthrottled, carburetted, hydrogen engine configurations with and without exhaust gas recirculation and water injection are presented. In addition to basic data gathering concerning performance and emissions, the test program conducted was formulated to address in detail the two major problems that must be overcome if hydrogen-fueled engines are to become viable: flashback and comparatively high NO/sub x/ emissions at high loads. In addition, the results of other hydrogen engine investigators were adjusted, using accepted methods, in order to make comparisons with the results of the present study. The comparisons revealed no major conflicts. In fact, with a few exceptions, there was found to be very good agreement between the results of the various studies.

  19. 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

  20. 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.

  1. 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

  2. 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)

  3. 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.

  4. 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.)

  5. High performance reliability fuel pellet

    International Nuclear Information System (INIS)

    Beuchel, P.H.; Lee, Y.C.

    1989-01-01

    A fuel pellet for a nuclear reactor fuel rod is described comprising: a substantially cylindrical central section; a convex first end section smoothly joined to one axial end of the central section at a first junction, the first junction approximating a smooth and continuous curved surface; a concave second end section joined to the central section at a second junction, the second junction approximating a smooth and continuous curved surface, wherein the curvature of the concave second end section is conformed to the curvature of the convex first end section

  6. CANDU fuel performance and development

    International Nuclear Information System (INIS)

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

    1978-12-01

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

  7. 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)

  8. 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

  9. 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.

  10. Fuel element performance computer modelling

    International Nuclear Information System (INIS)

    Locke, D.H.

    1978-01-01

    The meeting was attended by 88 participants from 17 countries. Altogether 47 papers were presented. The majority of the presentations contained a description of the equations and solutions used to describe and evaluate some of the physical processes taking place in water reactor fuel pins under irradiation. At the same time, particular attention was paid to the ''bench marking'' of the codes wherein solutions arrived at for particular experiments are compared with the results at the experiments

  11. 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.

  12. 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)

  13. 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.

  14. Second annual report of the Canadian nuclear fuel waste management program

    International Nuclear Information System (INIS)

    Boulton, J.

    1980-12-01

    This report, the second of a series of annual reports, reviews in general terms the progress which has been achieved in the research and development program for the safe, permanent disposal of Canada's nuclear fuel wastes. The report summarizes activities over the past year, in the areas of public interaction, used fuel storage and transportation, immobilization of used fuel and fuel reprocessing wastes, research and development associated with deep underground disposal, and environmental and safety assessment. (auth)

  15. Fourth annual report of the Canadian nuclear fuel waste management program

    International Nuclear Information System (INIS)

    Rosinger, E.L.J.; Dixon, R.S.

    1982-12-01

    This report, the fourth of a series of annual reports, reviews the progress that has been made in the research and development program for the safe management and disposal of Canada's nuclear fuel waste. The report summarizes activities over the past year in the following areas: public interaction, used fuel storage and transportation, immobilization of used fuel and fuel recycle waste, geoscience research associated with deep underground disposal, environmental research, environmental and safety assessment

  16. Third annual report of the Canadian nuclear fuel waste management program

    International Nuclear Information System (INIS)

    Dixon, R.S.; Rosinger, E.L.J.

    1981-12-01

    This report, the third of a series of annual reports, reviews the progress that has been made in the research and development program for the safe management and disposal of Canada's nuclear fuel wastes. The report summarizes activities over the past year, in the areas of public interaction, irradiated fuel storage and transportation, immobilization of irradiated fuel and fuel recycle wastes, research and development associated with deep underground disposal, and environmental and safety assessment

  17. 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

  18. 1981 Annual Status Report. Plutonium fuels and actinide programme

    International Nuclear Information System (INIS)

    1981-01-01

    In this 1981 report the work carried out by the European Institute for Transuranium elements is reviewed. Main topics are: operation limits of plutonium fuels: swelling of advanced fuels, oxide fuel transients, equation of state of nuclear materials; actinide cycle safety: formation of actinides (FACT), safe handling of plutonium fuel (SHAPE), aspects of the head-end processing of carbide fuel (RECARB); actinide research: crystal chemistry, solid state studies, applied actinide research

  19. 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.

  20. 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

  1. 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.

  2. Proceedings of the fourth annual fuel cells contractors review meeting

    International Nuclear Information System (INIS)

    Huber, W.J.

    1992-07-01

    Objective of the program was to develop the essential technology for private sector commercialization of various fuel cell electrical generation systems, which promise high fuel efficiencies (40--60%), possibilities for cogeneration, modularity, possible urban siting, and low emissions. Purpose of this meeting was to provide the R and D participants in the DOE/Fossil Energy-sponsored Fuel Cells Program with a forum. With the near commercialization of phosphoric acid fuel cells, major emphasis was on molten carbonate and solid oxide fuel cells. 22 papers were given in 3 formal sessions: molten carbonate fuel cells; solid oxide fuel cells; and systems and phosphoric acid. In addition, the proceedings also include a welcome to METC address and comments on the Fuel Cells program from the viewpoint of EPRI and DOE's vehicular fuel cell program. Separate abstracts have been prepared

  3. 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)

  4. KWIKPLAN: a computer program for projecting the annual requirements of nuclear fuel cycle operations

    International Nuclear Information System (INIS)

    Salmon, R.; Kee, C.W.

    1977-06-01

    The computer code KWIKPLAN was written to facilitate the calculation of projected nuclear fuel cycle activities. Using given projections of power generation, the code calculates annual requirements for fuel fabrication, fuel reprocessing, uranium mining, and plutonium use and production. The code uses installed capacity projections and mass flow data for six types of reactors to calculate projected fuel cycle activities and inventories. It calculates fissile uranium and plutonium flows and inventories after allowing for an economy with limited reprocessing capacity and a backlog of unreprocessed fuel. All calculations are made on a quarterly basis; printed and punched output of the projected fuel cycle activities are made on an annual basis. Since the punched information is used in another code to determine waste inventories, the code punches a table from which the effective average burnup can be calculated for the fuel being reprocessed

  5. 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

  6. 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)

  7. 1982 Annual Status Report Plutonium Fuels and Actinide Programme

    International Nuclear Information System (INIS)

    Lindner, R.

    1983-01-01

    The programme of the Transuranium Institute has long included work on advanced fuels for fast breeder reactors. Study of the swelling of carbide and nitride fuels is now nearing completion, the retention of fission gases in bubbles of different sizes in the fuel having been quantified as function of burn-up and temperature. An important step forward has been achieved in the studies of the Equation of State of Nuclear Fuels up to 5000 K. Formation of some of the less abundant isotopes in PWR fuel has been determined experimentally. Aerosol formation during the fabrication of plutonium containing fuels, part of the activity Safe Handling of Plutonium Fuel has been studied. Head-End Processing of carbide fuels has continued experiments with high burn up mixed carbides. In the field of actinide research the preparation and characterisation of pure specimens is carried out. Effect of actinides on the properties of waste glasses is investigated

  8. 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

  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. 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)

  11. 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)

  12. Annual Report: Advanced Energy Systems Fuel Cells (30 September 2013)

    Energy Technology Data Exchange (ETDEWEB)

    Gerdes, Kirk; Richards, George

    2014-04-16

    The comprehensive research plan for Fuel Cells focused on Solid State Energy Conversion Alliance (SECA) programmatic targets and included objectives in two primary and focused areas: (1) investigation of degradation modes exhibited by the anode/electrolyte/cathode (AEC), development of computational models describing the associated degradation rates, and generation of a modeling tool predicting long term AEC degradation response; and (2) generation of novel electrode materials and microstructures and implementation of the improved electrode technology to enhance performance. In these areas, the National Energy Technology Laboratory (NETL) Regional University Alliance (RUA) team has completed and reported research that is significant to the SECA program, and SECA continued to engage all SECA core and SECA industry teams. Examination of degradation in an operational solid oxide fuel cell (SOFC) requires a logical organization of research effort into activities such as fundamental data gathering, tool development, theoretical framework construction, computational modeling, and experimental data collection and validation. Discrete research activity in each of these categories was completed throughout the year and documented in quarterly reports, and researchers established a framework to assemble component research activities into a single operational modeling tool. The modeling framework describes a scheme for categorizing the component processes affecting the temporal evolution of cell performance, and provides a taxonomical structure of known degradation processes. The framework is an organizational tool that can be populated by existing studies, new research completed in conjunction with SECA, or independently obtained. The Fuel Cell Team also leveraged multiple tools to create cell performance and degradation predictions that illustrate the combined utility of the discrete modeling activity. Researchers first generated 800 continuous hours of SOFC experimental

  13. 2015 DOE Hydrogen and Fuel Cells Program Annual Merit Review and Peer Evaluation Report

    Energy Technology Data Exchange (ETDEWEB)

    none,

    2015-10-01

    This report summarizes comments from the Peer Review Panel at the 2015 DOE Hydrogen and Fuel Cells Program Annual Merit Review, held on June 8-12, 2015, in Arlington, Virginia. It covers the program areas of hydrogen production and delivery; hydrogen storage; fuel cells; manufacturing R&D; technology validation; safety, codes, and standards; market transformation; and systems analysis.

  14. 2011 DOE Hydrogen and Fuel Cells Program Annual Merit Review and Peer Evaluation Report

    Energy Technology Data Exchange (ETDEWEB)

    none,

    2011-09-01

    This report summarizes comments from the Peer Review Panel at the 2011 DOE Hydrogen and Fuel Cells Program Annual Merit Review, held on May 9-13, 2011, in Arlington, Virginia. It covers the program areas of hydrogen production and delivery; hydrogen storage; fuel cells; manufacturing R&D; technology validation; safety, codes, and standards; education; market transformation; and systems analysis.

  15. 2013 DOE Hydrogen and Fuel Cells Program Annual Merit Review and Peer Evaluation Report

    Energy Technology Data Exchange (ETDEWEB)

    none,

    2013-10-01

    This report summarizes comments from the Peer Review Panel at the 2013 DOE Hydrogen and Fuel Cells Program Annual Merit Review, held on May 13-17, 2013, in Arlington, Virginia. It covers the program areas of hydrogen production and delivery; hydrogen storage; fuel cells; manufacturing R&D; technology validation; safety, codes, and standards; market transformation; and systems analysis.

  16. 2014 DOE Hydrogen and Fuel Cells Program Annual Merit Review and Peer Evaluation Report

    Energy Technology Data Exchange (ETDEWEB)

    none,

    2014-10-01

    This report summarizes comments from the Peer Review Panel at the 2014 DOE Hydrogen and Fuel Cells Program Annual Merit Review, held on June 16-20, 2014, in Washington, DC. It covers the program areas of hydrogen production and delivery; hydrogen storage; fuel cells; manufacturing R&D; technology validation; safety, codes, and standards; market transformation; and systems analysis.

  17. 2012 DOE Hydrogen and Fuel Cells Program Annual Merit Review and Peer Evaluation Report

    Energy Technology Data Exchange (ETDEWEB)

    none,

    2012-09-01

    This report summarizes comments from the Peer Review Panel at the 2012 DOE Hydrogen and Fuel Cells Program Annual Merit Review, held on May 14-18, 2012, in Arlington, Virginia. It covers the program areas of hydrogen production and delivery; hydrogen storage; fuel cells; manufacturing R&D; technology validation; safety, codes, and standards; education; market transformation; and systems analysis.

  18. 40 CFR 80.604 - What are the annual reporting requirements for refiners and importers of NRLM diesel fuel?

    Science.gov (United States)

    2010-07-01

    ... Motor Vehicle Diesel Fuel; Nonroad, Locomotive, and Marine Diesel Fuel; and ECA Marine Fuel... requirements for refiners and importers of NRLM diesel fuel? 80.604 Section 80.604 Protection of Environment... importers of NRLM diesel fuel? Beginning with the annual compliance period that begins June 1, 2007, or the...

  19. 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

  20. 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

  1. 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.

  2. Proceedings of the third annual fuel cells contractors review meeting

    Energy Technology Data Exchange (ETDEWEB)

    Huber, W.J. (ed.)

    1991-06-01

    The overall objective of this program is to develop the essential technology for private sector characterization of the various fuel cell electrical generation systems. These systems promise high fuel to electricity efficiencies (40 to 60 percent), distinct possibilities for cogeneration applications, modularity of design, possibilities of urban siting, and environmentally benign emissions. The purpose of this meeting was to provide the research and development (R D) participants in the DOE/Fossil Energy-sponsored Fuel Cells Program with the opportunity to present key results of their research and to establish closer business contacts. Major emphasis was on phosphoric acid, molten carbonate, and solid oxide technology efforts. Research results of the coal gasification and gas stream cleanup R D activities pertinent to the Fuel Cells Program were also highlighted. Two hundred seventeen attendees from industry, utilities, academia, and Government participated in this 2-day meeting. Twenty-three papers were given in three formal sessions: molten carbonate fuel cells R D (9 papers), solid oxide fuel cells (8 papers), phosphoric acid fuel cells R D (6 papers). In addition to the papers and presentations, these proceedings also include comments on the Fuel Cells Program from the viewpoint of DOE/METC Fuel Cell Overview by Rita A. Bajura, DOE/METC Perspective by Manville J. Mayfield, Electric Power Research Institute by Daniel M. Rastler, Natural Gas by Hugh D. Guthrie, and Transportation Applications by Pandit G. Patil.

  3. 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

  4. 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

  5. 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

  6. 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)

  7. 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.)

  8. First annual report of the Canadian nuclear fuel waste management program

    International Nuclear Information System (INIS)

    Boulton, J.; Gibson, A.R.

    1979-12-01

    The research and development program for the safe, permanent disposal of Canada's nuclear fuel wastes has been established. This report, the first of a series of annual reports, reviews in general terms the progress which has been achieved. After briefly reviewing the rationale and organization of the program, the report summarizes activities in the area of public information, used fuel storage and transportation, immobilization of used fuel and fuel reprocessing wastes, research and development associated with deep underground disposal, and environmental and safety assessment. (auth) [fr

  9. 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)

  10. 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

  11. 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)

  12. 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

  13. 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

  14. 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

  15. 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

  16. 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)

  17. 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

  18. 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.

  19. 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)

  20. Independent Spent Fuel Storage Installations (ISFSI). Annual report, FY 1978

    International Nuclear Information System (INIS)

    Zima, G.E.

    1979-03-01

    The prime objective of the subject program is the identification of technical aspects of the design, operation and maintenance of independent spent fuel storage installations which could contribute to technical bases for Regulations and Regulatory Guides issued by NRC for these facilities. Activities on the various tasks of the program for the FY 1978 period are discussed in this report

  1. 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

  2. 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)

  3. DOE Hydrogen and Fuel Cells Program 2017 Annual Merit Review and Peer Evaluation Report

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2017-10-16

    The fiscal year 2017 U.S. Department of Energy (DOE) Hydrogen and Fuel Cells Program Annual Merit Review and Peer Evaluation Meeting (AMR), in conjunction with DOE's Vehicle Technologies Office AMR, was held from June June 5-9, 2017, in Washington, D.C. This report is a summary of comments by AMR peer reviewers about the hydrogen and fuel cell projects funded by DOE's Office of Energy Efficiency and Renewable Energy.

  4. DOE Hydrogen and Fuel Cells Program 2016 Annual Merit Review and Peer Evaluation Report

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2016-11-01

    The fiscal year 2016 U.S. Department of Energy (DOE) Hydrogen and Fuel Cells Program Annual Merit Review and Peer Evaluation Meeting (AMR), in conjunction with DOE's Vehicle Technologies Office AMR, was held from June 6-10, 2016, in Washington, D.C. This report is a summary of comments by AMR peer reviewers about the hydrogen and fuel cell projects funded by DOE's Office of Energy Efficiency and Renewable Energy.

  5. 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

  6. 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.

  7. 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)

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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.

  13. 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.

  14. 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.

  15. 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

  16. Hydrogen engine performance analysis project. Second annual report

    Energy Technology Data Exchange (ETDEWEB)

    Adt, Jr., R. R.; Swain, M. R.; Pappas, J. M.

    1980-01-01

    Progress in a 3 year research program to evaluate the performance and emission characteristics of hydrogen-fueled internal combustion engines is reported. Fifteen hydrogen engine configurations will be subjected to performance and emissions characterization tests. During the first two years, baseline data for throttled and unthrottled, carburetted and timed hydrogen induction, Pre IVC hydrogen-fueled engine configurations, with and without exhaust gas recirculation (EGR) and water injection, were obtained. These data, along with descriptions of the test engine and its components, the test apparatus, experimental techniques, experiments performed and the results obtained, are given. Analyses of other hydrogen-engine project data are also presented and compared with the results of the present effort. The unthrottled engine vis-a-vis the throttled engine is found, in general, to exhibit higher brake thermal efficiency. The unthrottled engine also yields lower NO/sub x/ emissions, which were found to be a strong function of fuel-air equivalence ratio. (LCL)

  17. Spent fuel and fuel pool component integrity. Annual report, FY 1979

    International Nuclear Information System (INIS)

    Johnson, A.B. Jr.; Bailey, W.J.; Schreiber, R.E.; Kustas, F.M.

    1980-05-01

    International meetings under the BEFAST program and under INFCE Working Group No. 6 during 1978 and 1979 continue to indicate that no cases of fuel cladding degradation have developed on pool-stored fuel from water reactors. A section from a spent fuel rack stand, exposed for 1.5 y in the Yankee Rowe (PWR) pool had 0.001- to 0.003-in.-deep (25- to 75-μm) intergranular corrosion in weld heat-affected zones but no evidence of stress corrosion cracking. A section of a 304 stainless steel spent fuel storage rack exposed 6.67 y in the Point Beach reactor (PWR) spent fuel pool showed no significant corrosion. A section of 304 stainless steel 8-in.-dia pipe from the Three Mile Island No. 1 (PWR) spent fuel pool heat exchanger plumbing developed a through-wall crack. The crack was intergranular, initiating from the inside surface in a weld heat-affected zone. The zone where the crack occurred was severely sensitized during field welding. The Kraftwerk Union (Erlangen, GFR) disassembled a stainless-steel fuel-handling machine that operated for 12 y in a PWR (boric acid) spent fuel pool. There was no evidence of deterioration, and the fuel-handling machine was reassembled for further use. A spent fuel pool at a Swedish PWR was decontaminated. The procedure is outlined in this report

  18. Annual Performance Reports: 2002-2003 State Assessment Data

    Science.gov (United States)

    Thurlow, Martha L.; Moen, Ross E.; Wiley, Hilda I.

    2005-01-01

    States and other educational entities receiving Part B funding under the Individuals with Disabilities Education Act (IDEA) submitted their Annual Performance Reports to the U.S. Secretary of Education on or before March 31, 2004. These reports contained information on a variety of indicators, including assessment participation and performance…

  19. Fuel cells for transportation program: FY1997 national laboratory annual report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-31

    The Department of Energy (DOE) Fuel Cells for Transportation Program is structured to effectively implement the research and development (R and D) required for highly efficient, low or zero emission fuel cell power systems to be a viable replacement for the internal combustion engine in automobiles. The Program is part of the Partnership for a New Generation of Vehicles (PNGV), a government-industry initiative aimed at development of an 80 mile-per-gallon vehicle. This Annual Report summarizes the technical accomplishments of the laboratories during 1997. Participants include: Argonne National Laboratory (ANL), Brookhaven National Laboratory (BNL), Lawrence Berkeley National Laboratory (LBNL), Los Alamos National Laboratory (LANL), Oak Ridge National Laboratory (ORNL), Pacific Northwest National Laboratory (PNNL), and the National Renewable Energy Laboratory (NREL). During 1997, the laboratory R and D included one project on solid oxide fuel cells; this project has since been terminated to focus Department resources on PEM fuel cells. The technical component of this report is divided into five key areas: fuel cell stack research and development; fuel processing; fuel cell modeling, testing, and evaluation; direct methanol PEM fuel cells; and solid oxide fuel cells.

  20. 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

  1. 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...

  2. 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)

  3. Predicting fuel performance for SP-100 conditions

    International Nuclear Information System (INIS)

    Baars, R.E.

    1985-01-01

    This paper reports on methods for analyzing fuel designs proposed for the thermionic and thermoelectric concepts for SP-100 application. The proposed fuel design for the thermionic concept consisted of fully-enriched oxide fuel clad in chemical vapor deposition (CVD) tungsten, which also served as the emitter for the thermionic fuel element (TFE). The fuel density was 95% of theoretical with the linear heat rate flattened radially by removing fuel from the center of the fuel pellet. The fuel inner diameter varied from approx.0.45 in. at the core center to zero at the edge of the core. The as-fabricated gap between fuel and emitter was 10 mils radial. The emitter thickness was 80 mils, and the outer diameter was 1.099 in. The LIFE-4 code was used for evaluation of this concept after extensive review of the code and development of a procedure that corrects certain deficiencies noted in analysis of several tests

  4. 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.

  5. 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

  6. 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.

  7. 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)

  8. 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

  9. 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

  10. Fuel performance evaluation through iodine activity monitoring

    International Nuclear Information System (INIS)

    Anantharaman, K.; Chandra, R.

    1995-01-01

    The objective of the failed fuel detection system is to keep a watch on fuel behaviour during operation. This paper describes the evaluation of fuel behaviour by monitoring the activities of various isotopes of iodine both during steady state and during a reactor shutdown. The limitations of this approach also has been explained. The monitoring of tramp uranium for different types of release, namely fixed contamination and continuous release from fuel, is also presented. (author)

  11. The manufacturing role in fuel performance

    International Nuclear Information System (INIS)

    Barr, A.P.

    1997-01-01

    Manufacturing companies have been involved in the CANDU fuel industry for more than 40 years. Early manufacturing contributions were the development of materials and processes used to fabricate the CANDU fuel bundle. As CANDU reactors were commissioned, the manufacturing contribution has been to produce economical, high quality fuel for the CANDU market. (author)

  12. Semi-annual status report of the Canadian Nuclear Fuel Waste Management Program, April 1--September 30, 1991

    Energy Technology Data Exchange (ETDEWEB)

    Wright, E. D. [comp.

    1992-02-01

    This report is the eleventh in a series of semi-annual status reports on the research and development program for the safe management and disposal of Canada's nuclear fuel waste. it describes progress achieved in the three major subprograms, engineered systems, natural systems and performance assessment, from 1991 April 1 to September 30. It also gives a brief description of the activities being carried out in preparation for the public and governmental review of the disposal concept. Since 1987, this program has been jointly funded by AECL and Ontario Hydro under the auspices of the CANDU Owners Group (COG).

  13. High Performance Fuel Technology Development(I)

    International Nuclear Information System (INIS)

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

    2010-04-01

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

  14. Addressing forecast uncertainty impact on CSP annual performance

    Science.gov (United States)

    Ferretti, Fabio; Hogendijk, Christopher; Aga, Vipluv; Ehrsam, Andreas

    2017-06-01

    This work analyzes the impact of weather forecast uncertainty on the annual performance of a Concentrated Solar Power (CSP) plant. Forecast time series has been produced by a commercial forecast provider using the technique of hindcasting for the full year 2011 in hourly resolution for Ouarzazate, Morocco. Impact of forecast uncertainty has been measured on three case studies, representing typical tariff schemes observed in recent CSP projects plus a spot market price scenario. The analysis has been carried out using an annual performance model and a standard dispatch optimization algorithm based on dynamic programming. The dispatch optimizer has been demonstrated to be a key requisite to maximize the annual revenues depending on the price scenario, harvesting the maximum potential out of the CSP plant. Forecasting uncertainty affects the revenue enhancement outcome of a dispatch optimizer depending on the error level and the price function. Results show that forecasting accuracy of direct solar irradiance (DNI) is important to make best use of an optimized dispatch but also that a higher number of calculation updates can partially compensate this uncertainty. Improvement in revenues can be significant depending on the price profile and the optimal operation strategy. Pathways to achieve better performance are presented by having more updates both by repeatedly generating new optimized trajectories but also more often updating weather forecasts. This study shows the importance of working on DNI weather forecasting for revenue enhancement as well as selecting weather services that can provide multiple updates a day and probabilistic forecast information.

  15. 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.

  16. 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)

  17. 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)

  18. 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)

  19. 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

  20. 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.

  1. 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)

  2. 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.

  3. Development of High Performance Hybrid Fuels

    Data.gov (United States)

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

  4. 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)

  5. 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)

  6. 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.

  7. 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.

  8. Computer simulation of fuel element performance

    Energy Technology Data Exchange (ETDEWEB)

    Sukhanov, G I

    1979-01-01

    The review presents reports made at the Conference on the Bahaviour and Production of Fuel for Water Reactors on March 13-17, 1979. Discussed at the Conference are the most developed and tested calculation models specially evolved to predict the behaviour of fuel elements of water reactors. The following five main aspects of the problem are discussed: general conceptions and programs; mechanical mock-ups and their applications; gas release, gap conductivity and fuel thermal conductivity; analysis of nonstationary processes; models of specific phenomena. The review briefly describes the physical principles of the following models and programs: the RESTR, providing calculation of the radii of zones of columnar and equiaxial grains as well as the radius of the internal cavity of the fuel core; programs for calculation of fuel-can interaction, based on the finite elements method; a model predicting the behaviour of the CANDU-PHW fuel elements in transient conditions. General results are presented of investigations of heat transfer through a can-fuel gap and thermal conductivity of UO/sub 2/ with regard for cracking and gas release of the fuel. Many programs already suit the accepted standards and are intensively tested at present.

  9. 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)

  10. SCANAIR: A transient fuel performance code

    International Nuclear Information System (INIS)

    Moal, Alain; Georgenthum, Vincent; Marchand, Olivier

    2014-01-01

    Highlights: • Since the early 1990s, the code SCANAIR is developed at IRSN. • The software focuses on studying fast transients such as RIA in light water reactors. • The fuel rod modelling is based on a 1.5D approach. • Thermal and thermal-hydraulics, mechanical and gas behaviour resolutions are coupled. • The code is used for safety assessment and integral tests analysis. - Abstract: Since the early 1990s, the French “Institut de Radioprotection et de Sûreté Nucléaire” (IRSN) has developed the SCANAIR computer code with the view to analysing pressurised water reactor (PWR) safety. This software specifically focuses on studying fast transients such as reactivity-initiated accidents (RIA) caused by possible ejection of control rods. The code aims at improving the global understanding of the physical mechanisms governing the thermal-mechanical behaviour of a single rod. It is currently used to analyse integral tests performed in CABRI and NSRR experimental reactors. The resulting validated code is used to carry out studies required to evaluate margins in relation to criteria for different types of fuel rods used in nuclear power plants. Because phenomena occurring during fast power transients are complex, the simulation in SCANAIR is based on a close coupling between several modules aimed at modelling thermal, thermal-hydraulics, mechanical and gas behaviour. During the first stage of fast power transients, clad deformation is mainly governed by the pellet–clad mechanical interaction (PCMI). At the later stage, heat transfers from pellet to clad bring the cladding material to such high temperatures that the boiling crisis might occurs. The significant over-pressurisation of the rod and the fact of maintaining the cladding material at elevated temperatures during a fairly long period can lead to ballooning and possible clad failure. A brief introduction describes the context, the historical background and recalls the main phenomena involved under

  11. SCANAIR: A transient fuel performance code

    Energy Technology Data Exchange (ETDEWEB)

    Moal, Alain, E-mail: alain.moal@irsn.fr; Georgenthum, Vincent; Marchand, Olivier

    2014-12-15

    Highlights: • Since the early 1990s, the code SCANAIR is developed at IRSN. • The software focuses on studying fast transients such as RIA in light water reactors. • The fuel rod modelling is based on a 1.5D approach. • Thermal and thermal-hydraulics, mechanical and gas behaviour resolutions are coupled. • The code is used for safety assessment and integral tests analysis. - Abstract: Since the early 1990s, the French “Institut de Radioprotection et de Sûreté Nucléaire” (IRSN) has developed the SCANAIR computer code with the view to analysing pressurised water reactor (PWR) safety. This software specifically focuses on studying fast transients such as reactivity-initiated accidents (RIA) caused by possible ejection of control rods. The code aims at improving the global understanding of the physical mechanisms governing the thermal-mechanical behaviour of a single rod. It is currently used to analyse integral tests performed in CABRI and NSRR experimental reactors. The resulting validated code is used to carry out studies required to evaluate margins in relation to criteria for different types of fuel rods used in nuclear power plants. Because phenomena occurring during fast power transients are complex, the simulation in SCANAIR is based on a close coupling between several modules aimed at modelling thermal, thermal-hydraulics, mechanical and gas behaviour. During the first stage of fast power transients, clad deformation is mainly governed by the pellet–clad mechanical interaction (PCMI). At the later stage, heat transfers from pellet to clad bring the cladding material to such high temperatures that the boiling crisis might occurs. The significant over-pressurisation of the rod and the fact of maintaining the cladding material at elevated temperatures during a fairly long period can lead to ballooning and possible clad failure. A brief introduction describes the context, the historical background and recalls the main phenomena involved under

  12. 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

  13. 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.)

  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. Annual Summary of the Integrated Disposal Facility Performance Assessment 2012

    Energy Technology Data Exchange (ETDEWEB)

    Khaleel, R. [INTERA, Austin, TX (United States); Nichols, W. E. [CH2M HILL Plateau Remediation Company, Richland, WA (United States)

    2012-12-27

    An annual summary of the adequacy of the Hanford Immobilized Low-Activity Waste (ILAW) Performance Assessment (PA) is required each year (DOE O 435.1 Chg 1,1 DOE M 435.1-1 Chg 1;2 and DOE/ORP-2000-013). The most recently approved PA is DOE/ORP-2000-24.4 The ILAW PA evaluated the adequacy of the ILAW disposal facility, now referred to as the Integrated Disposal Facility (IDF), for the safe disposal of vitrified Hanford Site tank waste.

  16. 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

  17. 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

  18. 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.

  19. 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

  20. Fuel management for off-load annual refuelling of the D-HHT 600 MW(e) reference core

    Energy Technology Data Exchange (ETDEWEB)

    Hansen, U

    1973-03-16

    The reference design for the Dragon-HHT reactor has been optimised for on-load continuous refuelling. The possiblity to operate the reactor on a discontinuous annual reloading schedule might prove of interest and/or necessity. In this paper the influence of an annual 4-batch fuel management scheme on the core physics and fuel cycle economics is investigated. The results of the present investigation give a good indication of the relative merits of the two fuel management schemes. Although a broader parameter survey and a more detailed scrutinising of special cases would be desirable, we feel that the main conclusions are correct and that the principle differences have been elicited.

  1. 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.

  2. 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.

  3. 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

  4. Implications of manufacturing deviations on fuel performance

    International Nuclear Information System (INIS)

    Chellapandi, P.; Clement Ravi Chandar, S.; Chetal, S.C.; Baldev Raj

    2009-01-01

    Prototype Fast Breeder Reactor (PFBR) core consists of 181 Fuel subassembly (FSA), 120 blanket SA and shielding SA of steel and B 4 C besides 9 Control Safety Rod SA and 3 Diverse Safety Rod SA. All the subassemblies (SA) stand vertically on the grid plate. PFBR FSA consists of 217 fuel pins of 2540 mm length arranged in a triangular pitch standing vertically on the rails inside a hexagonal duct. The bottom of the hexagon is screwed and welded to a cylindrical foot of length 600 mm. The SA foot has radial slots that provide an entry to the coolant sodium and also houses flow control devices. While the foot of the A handling head is welded to the top of hexagon, provides an aid for fuel handling machine to insert and withdraw the FSA from the grid plate. The length of the FSA is 4500 mm.

  5. 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)

  6. 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.

  7. 11. annual report of the technical advisory committee on the nuclear fuel waste management program

    International Nuclear Information System (INIS)

    1990-10-01

    The Eleventh Annual Report of the Technical Advisory Committee (TAC) assesses the scientific and technical progress made within the Canadian Nuclear Fuel Waste Management Program (NFWMP) during the period July 1989 to June 1990. The Committee notes that the general concept of a multibarrier system involving geologic media and engineered systems is based on known technologies and current scientific knowledge, and has gained strong international scientific and engineering support as currently the most feasible and practical. TAC continues to endorse the full investigation of the concept of nuclear waste disposal deep in plutonic formations, such as those in the Canadian Shield

  8. 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)

  9. 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)

  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. 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.

  13. 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)

  14. 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)

  15. 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

  16. 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.

  17. 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

  18. The Justification of Organizational Performance in Annual Report Narratives

    Directory of Open Access Journals (Sweden)

    Marcelo Sanches Pagliarussi

    2010-04-01

    Full Text Available In this paper, the nature of organizational discourse is theoretically underpinned by the concept of self-serving attributions, a type of causal reasoning that allows the writer to take credit for good news and avoid blame for bad news. We incorporated signaling theory to the extant theoretical framework for self-serving attributions in order to develop hypotheses for the expected levels of attributional bias in the justification of organizational performance. A sample of 49 companies was selected, both from a bad year and a good year regarding the capital market context. Each company’s Letter to Shareholders was content analyzed in order to test our propositions concerning the presence and intensity of self-serving attributions in that section of annual reports. The results partially corroborate the proposed theoretical hypotheses, but the sample size is an issue in terms of robustness. Nevertheless, the results indicate that companies attempt to create a positive corporate image to external stakeholders even when negative performance occurs in a clearly favorable external context. Moreover, we observed that companies with positive performance in a good external context blame negative effects on the environment in a proportion equivalent to that observed for companies with positive performance in a bad year.

  19. 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)

  20. 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.

  1. 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

  2. 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)

  3. 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

  4. 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

  5. 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

  6. 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)

  7. 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.

  8. 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.

  9. 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

  10. 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

  11. 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

  12. 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)

  13. 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

  14. Toxicological and performance aspects of oxygenated motor vehicle fuels

    National Research Council Canada - National Science Library

    National Research Council Staff; Commission on Life Sciences; Division on Earth and Life Studies; National Research Council; National Academy of Sciences

    .... Other questions have been raised about reduced fuel economy and engine performance and pollution of ground water due to the use of MTBE in gasoline. The book provides conclusions and recommendations about each major topic addressed in the government's report.

  15. User manual for the probabilistic fuel performance code FRP

    International Nuclear Information System (INIS)

    Friis Jensen, J.; Misfeldt, I.

    1980-10-01

    This report describes the use of the probabilistic fuel performance code FRP. Detailed description of both input to and output from the program are given. The use of the program is illustrated by an example. (author)

  16. 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)

  17. 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.

  18. An update on CANDU fuel channel performance

    International Nuclear Information System (INIS)

    Price, E.G.; Holt, R.A; Wong, H.; Gautheir, P.; Ellis, P.J.; Slade, J.

    1998-01-01

    Fuel channel components, particularly pressure tubes, are monitored to establish changes in their condition expected from their service environment. Such monitoring takes the form of the in-service NDE, in-service sampling and periodic tube removal for destructive examination. The pressure tubes in early CANDU 6 stations have now reached mid-life. The expected trends in behaviour of the various properties of the pressure tubes are being confirmed. Diametral deformation potentially limits the life of some pressure tubes, and fuel developments, such as the CANFLEX fuel bundle, will provide additional CHF margins. Corrosion and the associated deuterium pick-up are monitored to assess if any change in rates have occurred. Both have historically continued on an essentially linear trend with time. Recent data indicate that the rates of pick-up could be gradually increasing. Changes in mechanical properties, particularly toughness, have reached saturation levels. In the early reactors, the toughness is acceptable, assuring leak-before-break. The pressure tubes in later reactors have very low concentrations of chlorine and phosphorous and the level of hydrogen are also low, and these tubes are predicted to retain relativity high fracture toughness in-service. Debris fretting wear has occurred in CANDU reactors. Laboratory testing and experience show that no cracks have initiated from debris frets to date. Further debris fretting will have to be avoided by care in maintenance procedures, and existing defects must be monitored in the future as hydrogen concentration levels increase. Spacer location and repositioning (SLAR) has been necessary in earlier reactors with loose spacers. Procedures and equipment for SLAR have been developed and successfully applied. The efficiency of spacer relocation operations has improved with experience. (authors)

  19. 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.

  20. 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

  1. 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

  2. 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

  3. 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)

  4. 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)

  5. 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

  6. 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

  7. 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

  8. 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)

  9. 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)

  10. Empirical Study on Annual Energy-Saving Performance of Energy Performance Contracting in China

    Directory of Open Access Journals (Sweden)

    Hongquan Ruan

    2018-05-01

    Full Text Available A lack of trust in Energy Service Company (ESCo is the most critical factor affecting the development of Energy Performance Contracting (EPC in China, compared with other constraints. One cannot easily estimate the energy-saving performance of an EPC project. Under that condition, lack of trust may cause the Energy-Consuming Unit (ECU to suspect the energy-saving performance promised by the ESCo, thus leaving potentially profitable projects without necessary funding. Currently, specific studies taking an across-projects viewpoint on annual energy-saving performance of EPC projects in multiple subsectors, objectively and quantitatively, are lacking. This paper studies the regression relationships of annual energy-saving quantity in terms of revamping cost and the regression relationships of annual cost saving in terms of revamping cost. The regression results show that there are statistically significant correlations in the above relationships in the nine subsectors investigated. This is significant for ESCos and ECUs, because knowledge on energy-saving performance could contribute to EPC investment decisions and trust relationships between ESCos and ECUs. Then, a multiple linear regression model of revamping cost is set up to analyze its influencing factors. The model indicates that the subsector the sample belongs to, financing, registered capital of the ESCo, and contract period have significant effects on revamping cost. Thus, policy implications regarding innovation of EE promotion technology, clarifying ESCos’ exit mechanism, innovation of financing mechanism, and improving the market credit environment for promoting investment in EPC projects, are provided.

  11. 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)

  12. 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.

  13. 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

  14. U.S. Department of Energy Hydrogen and Fuel Cells Program: 2017 Annual Merit Review and Peer Evaluation Report

    Energy Technology Data Exchange (ETDEWEB)

    Popovich, Neil A [National Renewable Energy Laboratory (NREL), Golden, CO (United States)

    2017-10-18

    The fiscal year 2017 U.S. Department of Energy (DOE) Hydrogen and Fuel Cells Program Annual Merit Review and Peer Evaluation Meeting (AMR), in conjunction with DOE's Vehicle Technologies Office AMR, was held from June June 5-9, 2017, in Washington, D.C. This report is a summary of comments by AMR peer reviewers about the hydrogen and fuel cell projects funded by DOE's Office of Energy Efficiency and Renewable Energy.

  15. 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.

  16. 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.

  17. 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

  18. 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.

  19. 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

  20. 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)

  1. 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

  2. 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)

  3. 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

  4. 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.

  5. 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

  6. 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)

  7. U.S. Department of Energy Hydrogen and Fuel Cells Program 2011 Annual Merit Review and Peer Evaluation Report

    Energy Technology Data Exchange (ETDEWEB)

    Satypal, S.

    2011-09-01

    This document summarizes the comments provided by peer reviewers on hydrogen and fuel cell projects presented at the FY 2011 U.S. Department of Energy (DOE) Hydrogen Program and Vehicle Technologies Program Annual Merit Review and Peer Evaluation Meeting (AMR), held May 9-13, 2011 in Arlington, Virginia

  8. 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)

  9. 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)

  10. 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)

  11. 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

  12. 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

  13. 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)

  14. 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

  15. 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

  16. 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.

  17. 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)

  18. 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.

  19. 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

  20. 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

  1. 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.

  2. 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.

  3. 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)

  4. Forage selection and performance of sheep grazing dry annual range.

    NARCIS (Netherlands)

    Ridder, de N.; Benjamin, R.W.; Keulen, van H.

    1986-01-01

    During 114 days of grazing, sheep grazing a dry annual pasture in Israel selected the fine fraction available with a higher nutritive value. As this fraction became depleted and feed quality dropped, organic matter intake dropped from 1.73 to 0.75 kg/sheep/d. Sheep lost weight, body condition and

  5. Evaluation of the performance of a dual fuel engine

    Energy Technology Data Exchange (ETDEWEB)

    Sato, Toshiharu [The Japan Gas Association, Tokyo (Japan). NGV Project Dept.; Daisho, Yasuhiro; Saito, Takeshi [Waseda Univ., Tokyo (Japan)

    1998-12-31

    Dual fuel operation, in which natural gas is mixed with the diesel engine intake air and ignition is by diesel fuel spray, has the advantage that engine conversion is simple. Under high load it has the same high efficiency as a diesel engine and it can be switched to normal diesel operation for long distance running. Also, NO{sub x} and black smoke emissions can also be reduced. However, the disadvantages are to increase HC and CO emissions, to reduce efficiency under low load, and to emit the large amount of NO{sub x} under high load. Waseda University was commissioned by Tokyo Gas Co., Ltd. to conduct research program involving experimentation ragarding a dual fuel engine. It was then discovered that the most effective means of solving the problems mentioned above is Exhaust Gas Recirculation (EGR) and that the effect can be increased by heating the intake air. An old engine before the current emission standard was converted to dual fuel operation. It was found that these measures enables NO{sub x}, black smoke and CO{sub 2} to be reduced while high thermal efficiency was maintained. They did not reach the point of satisfying latest Japanese emission standard. But it seemed that good results would have been obtained, if a base engine with good emissions had been converted for dual fuel operation. The results of assessing the performance of the dual fuel engine at this time are reported here, centered on the effect of EGR and intake heating. (orig.)

  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. 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.

  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. Study on performance of blended fuel PPO - Diesel at generator

    Science.gov (United States)

    Prasetyo, Joni; Prasetyo, Dwi Husodo; Murti, S. D. Sumbogo; Adiarso, Priyanto, Unggul

    2018-02-01

    Bio-energy is renewable energy made from plant. Biomass-based energy sources are potentially CO2 neutral and recycle the same carbon atoms. In order to reduce pollution caused by fossil fuel combustion either for mechanical or electrical energy generation, the performance characteristic of purified palm oil blends are analyzed at various ratios. Bio-energy, Pure Plant Oil, represent a sustainable solution.A generator has been modified due to adapt the viscosity ofblended fuel, PPO - diesel, by pre-heating. Several PPO - diesel composition and injection timing were tested in order to investigate the characteristic of mixed fuel with and without pre-heating. The term biofuel refers to liquid or gaseous fuels for the internal combustion engines that are predominantly produced fro m biomass. Surprising result showed that BSFC of blended PPO - diesel was more efficient when injection timing set more than 15° BTDC. The mixed fuel produced power with less mixed fuel even though the calorie content of diesel is higher than PPO. The most efficient was 20% PPO in diesel with BSFC 296 gr fuel / kwh rather than 100% diesel with BSFC 309 gr fuel / kwh at the same injection timing 18° BTDC with pre-heating. The improvement of BSFC is caused by heating up of mixed fuel which it added calorie in the mixed fuel. Therefore, the heating up of blended PPO - diesel is not only to adapt the viscosity but also improving the efficiency of fuel usage representing by lower BSFC. In addition, torque of the 20% PPO was also as smooth as 100% diesel representing by almost the same torqueat injection timing 15° BTDC. The AIP Proceedings article template has many predefined paragraph styles for you to use/apply as you write your paper. To format your abstract, use the Microsoft Word template style: Abstract. Each paper must include an abstract. Begin the abstract with the word "Abstract" followed by a period in bold font, and then continue with a normal 9 point font.

  10. 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

  11. 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

  12. 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

  13. 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.

  14. 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.

  15. Toxicological and performance aspects of oxygenated motor vehicle fuels

    National Research Council Canada - National Science Library

    National Research Council Staff; Commission on Life Sciences; Division on Earth and Life Studies; National Research Council; National Academy of Sciences

    ... COMMITTEE ON TOXICOLOGICAL PERFORMANCE ASPECTS OXYGENATED MOTOR VEHICLE FUELS ENVIRONMENTAL STUDIES TOXICOLOGY COMMISSION LIFE SCIENCES NATIONAL RESEARCH COUNCIL AND OF BOARD ON AND ON NATIONAL ACADEMY PRESS Washington, D.C. 1996 i Copyrightoriginal retained, the be not from cannot book, paper original however, for version formatting, authoritative the t...

  16. 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.

  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. Fuel debris characterization and treatment technologies development for TEPCO's Fukushima Daiichi Nuclear Power Station. 2012 annual research and development report

    International Nuclear Information System (INIS)

    2014-03-01

    Since March 11, 2011, severe accidents occurred at Fukushima Daiichi Nuclear Power Station (1F NPS), the Government of Japan and Tokyo Electric Power Company (TEPCO) and all Japan's companies have been worked on the remediation. The first meeting of 'Government and TEPCO's Mid-to-Long Term Countermeasure Meeting' was held on December 16, 2011, and then' the Council for the Decommissioning of TEPCO's Fukushima Daiichi Nuclear Power Station' was established on February 8, 2013, and 19 research and development projects and Working team / Sub working team were launched. Japan Atomic Energy Agency (JAEA); Fukushima Project Teams in Nuclear Fuel Cycle Laboratories and Nuclear Science Research Institute are belonging to the projects of 'Fuel debris characterization (2-(3)-1)', 'Analysis of fuel debris (2-(3)-2)' and 'Treatment technology development of fuel debris (2-(3)-3)'. In the 2012 JFY, we carried out research and development on the 'Fuel debris characterization (2-(3)-1)' and 'Treatment technology development of fuel debris (2-(3)-3)', and obtained some results on the debris properties and debris treatment technologies. This document report annual research and development results of above two projects in 2012 JFY. In the Project of 'Fuel debris characterization (2-(3)-1)', the debris chemical form, phase state and composition were estimated by thermodynamic calculation with referring sever accident code results and the fuel debris properties which needed for developing the methods/devices for defueling was identified with information of Three Mile Island and sever accident study. As for investigation of reaction and products specific to 1F accident, fundamental data on the debris such as mechanical properties i.e. hardness, and effects of sea water, B 4 C, ratio of Zr content and O/M, and thermal properties as melting points, thermal conductivity, etc. were obtained by

  20. 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)

  1. 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.

  2. 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.)

  3. Ontario Hydro Pickering Generating Station fuel handling system performance

    International Nuclear Information System (INIS)

    Underhill, H.J.

    1986-01-01

    The report briefly describes the Pickering Nuclear Generating Station (PNGS) on-power fuel handling system and refuelling cycle. Lifetime performance parameters of the fuelling system are presented, including station incapability charged to the fuel handling system, cost of operating and maintenance, dose expenditure, events causing system unavailability, maintenance and refuelling strategy. It is concluded that the 'CANDU' on-power fuelling system, by consistently contributing less than 1% to the PNGS incapability, has been credited with a 6 to 20% increase in reactor capacity factor, compared to off-power fuelling schemes. (author)

  4. Performance of FFTF reference fuel and control assemblies

    International Nuclear Information System (INIS)

    Leggett, R.D.; Weber, E.T.

    1984-11-01

    This paper describes the performance of the reference fuel and control assemblies used in FFTF through the first four cycles of irradiation (446 equivalent full power days, EFPD). These assemblies performed flawlessly through the rigors of the Startup Testing Program, STP, (beginning in late 1979) with its cyclic operation and continued to do so throughout Cycles 1, 2, 3 and 4, the latter ending in April 1984

  5. Predictive Bias and Sensitivity in NRC Fuel Performance Codes

    Energy Technology Data Exchange (ETDEWEB)

    Geelhood, Kenneth J.; Luscher, Walter G.; Senor, David J.; Cunningham, Mitchel E.; Lanning, Donald D.; Adkins, Harold E.

    2009-10-01

    The latest versions of the fuel performance codes, FRAPCON-3 and FRAPTRAN were examined to determine if the codes are intrinsically conservative. Each individual model and type of code prediction was examined and compared to the data that was used to develop the model. In addition, a brief literature search was performed to determine if more recent data have become available since the original model development for model comparison.

  6. 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

  7. 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.

  8. Assessment of MARMOT. A Mesoscale Fuel Performance Code

    Energy Technology Data Exchange (ETDEWEB)

    Tonks, M. R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schwen, D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Y. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Chakraborty, P. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bai, X. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Fromm, B. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Yu, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Teague, M. C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Andersson, D. A. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-04-01

    MARMOT is the mesoscale fuel performance code under development as part of the US DOE Nuclear Energy Advanced Modeling and Simulation Program. In this report, we provide a high level summary of MARMOT, its capabilities, and its current state of validation. The purpose of MARMOT is to predict the coevolution of microstructure and material properties of nuclear fuel and cladding. It accomplished this using the phase field method coupled to solid mechanics and heat conduction. MARMOT is based on the Multiphysics Object-Oriented Simulation Environment (MOOSE), and much of its basic capability in the areas of the phase field method, mechanics, and heat conduction come directly from MOOSE modules. However, additional capability specific to fuel and cladding is available in MARMOT. While some validation of MARMOT has been completed in the areas of fission gas behavior and grain growth, much more validation needs to be conducted. However, new mesoscale data needs to be obtained in order to complete this validation.

  9. 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

  10. U.S. Department of Energy Hydrogen and Fuel Cells Program, 2013 Annual Merit Review and Peer Evaluation Report (Book)

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2013-10-01

    The fiscal year (FY) 2013 U.S. Department of Energy (DOE) Hydrogen and Fuel Cells Program Annual Merit Review and Peer Evaluation Meeting (AMR), in conjunction with DOE's Vehicle Technologies Office AMR, was held from May 13-16, 2013, at the Crystal City Marriott and Crystal Gateway Marriott in Arlington, Virginia. This report is a summary of comments by AMR peer reviewers about the hydrogen and fuel cell projects funded by DOE's Office of Energy Efficiency and Renewable Energy (EERE).

  11. Performance of advanced oxide fuel pins in EBR-II

    International Nuclear Information System (INIS)

    Lawrence, L.A.; Jensen, S.M.; Hales, J.W.; Karnesky, R.A.; Makenas, B.J.

    1986-05-01

    The effects of design and operating parameters on mixed-oxide fuel pin irradiation performance were established for the Hanford Engineering Development Laboratory (HEDL) advanced oxide EBR-II test series. Fourteen fuel pins breached in-reactor with reference 316 SS cladding. Seven of the breaches are attributed to FCMI. Of the remaining seven breached pins, three are attributed to local cladding over-temperatures similar to the breach mechanism for the reference oxide pins irradiated in EBR-II. FCCI was found to be a contributing factor in two high burnup, i.e., 11.7 at. % breaches. The remaining two breaches were attributed to mechanical interaction of UO 2 fuel and fission products accumulated in the lower cladding insulator gap, and a loss of cladding ductility possibly due to liquid metal embrittlement. Fuel smear density appears to have the most significant impact on lifetime. Quantitative evaluations of cladding diameter increases attributed to FCMI, established fuel smear density, burnup, and cladding thickness-to-diameter ratio as the major parameters influencing the extent of cladding strain

  12. Research on Elemental Technology of Advanced Nuclear Fuel Performance Verification

    International Nuclear Information System (INIS)

    Kim, Yong Soo; Lee, Dong Uk; Jean, Sang Hwan; Koo, Min

    2003-04-01

    Most of current properties models and fuel performance models used in the performance evaluation codes are based on the in-pile data up to 33,000 MWd/MtU. Therefore, international experts are investigating the properties changes and developing advanced prediction models for high burn-up application. Current research is to develop high burn-up fission gas release model for the code and to support the code development activities by collecting data and models, reviewing/assessing the data and models together, and benchmarking the selected models against the appropriate in-pile data. For high burn-up applications, two stage two step fission gas release model is developed based on the real two diffusion process in the grain lattice and grain boundaries of the fission gases and the observation of accelerated release rate in the high burn-up. It is found that the prediction of this model is in excellent agreement with the in-pile measurement results, not only in the low burn-up but also in the high burn-up. This research is found that the importance of thermal conductivity of oxide fuel, especially in the high burn-up, is focused again. It is found that even the temperature dependent models differ from one to another and most of them overestimate the conductivity in the high burn-up. An in-pile data benchmarking of high LHGR fuel rod shows that the difference can reach 30%∼40%, which predicts 400 .deg. C lower than the real fuel centerline temperature. Recent models on the thermal expansion and heat capacity of oxide fuel are found to be well-defined. Irradiation swelling of the oxide fuel are now well-understood that in most cases in LWRs solid fission product swelling is dominant. Thus, the accumulation of in-pile data can enhance the accuracy of the model prediction, rather than theoretical modeling works. Thermo-physical properties of Zircaloy cladding are also well-defined and well-understood except the thermal expansion. However, it turns out that even the

  13. Nondestructive analysis of irradiated fuels

    International Nuclear Information System (INIS)

    Dudey, N.D.; Frick, D.C.

    1977-01-01

    The principal nondestructive examination techniques presently used to assess the physical integrity of reactor fuels and cladding materials include gamma-scanning, profilometry, eddy current, visual inspection, rod-to-rod spacing, and neutron radiography. LWR fuels are generally examined during annual refueling outages, and are conducted underwater in the spent fuel pool. FBR fuels are primarily examined in hot cells after fuel discharge. Although the NDE techniques are identical, LWR fuel examinations emphasize tests to demonstrate adherence to technical specification and reliable fuel performance; whereas, FBR fuel examinations emphasize aspects more related to the relative performance of different types of fuel and cladding materials subjected to variable irradiation conditions

  14. Nuclear power performance and safety. V.5. Nuclear fuel cycle

    International Nuclear Information System (INIS)

    1988-01-01

    The International Conference on Nuclear Power Performance and Safety, organized by the International Atomic Energy Agency, was held at the Austria Centre Vienna (ACV) in Vienna, Austria, from 28 September to 2 October 1987. The objective of the Conference was to promote an exchange of worldwide information on the current trends in the performance and safety of nuclear power and its fuel cycle, and to take a forward look at the expectations and objectives for the 1990s. Policy decisions for waste management have already been taken in many countries and the 1990s should be a period of demonstration and implementation of these policies. As ilustrated by data presented from a number of countries, many years of experience in radioactive waste management have been achieved and the technology exists to implement the national plans and policies that have been developed. The establishment of criteria, the development of safety performance methodology and site investigation work are key activities essential to the successful selection, characterization and construction of geological repositories for the final disposal of radioactive waste. Considerable work has been done in these areas over the last ten years and will continue into the 1990s. However, countries that are considering geological disposal for high level waste now recognize the need for relating the technical aspects to public understanding and acceptance of the concept and decision making activities. The real challenge for the 1990s in waste disposal will be successfully to integrate technological activities within a process which responds to institutional and public concern. Volume 5 of the Proceedings comprehends the contributions on waste management in the 1990s. Decontamination and decommissioning, waste management, treatment and disposal, nuclear fuel cycle - present and future. Enrichment services and advanced reactor fuels, improvements in reactor fuel utilization and performance, spent fuel management

  15. 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)

  16. 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

  17. Radionuclide release rates from spent fuel for performance assessment modeling

    International Nuclear Information System (INIS)

    Curtis, D.B.

    1994-01-01

    In a scenario of aqueous transport from a high-level radioactive waste repository, the concentration of radionuclides in water in contact with the waste constitutes the source term for transport models, and as such represents a fundamental component of all performance assessment models. Many laboratory experiments have been done to characterize release rates and understand processes influencing radionuclide release rates from irradiated nuclear fuel. Natural analogues of these waste forms have been studied to obtain information regarding the long-term stability of potential waste forms in complex natural systems. This information from diverse sources must be brought together to develop and defend methods used to define source terms for performance assessment models. In this manuscript examples of measures of radionuclide release rates from spent nuclear fuel or analogues of nuclear fuel are presented. Each example represents a very different approach to obtaining a numerical measure and each has its limitations. There is no way to obtain an unambiguous measure of this or any parameter used in performance assessment codes for evaluating the effects of processes operative over many millennia. The examples are intended to suggest by example that in the absence of the ability to evaluate accuracy and precision, consistency of a broadly based set of data can be used as circumstantial evidence to defend the choice of parameters used in performance assessments

  18. 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.

  19. 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

  20. Fuel performance at high burnup for water reactors

    International Nuclear Information System (INIS)

    1991-02-01

    The present meeting was scheduled by the International Atomic Energy Agency, upon proposal of the Members of the International Working Group on Water Reactor Fuel Performance and Technology. The purpose of this meeting was to review the ''state-of-the-art'' in the area of Fuel Performance at High Burnup for Water Reactors. Previous IAEA meetings on this topic were held in Mol in 1981 and 1984 and on related topics in Stockholm and Lyon in 1987. Fifty-five participants from 16 countries and two international organizations attended the meeting and 28 papers were presented and discussed. The papers were presented in five sub-sessions and during the meeting, working groups composed of the session chairmen and paper authors prepared the summary of each session with conclusions and recommendations for future work. A separate abstract was prepared for each of these papers. Refs, figs and tabs

  1. Business cycles and the financial performance of fuel cell companies

    International Nuclear Information System (INIS)

    Henriques, I.; Sadorsky, P.

    2005-01-01

    Fuel cells are expected to play a major role in a hydrogen powered world. They will provide power to homes, modes of transportation and appliances. Hydrogen is the most abundant element in nature, but it must be extracted in order to be usable. It can be produced from oil, natural gas and coal or from renewable sources such as biomass, thermal or nuclear reactions. Fuel cells running on hydrogen extracted from non renewable resources have an efficiency of 30 per cent, which is twice as efficient as an internal combustion engine. The greatest barrier to mass commercialization is the cost of making hydrogen-powered auto engines. Also, an infrastructure must be developed to refill hydrogen cars. One solution is to build a hydrogen highway using the existing natural gas grid to produce hydrogen and sell it at existing filling stations. The cost of building 12,000 refueling pumps in urban areas which will provide access to 70 per cent of America's population is estimated at $10 to $15 billion. This paper described the vector autoregression (VAR) model which empirically examines the relationship between financial performance of fuel cell companies and business cycles. It was used to measure how sensitive the financial performance of fuel cell companies are to changes in macroeconomic activity. A four variable VAR model was developed to examine the relationship between stock prices, oil prices and interest rates. It was shown that the stock prices of fuel cell companies are affected by shocks to technology stock prices and oil prices, with the former having a longer lasting impact. These results add to the growing literature that oil price movements are not as important as once thought. 15 refs., 3 tabs., 3 figs

  2. 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

  3. Preliminary analysis on incore performance of nuclear fuel: pt. 4

    International Nuclear Information System (INIS)

    Noh, S.K.; Chang, M.H.; Lee, C.C.; Chung, Y.H.; Kuk, K.Y.; Park, C.Y.; Lee, S.K.

    1981-01-01

    An analysis has been performed for thermal hydraulic design parameters of Wolsung-1 reactor core in steady state with the help of a computer code COBRA-IV-I. The design parameters are coolant enthalpy, flow velocity, coolant quality, pressure and fuel temperature distribution. The maximum power channel has been taken into account in this work. The results appear to be reasonably agreeable with data from PSR'S, with the maximum difference between this work and PSR'S being 4.3%

  4. 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.

  5. Δ14C level of annual plants and fossil fuel derived CO2 distribution across different regions of China

    International Nuclear Information System (INIS)

    Xi, X.T.; Ding, X.F.; Fu, D.P.; Zhou, L.P.; Liu, K.X.

    2013-01-01

    The 14 C level in annual plants is a sensitive tracer for monitoring fossil fuel derived CO 2 in the atmosphere. Corn leave samples were selected from different regions of China, including high mountains in the Tibetan Plateau, grassland in Inner Mongolia, and inland and coastal cities during the summer of 2010. The 14 C/ 12 C ratio of the samples was measured with the NEC compact AMS system at the Institute of Heavy Ion Physics, Peking University. The fossil fuel derived CO 2 was estimated by comparing the measured Δ 14 C values of corn leave samples to background atmospheric Δ 14 C level. The influences of topography, meteorological conditions and carbon cycling processes on the fossil fuel derived CO 2 concentration are considered when interpreting the data. Our results show a clear association of the low Δ 14 C values with regions where human activities are intensive.

  6. Performance analysis of a stationary fuel cell thermoelectric cogeneration system

    Energy Technology Data Exchange (ETDEWEB)

    Kuo, J.K.; Hwang, J.J.; Lin, C.H. [Department of Greenergy, National University of Tainan, Tainan, 70005 (China)

    2012-12-15

    The main purpose of our study was to use an experimental method and system dynamic simulation technology to examine a proton exchange membrane fuel cell thermoelectric cogeneration system that provides both high-quality electric power and heated water. In the second part of our study, we experimentally verified the development of key components of the fuel cell and conducted a comprehensive analysis of the subsystems, including the fuel cell module, hydrogen supply subsystem, air supply subsystem, humidifier subsystem, and heat recovery subsystem. Finally, we integrated all of the subsystems into a PEM fuel cell thermoelectric cogeneration system and performed efficiency tests and analysis of power generation, heat recovery, and thermoelectric cogeneration. After comparing this system's efficiency results using simulation and experimentation, we determined that the accuracy of the simulation values when compared to the experimental values was >95%, showing that this system's simulation nearly approached the efficiency of the actual experiment, including more than 53% for power generation efficiency, more than 39% for heat recovery efficiency, and more than 93% for thermoelectric cogeneration combined efficiency. (Copyright copyright 2012 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  7. Advanced fast reactor fuels program. Second annual progress report, July 1, 1975--September 30, 1976

    International Nuclear Information System (INIS)

    Baker, R.D.

    1978-12-01

    Results of steady-state (EBR-II) irradiation testing, off-normal irradiation design and testing, fuel-cladding compatibility, and chemical stability of uranium--plutonium carbide and nitride fuels are presented

  8. Fuel Performance Experiments and Modeling: Fission Gas Bubble Nucleation and Growth in Alloy Nuclear Fuels

    International Nuclear Information System (INIS)

    McDeavitt, Sean; Shao, Lin; Tsvetkov, Pavel; Wirth, Brian; Kennedy, Rory

    2014-01-01

    Advanced fast reactor systems being developed under the DOE's Advanced Fuel Cycle Initiative are designed to destroy TRU isotopes generated in existing and future nuclear energy systems. Over the past 40 years, multiple experiments and demonstrations have been completed using U-Zr, U-Pu-Zr, U-Mo and other metal alloys. As a result, multiple empirical and semi-empirical relationships have been established to develop empirical performance modeling codes. Many mechanistic questions about fission as mobility, bubble coalescience, and gas release have been answered through industrial experience, research, and empirical understanding. The advent of modern computational materials science, however, opens new doors of development such that physics-based multi-scale models may be developed to enable a new generation of predictive fuel performance codes that are not limited by empiricism.

  9. Fuel Performance Experiments and Modeling: Fission Gas Bubble Nucleation and Growth in Alloy Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

    McDeavitt, Sean [Texas A & M Univ., College Station, TX (United States); Shao, Lin [Texas A & M Univ., College Station, TX (United States); Tsvetkov, Pavel [Texas A & M Univ., College Station, TX (United States); Wirth, Brian [Univ. of Tennessee, Knoxville, TN (United States); Kennedy, Rory [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-04-07

    Advanced fast reactor systems being developed under the DOE's Advanced Fuel Cycle Initiative are designed to destroy TRU isotopes generated in existing and future nuclear energy systems. Over the past 40 years, multiple experiments and demonstrations have been completed using U-Zr, U-Pu-Zr, U-Mo and other metal alloys. As a result, multiple empirical and semi-empirical relationships have been established to develop empirical performance modeling codes. Many mechanistic questions about fission as mobility, bubble coalescience, and gas release have been answered through industrial experience, research, and empirical understanding. The advent of modern computational materials science, however, opens new doors of development such that physics-based multi-scale models may be developed to enable a new generation of predictive fuel performance codes that are not limited by empiricism.

  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. 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

  12. 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)

  13. Performance evaluation of integrated fuel processor for residential PEMFCs application

    International Nuclear Information System (INIS)

    Yu Taek Seo; Dong Joo Seo; Young-Seog Seo; Hyun-Seog Roh; Wang Lai Yoon; Jin Hyeok Jeong

    2006-01-01

    KIER has been developing the natural gas fuel processor to produce hydrogen rich gas for residential PEMFCs system. To realize a compact and high efficiency, the unit processes of steam reforming, water gas shift, and preferential oxidation are chemically and physically integrated in a package. Current fuel processor designed for 1 kW class PEMFCs shows thermal efficiency of 78% as a HHV basis with methane conversion of 90% at rated load operation. CO concentration below 10 ppm in the produced gas is achieved with preferential oxidation unit using Pt and Ru based catalyst under the condition of [O 2 ]/[CO]=2.0. The partial load operation have been carried out to test the performance of fuel processor from 40% to 80% load, showing stable methane conversion and CO concentration below 10 ppm. The durability test for the daily start-stop and 8 hr operation procedure is under investigation and shows no deterioration of its performance after 40 start-stop cycles. (authors)

  14. 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

  15. 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)

  16. Performance evaluation of direct methanol fuel cells for portable applications

    Energy Technology Data Exchange (ETDEWEB)

    Rashidi, R.; Dincer, I.; Naterer, G.F. [Faculty of Engineering and Applied Science, University of Ontario Institute of Technology, 2000 Simcoe Street North, Oshawa, Ontario (Canada); Berg, P. [Faculty of Science, University of Ontario Institute of Technology, 2000 Simcoe Street North, Oshawa, Ontario (Canada)

    2009-02-15

    This study examines the feasibility of powering a range of portable devices with a direct methanol fuel cell (DMFC). The analysis includes a comparison between a Li-ion battery and DMFC to supply the power for a laptop, camcorder and a cell phone. A parametric study of the systems for an operational period of 4 years is performed. Under the assumptions made for both the Li-ion battery and DMFC system, the battery cost is lower than the DMFC during the first year of operation. However, by the end of 4 years of operational time, the DMFC system would cost less. The weight and cost comparisons show that the fuel cell system occupies less space than the battery to store a higher amount of energy. The weight of both systems is almost identical. Finally, the CO{sub 2} emissions can be decreased by a higher exergetic efficiency of the DMFC, which leads to improved sustainability. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-03-01

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

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  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. 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.

  2. Analysis performance of proton exchange membrane fuel cell (PEMFC)

    Science.gov (United States)

    Mubin, A. N. A.; Bahrom, M. H.; Azri, M.; Ibrahim, Z.; Rahim, N. A.; Raihan, S. R. S.

    2017-06-01

    Recently, the proton exchange membrane fuel cell (PEMFC) has gained much attention to the technology of renewable energy due to its mechanically ideal and zero emission power source. PEMFC performance reflects from the surroundings such as temperature and pressure. This paper presents an analysis of the performance of the PEMFC by developing the mathematical thermodynamic modelling using Matlab/Simulink. Apart from that, the differential equation of the thermodynamic model of the PEMFC is used to explain the contribution of heat to the performance of the output voltage of the PEMFC. On the other hand, the partial pressure equation of the hydrogen is included in the PEMFC mathematical modeling to study the PEMFC voltage behaviour related to the input variable input hydrogen pressure. The efficiency of the model is 33.8% which calculated by applying the energy conversion device equations on the thermal efficiency. PEMFC’s voltage output performance is increased by increasing the hydrogen input pressure and temperature.

  3. Optimum Performance of Direct Hydrogen Hybrid Fuel Cell Vehicles

    OpenAIRE

    Zhao, Hengbing; Burke, Andy

    2009-01-01

    Proton Exchange Membrane fuel cell (PEMFC) technology is one of the most attractive candidates for transportation applications due to its inherently high efficiency and high power density. However, the fuel cell system efficiency can suffer because of the need for forced air supply and water-cooling systems. Hence the operating strategy of the fuel cell system can have a significant impact on the fuel cell system efficiency and thus vehicle fuel economy. The key issues are how the fuel cell b...

  4. U.S. Department of Energy Hydrogen and Fuel Cells Program 2012 Annual Merit Review and Peer Evaluation Report: May 14-18, 2012, Arlington, VA

    Energy Technology Data Exchange (ETDEWEB)

    2012-09-01

    This document summarizes the comments provided by peer reviewers on hydrogen and fuel cell projects presented at the fiscal year (FY) 2012 U.S. Department of Energy (DOE) Hydrogen and Fuel Cells Program and Vehicle Technologies Program Annual Merit Review and Peer Evaluation Meeting (AMR), held May 14-18, 2012, in Arlington, VA.

  5. 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)

  6. 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.

  7. Performance evaluation of an open-cathode PEM fuel cell stack under ambient conditions: Case study of United Arab Emirates

    International Nuclear Information System (INIS)

    Al-Zeyoudi, Hend; Sasmito, Agus P.; Shamim, Tariq

    2015-01-01

    Highlights: • Performance evaluation of open-cathode PEM fuel cell stacks with forced air-convection. • Stack performance can vary up to 40% from winter to summer. • Hot and arid condition leads to membrane drying and performance deterioration. • Anode humidification improves the stack performance up to 40% during summer. - Abstract: The open-cathode polymer electrolyte membrane (PEM) fuel cell stack has been a promising candidate as a sustainable energy conversion system for replacing fossil fuel-based energy conversion devices in portable and automotive applications. As the ambient air is directly used to provide both oxidant and cooling, the complex cooling loop can be avoided which reduces the complexity and cost. However, the stack performance is highly affected by ambient conditions, i.e., ambient temperature and humidity. In this study, the effect of monthly ambient air conditions (temperature and humidity) is evaluated with respect to the stack’s power production performance as well as thermal, water and gas management by employing a validated three-dimensional open-cathode PEM fuel cell stack model. The annual climate data from the hot and arid environment of Abu Dhabi, United Arab Emirates (UAE) are used as a case study. The objective is to develop a better fundamental understanding of the interactions of physical phenomena in a fuel cell stack, which can assist in improving the performance and operation of an open-cathode PEM fuel cell-powered vehicle. The results indicate that the stack performance can vary significantly (up to 40%) from winter to summer, especially at high operating currents, with significant changes in the stack temperature and the water content at the membrane. Moreover, the anode humidification results in a significant improvement in the stack performance (up to 40%) in hot and dry conditions. However, a careful balance has to be struck between the humidifier parasitic load and the stack power.

  8. 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

  9. Irradiation performance of HTGR fuel in HFIR experiment HRB-13

    International Nuclear Information System (INIS)

    Tiegs, T.N.

    1982-03-01

    Irradiation capsule HRB-13 tested High-Temperature Gas-Cooled Reactor (HTGR) fuel under accelerated conditions in the High Flux Isotope Reactor (HFIR) at ORNL. The ORNL part of the capsule was designed to provide definitive results on how variously misshapen kernels affect the irradiation performance of weak-acid-resin (WAR)-derived fissile fuel particles. Two batches of WAR fissile fuel particles were Triso-coated and shape-separated into four different fractions according to their deviation from spericity, which ranged from 9.6 to 29.7%. The fissile particles were irradiated for 7721 h. Heavy-metal burnups ranged from 80 to 82.5% FIMA (fraction of initial heavy-metal atoms). Fast neutron fluences (>0.18 MeV) ranged from 4.9 x 10 25 neutrons/m 2 to 8.5 x 10 25 neutrons/m 2 . Postirradiation examination showed that the two batches of fissile particles contained chlorine, presumably introduced during deposition of the SiC coating

  10. Performance evaluation of the Loviisa advanced type fuel rods

    International Nuclear Information System (INIS)

    Ranta-Puska, K.; Pihlatie, M.

    2001-01-01

    The fuel vendor TVEL has supplied to Loviisa WWER-440 power plant six lead assemblies of an advanced type which have profiling of the fuel enrichment, demountability of the assembly and a reduced shroud wall thickness. The pool side examination programme of these assemblies is underway including visual inspections, diameter and length measurements between operation cycles, and end-of-life fission gas release measurements, determined from 85 Kr activity in the plenum. Complementary evaluations and testing of models are done with the ENIGMA fuel performance code. The diameters of the corner rods have decreased to 30 μm during the first cycle and 40 to 70 μm after two cycles (with rod burnups of 24-30 MWd/kgU). The extent of creep-down is generally as expected, and agrees with the creep model adjusted for Russian Zr1%Nb cladding type and the Loviisa coolant and neutron flux conditions. The gap closure and reversed hoop strain are to be awaited during the third cycle so the new data will be an interesting validation exercise for the model and ENIGMA. Calculated temperatures stay low, and therefore low fission gas release fractions are anticipated as well

  11. 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

  12. Annual health, safety and environmental performance report for 1992

    International Nuclear Information System (INIS)

    Orman, R.F.; Richards, S.

    1993-12-01

    This report summarizes the safety and environmental record of the operations of Atomic Energy of Canada Limited (AECL) during 1992. An introduction highlights the results and describes the facilities and organizational systems. Subsequent sections indicate the performance of the company with respect to personnel radiation exposures, occupational injuries, the handling of wastes and the release of materials into the environment. Programs in health, safety and environmental protection are presented, along with site remediation and emergency preparedness practices

  13. Performance of the Defense Acquisition System, 2013 Annual Report

    Science.gov (United States)

    2013-06-28

    causes in the 18 analyses conducted by PARCA over the last 3 years (see Bliss , 2012a; Bliss , 2013). 34 Performance of the Defense Acquisition...assumptions (Arena et al., 2012; Bliss , 2012b). Framing assumptions are any explicit or implicit assumptions central in shaping cost, schedule, and...contracts starting after 1980 (n = 367) Cy cl e tim e in y ea rs All development contract cycle times increased significantly after 1980 59

  14. Annual health, safety and environmental performance report for 1992

    International Nuclear Information System (INIS)

    Orman, R.F.; Richards, S.

    1993-12-01

    This report summarizes the safety and environmental record of the operations of Atomic Energy of Canada Limited (AECL) during 1992. an introduction highlights the results and describes the facilities and organizational systems. Subsequent sections indicate the performance of the company with respect to personnel radiation exposures, occupational injuries, the handling of wastes and the release of materials into the environment. Programs in health, safety and environmental protection are presented, along with site remediation and emergency preparedness practices

  15. Performance assessment of advanced engineering workstations for fuel management applications

    International Nuclear Information System (INIS)

    Turinsky, P.J.

    1989-07-01

    The purpose of this project was to assess the performance of an advanced engineering workstation [AEW] with regard to applications to incore fuel management for LWRs. The attributes of most interest to us that define an AEW are parallel computational hardware and graphics capabilities. The AEWs employed were super microcomputers manufactured by MASSCOMP, Inc. These computers utilize a 32-bit architecture, graphics co-processor, multi-CPUs [up to six] attached to common memory and multi-vector accelerators. 7 refs., 33 figs., 4 tabs

  16. Better fuel handling system performance through improved elastomers and seals

    Energy Technology Data Exchange (ETDEWEB)

    Wensel, R G; Metcalfe, R [Atomic Energy of Canada Ltd., Chalk River, ON (Canada)

    1997-12-31

    In the area of elastomers, tests have identified specific compounds that perform well in each class of CANDU service. They offer gains in service life, sometimes by factors of ten or more. Moreover, the aging characteristics of these specific compounds are being thoroughly investigated, whereas many elastomers used previously were either non-specific or their aging was unknown. In this paper the benefits of elastomer upgrading, as well as the deficiencies of current station elastomer practices, are discussed in the context of fuel handling equipment. Guidelines for procurement, storage, handling and condition monitoring of elastomer seals are outlined. (author). 3 figs.

  17. Better fuel handling system performance through improved elastomers and seals

    International Nuclear Information System (INIS)

    Wensel, R.G.; Metcalfe, R.

    1996-01-01

    In the area of elastomers, tests have identified specific compounds that perform well in each class of CANDU service. They offer gains in service life, sometimes by factors of ten or more. Moreover, the aging characteristics of these specific compounds are being thoroughly investigated, whereas many elastomers used previously were either non-specific or their aging was unknown. In this paper the benefits of elastomer upgrading, as well as the deficiencies of current station elastomer practices, are discussed in the context of fuel handling equipment. Guidelines for procurement, storage, handling and condition monitoring of elastomer seals are outlined. (author). 3 figs

  18. EPRI fuel performance data base: user's manual. Final report

    International Nuclear Information System (INIS)

    Simpson, J.; Lee, S.; Rumble, E.

    1980-10-01

    This user's manual provides instructions for accessing the data in the EPRI fuel performance data base (FPDB) and manipulating that data to solve specific problems that the user wishes to specify. The user interacts with the FPDB through the Relational Information Management System (RIMS) computer program. The structure and format of the FPDB and the general syntax of the data base commands are described. Instructions follow for the use of each command. Appendixes provide more detailed information about the FPDB and its software. The FPDB currently resides on a PRIME-750 computer

  19. Performance of a passive direct ethanol fuel cell

    Science.gov (United States)

    Pereira, J. P.; Falcão, D. S.; Oliveira, V. B.; Pinto, A. M. F. R.

    2014-06-01

    Ethanol emerges as an attractive fuel since it is less toxic and has higher energy density than methanol and can be produced from biomass. Direct ethanol fuel cells (DEFCs) appear as a good choice for producing sustainable energy for portable applications. However, they are still far from attaining acceptable levels of power output, since their performance is affected by the slow electrochemical ethanol oxidation and water and ethanol crossover. In the present work, an experimental study on the performance of a passive DEFC is described. Tailored MEAs (membrane electrode assembly) with different catalyst loadings, anode diffusion layers and membranes were tested in order to select optimal working conditions at high ethanol concentrations and low ethanol crossover. The performance increased with an increase of membrane and anode diffusion layer thicknesses and anode catalyst loading. A maximum power density of 1.33 mW cm-2, was obtained using a Nafion 117 membrane, 4 mg cm-2 of Pt-Ru and 2 mg cm-2 of Pt on the anode and cathode catalyst layers, ELAT as anode diffusion layer, carbon cloth as cathode diffusion layer and an ethanol concentration of 2 M. As far as the authors are aware this is the first work reporting an experimental optimization of passive DEFCs.

  20. Fueling strategies to optimize performance: training high or training low?

    Science.gov (United States)

    Burke, L M

    2010-10-01

    Availability of carbohydrate as a substrate for the muscle and central nervous system is critical for the performance of both intermittent high-intensity work and prolonged aerobic exercise. Therefore, strategies that promote carbohydrate availability, such as ingesting carbohydrate before, during and after exercise, are critical for the performance of many sports and a key component of current sports nutrition guidelines. Guidelines for daily carbohydrate intakes have evolved from the "one size fits all" recommendation for a high-carbohydrate diets to an individualized approach to fuel needs based on the athlete's body size and exercise program. More recently, it has been suggested that athletes should train with low carbohydrate stores but restore fuel availability for competition ("train low, compete high"), based on observations that the intracellular signaling pathways underpinning adaptations to training are enhanced when exercise is undertaken with low glycogen stores. The present literature is limited to studies of "twice a day" training (low glycogen for the second session) or withholding carbohydrate intake during training sessions. Despite increasing the muscle adaptive response and reducing the reliance on carbohydrate utilization during exercise, there is no clear evidence that these strategies enhance exercise performance. Further studies on dietary periodization strategies, especially those mimicking real-life athletic practices, are needed. © 2010 John Wiley & Sons A/S.

  1. Annual health, safety and environmental performance report for 1993

    Energy Technology Data Exchange (ETDEWEB)

    Gallapher, J D; Wright, M G

    1994-05-01

    This report summarizes the occupational health and safety and the environmental protection record of the operations of Atomic Energy of Canada Limited (AECL) during 1993. An introduction highlights the results and describes the facilities and organizational systems. Subsequent sections indicate the performance of the company with respect to personnel radiation exposures, occupational injuries, the handling of wastes, and the release of materials into the environment. Programs in health, safety and environmental protection are presented, along with site remediation and emergency preparedness practices. (author). 14 figs.

  2. Annual health, safety and environmental performance report for 1993

    International Nuclear Information System (INIS)

    Gallapher, J.D.; Wright, M.G.

    1994-05-01

    This report summarizes the occupational health and safety and the environmental protection record of the operations of Atomic Energy of Canada Limited (AECL) during 1993. An introduction highlights the results and describes the facilities and organizational systems. Subsequent sections indicate the performance of the company with respect to personnel radiation exposures, occupational injuries, the handling of wastes, and the release of materials into the environment. Programs in health, safety and environmental protection are presented, along with site remediation and emergency preparedness practices. (author). 14 figs

  3. Enhanced Portfolio Performance Using a Momentum Approach to Annual Rebalancing

    Directory of Open Access Journals (Sweden)

    Michael D. Mattei

    2018-02-01

    Full Text Available After diversification, periodic portfolio rebalancing has become one of the most widely practiced methods for reducing portfolio risk and enhancing returns. Most of the rebalancing strategies found in the literature are generally regarded as contrarian approaches to rebalancing. A recent article proposed a rebalancing approach that incorporates a momentum approach to rebalancing. The momentum approach had a better risk adjusted return than either the traditional approach or a Buy-and-Hold approach. This article identifies an improvement to the momentum approach and then examines the impact of transactions costs and taxes on the portfolio performance of four active rebalancing approaches.

  4. Annual index & ranking for institutional Open Access performance

    DEFF Research Database (Denmark)

    Grigorov, Ivo

    A simple institutional index based on % of peer-review publications that are accessible through (gold and/or green) OA via the internet, can be a very powerful tool for both administrators and funders to see where their efforts and institutions rank in the global OA effort. Credible answers...... to questions like: "Who are the leaders?", "Who are the latest performers?", "What publicly-funded institutes researching into topical themes with socio-economic consequences, could do better at making their research accessible to stakeholders?" can also stimulate non-commited institutes to invest more in OA...

  5. Energy harvesting influences electrochemical performance of microbial fuel cells

    Science.gov (United States)

    Lobo, Fernanda Leite; Wang, Xin; Ren, Zhiyong Jason

    2017-07-01

    Microbial fuel cells (MFCs) can be effective power sources for remote sensing, wastewater treatment and environmental remediation, but their performance needs significant improvement. This study systematically analyzes how active harvesting using electrical circuits increased MFC system outputs as compared to passive resistors not only in the traditional maximal power point (MPP) but also in other desired operating points such as the maximum current point (MCP) and the maximum voltage point (MVP). Results show that active harvesting in MPP increased power output by 81-375% and active harvesting in MCP increased Coulombic efficiency by 207-805% compared with resisters operated at the same points. The cyclic voltammograms revealed redox potential shifts and supported the performance data. The findings demonstrate that active harvesting is a very effective approach to improve MFC performance across different operating points.

  6. Annual report 2003

    International Nuclear Information System (INIS)

    2003-01-01

    Delivering products and services to nuclear power plants operators, AREVA operates in every sector of the civilian nuclear power and fuel cycle industry. This annual report 2003 provides, in seven chapters, information on persons responsible for the annual report and for auditing the financial statements, general information on the company and share capital (statute, capital, share trading, dividends), information on company operations, changes and future prospects, assets, financial position and financial performance, corporate governance, recent developments and future prospects. (A.L.B.)

  7. Annual report of the working group 'fuel pin and fuel element mechanics' of the Institut fuer Reaktortechnik (IRT) of the Technische Hochschule Darmstadt for the Fast Breeder Project

    International Nuclear Information System (INIS)

    Fabian, H.; Humbach, W.; Lassmann, K.; Mueller, J.J.; Preusser, T.; Schmelz, K.

    1978-09-01

    This report comprises six single lectures given at an information meeting organized by the Institut fuer Reaktortechnik der Technischen Hochschule Darmstadt (IRT) in Darmstadt on April 24, 1978. The lectures are an account of work performed at IRT on the mechanics of fuel pins and fuel elements and supported by the Fast Breeder Project (PSB) of KfK. These activities can be broken down into studies of the integral fuel pin (URANUS computer code) and into multidimensional studies of the fuel pin using the finite-element method (FINEL and ZIDRIG computer codes). Moreover, a report is presented of the status of the test facility for simulation of out-of-pile cladding tube loads and of the IRT project on the simulation and analysis of radiation damage. (orig./GL) [de

  8. 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)

  9. Thermal performance of fresh mixed-oxide fuel in a fast flux LMR [liquid metal reactor

    International Nuclear Information System (INIS)

    Ethridge, J.L.; Baker, R.B.

    1985-01-01

    A test was designed and irradiated to provide power-to-melt (heat generation rate necessary to initiate centerline fuel melting) data for fresh mixed-oxide UO 2 -PuO 2 fuel irradiated in a fast neutron flux under prototypic liquid metal reactor (LMR) conditions. The fuel pin parameters were selected to envelope allowable fabrication ranges and address mass production of LMR fuel using sintered-to-size techniques. The test included fuel pins with variations in fabrication technique, pellet density, fuel-to-cladding gap, Pu concentration, and fuel oxygen-to-metal ratios. The resulting data base has reestablished the expected power-to-melt in mixed-oxide fuels during initial reactor startup when the fuel temperatures are expected to be the highest. Calibration of heat transfer models of fuel pin performance codes with these data are providing more accurate capability for predicting steady-state thermal behavior of current and future mixed-oxide LMR fuels

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

    Energy Technology Data Exchange (ETDEWEB)

    Demkowicz, Paul A., E-mail: paul.demkowicz@inl.gov [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-6188 (United States); Hunn, John D. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6093 (United States); Ploger, Scott A. [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-6188 (United States); Morris, Robert N.; Baldwin, Charles A. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6093 (United States); Harp, Jason M.; Winston, Philip L. [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-6188 (United States); Gerczak, Tyler J. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6093 (United States); Rooyen, Isabella J. van [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-6188 (United States); Montgomery, Fred C.; Silva, Chinthaka M. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6093 (United States)

    2016-09-15

    Highlights: • Post-irradiation examination was performed on AGR-1 coated particle fuel. • Cesium release from the particles was very low in the absence of failed SiC layers. • Silver release was often substantial, and varied considerably with temperature. • Buffer and IPyC layers were found to play a key role in TRISO coating behavior. • Fission products palladium and silver were found in the SiC layer of particles. - Abstract: The AGR-1 experiment contained 72 low-enriched uranium oxide/uranium carbide TRISO coated particle fuel compacts in six capsules irradiated to burnups of 11.2 to 19.6% FIMA, with zero TRISO coating failures detected during the irradiation. The irradiation performance of the fuel including the extent of fission product release and the evolution of kernel and coating microstructures was evaluated based on detailed examination of the irradiation capsules, the fuel compacts, and individual particles. Fractional release of {sup 110m}Ag from the fuel compacts was often significant, with capsule-average values ranging from 0.01 to 0.38. Analysis of silver release from individual compacts indicated that it was primarily dependent on fuel temperature history. Europium and strontium were released in small amounts through intact coatings, but were found to be significantly retained in the outer pyrocarbon and compact matrix. The capsule-average fractional release from the compacts was 1 × 10{sup −4} to 5 × 10{sup −4} for {sup 154}Eu and 8 × 10{sup −7} to 3 × 10{sup −5} for {sup 90}Sr. The average {sup 134}Cs fractional release from compacts was <3 × 10{sup −6} when all particles maintained intact SiC. An estimated four particles out of 2.98 × 10{sup 5} in the experiment experienced partial cesium release due to SiC failure during the irradiation, driving {sup 134}Cs fractional release in two capsules to approximately 10{sup −5}. Identification and characterization of these particles has provided unprecedented insight into

  11. Experimental Validation of a Domestic Stratified Hot Water Tank Model in Modelica for Annual Performance Assessment

    DEFF Research Database (Denmark)

    Carmo, Carolina; Dumont, Olivier; Nielsen, Mads Pagh

    2015-01-01

    The use of stratified hot water tanks in solar energy systems - including ORC systems - as well as heat pump systems is paramount for a better performance of these systems. However, the availability of effective and reliable models to predict the annual performance of stratified hot water tanks...

  12. Irradiation performance of full-length metallic IFR fuels

    International Nuclear Information System (INIS)

    Tsai, H.; Neimark, L.A.

    1992-07-01

    An assembly irradiation of 169 full-length U-Pu-Zr metallic fuel pins was successfully completed in FFTF to a goal burnup of 10 at.%. All test fuel pins maintained their cladding integrity during the irradiation. Postirradiation examination showed minimal fuel/cladding mechanical interaction and excellent stability of the fuel column. Fission-gas release was normal and consistent with the existing data base from irradiation testing of shorter metallic fuel pins in EBR-II

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-01

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

  14. Performance Analysis and Development Strategies for Solid Oxide Fuel Cells

    International Nuclear Information System (INIS)

    Ivers-Tiffee, E; Leonide, A; Weber, A

    2011-01-01

    Solid oxide fuel cells (SOFC) are of great interest for a diverse range of applications. Within the past 10 years, an increase in power density by one order of magnitude, a lowering of the operating temperature by 200 K, and degradation rates lowered by a factor of 10 have been achieved on the cell and stack level. However, there is still room for further enhancement of the overall performance by suitably tailoring the cell components on a micro- and nanostructural level. The efficiency of the electrochemically active single cell is characterized by the linear ohmic losses within the electrolyte and by nonlinear polarization losses at the electrode-electrolyte interfaces. Both depend on material composition and operation conditions (temperature and time, fuel utilisation and gas composition). The area-specific resistance (ASR) is considered as the figure of merit for overall performance. ASR values of anode supported cells (ASC) were determined by means of impedance spectroscopy and subsequently separated into ohmic losses (mainly electrolyte) and nonlinear polarisation losses resulting from gas diffusion and activation polarization in the cathode and anode. The efficiencies of ASCs will be discussed for various material combinations in the temperature range of technological interest (between 550 deg. C and 850 deg. C).

  15. Solid oxide fuel cell performance under severe operating conditions

    DEFF Research Database (Denmark)

    Koch, Søren; Hendriksen, P.V.; Mogensen, Mogens Bjerg

    2006-01-01

    The performance and degradation of Solid Oxide Fuel Cells (SOFC) were studied under severe operating conditions. The cells studied were manufactured in a small series by ECN, in the framework of the EU funded CORE-SOFC project. The cells were of the anode-supported type with a double layer LSM...... cathode. They were operated at 750 °C or 850 °C in hydrogen with 5% or 50% water at current densities ranging from 0.25 A cm–2 to 1 A cm–2 for periods of 300 hours or more. The area specific cell resistance, corrected for fuel utilisation, ranged between 0.20 Ω cm2 and 0.34 Ω cm2 at 850 °C and 520 m......V, and between 0.51 Ω cm2 and 0.92 Ω cm2 at 750 °C and 520 mV. The degradation of cell performance was found to be low (ranging from 0 to 8%/1,000 hours) at regular operating conditions. Voltage degradation rates of 20 to 40%/1,000 hours were observed under severe operating conditions, depending on the test...

  16. Performance of fuel cell for energy supply of passive house

    Energy Technology Data Exchange (ETDEWEB)

    Badea, G.; Felseghi, R. A., E-mail: Raluca.FELSEGHI@insta.utcluj.ro; Mureşan, D.; Naghiu, G. [Technical University of Cluj-Napoca, Building Services Engineering Department, Bd. December 21, no. 128-130, 400600, Cluj-Napoca (Romania); Răboacă, S. M. [National R& D Institute for Cryogenic and Isotopic Technologies, str. Uzinei, no. 4, Rm. Vălcea, 240050 (Romania); Aşchilean, I. [SC ACI Cluj SA, Avenue Dorobanţilor, no. 70, 400609, Cluj-Napoca (Romania)

    2015-12-23

    Hydrogen technology and passive house represent two concepts with a remarkable role for the efficiency and decarbonisation of energy systems in the residential buildings area. Through design and functionality, the passive house can make maximum use of all available energy resources. One of the solutions to supply energy of these types of buildings is the fuel cell, using this technology integrated into a system for generating electricity from renewable primary sources, which take the function of backup power (energy reserve) to cover peak load and meteorological intermittents. In this paper is presented the results of the case study that provide an analysis of the energy, environmental and financial performances regarding energy supply of passive house by power generation systems with fuel cell fed with electrolytic hydrogen produced by harnessing renewable energy sources available. Hybrid systems have been configured and operate in various conditions of use for five differentiated locations according to the main areas of solar irradiation from the Romanian map. Global performance of hybrid systems is directly influenced by the availability of renewable primary energy sources - particular geo-climatic characteristics of the building emplacement.

  17. 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

  18. 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.

  19. 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.

  20. Irradiation performance updates on Korean advanced fuels for PWRs

    International Nuclear Information System (INIS)

    Jang, Y.K.; Jeon, K.L.; Kim, Y.H.; Yoo, J.S.; Kim, J.I.; Shin, J.C.; Chung, J.G.; Park, J.R.; Chung, S.K.; Kim, T.W.; Yoon, Y.B.; Park, K.M.; Yoo, M.J.; Kim, M.S.; Lee, T.H.

    2010-01-01

    The developments of advanced nuclear fuels for PWRs were started in 1999 and in 2001, respectively: PLUS7 TM for eight operating optimized power reactors of 1000 MWe class (OPR1000) and four advanced power reactors of 1400 MWe class (APR1400) under construction, and 16ACE7 TM and 17ACE7 TM for an operating 16x16 Westinghouse type plant and six operating 17x17 Westinghouse type plants. The design targets were as follows: batch average burnup up to 55 GWD/MTU, over 10% thermal margin increase, improvement of the mechanical integrity of higher seismic capability, higher debris or grid fretting wear performance, higher control rod insertion capability, increase of neutron economy, improvement of manufacturability, solving incomplete rod insertion (IRI) issue and top nozzle screw failure issue, etc. in comparison of the existing nuclear fuels. The irradiation tests using each four LTAs (Lead Test Assemblies) during 3 cycles were completed in three Korean nuclear reactors until 2009. The eight irradiation performance items which are assembly growth, rod growth, grid width growth, assembly bow, rod bow, assembly twist, rod diameter and cladding oxidation were examined in pool-side after each cycle and evaluated. The irradiation tests could be continued by expecting the good performances for next cycle from the previous cycle. After 2 cycle irradiations, the region implementation could be started in 15 nuclear power plants. Even though the verifications using the LTAs were completed, each surveillance program was launched and the irradiation performance data were being updated during region implementation. In addition to pool-side examinations (PSEs) by assembly-wise during irradiation tests, six rod-wise performance items were also examined in pool-side using each LTA after discharge. All performance items met their design criteria as a result of the evaluation. Even though the interesting ones among the irradiation performance parameters were assembly and grid growths

  1. Performance testing of thermal analysis codes for nuclear fuel casks

    International Nuclear Information System (INIS)

    Sanchez, L.C.

    1987-01-01

    In 1982 Sandia National Laboratories held the First Industry/Government Joint Thermal and Structural Codes Information Exchange and presented the initial stages of an investigation of thermal analysis computer codes for use in the design of nuclear fuel shipping casks. The objective of the investigation was to (1) document publicly available computer codes, (2) assess code capabilities as determined from their user's manuals, and (3) assess code performance on cask-like model problems. Computer codes are required to handle the thermal phenomena of conduction, convection and radiation. Several of the available thermal computer codes were tested on a set of model problems to assess performance on cask-like problems. Solutions obtained with the computer codes for steady-state thermal analysis were in good agreement and the solutions for transient thermal analysis differed slightly among the computer codes due to modeling differences

  2. 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.

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

    International Nuclear Information System (INIS)

    Felt, R.E.

    1980-05-01

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

  4. 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

  5. Annual Status Report (FY2016) Performance Assessment for the Environmental Restoration Disposal Facility

    Energy Technology Data Exchange (ETDEWEB)

    Casbon, M. A. [CH2M HILL Plateau Remediation Company, Richland, WA (United States); Nichols, W. E. [CH2M HILL Plateau Remediation Company, Richland, WA (United States)

    2017-03-15

    DOE O 435.1, Radioactive Waste Management, and DOE M 435.1-1, Radioactive Waste Management Manual, require that a determination of continued adequacy of the performance assessment (PA), composite analysis (CA), and disposal authorization statement (DAS) be made on an annual basis, and it must consider the results of data collection and analysis from research, field studies, and monitoring. Annual summaries of low-level waste (LLW) disposal operations must be prepared with respect to the conclusions and recommendations of the PA and CA, and a determination of the need to revise the PA or CA must be made. The annual summary requirement provides a structured approach for demonstrating the continued adequacy of the PA and CA in demonstrating a reasonable expectation that the performance objectives will be met. This annual summary addresses only the status of the Environmental Restoration Disposal Facility (ERDF) PA (CP-60089, Performance Assessment for the Environmental Restoration Disposal Facility, Hanford Site, Washington, formerly WCH-520 Rev. 1)1. The CA for ERDF is supported by DOE/RL-2016-62, Annual Status Report (FY 2016): Composite Analysis of Low Level Waste Disposal in the Central Plateau at the Hanford Site. The ERDF PA portion of the CA document is found in Section 3.1.4, and the ERDF operations portion is found in Section 3.3.3.2 of that document.

  6. Annual report and accounts 1980/81. [UK; Solid Fuels Avisory Service; book

    Energy Technology Data Exchange (ETDEWEB)

    1981-01-01

    The work of the Solid Fuel Advisory Service during 1980/81 is described under the headings: sales promotion, market research, service to the public, technical, chimneys and training. The accounts are then presented.

  7. Performance evaluation of open core gasifier on multi-fuels

    Energy Technology Data Exchange (ETDEWEB)

    Bhoi, P.R.; Singh, R.N.; Sharma, A.M.; Patel, S.R. [Thermo Chemical Conversion Division, Sardar Patel Renewable Energy Research Institute (SPRERI), Vallabh Vidyanagar 388 120, Gujarat (India)

    2006-06-15

    Sardar Patel renewable energy research institute (SPRERI) has designed and developed open core, throat-less, down draft gasifier and installed it at the institute. The gasifier was designed for loose agricultural residues like groundnut shells. The purpose of the study is to evaluate the gasifier on multi-fuels such as babul wood (Prosopis juliflora), groundnut shell briquettes, groundnut shell, mixture of wood (Prosopis juliflora) and groundnut shell in the ratio of 1:1 and cashew nut shell. The gasifier performance was evaluated in terms of fuel consumption rate, calorific value of producer gas and gasification efficiency. Gasification efficiency of babul wood (Prosopis juliflora), groundnut shell briquettes, groundnut shell, mixture of Prosopis juliflora and groundnut shell in the ratio of 1:1 and cashew nut shell were 72%, 66%, 70%, 64%, 70%, respectively. Study revealed that babul wood (Prosopis juliflora), groundnut shell briquettes, groundnut shell, mixture of wood (Prosopis juliflora) and groundnut shell in the ratio of 1:1 and cashew nut shell were satisfactorily gasified in open core down draft gasifier. The study also showed that there was flow problem with groundnut shell. (author)

  8. Performance assessment of DOE spent nuclear fuel and surplus plutonium

    International Nuclear Information System (INIS)

    Duguid, J.O.; Vallikat, V.; McNeish, J.

    1998-01-01

    Yucca Mountain, in southern Nevada, is under consideration by the US Department of Energy (DOE) as a potential site for the disposal of the nation's radioactive wastes in a geologic repository. The wastes consist of commercial spent fuel, DOE spent nuclear fuel (SNF), high level waste (HLW), and surplus plutonium. The DOE was mandated by Congress in the fiscal 1997 Energy and Water Appropriations Act to complete a viability assessment (VA) of the repository in September of 1998. The assessment consists of a preliminary design concept for the critical elements of the repository, a total system performance assessment (TSPA), a plan and cost estimate for completion of the license application, and an estimate of the cost to construct and operate the repository. This paper presents the results of the sensitivity analyses that were conducted to examine the behavior of DOE SNF and plutonium waste forms in the environment of the base case repository that was modeled for the TSPA-VA. Fifteen categories of DOE SNF and two Plutonium waste forms were examined and their contribution to radiation dose to humans was evaluated

  9. Reduced enrichment fuels for Canadian research reactors - Fabrication and performance

    Energy Technology Data Exchange (ETDEWEB)

    Wood, J C; Foo, M T; Berthiaume, L C; Herbert, L N; Schaefer, J D; Hawley, D [Atomic Energy of Canada Limited, Chalk River Nuclear Laboratories, Chalk River, ON KOJ 1JO (Canada)

    1985-07-01

    Our facilities have been upgraded to manufacture fuel rods comprising dispersions of U{sub 3}Si in aluminum, to complement the dispersions of U{sub 3}Si alloyed with 1.5 and 3.0 wt% Al fabricated and tested previously. Further advances have been made in process optimization particularly in core extrusion where production rate has been doubled while maintaining high quality standards. Our mini-element irradiations of Al-61.5 wt% (U,3.5 wt% Si, 1.5 wt% Al) and Al-62.4 wt% (U,3.2 wt% Si, 30 wt% Al) have been completed successfully up to the terminal burnup of 93 atomic percent. Fuel core swelling remained marginally below 1% per 10 atomic percent burnup over the whole irradiation. Also mini-elements containing Al-72.4 wt% USiAl and Al-73.4 wt% USi*Al have been irradiated to 82 atomic percent burnup, their swelling rate marginally exceeding 1% per 10 atomic percent burnup. Three full-size 12-element NRU assemblies containing Al-62.4 wt% USi*Al have been fabricated and installed in the NRU reactor where they have performed normally without problems. The cores for four more full-size 12-element NRU assemblies containing Al-61.0 wt% U{sub 3}Si have been manufactured. (author)

  10. Reduced enrichment fuels for Canadian research reactors - Fabrication and performance

    International Nuclear Information System (INIS)

    Wood, J.C.; Foo, M.T.; Berthiaume, L.C.; Herbert, L.N.; Schaefer, J.D.; Hawley, D.

    1985-01-01

    Our facilities have been upgraded to manufacture fuel rods comprising dispersions of U 3 Si in aluminum, to complement the dispersions of U 3 Si alloyed with 1.5 and 3.0 wt% Al fabricated and tested previously. Further advances have been made in process optimization particularly in core extrusion where production rate has been doubled while maintaining high quality standards. Our mini-element irradiations of Al-61.5 wt% (U,3.5 wt% Si, 1.5 wt% Al) and Al-62.4 wt% (U,3.2 wt% Si, 30 wt% Al) have been completed successfully up to the terminal burnup of 93 atomic percent. Fuel core swelling remained marginally below 1% per 10 atomic percent burnup over the whole irradiation. Also mini-elements containing Al-72.4 wt% USiAl and Al-73.4 wt% USi*Al have been irradiated to 82 atomic percent burnup, their swelling rate marginally exceeding 1% per 10 atomic percent burnup. Three full-size 12-element NRU assemblies containing Al-62.4 wt% USi*Al have been fabricated and installed in the NRU reactor where they have performed normally without problems. The cores for four more full-size 12-element NRU assemblies containing Al-61.0 wt% U 3 Si have been manufactured. (author)

  11. Diesel engine performance as influenced by fuel temperature

    Energy Technology Data Exchange (ETDEWEB)

    Sumner, H.R.; Best, W.D.; Monroe, G.E.

    1986-11-01

    The effects of diesel fuel temperature on the efficiency of a 4.4-L diesel engine were studied. Fuel temperatures of 41, 67, and 81 C were used with engine loads of 0 to 100% of full load at three engine frequencies. Regression equations were developed that predicted fuel economy as a function of PTO power at three engine frequencies. An increase in engine fuel temperature did not improve fuel economy, but did result in reduced fuel mass flow through the injector pump and reduced maximum PTO power. Reducing engine frequency improved fuel economy and supported the 'throttle back shift up' technique for saving fuel. 4 figs., 1 tab., 11 refs.

  12. 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)

  13. Annual energy performance of R744 and R410A heat pumping systems

    International Nuclear Information System (INIS)

    Jin, Zhequan; Eikevik, Trygve M.; Nekså, Petter; Hafner, Armin; Wang, Ruzhu

    2017-01-01

    Highlights: • Annual energy performance of R744 and R410A heat pumping systems is compared. • Several dynamic models of heat pumping systems are developed. • Annual energy efficiency of R744 hybrid ground-coupled heat pumping system. • The seasonal COPc and COPh of R744 hybrid system are 3.55 and 3.32. • The superiority of R744 system on the integration of two heat sinks is discussed. - Abstract: This work compares the annual energy performance of heat pumping systems using R744 and R410A as refrigerant. Focus is the annual energy efficiency of R744 hybrid ground-coupled heat pumping system. The hybrid system uses both ambient air and ground as heat sinks in the cooling mode. This is important to eliminate the underground heat accumulation phenomenon in warm climates. Several quasi-steady state models of heat pumping systems, using R744 and R410A, have been developed. Simulation results show that the annual COP_c and COP_h of an R744 hybrid system reaches 3.55 and 3.32, and its cooling performance is 42% better than for a R744 ASHP and 23% better than for a R744 GCHP system. The annual energy performance factor of a R410A ASHP system is better than for a R744 hybrid system, but the COP_c for the R410A system will be lower when the ambient temperature is higher than 30 °C.

  14. Performance Evaluation of Solid Oxide Fuel Cell by Computer ...

    African Journals Online (AJOL)

    The search for sustainable energy source that can compete with the existing one led to the discovery and acceptance of fuel cell technologies as a perfect replacement for fossil fuel. The ability of Solid Oxide Fuel Cells (SOFC) to capture the heat generation during the process of energy generation from electrochemical ...

  15. 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)

  16. 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.

  17. 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.

  18. Performance of denitrifying microbial fuel cell with biocathode over nitrite

    Directory of Open Access Journals (Sweden)

    Zhao eHuimin

    2016-03-01

    Full Text Available Microbial fuel cell (MFC with nitrite as an electron acceptor in cathode provided a new technology for nitrogen removal and electricity production simultaneously. The influences of influent nitrite concentration and external resistance on the performance of denitrifying MFC were investigated. The optimal effectiveness were obtained with the maximum total nitrogen (TN removal rate of 54.80±0.01 g m-3 d-1. It would be rather desirable for the TN removal than electricity generation at lower external resistance. Denaturing gradient gel electrophoresis suggested that Proteobacteria was the predominant phylum, accounting for 35.72%. Thiobacillus and Afipia might benefit to nitrite removal. The presence of nitrifying Devosia indicated that nitrite was oxidized to nitrate via a biochemical mechanism in the cathode. Ignavibacterium and Anaerolineaceae was found in the cathode as a heterotrophic bacterium with sodium acetate as substrate, which illustrated that sodium acetate in anode was likely permeated through proton exchange membrane to the cathode .

  19. Multiscale Modeling and Uncertainty Quantification for Nuclear Fuel Performance

    Energy Technology Data Exchange (ETDEWEB)

    Estep, Donald [Colorado State Univ., Fort Collins, CO (United States); El-Azab, Anter [Florida State Univ., Tallahassee, FL (United States); Pernice, Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States); Peterson, John W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Polyakov, Peter [Univ. of Wyoming, Laramie, WY (United States); Tavener, Simon [Colorado State Univ., Fort Collins, CO (United States); Xiu, Dongbin [Purdue Univ., West Lafayette, IN (United States); Univ. of Utah, Salt Lake City, UT (United States)

    2017-03-23

    In this project, we will address the challenges associated with constructing high fidelity multiscale models of nuclear fuel performance. We (*) propose a novel approach for coupling mesoscale and macroscale models, (*) devise efficient numerical methods for simulating the coupled system, and (*) devise and analyze effective numerical approaches for error and uncertainty quantification for the coupled multiscale system. As an integral part of the project, we will carry out analysis of the effects of upscaling and downscaling, investigate efficient methods for stochastic sensitivity analysis of the individual macroscale and mesoscale models, and carry out a posteriori error analysis for computed results. We will pursue development and implementation of solutions in software used at Idaho National Laboratories on models of interest to the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program.

  20. 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

  1. 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.

  2. 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.

  3. U.S. Department of Energy FY 1994 and 1995 annual performance report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-12-31

    This is the Department of Energy`s first Annual Performance Report. The topics of the report include a new era for the US DOE, sustainable energy, science and technology, national security--reducing the global nuclear danger, environmental quality, economic productivity through a competitive economy and the critical success factors--assessing the way the US DOE does business.

  4. Annual Status Report (FY2014) 200 West and 200 East Performance Assessments

    Energy Technology Data Exchange (ETDEWEB)

    Nichols, W. E. [CH2M Hill Plateau Remediation Company, Richland, WA (United States); Khaleel, R. [INTERA, Inc., Austin, TX (United States); Mehta, S. [CH2M Hill Plateau Remediation Company, Richland, WA (United States)

    2015-03-24

    This annual review provides the projected dose estimates of radionuclide inventories disposed in the active Low-Level Waste Burial Grounds (LLBGs) since September 26, 1988. These estimates are calculated using the original does methodology developed in the performance assessment analyses.

  5. BWR fuel assembly with improved spacer and fuel bundle design for enhanced thermal-hydraulic performance

    International Nuclear Information System (INIS)

    Mildrum, C.M.; Taleyarkhan, R.P.

    1987-01-01

    In a fuel assembly having a bundle of elongated fuel rods disposed in side-by-side relationship so as to form an array of spaced fuel rods, an outer tubular flow channel surrounding the fuel rods so as to direct flow of coolant/moderator fluid along the fuel rods, a hollow water cross extending centrally through and interconnected with the outer flow channel so as to divide the channel into separate compartments and the bundle of fuelrods into a plurality of mini-bundles thereof being disposed in the compartments, and spacers axially displaced along the fuel rods in each of the mini-bundles thereof. Each spacer is composed of inner and outer means which together define spacer cells at corner, side and interior locations of the spacer and have respective protrusions formed thereon which extend into cells so as to maintain the fuel rods received through the spacer cells in laterally spaced relationships. The improvement is described which comprises: (a) a generally uniform poison coating within at least a majority of the fuel rods; (b) a predetermined pattern of fuel enrichment with respect to the fuel rods of each mini-bundle thereof which together with the uniform poison coating within the fuel rods ensures that the packing powers of the fuel rods in the corner and side cells of the spacers are less than the peaking power of a leading one of the fuel rods in the interior cells of the spacers; and (c) each of the fuel rods being received through the cells of each spacer having a diametric size smaller than that of each of the fuel rods received through the side and interior cells of each spacer, the diametric sizes of each of the fuel rods received through the side and interior cells of each spacer being generally equal

  6. 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.

  7. Activities promoting the achievement of high nuclear fuel performance indicators

    International Nuclear Information System (INIS)

    Naev, I.; Tomov, A.

    2011-01-01

    This presentation begins with brief general information about Kozloduy Nuclear Power Plant and organization activities about fresh fuel delivery assurance. The TVSA implementation, fuel cycle, fresh fuel standard entrance inspection and additional fresh fuel inspection are briefly described. Activities concerning core refueling, radiochemistry analysis, control rods drop time, measurement of the distance between the reactor flange and PTU flange, specific items for core unloading and a comparison between the two variants for operations scope with full and without full core unloading are presented. The core unloading - results and next steps, final core design (Unit 6, 2010), preparing for core loading (Unit 6, 2010) , core loading (Unit 6, 2010), after loading core inspection (Unit 6, 2010), core inspection, reactor assembling (Unit 6, 2010), fuel control during reactor startup, fuel control during operation period and fuel assembly data base are also discussed

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  9. Correlation of creep and swelling with fuel pin performance

    International Nuclear Information System (INIS)

    Jackson, R.J.; Washburn, D.F.; Garner, F.A.; Gilbert, E.R.

    1975-09-01

    The HEDL PNL-11 experiment described was one in a series of fueled subassemblies irradiated in EBR-II to demonstrate the adequacy of the FFTF fuel pin design. The cladding material, dimensions, and fuel density are prototypic of FFTF. Because neutron flux in EBR-II is lower than in FFTF, the uranium enrichment is higher in these experimental fuel pins, irradiated in EBR-II, than the FFTF enrichment for comparable linear heat rates. Some pertinent oprating conditions for the center fuel pin in this experiment are listed. This 37-pin subassembly represents, at 110,000 MWd/MTM, the highest burnup yet attained by a prototypic FFTF subassembly. Similarly, this is the highest fluence presently attained by prototypic fuel pins. A cladding breach occurred in one fuel pin which is presently being examined. Results are presented and discussed

  10. Alternate-Fueled Combustor-Sector Performance—Part A: Combustor Performance and Part B: Combustor Emissions

    OpenAIRE

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

    2012-01-01

    Alternate aviation fuels for military or commercial use are required to satisfy MIL-DTL-83133F or ASTM D 7566 standards, respectively, and are classified as “drop-in’’ fuel replacements. To satisfy legacy issues, blends to 50% alternate fuel with petroleum fuels are acceptable. 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...

  11. 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.

  12. An experimental assessment on the influence of high octane fuels on biofuel based dual fuel engine performance, emission, and combustion

    Directory of Open Access Journals (Sweden)

    Masimalai Senthilkumar

    2017-01-01

    Full Text Available This paper presents an experimental study on the effect of different high octane fuels (such as eucalyptus oil, ethanol, and methanol on engine’s performance behaviour of a biofuel based dual fuel engine. A single cylinder Diesel engine was modified and tested under dual fuel mode of operation. Initially the engine was run using neat diesel, neat mahua oil as fuels. In the second phase, the engine was operated in dual fuel mode by using a specially designed variable jet carburettor to supply the high octane fuels. Engine trials were made at 100% and 40% loads (power outputs with varying amounts of high octane fuels up-to the maximum possible limit. The performance and emission characteristics of the engine were obtained and analysed. Results indicated significant improvement in brake thermal efficiency simultaneous reduction in smoke and NO emissions in dual fuel operation with all the inducted fuels. At 100% load the brake thermal efficiency increased from 25.6% to a maximum of 32.3, 30.5, and 28.4%, respectively, with eucalyptus oil, ethanol, and methanol as primary fuels. Smoke was reduced drastically from 78% with neat mahua oil a minimum of 41, 48, and 53%, respectively, with eucalyptus oil, ethanol, and methanol at the maximum efficiency point. The optimal energy share for the best engine behaviour was found to be 44.6, 27.3, and 23.2%, respectively, for eucalyptus oil, ethanol, and methanol at 100% load. Among the primary fuels tested, eucalyptus oil showed the maximum brake thermal efficiency, minimum smoke and NO emissions and maximum energy replacement for the optimal operation of the engine.

  13. 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)

  14. 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.

  15. Development of Improved Models and Designs for Coated-Particle Gas Reactor Fuels (I-NERI Annual Report)

    International Nuclear Information System (INIS)

    Petti, David Andrew; Maki, John Thomas; Languille, Alain; Martin, Philippe; Ballinger, Ronald

    2002-01-01

    The objective of this INERI project is to develop improved fuel behavior models for gas reactor coated particle fuels and to develop improved coated-particle fuel designs that can be used reliably at very high burnups and potentially in fast gas-cooled reactors. Thermomechanical, thermophysical, and physiochemical material properties data were compiled by both the US and the French and preliminary assessments conducted. Comparison between U.S. and European data revealed many similarities and a few important differences. In all cases, the data needed for accurate fuel performance modeling of coated particle fuel at high burnup were lacking. The development of the INEEL fuel performance model, PARFUME, continued from earlier efforts. The statistical model being used to simulate the detailed finite element calculations is being upgraded and improved to allow for changes in fuel design attributes (e.g. thickness of layers, dimensions of kernel) as well as changes in important material properties to increase the flexibility of the code. In addition, modeling of other potentially important failure modes such as debonding and asphericity was started. A paper on the status of the model was presented at the HTR-2002 meeting in Petten, Netherlands in April 2002, and a paper on the statistical method was submitted to the Journal of Nuclear Material in September 2002. Benchmarking of the model against Japanese and an older DRAGON irradiation are planned. Preliminary calculations of the stresses in a coated particle have been calculated by the CEA using the ATLAS finite element model. This model and the material properties and constitutive relationships will be incorporated into a more general software platform termed Pleiades. Pleiades will be able to analyze different fuel forms at different scales (from particle to fuel body) and also handle the statistical variability in coated particle fuel. Diffusion couple experiments to study Ag and Pd transport through SiC were

  16. Operational requirements of spherical HTR fuel elements and their performance

    International Nuclear Information System (INIS)

    Roellig, K.; Theymann, W.

    1985-01-01

    The German development of spherical fuel elements with coated fuel particles led to a product design which fulfils the operational requirements for all HTR applications with mean gas exit temperatures from 700 deg C (electricity and steam generation) up to 950 deg C (supply of nuclear process heat). In spite of this relatively wide span for a parameter with strong impact on fuel element behaviour, almost identical fuel specifications can be used for the different reactor purposes. For pebble bed reactors with relatively low gas exit temperatures of 700 deg C, the ample design margins of the fuel elements offer the possibility to enlarge the scope of their in-service duties and, simultaneously, to improve fuel cycle economics. This is demonstrated for the HTR-500, an electricity and steam generating 500 MWel eq plant presently proposed as follow-up project to the THTR-300. Due to the low operating temperatures of the HTR-500 core, the fuel can be concentrated in about 70% of the pebbles of the core thus saving fuel cycle costs. Under all design accident conditions fuel temperatures are maintained below 1250 deg C. This allows a significant reduction in the engineered activity barriers outside the primary circuit, in particular for the loss of coolant accident. Furthermore, access to major primary circuit components and the reuse of the fuel elements after any design accident are possible. (author)

  17. 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)

  18. 1997-1998 Annual Review of the 200 West and 200 East area performance assessments

    International Nuclear Information System (INIS)

    WOOD, M.I.

    1999-01-01

    An annual review of the 200 West and 200 East Area Performance Assessment (PA) analyses for fiscal year 1998 was completed. Burial ground disposal operations were found to be compliant with performance objectives in DOE Order 5820.2A. Other newly generated information and analyses relevant to PA assumptions and results were summarized. This report was initially submitted to the Department of Energy-Richland Office (DOE-RL) as a letter report in October, 1998

  19. 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

  20. 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

  1. Early-in-life thermal performance of UO2--PuO2 fast reactor fuel

    International Nuclear Information System (INIS)

    Baker, R.B.; Leggett, R.D.

    1979-01-01

    Results from the combined analyses of two thermal performance tests, HEDL P-19 and HEDL P-20 are described. The tests were designed to provide data on the power required to cause incipient fuel melting early in life under conditions prototypic of FFTF driver fuel pins and similar FBR fuel systems

  2. 76 FR 3517 - Standards of Performance for Fossil-Fuel-Fired, Electric Utility, Industrial-Commercial...

    Science.gov (United States)

    2011-01-20

    ... Standards of Performance for Fossil-Fuel-Fired, Electric Utility, Industrial-Commercial-Institutional, and... following: Category NAICS \\1\\ Examples of regulated entities Industry 221112 Fossil fuel-fired electric utility steam generating units. Federal Government 22112 Fossil fuel-fired electric utility steam...

  3. Annual report of Power Reactor and Nuclear Fuel Development Corporation, fiscal year 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-09-01

    The experimental FBR `Joyo` has continued the irradiation operation at 100 MWt. After the 11th periodic inspection, the 30th cycle operation was carried out. The cumulative operation time as of the end of the fiscal year was 51,630 hours, and the cumulative heat output was about 4.2 billion kWh. The prototype FBR `Monju` has succeeded in electric power generation in August, 1995, but the sodium leak accident occurred in December, 1995. The elucidation of the cause of the sodium leak accident and the total inspection for the safety have been carried out. As for FBRs, the research and development of the reactor physics, the design of a large FBR, the equipment systems, the fuel and materials, the structures and the safety have been advanced. The ATR `Fugen` Power Station has continued the operation smoothly, and as of the end of the fiscal year, the total generated electric power was about 17.3 billion kWh, and the capacity factor was 66.3%. It boasts about the result of using MOX fuel. The exploration of uranium resources, the development of uranium conversion, uranium enrichment and plutonium fuel, the reprocessing of spent fuel, the development of environmental technology for radioactive waste, creative and innovative research and development, safety control and safety research and others are reported. (K.I.)

  4. Annual and activity report 2005 - INB - Brazilian Nuclear Industries. Nuclear fuel: technology for the essential

    International Nuclear Information System (INIS)

    2005-01-01

    This document reports the activities of Brazilian Nuclear Industry company during 2005 as follows: uranium isotope enrichment; production of nuclear fuel; mineral resources; finance and administration; planning and sales; quality, safety and environment, communication and social action; economic and financial management

  5. 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

  6. 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.

  7. U.S. Department of Energy Hydrogen and Fuel Cells Program 2016 Annual Merit Review and Peer Evaluation Report: June 6-10, 2016, Washington, DC

    Energy Technology Data Exchange (ETDEWEB)

    Popovich, Neil

    2016-10-01

    The fiscal year 2016 U.S. Department of Energy (DOE) Hydrogen and Fuel Cells Program Annual Merit Review and Peer Evaluation Meeting (AMR), in conjunction with DOE's Vehicle Technologies Office AMR, was held from June 6-10, 2015, in Washington, D.C.. This report is a summary of comments by AMR peer reviewers about the hydrogen and fuel cell projects funded by DOE's Office of Energy Efficiency and Renewable Energy.

  8. U.S. Department of Energy Hydrogen and Fuel Cells Program 2015 Annual Merit Review and Peer Evaluation Report: June 8-12, 2015, Arlington, Virginia

    Energy Technology Data Exchange (ETDEWEB)

    Popovich, Neil

    2015-10-01

    The fiscal year 2015 U.S. Department of Energy (DOE) Hydrogen and Fuel Cells Program Annual Merit Review and Peer Evaluation Meeting (AMR), in conjunction with DOE's Vehicle Technologies Office AMR, was held from June 8-12, 2015, in Arlington, Virginia. This report is a summary of comments by AMR peer reviewers about the hydrogen and fuel cell projects funded by DOE's Office of Energy Efficiency and Renewable Energy.

  9. Optimal setpoint generation for improved fuel temperature performance

    International Nuclear Information System (INIS)

    Johns, R.M.; Edwards, R.M.

    1995-01-01

    Nuclear power plant systems feature a high degree of non-linearity and high noise level, and the performance of conventional control systems may degrade when power plants operate under a wide range of conditions, such as startup, test, shutdown, etc. The conventional control system is not intended for nuclear power plant full-range operation. This is the reason that, at present, nuclear power plants rely on manual operations for most wide-range control and only use automatic control around nominal conditions. The availability of new powerful control techniques and mathematical tools has motivated an expanding research effort toward the development of the advanced hybrid feedforward-feedback control system. The planned command input is based on the analysis of a system model in some form in order to improve the performance of the overall system. The use of a feedforward optimal controller to improve the fuel temperature response to a step change in desired reactor power is being demonstrated. The Penn State TRIGA reactor is used as the basis of the reactor model so that validation of the controller may be shown

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

    OpenAIRE

    Teresa J. Leo; Miguel A. Raso; Emilio Navarro; Eleuterio Mora

    2013-01-01

    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 opera...

  11. Irradiation performance of HTGR fuel rods in HFIR experiments HRB-11 and -12

    International Nuclear Information System (INIS)

    Homan, F.J.; Tiegs, T.N.; Kania, M.J.; Long, E.L. Jr.; Thoms, K.R.; Robbins, J.M.; Wagner, P.

    1980-06-01

    Capsules HRB-11 and -12 were irradiated in support of development of weak-acid-resin-derived recycle fuel for the high-enriched uranium (HEU) fuel cycle for the HTGR. Fissil fuel particles with initial oxygen-to-metal ratios between 1.0 and 1.7 performed acceptably to full burnup for HEU fuel. Particles with ratios below 1.0 showed excessive chemical interaction between rare earth fission products and the SiC layer

  12. 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

  13. 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.

  14. 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...

  15. Effect of variation in LPG composition on emissions and performance in a dual fuel diesel engine

    Energy Technology Data Exchange (ETDEWEB)

    H.E. Saleh [Mattaria, Helwan University, Cairo (Egypt). Department of Mechanical Power Engineering

    2008-10-15

    This paper investigates the effect of variation in LPG composition on emissions and performance characteristics in a dual fuel engine run on diesel fuel and five gaseous fuel of LPG with different composition. To quantify the best LPG composition for dual fuel operation especially in order to improve the exhaust emissions quality while maintaining high thermal efficiency comparable to a conventional diesel engine, a two-cylinder, naturally aspirated, four-stroke, DI diesel engine converted to run as pilot-injected dual fuel engine. The tests and data collection were performed under various conditions of load at constant engine speed. From the results, it is observed that the exhaust emissions and fuel conversion efficiency of the dual fuel engine are found to be affected when different LPG composition is used as higher butane content lead to lower NOx levels while higher propane content reduces CO levels. Fuel No. 3 (70% propane, 30% butane) with mass fraction 40% substitution of the diesel fuel was the best LPG composition in the dual fuel operation except that at part loads. Also, tests were made for fuel No. 3-diesel blend in the dual fuel operation at part loads to improve the engine performances and exhaust emissions by using the Exhaust Gas Recirculation (EGR) method. 26 refs., 15 figs., 5 tabs.

  16. 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)

  17. 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%.

  18. Study on the performance of fuel elements with carbide and carbide-nitride fuel

    International Nuclear Information System (INIS)

    Golovchenko, Yu.M.; Davydov, E.F.; Maershin, A.A.

    1985-01-01

    Characteristics, test conditions and basic results of material testing of fuel elements with carbide and carbonitride fuel irradiated in the BOR-60 reactor up to 3-10% burn-up at specific power rate of 55-70 kW/m and temperatures of the cladding up to 720 deg C are described. Increase of cladding diameter is stated mainly to result from pressure of swelling fuel. The influence of initial efficient porosity of the fuel on cladding deformation and fuel stoichiometry on steel carbonization is considered. Utilization of carbide and carbonitride fuel at efficient porosity of 20% at the given test modes is shown to ensure their operability up to 10% burn-up

  19. Fuel cell mining vehicles: design, performance and advantages

    International Nuclear Information System (INIS)

    Betournay, M.C.; Miller, A.R.; Barnes, D.L.

    2003-01-01

    The potential for using fuel cell technology in underground mining equipment was discussed with reference to the risks associated with the operation of hydrogen vehicles, hydrogen production and hydrogen delivery systems. This paper presented some of the initiatives for mine locomotives and fuel cell stacks for underground environments. In particular, it presents the test results of the first applied industrial fuel cell vehicle in the world, a mining and tunneling locomotive. This study was part of an international initiative managed by the Fuel Cell Propulsion Institute which consists of several mining companies, mining equipment manufacturers, and fuel cell technology developers. Some of the obvious benefits of fuel cells for underground mining operations include no exhaust gases, lower electrical costs, significantly reduced maintenance, and lower ventilation costs. Another advantage is that the technology can be readily automated and computer-based for tele-remote operations. This study also quantified the cost and operational benefits associated with fuel cell vehicles compared to diesel vehicles. It is expected that higher vehicle productivity could render fuel cell underground vehicles cost-competitive. 6 refs., 1 tab

  20. MOX fuel fabrication, in reactor performance and improvement

    International Nuclear Information System (INIS)

    Vliet, J. van; Deramaix, P.; Nigon, J.L.; Fournier, W.

    1998-01-01

    In Europe, MOX fuel for light water reactors (LWRs) has first been manufactured in Belgium and Germany. Belgonucleaire (BN) loaded the first MOX assembly in the BR3 Pressurised Water Reactor (PWR) in 1963. In June 1998, more than 750 tHM LWR MOX fuel assemblies were manufactured on a industrial scale in Europe without any particular difficulty relating to fuel fabrication, reactor operation or fuel behaviour. So, today plutonium recycling through MOX fuel is a mature industry, with successful operational experience and large-scale fabrication plants. In this field, COGEMA and BELGONUCLEAIRE are the main actors by operating simultaneously three complete multidesign fuel production plants: MELOX plant (in Marcoule), CADARACHE plant and P0 plant (in Dessel, Belgium). Present MOX production capacity available to COGEMA and BN fits 175 tHM per year and is to be extended to reach about 325 tHM in the year 2000. This will represent 75% of the total MOX fabrication capacity in Europe. The industrial mastery and the high production level in MOX fabrication assured by high technology processes confer to these companies a large expertise for Pu recycling. This allows COGEMA and BN to be major actors in Pu-based fuels in the coming second nuclear era with advanced fuel cycles. (author)

  1. Recent Advances in High-Performance Direct Methanol Fuel Cells

    Science.gov (United States)

    Narayanan, S. R.; Chun, W.; Valdez, T. I.; Jeffries-Nakamura, B.; Frank, H.; Surumpudi, S.; Halpert, G.; Kosek, J.; Cropley, C.; La Conti, A. B.; hide

    1996-01-01

    Direct methanol fuel cells for portable power applications have been advanced significantly under DARPA- and ARO-sponsored programs over the last five years. A liquid-feed, direct methanol fuel cell developed under these programs, employs a proton exchange membrane as electrolyte and operates on aqueous solutions of methanol with air or oxygen as the oxidant.

  2. Technical Advisory Committee on the nuclear fuel waste management program : thirteenth annual report

    International Nuclear Information System (INIS)

    Shemilt, L.W.

    1993-03-01

    Since the last reporting period by the Technical Advisory Committee (TAC) the emphasis of the work in the Canadian Nuclear Fuel Waste Management Program (CNFWMP) has been on the writing of the Environmental Impact Statement (EIS) and the associated set of nine primary reference documents as well as supporting documents. These are in preparation for submission to the Environmental Assessment Review Panel who will lead the national evaluation of the disposal concept under the auspices of the Federal Environmental Assessment Review Office (FEARO). The disposal concept developed over the last fourteen years by Atomic Energy of Canada Limited (AECL) and anticipated to be presented by means of the EIS in 1994, is based on a multiple system of natural and man-made barriers wherein nuclear waste is first enclosed in corrosion-resistant containers, designed to last at least 500 years, and then placed in a vault excavated 500 - 1000 m deep in granitic rocks of the Canadian Shield. After container emplacement either in or on the floor of the vault, and with a surrounding buffer material of a bentonite clay/sand mixture, the vault will be backfilled and sealed with crushed rock, buffer and sand, as will be the shafts and exploratory boreholes. The case study being presented by AECL to demonstrate the safety of this concept and the technology to implement it, relies on computer simulations of a hypothetical disposal site with geological characteristics similar to those at the Underground Research Laboratory (URL) in the Whiteshell Research Area (WRA) located in Manitoba. The preliminary simulation results suggest that safe containment can be achieved provided that the waste is surrounded by a sparsely-fractured zone of rock wherein movement of contaminants carried by groundwater is modelled as a diffusive as opposed to a advective process. The principal focus of work during the past year within the environmental and safety assessment has been to complete the Post

  3. Performance of refractory alloy-clad fuel pins

    International Nuclear Information System (INIS)

    Dutt, D.S.; Cox, C.M.; Millhollen, M.K.

    1984-12-01

    This paper discusses objectives and basic design of two fuel-cladding tests being conducted in support of SP-100 technology development. Two of the current space nuclear power concepts use conventional pin type designs, where a coolant removes the heat from the core and transports it to an out-of-core energy conversion system. An extensive irradiation testing program was conducted in the 1950's and 1960's to develop fuel pins for space nuclear reactors. The program emphasized refractory metal clad uranium nitride (UN), uranium carbide (UC), uranium oxide (UO 2 ), and metal matrix fuels (UCZr and BeO-UO 2 ). Based on this earlier work, studies presented here show that UN and UO 2 fuels in conjunction with several refractory metal cladding materials demonstrated high potential for meeting space reactor requirements and that UC could serve as an alternative but higher risk fuel

  4. Performance of HT9 clad metallic fuel at high temperature

    International Nuclear Information System (INIS)

    Pahl, R.G.; Lahm, C.E.; Hayes, S.L.

    1992-01-01

    Steady-state testing of HT9 clad metallic fuel at high temperatures was initiated in EBR-II in November of 1987. At that time U-10 wt. % Zr fuel clad with the low-swelling ferritic/martensitic alloy HT9 was being considered as driver fuel options for both EBR-II and FFTF. The objective of the X447 test described here was to determine the lifetime of HT9 cladding when operated with metallic fuel at beginning of life inside wall temperatures approaching ∼660 degree C. Though stress-temperature design limits for HT9 preclude its use for high burnup applications under these conditions due to excessive thermal creep, the X447 test was carried out to obtain data on high temperature breach phenomena involving metallic fuel since little data existed in that area

  5. 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)

  6. Perform Thermodynamics Measurements on Fuel Cycle Case Study Systems

    Energy Technology Data Exchange (ETDEWEB)

    Martin, Leigh R. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-09-01

    This document was prepared to meet FCR&D level 3 milestone M3FT-14IN0304022, “Perform Thermodynamics Measurements on Fuel Cycle Case Study Systems.” This work was carried out under the auspices of the Thermodynamics and Kinetics FCR&D work package. This document reports preliminary work in support of determining the thermodynamic parameters for the ALSEP process. The ALSEP process is a mixed extractant system comprised of a cation exchanger 2-ethylhexyl-phosphonic acid mono-2-ethylhexyl ester (HEH[EHP]) and a neutral solvating extractant N,N,N’,N’-tetraoctyldiglycolamide (TODGA). The extractant combination produces complex organic phase chemistry that is challenging for traditional measurement techniques. To neutralize the complexity, temperature dependent solvent extraction experiments were conducted with neat TODGA and scaled down concentrations of the ALSEP formulation to determine the enthalpies of extraction for the two conditions. A full set of thermodynamic data for Eu, Am, and Cm extraction by TODGA from 3.0 M HNO3 is reported. These data are compared to previous extraction results from a 1.0 M HNO3 aqueous medium, and a short discussion of the mixed HEH[EHP]/TODGA system results is offered.

  7. Software Design Document for the AMP Nuclear Fuel Performance Code

    International Nuclear Information System (INIS)

    Philip, Bobby; Clarno, Kevin T.; Cochran, Bill

    2010-01-01

    The purpose of this document is to describe the design of the AMP nuclear fuel performance code. It provides an overview of the decomposition into separable components, an overview of what those components will do, and the strategic basis for the design. The primary components of a computational physics code include a user interface, physics packages, material properties, mathematics solvers, and computational infrastructure. Some capability from established off-the-shelf (OTS) packages will be leveraged in the development of AMP, but the primary physics components will be entirely new. The material properties required by these physics operators include many highly non-linear properties, which will be replicated from FRAPCON and LIFE where applicable, as well as some computationally-intensive operations, such as gap conductance, which depends upon the plenum pressure. Because there is extensive capability in off-the-shelf leadership class computational solvers, AMP will leverage the Trilinos, PETSc, and SUNDIALS packages. The computational infrastructure includes a build system, mesh database, and other building blocks of a computational physics package. The user interface will be developed through a collaborative effort with the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Capability Transfer program element as much as possible and will be discussed in detail in a future document.

  8. 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

  9. 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

  10. IAMBUS, a computer code for the design and performance prediction of fast breeder fuel rods

    International Nuclear Information System (INIS)

    Toebbe, H.

    1990-05-01

    IAMBUS is a computer code for the thermal and mechanical design, in-pile performance prediction and post-irradiation analysis of fast breeder fuel rods. The code deals with steady, non-steady and transient operating conditions and enables to predict in-pile behavior of fuel rods in power reactors as well as in experimental rigs. Great effort went into the development of a realistic account of non-steady fuel rod operating conditions. The main emphasis is placed on characterizing the mechanical interaction taking place between the cladding tube and the fuel as a result of contact pressure and friction forces, with due consideration of axial and radial crack configuration within the fuel as well as the gradual transition at the elastic/plastic interface in respect to fuel behavior. IAMBUS can be readily adapted to various fuel and cladding materials. The specific models and material correlations of the reference version deal with the actual in-pile behavior and physical properties of the KNK II and SNR 300 related fuel rod design, confirmed by comparison of the fuel performance model with post-irradiation data. The comparison comprises steady, non-steady and transient irradiation experiments within the German/Belgian fuel rod irradiation program. The code is further validated by comparison of model predictions with post-irradiation data of standard fuel and breeder rods of Phenix and PFR as well as selected LWR fuel rods in non-steady operating conditions

  11. Performance Degradation Tests of Phosphoric Acid Doped PBI Membrane Based High Temperature PEM Fuel Cells

    DEFF Research Database (Denmark)

    Zhou, Fan; Araya, Samuel Simon; Grigoras, Ionela

    2014-01-01

    Degradation tests of two phosphoric acid (PA) doped PBI membrane based HT-PEM fuel cells were reported in this paper to investigate the effects of start/stop and the presence of methanol in the fuel to the performance degradation. Continuous tests with H2 and simulated reformate which was composed...... of H2, water steam and methanol as the fuel were performed on both single cells. 12-h-startup/12-h-shutdown dynamic tests were performed on the first single cell with pure dry H2 as the fuel and on the second single cell with simulated reformate as the fuel. Along with the tests electrochemical...... techniques such as polarization curves and electrochemical impedance spectroscopy (EIS) were employed to study the degradation mechanisms of the fuel cells. Both single cells showed an increase in the performance in the H2 continuous tests, because of a decrease in the ORR kinetic resistance probably due...

  12. 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

  13. 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

  14. Plant Performance of Solid Oxide Fuel Cell Systems Fed by Alternative Fuels

    DEFF Research Database (Denmark)

    Rokni, Masoud

    2016-01-01

    Different plant design for several fuel types such as natural gas, methanol, ethanol, DME, ammonia and pure hydrogen are presented and analysed. Anode recirculation which is an important issue in SOFC plants are also explored and studied. It is shown that depending on type of the fuel whether fuel...... recycle increases plant efficiency only if fuel utilization factor is low. Other important issues such as why plant efficiency is lower when it is fed with hydrogen or biogas compared to when it is fed by other fuels such as methanol, ethanol, DME and ammonia will also be discussed and explained....... For example, plant efficiency of 45%, 54% and 50.5% can be achieved if the hydrogen, ethanol and methanol are used respectively....

  15. Analysis of WWER-440 fuel performance under normal operating conditions

    Energy Technology Data Exchange (ETDEWEB)

    Gunduz, Oe; Koese, S; Akbas, T [Atomenerjisi Komisyonu, Ankara (Turkey); Colak, Ue [Ankara Nuclear Research and Training Center (Turkey)

    1994-12-31

    FRAPCON-2 code originally developed for LWR fuel behaviour simulation is used to analyse the WWER-440 fuel rod behaviour at normal operational conditions. The code is capable of utilizing different models for mechanical analysis and gas release calculations. Heat transfer calculations are accomplished through a collocation technique by the method of weighted residuals. Temperature and burnup element properties are evaluated using MATPRO package. As the material properties of Zr-1%Nb used as cladding in WWER-440s are not provided in the code, Zircaloy-4 is used as a substitute for Zr-1%Nb. Mac-Donald-Weisman model is used for gas release calculation. FRACAS-1 and FRACAS-2 models are used in the mechanical calculations. It is assumed that the reactor was operated for 920 days (three consecutive cycles), the burnup being 42000 Mwd/t U. Results of the fuel rod behaviour analysis are given for three axial nodes: bottom node, central node and top node. The variations of the following characteristic fuel rod parameters are studied through the prescribed power history: unmoved gap thickness, gap heat transfer coefficient, fuel axial elongation, cladding axial elongation, fuel centerline temperature and ZrO-thickness at cladding surface. The value of each parameter is calculated as a function of the effective power days for the three nodes by using FRACAS-1 and FRACAS-2 codes for comparison.The results show that calculations with deformable pellet approximation with FRACAS-II model could provide better information for the behaviour of a typical fuel rod. Calculations indicate that fuel rod failure is not observed during the operation. All fuel rod parameters investigated are found to be within the safety limits. It is concluded, however, that for better assessment of reactor safety these calculations should be extended for transient conditions such as LOCA. 1 tab., 10 figs., 4 refs.

  16. Comprehensive assessment of firm financial performance using financial ratios and linguistic analysis of annual reports

    OpenAIRE

    Renáta Myšková; Petr Hájek

    2017-01-01

    Indicators of financial performance, especially financial ratio analysis, have become important financial decision-support information used by firm management and other stakeholders to assess financial stability and growth potential. However, additional information may be hidden in management communication. The article deals with the analysis of the annual reports of U.S. firms from both points of view, a financial one based on a set of financial ratios, and a linguistic one based on the anal...

  17. 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

  18. 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)

  19. 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

  20. Saclay Center of Nuclear Studies, Direction of Materials and Nuclear Fuels, Department of Physico-Chemistry, Division of Physical Chemistry. 1968 Annual report

    International Nuclear Information System (INIS)

    Schmidt, M.; Clerc, M.; Le Calve, J.; Bourene, M.; Lesigne, B.; Gillois, M.; Devillers, C.; Arvis, M.; Gilles, L.; Moreau, M.; Sutton, J.; Faraggi, M.; Desalos, J.; Tran Dinh Son; Barat, F.; Hickel, B.; Chachaty, C.; Forchioni, A.; Shiotani, M.; Larher, Y.; Maurice, P.; Le Bail, H.; Nenner, T.

    1969-03-01

    This document is the 1968 annual report of research activities at the Physico-Chemistry Department (Physical Chemistry Division), part of the Directorate of Materials and Nuclear Fuels of the CEA Saclay center of nuclear studies. The report is divided into two main parts: radiolysis and photolysis studies (gaseous phase, condensed phase), and general physico-chemical studies (sorption, molecular jets)

  1. Performance Specification Shippinpark Pressurized Water Reactor Fuel Drying and Canister Inerting System for PWR Core 2 Blanket Fuel Assemblies Stored within Shippingport Spent Fuel Canisters

    International Nuclear Information System (INIS)

    JOHNSON, D.M.

    2000-01-01

    This specification establishes the performance requirements and basic design requirements imposed on the fuel drying and canister inerting system for Shippingport Pressurized Water Reactor (PWR) Core 2 blanket fuel assemblies (BFAs) stored within Shippingport spent fuel (SSFCs) canisters (fuel drying and canister inerting system). This fuel drying and canister inerting system is a component of the U.S. Department of Energy, Richland Operations Office (RL) Spent Nuclear Fuels Project at the Hanford Site. The fuel drying and canister inerting system provides for removing water and establishing an inert environment for Shippingport PWR Core 2 BFAs stored within SSFCs. A policy established by the U.S. Department of Energy (DOE) states that new SNF facilities (this is interpreted to include structures, systems and components) shall achieve nuclear safety equivalence to comparable U.S. Nuclear Regulatory Commission (NRC)-licensed facilities. This will be accomplished in part by applying appropriate NRC requirements for comparable NRC-licensed facilities to the fuel drying and canister inerting system, in addition to applicable DOE regulations and orders

  2. State and Alternative Fuel Provider Fleets - Fleet Compliance Annual Report: Model Year 2015, Fiscal Year 2016

    Energy Technology Data Exchange (ETDEWEB)

    2016-12-01

    The U.S. Department of Energy (DOE) regulates covered state government and alternative fuel provider fleets, pursuant to the Energy Policy Act of 1992 (EPAct), as amended. Covered fleets may meet their EPAct requirements through one of two compliance methods: Standard Compliance or Alternative Compliance. For model year (MY) 2015, the compliance rate with this program for the more than 3011 reporting fleets was 100%. More than 294 fleets used Standard Compliance and exceeded their aggregate MY 2015 acquisition requirements by 8% through acquisitions alone. The seven covered fleets that used Alternative Compliance exceeded their aggregate MY 2015 petroleum use reduction requirements by 46%.

  3. Thirteenth annual report of the Technical Advisory Committee on the Nuclear Fuel Waste Management Program

    Energy Technology Data Exchange (ETDEWEB)

    1993-05-15

    This report details activities since the last reporting period by the Technical Advisory Committee (TAC). The emphasis of the work in the Canadian Nuclear Fuel Waste Management Program (CNFWMP) has been on the writing of the Environmental Impact Statement (EIS) and the associated set of the primary reference document as well as supporting documents. These are in preparation for submission to the Environmental Assessment Review Panel who will lead the national evaluation of the disposal concept under the auspices of the Federal Environmental Assessment Review Office (FEARO).

  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. Load-following performance and assessment of CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Tayal, M.; Floyd, M.; Rattan, D.; Xu, Z.; Manzer, A.; Lau, J. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Kohn, E. [Ontario Power Generation, Fuel and Fuel Channel Analysis Dept., Toronto, Ontario (Canada)

    1999-09-01

    Load following of nuclear reactors is now becoming an economic necessity in some countries. When nuclear power stations are operated in a load-following mode, the reactor and the fuel may be subjected to step changes in power on a weekly, daily, or even hourly basis, depending on the grid's needs. This paper updates the previous surveys of load-following capability of CANDU fuel, focusing mainly on the successful experience at the Bruce B station. As well, initial analytical assessments are provided that illustrate the capability of CANDU fuel to survive conditions other than those for which direct in-reactor evidence is available. (author)

  6. 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)

  7. 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)

  8. 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.

  9. Computer simulation of the behaviour and performance of a CANDU fuel rod

    International Nuclear Information System (INIS)

    Marino, A.C.

    1997-01-01

    At the Argentine Atomic Energy Commission (Comision Nacional de Energia Atomica, CNEA) the BACO code (for 'BArra COmbustible', fuel rod) was developed. It allows the simulation of the thermo-mechanical performance of a cylindrical fuel rod in a Pressurized Heavy Water Reactor (PHWR). The standard present version of the code (2.30), is a powerful tool for a relatively easy and complete evaluation of fuel behaviour predictions. Input parameters and, therefore, output ones may include statistical dispersion. As a demonstration of BACO capabilities we include a review of CANDU fuel applications, and the calculation and a parametric analysis of a characteristic CANDU fuel. (author)

  10. Firm Characteristics and Performance Disclosure in Annual Reports of Nigerian Banks using the Balanced Scorecard

    Directory of Open Access Journals (Sweden)

    Solabomi Ajibolade

    2017-05-01

    Full Text Available This study investigated the influence of four firm characteristics (size, organisational structure, age and systemic importance on extent of performance disclosures by Nigerian banks using the balanced scorecard (BSC model. The population of the study comprised of publicly-listed banks in Nigeria, in operation from 2012 to 2014. Using a self-designed disclosure checklist, the annual reports of a sample of 15 publicly quoted banks in Nigeria were content-analysed for performance disclosure for the period 2012-2014. Descriptive statistics, t-test and Analysis of Variance (ANOVA were applied in data analysis, deducing inference at 5% significance level. It was observed that firms did not significantly differ in the extent of performance disclosure in each of the four BSC perspectives on one hand, and the overall BSC measure on the other hand, on the account of the four firm attributes examined. Considering that annual reports are mainstream amongst the media used to communicate firm performance to the public, it was recommended that preparers of such documents should consider disclosing financial and non-financial performance; this will not only provide a comprehensive basis to judge organisational performance, but will also assist in diffusing the clout created by asymmetry of information between preparers and users of performance reports.

  11. Annual Report: DOE Advanced Combustion Systems & Fuels R&D; Light-Duty Diesel Combustion

    Energy Technology Data Exchange (ETDEWEB)

    Busch, Stephen [Sandia National Lab. (SNL-CA), Livermore, CA (United States)

    2017-11-01

    Despite compliance issues in previous years, automakers have demonstrated that the newest generation of diesel power trains are capable of meeting all federal and state regulations (EPA, 2016). Diesels continue to be a cost-effective, efficient, powerful propulsion source for many light- and medium-duty vehicle applications (Martec, 2016). Even modest reductions in the fuel consumption of light- and medium duty diesel vehicles in the U.S. will eliminate millions of tons of CO2 emissions per year. Continued improvement of diesel combustion systems will play an important role in reducing fleet fuel consumption, but these improvements will require an unprecedented scientific understanding of how changes in engine design and calibration affect the mixture preparation, combustion, and pollutant formation processes that take place inside the cylinder. The focus of this year’s research is to provide insight into the physical mechanisms responsible for improved thermal efficiency observed with a stepped-lip piston. Understanding how piston design can influence efficiency will help engineers develop and optimize new diesel combustion systems.

  12. Annual Performance Assessment and Composite Analysis Review for the ICDF Landfill FY 2008

    International Nuclear Information System (INIS)

    Koslow, Karen; Rood, Arthur

    2009-01-01

    This report addresses low-level waste disposal operations at the Idaho Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) Disposal Facility (ICDF) landfill from the start of operations in Fiscal Year 2003 through Fiscal Year 2008. The ICDF was authorized in the Operable Unit 3-13 Record of Decision for disposal of waste from the Idaho National Laboratory Site CERCLA environmental restoration activities. The ICDF has been operating since 2003 in compliance with the CERCLA requirements and the waste acceptance criteria developed in the CERCLA process. In developing the Operable Unit 3-13 Record of Decision, U.S. Department of Energy Order (DOE) 435.1, 'Radioactive Waste Management', was identified as a 'to be considered' requirement for the ICDF. The annual review requirement under DOE Order 435.1 was determined to be an administrative requirement and, therefore, annual reviews were not prepared on an annual basis. However, the landfill has been operating for 5 years and, since the waste forms and inventories disposed of have changed from what was originally envisioned for the ICDF landfill, the ICDF project team has decided that this annual review is necessary to document the changes and provide a basis for any updates in analyses that may be necessary to continue to meet the substantive requirements of DOE Order 435.1. For facilities regulated under DOE Order 435.1-1, U.S. DOE Manual 435.1-1, 'Radioactive Waste Management', IV.P.(4)(c) stipulates that annual summaries of low-level waste disposal operations shall be prepared with respect to the conclusions and recommendations of the performance assessment and composite analysis. Important factors considered in this review include facility operations, waste receipts, and results from monitoring and research and development programs. There have been no significant changes in operations at the landfill in respect to the disposal geometry, the verification of waste characteristics, and the

  13. Annual Performance Assessment and Composite Analysis Review for the ICDF Landfill FY 2008

    Energy Technology Data Exchange (ETDEWEB)

    Karen Koslow

    2009-08-31

    This report addresses low-level waste disposal operations at the Idaho Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) Disposal Facility (ICDF) landfill from the start of operations in Fiscal Year 2003 through Fiscal Year 2008. The ICDF was authorized in the Operable Unit 3-13 Record of Decision for disposal of waste from the Idaho National Laboratory Site CERCLA environmental restoration activities. The ICDF has been operating since 2003 in compliance with the CERCLA requirements and the waste acceptance criteria developed in the CERCLA process. In developing the Operable Unit 3-13 Record of Decision, U.S. Department of Energy Order (DOE) 435.1, 'Radioactive Waste Management', was identified as a 'to be considered' requirement for the ICDF. The annual review requirement under DOE Order 435.1 was determined to be an administrative requirement and, therefore, annual reviews were not prepared on an annual basis. However, the landfill has been operating for 5 years and, since the waste forms and inventories disposed of have changed from what was originally envisioned for the ICDF landfill, the ICDF project team has decided that this annual review is necessary to document the changes and provide a basis for any updates in analyses that may be necessary to continue to meet the substantive requirements of DOE Order 435.1. For facilities regulated under DOE Order 435.1-1, U.S. DOE Manual 435.1-1, 'Radioactive Waste Management', IV.P.(4)(c) stipulates that annual summaries of low-level waste disposal operations shall be prepared with respect to the conclusions and recommendations of the performance assessment and composite analysis. Important factors considered in this review include facility operations, waste receipts, and results from monitoring and research and development programs. There have been no significant changes in operations at the landfill in respect to the disposal geometry, the verification of

  14. Gamification in thoracic surgical education: Using competition to fuel performance.

    Science.gov (United States)

    Mokadam, Nahush A; Lee, Richard; Vaporciyan, Ara A; Walker, Jennifer D; Cerfolio, Robert J; Hermsen, Joshua L; Baker, Craig J; Mark, Rebecca; Aloia, Lauren; Enter, Dan H; Carpenter, Andrea J; Moon, Marc R; Verrier, Edward D; Fann, James I

    2015-11-01

    In an effort to stimulate residents and trainers to increase their use of simulation training and the Thoracic Surgery Curriculum, a gamification strategy was developed in a friendly but competitive environment. "Top Gun." Low-fidelity simulators distributed annually were used for the technical competition. Baseline and final video assessments were performed, and 5 finalists were invited to compete in a live setting from 2013 to 2015. "Jeopardy." A screening examination was devised to test knowledge contained in the Thoracic Surgery Curriculum. The top 6 2-member teams were invited to compete in a live setting structured around the popular game show Jeopardy. "Top Gun." Over 3 years, there were 43 baseline and 34 final submissions. In all areas of assessment, there was demonstrable improvement. There was increasing evidence of simulation as seen by practice and ritualistic behavior. "Jeopardy." Sixty-eight individuals completed the screening examination, and 30 teams were formed. The largest representation came from the second-year residents in traditional programs. Contestants reported an average in-training examination percentile of 72.9. Finalists reported increased use of the Thoracic Surgery Curriculum by an average of 10 hours per week in preparation. The live competition was friendly, engaging, and spirited. This gamification approach focused on technical and cognitive skills, has been successfully implemented, and has encouraged the use of simulators and the Thoracic Surgery Curriculum. This framework may capitalize on the competitive nature of our trainees and can provide recognition of their achievements. Copyright © 2015 The American Association for Thoracic Surgery. Published by Elsevier Inc. All rights reserved.

  15. FCI: remedy development for the fuel performance improvement program

    International Nuclear Information System (INIS)

    Buckman, F.W.; Crouthamel, C.E.; Freshley, M.D.

    1979-01-01

    Out-of-reactor experiments and irradiations are being utilized to develop and demonstrate the efficacy of specific advanced fuel designs to improve FCI behavior. The advanced light water reactor fuel designs being evaluated combine annular pellets, graphite coating on the inner surface of the cladding, and helium pressurization. A sphere-pac fuel design is also being developed. Characterization of the graphite coatings includes studies of composition, application methods, thickness control, moisture control, thermal conductivity, compatibility with the zircaloy cladding, strain-to-failure, and friction and wear characteristics. Rods of the different fuel designs, as well as reference rods, are being irradiated in the Halden Boiling Water Reactor and the Big Rock Point Reactor to accumulate burnup prior to ramping tests

  16. The manufacture, quality control and performance of KANUPP fuel

    International Nuclear Information System (INIS)

    Butt, M.I.; Salim, M.; Ahmad, I.

    1989-01-01

    KANUPP is a 137 MWe CANDU reactor. The fuel material is high-density sintered pellets (95-97% T.D.) of natural UO 2 in Zircaloy 4 sheaths. Reactor-grade UO 2 powder is precompacted, granulated, blended with 0.2% zinc stearate, and compacted into green pellets. The pellets are sintered in a reducing atmosphere, then finished by grinding, culled, and loaded into Zr-4 tubes. The welded elements are assembled into a fuel bundle. Quality control and quality assurance procedures are followed during all stages of manufacturing. The entire core of KANUPP now consists of locally manufactured fuel. Several bundles have already achieved the design burnup (8650 MWD/TU). There have never been any failures of these fuel bundles. (6 refs., 5 figs., 8 tabs.)

  17. 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.

  18. BWR fuel performance under advanced water chemistry conditions – a delicate journey towards zero fuel failures – a review

    International Nuclear Information System (INIS)

    Hettiarachchi, S.

    2015-01-01

    Boiling Water Reactors (BWRs) have undergone a variety of chemistry evolutions over the past few decades as a result of the need to control stress corrosion cracking of reactor internals, radiation fields and personnel exposure. Some of the advanced chemistry changes include hydrogen addition, zinc addition, iron reduction using better filtration technologies, and more recently noble metal chemical addition to many of the modern day operating BWRs. These water chemistry evolutions have resulted in changes in the crud distribution on fuel cladding material, Co-60 levels and the Rod oxide thickness (ROXI) measurements using the conventional eddy current techniques. A limited number of Post-Irradiation Examinations (PIE) of fuel rods that exhibited elevated oxide thickness using eddy current techniques showed that the actual oxide thickness by metallography is much lower. The difference in these observations is attributed to the changing magnetic properties of the crud affecting the rod oxide thickness measurement by the eddy current technique. This paper will review and summarize the BWR fuel cladding performance under these advanced and improved water chemistry conditions and how these changes have affected the goal to reach zero fuel failures. The paper will also provide a brief summary of some of the results of hot cell PIE, results of crud composition evaluation, crud spallation, oxide thickness measurements, hydrogen content in the cladding and some fuel failure observations. (author) Key Words: Boiling Water Reactor, Fuel Performance, Hydrogen Addition, Zinc Addition, Noble Metal Chemical Addition, Zero Leakers

  19. Reforming petroleum-based fuels for fuel cell vehicles : composition-performance relationships

    International Nuclear Information System (INIS)

    Kopasz, J. P.; Miller, L. E.; Ahmed, S.; Devlin, P. R.; Pacheco, M.

    2001-01-01

    Onboard reforming of petroleum-based fuels, such as gasoline, may help ease the introduction of fuel cell vehicles to the marketplace. Although gasoline can be reformed, it is optimized to meet the demands of ICEs. This optimization includes blending to increase the octane number and addition of oxygenates and detergents to control emissions. The requirements for a fuel for onboard reforming to hydrogen are quite different than those for combustion. Factors such as octane number and flame speed are not important; however, factors such as hydrogen density, catalyst-fuel interactions, and possible catalyst poisoning become paramount. In order to identify what factors are important in a hydrocarbon fuel for reforming to hydrogen and what factors are detrimental, we have begun a program to test various components of gasoline and blends of components under autothermal reforming conditions. The results indicate that fuel composition can have a large effect on reforming behavior. Components which may be beneficial for ICEs for their octane enhancing value were detrimental to reforming. Fuels with high aromatic and naphthenic content were more difficult to reform. Aromatics were also found to have an impact on the kinetics for reforming of paraffins. The effects of sulfur impurities were dependent on the catalyst. Sulfur was detrimental for Ni, Co, and Ru catalysts. Sulfur was beneficial for reforming with Pt catalysts, however, the effect was dependent on the sulfur concentration

  20. The performances of the LPG-fueled vehicle

    International Nuclear Information System (INIS)

    Anon.

    1996-01-01

    Todays, LPG-fueled vehicles are simply equipped with transformed classical petrol engines. The LPG consumption is greater but the emission of pollutants is greatly reduced (from 40 to 98% for the CO and from 30 to 80% for the HC and NOx, depending on the temperature of use). This short paper summarizes the environmental advantages of the LPG-fueled vehicles and the forthcoming technological evolutions expected in Europe, Japan and the USA. (J.S.)

  1. 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)

  2. Improving the performance of solid oxide fuel cell systems

    OpenAIRE

    Halinen, Matias

    2015-01-01

    Solid oxide fuel cell (SOFC) systems can provide power production at a high electrical efficiency and with very low emissions. Furthermore, they retain their high electrical efficiency over a wide range of output power and offer good fuel flexibility, which makes them well suited for a range of applications. Currently SOFC systems are under investigation by researchers as well as being developed by industrial manufacturers. The first commercial SOFC systems have been on the market for some...

  3. Current Status of the LIFE Fast Reactors Fuel Performance Codes

    International Nuclear Information System (INIS)

    Yacout, A.M.; Billone, M.C.

    2013-01-01

    The LIFE-4 (Rev. 1) code was calibrated and validated using data from (U,Pu)O2 mixed-oxide fuel pins and UO2 blanket rods which were irradiation tested under steady-state and transient conditions. – It integrates a broad material and fuel-pin irradiation database into a consistent framework for use and extrapolation of the database to reactor design applications. – The code is available and running on different computer platforms (UNIX & PC) – Detailed documentations of the code’s models, routines, calibration and validation data sets are available. LIFE-METAL code is based on LIFE4 with modifications to include key phenomena applicable to metallic fuel, and metallic fuel properties – Calibrated with large database from irradiations in EBR-II – Further effort for calibration and detailed documentation. Recent activities with the codes are related to reactor design studies and support of licensing efforts for 4S and KAERI SFR designs. Future activities are related to re-assessment of the codes calibration and validation and inclusion of models for advanced fuels (transmutation fuels)

  4. Evaluation of safety, performance and emissions of synthetic fuel blends in a Cessna Citation II

    NARCIS (Netherlands)

    Snijders, T.A.; Melkert, J.A.

    2011-01-01

    Prior to being used in aviation, alternative fuels have to be tested thoroughly to ensure safe operation. At Delft University of Technology, a test programme was performed to evaluate the safety, performance and emissions of synthetic fuel blends. During test preparations, compatibility of the

  5. Specification of PWR UO2 pellet design parameters with the fuel performance code FRAPCON-1

    International Nuclear Information System (INIS)

    Silva, A.T.; Marra Neto, A.

    1988-08-01

    UO 2 pellet design parameters are analysed to verify their influence in the fuel basic properties and in its performance under irradiation in pressurized water reactors. Three groups of parameters are discussed: 1) content of fissionable and impurity materials; 2) stoichiometry; 3) density pore morpholoy, and microstructure. A methodology is applied with the fuel performance program FRAPCON-1 to specify these parameters. (author [pt

  6. Effect of time-varying humidity on the performance of a polymer electrolyte membrane fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Noorani, Shamsuddin [Department of Mechanical Engineering, University of Michigan-Dearborn (United States); Shamim, Tariq [Mechanical Engineering, Masdar Institute of Science and Technology (United Arab Emirates)], E-mail: tshamim@masdar.ac.ae

    2011-07-01

    In the energy sector, fuel cells constitute a promising solution for the future due to their energy-efficient and environment-friendly characteristics. However, the performance of fuel cells is very much affected by the humidification level of the reactants, particularly in hot regions. The aim of this paper is to develop a better understanding of the effect of driving conditions on the performance of fuel cells. A macroscopic single-fuel-cell-based, one dimensional, isothermal model was used on a polymer electrolyte membrane fuel cell to carry out a computational study of the impact of humidity conditions which vary over time. It was found that the variation of humidity has a significant effect on water distribution but a much lower impact on power and current densities. This paper provided useful information on fuel cells' performance under varying conditions which could be used to improve their design for mobile applications.

  7. 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

  8. Performance Analysis of Air Breathing Proton Exchange Membrane Fuel Cell Stack (PEMFCS) At Different Operating Condition

    Science.gov (United States)

    Sunil, V.; Venkata siva, G.; Yoganjaneyulu, G.; Ravikumar, V. V.

    2017-08-01

    The answer for an emission free power source in future is in the form of fuel cells which combine hydrogen and oxygen producing electricity and a harmless by product-water. A proton exchange membrane (PEM) fuel cell is ideal for automotive applications. A single cell cannot supply the essential power for any application. Hence PEM fuel cell stacks are used. The effect of different operating parameters namely: type of convection, type of draught, hydrogen flow rate, hydrogen inlet pressure, ambient temperature and humidity, hydrogen humidity, cell orientation on the performance of air breathing PEM fuel cell stack was analyzed using a computerized fuel cell test station. Then, the fuel cell stack was subjected to different load conditions. It was found that the stack performs very poorly at full capacity (runs only for 30 min. but runs for 3 hours at 50% capacity). Hence, a detailed study was undertaken to maximize the duration of the stack’s performance at peak load.

  9. 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.

  10. An experimental investigation of Perkins A63544 diesel engine performance using D-Series fuel

    International Nuclear Information System (INIS)

    Hassan-beygi, Seyed Reza; Istan, Vahideh; Ghobadian, Barat; Aboonajmi, Mohammad

    2013-01-01

    Highlights: • The performance of a diesel engine was evaluated using newly developed D-Series fuel. • The specifications of D-Series fuel were in the range of ASTM D-6751-09 standard. • The D-Series fuel did not change the engine power and torque significantly except the D 65 B 25 E 10 fuel blend. • The D-Series fuel blends increased the engine specific fuel consumption compare with neat-diesel fuel. • The D 93 B 5 E 2 fuel blend could be suggested as an appropriate alternative for neat petro-diesel fuel. - Abstract: This paper reports the results of an investigation using a newly developed fuel mixture called ‘D-Series fuel’ on a Perkins A63544 direct injection diesel engine. The biodiesel and bioethanol fuels were added to diesel fuel in a manner that specifications of the formed mixture did not change considerably. The performance of the engine under test was then evaluated without any modification or change in engine components and systems using the D-Series fuel. The obtained data was statistically analyzed using two factors completely randomized design to study the effects of the engine speeds and fuel blend types on the engine power, torque, and specific fuel consumption. The analysis of variance showed that the engine speeds and fuel types had statistically significant effects at 1% probability level (P 65 B 25 E 10 , 65% diesel, 25% biodiesel and 10% bioethanol, blend which decreased the engine power. The engine torque was decreased with increasing the engine speed for all the fuel blends in range of 319–296 N m. The maximum torque reduction was about 25 N m for neat petro-diesel fuel. The engine torque was decreased significantly (P 93 B 5 E 2 fuel blend could be suggested as an appropriate alternative for neat petro-diesel fuel, though the D 86 B 10 E 4 and D 79 B 15 E 6 blends could be also suggested for greater ratios of biodiesel and bioethanol application in D-Series fuel application

  11. Biological production of liquid fuels from biomass. Annual report, September 1, 1978-August 31, 1979

    Energy Technology Data Exchange (ETDEWEB)

    Pye, E.K.; Humphrey, A.E.

    1979-01-01

    The production of liquid fuels from renewable resources such as poplar wood and lignocellulosic wastes from a refuse hydropulper were studied. The particular scheme being studied involves the conversion of a cellulosic residue, resulting from a solvent delignified lignocellulosic feed, into either high concentration sugar syrups or into ethyl and/or butyl alcohol. The process is aimed at achieving total raw material utilization and maximization of high value by-product recovery. Specific goals of the investigation are the demonstration of the process technical feasibility and economic practicality and its optimization for maximum economic yield and efficiency. The construction of a pilot apparatus for solvent delignifying 150g samples of lignocellulosic feeds has been completed. Also, an analysis method for characterizing the delignified product has been selected and tested. Delignified samples are now being prepared and tested for their extent of delignification and susceptibility to enzyme hydrolysis.

  12. FRAPCON-3: Modifications to fuel rod material properties and performance models for high-burnup application

    International Nuclear Information System (INIS)

    Lanning, D.D.; Beyer, C.E.; Painter, C.L.

    1997-12-01

    This volume describes the fuel rod material and performance models that were updated for the FRAPCON-3 steady-state fuel rod performance code. The property and performance models were changed to account for behavior at extended burnup levels up to 65 Gwd/MTU. The property and performance models updated were the fission gas release, fuel thermal conductivity, fuel swelling, fuel relocation, radial power distribution, solid-solid contact gap conductance, cladding corrosion and hydriding, cladding mechanical properties, and cladding axial growth. Each updated property and model was compared to well characterized data up to high burnup levels. The installation of these properties and models in the FRAPCON-3 code along with input instructions are provided in Volume 2 of this report and Volume 3 provides a code assessment based on comparison to integral performance data. The updated FRAPCON-3 code is intended to replace the earlier codes FRAPCON-2 and GAPCON-THERMAL-2. 94 refs., 61 figs., 9 tabs

  13. Performance limits of coated particle fuel. Part III. Fission product migration in HTR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Nabielek, H.; Hick, H.; Wagner-Loffler, M.; Voice, E. H.

    1974-06-15

    A general introduction and literature survey to the physics and mathematics of fission product migration in HTR fuel is given as well as a review of available experimental results and their evaluation in terms of models and materials data.

  14. 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

  15. Numerical study of assembly pressure effect on the performance of proton exchange membrane fuel cell

    Energy Technology Data Exchange (ETDEWEB)

    Taymaz, Imdat; Benli, Merthan [Department of Mechanical Engineering, University of Sakarya, 54187 Adapazari (Turkey)

    2010-05-15

    The performance of the fuel cell is affected by many parameters. One of these parameters is assembly pressure that changes the mechanical properties and dimensions of the fuel cell components. Its first duty, however, is to prevent gas or liquid leakage from the cell and it is important for the contact behaviors of fuel cell components. Some leakage and contact problems can occur on the low assembly pressures whereas at high pressures, components of the fuel cell, such as bipolar plates (BPP), gas diffusion layers (GDL), catalyst layers, and membranes, can be damaged. A finite element analysis (FEA) model is developed to predict the deformation effect of assembly pressure on the single channel PEM fuel cell in this study. Deformed fuel cell single channel model is imported to three-dimensional, computational fluid dynamics (CFD) model which is developed for simulating proton exchange membrane (PEM) fuel cells. Using this model, the effect of assembly pressure on fuel cell performance can be calculated. It is found that, when the assembly pressure increases, contact resistance, porosity and thickness of the gas diffusion layer (GDL) decreases. Too much assembly pressure causes GDL to destroy; therefore, the optimal assembly pressure is significant to obtain the highest performance from fuel cell. By using the results of this study, optimum fuel cell design and operating condition parameters can be predicted accordingly. (author)

  16. Analysis of UO{sub 2}-BeO fuel under transient using fuel performance code

    Energy Technology Data Exchange (ETDEWEB)

    Gomes, Daniel S.; Abe, Alfredo Y.; Muniz, Rafael O.R.; Giovedi, Claudia, E-mail: dsgomes@ipen.br, E-mail: alfredo@ctmsp.mar.mil.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), São Paulo, SP (Brazil). Departamento de Engenharia Naval e Oceânica

    2017-11-01

    Recent research has appointed the need to replace the classic fuel concept, used in light water reactors. Uranium dioxide has a weak point due to the low thermal conductivity, that produce high temperatures on the fuel. The ceramic composite fuel formed of uranium dioxide (UO{sub 2}), with the addition of beryllium oxide (BeO), presents high thermal conductivity compared with UO{sub 2}. The oxidation of zirconium generates hydrogen gas that can create a detonation condition. One of the preferred options are the ferritic alloys formed of iron-chromium and aluminum (FeCrAl), that should avoid the hydrogen release due to oxidation. In general, the FeCrAl alloys containing 10 - 20Cr, 3 - 5Al, and 0 - 0.12Y in weight percent. The FeCrAl alloys should exhibit a slow oxidation kinetics due to chemical composition. Resistance to oxidation in the presence of steam is improved as a function of the content of chromium and aluminum. In this way, the thermal and mechanical properties of the UO{sub 2}-BeO-10%vol, composite fuel were coupled with FeCrAl alloys and added to the fuel codes. In this work, we examine the fuel rod behavior of UO{sub 2}-10%vol-BeO/FeCrAl, including a simulated transient of reactivity. The fuels behavior shown reduced temperature with UO{sub 2}-BeO/Zr, UO{sub 2}-BeO/FeCrAl also were compared with UO{sub 2}/Zr system. The case reactivity initiated accident analyzed, reproducing the fuel rod called VA-1 using UO{sub 2}/Zr alloys and compared with UO{sub 2}-BeO/FeCrAl. (author)

  17. Fuel Composition and Performance Analysis of Endothermically Heated Fuels for Pulse Detonation Engines

    Science.gov (United States)

    2009-03-01

    exchanger was constructed on an inner 2 in Inconel 625 schedule 10 pipe and an outer 2 ½ in Inconel 600 schedule 40 pipe 0.91 m (36 in) in length. The...switched to positions two and three for the remainder of the experiments. 46 The detonation tubes are fabricated from inconel and include heat...and four. Fuel Heating System 47 The fuel heating system centers around two pairs of inconel heat exchangers. The first pair was developed in

  18. 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)

  19. Seasonal Changes in Atmospheric Noise Levels and the Annual Variation in Pigeon Homing Performance

    Science.gov (United States)

    Hagstrum, J. T.; McIsaac, H. P.; Drob, D. P.

    2015-12-01

    The remarkable navigational ability of homing pigeons (Columba livia) is influenced by a number of factors, an unknown one of which causes the "Wintereffekt"1 or annual variation in homing performance. Minima in homeward orientation and return speeds have been observed in winter, with maxima in summer, during repetitive pigeon releases from single sites near experimental lofts in Wilhelmshaven, Göttingen, and Munich, Germany, and near Pisa, Italy1-4. Overall the annual variation is more pronounced in northern Germany than Italy4, and both mature and juvenile cohorts respond to this seasonal factor. Older, more experienced pigeons are better at compensating for its effects than naïve ones, but are still affected after numerous releases. The narrow low-frequency band of atmospheric background noise (microbaroms; 0.1-0.3 Hz) also varies with an annual cycle that generally has higher amplitudes in winter than in summer depending on location5. In addition, homing pigeons, and possibly other birds, apparently use infrasonic signals of similar frequency as navigational cues6, and a seasonal variation in background noise levels could cause corresponding changes in signal-to-noise ratios and thus in homing performance. The annual variation in homing performance, however, was not observed during long-term pigeon releases at two sites in eastern North America. The annual and geographic variability in homing performance in the northern hemisphere can be explained to a first order by seasonal changes in infrasonic noise sources related to ocean storm activity, and to the direction and intensity of stratospheric winds. In addition, increased dispersion in departure bearings of individual birds for some North American releases were likely caused by additional infrasonic noise associated with severe weather events during tornado and Atlantic hurricane seasons. 1Kramer, G. & von Saint Paul, U., J. Ornithol. 97, 353-370 (1956); 2Wallraff, H. G., Z. Tierpsychol. 17, 82-113 (1960

  20. Method of performing shutdown reactivity measurements in spent nuclear fuel storage pools

    International Nuclear Information System (INIS)

    Levine, S.H.; Schultz, M.A.; Chang, D.

    1981-01-01

    The objective of this paper is to develop a device to measure the k/infinity/ of a spent fuel assembly used in light water reactors. A subcritical assembly having a cross configuration is designed to allow measurement of the k/sub //infinity/ of a spent fuel assembly by comparing the change in its multiplication with that of a fuel assembly of known k/infinity/. Calculations have been performed using nucleonic codes to develop polynomial equations that relate the k/infinity/ of the spent fuel assembly to measured data. The measurements involve taking count rates with the spent fuel assembly in the center position of the subcritical assembly, and the measured data are the count rate ratio of the spent fuel assembly over the count rate taken with a fuel assembly of known k/infinity/. The polynomial equations are easy to program on a microcomputer, which, together with the subcritical assembly, form the k/infinity/ meter. 9 refs