WorldWideScience

Sample records for fuel performance meeting

  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 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

  3. 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

  4. 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

  5. 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)

  6. 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

  7. 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.

  8. 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

  9. Specialists' meeting on fuel element performance computer modelling, Preston, United Kingdom, 15-19 March 1982

    International Nuclear Information System (INIS)

    1983-03-01

    The 46 papers of the meeting concerned with computer models of Water Reactor fuel elements cover practically all aspects of behavior of fuel elements in normal operation and in accident condition. Each session of the meeting produced a critical evaluation of one of the 5 topics into which the subject area had been divided. The sessions' report summarize the papers and make recommendations for further work. Separate abstracts were prepared for all the papers presented at this meeting

  10. International Working Group on Water Reactor Fuel Performance and Technology. Summary report of the 14. plenary meeting. Working material

    International Nuclear Information System (INIS)

    1997-01-01

    The fourteenth Plenary Meeting of the International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) was held at IAEA Headquarters, Vienna, from 21 to 23 May 1997. Twenty-seven participants, from twenty two Member States and two international organizations, attended the meeting. These presentations generally gave: The general situation of the nuclear industry in the country; Fuel fabrication; Fuel performance, high burnup fuel (including MOX) operational experience; Status and trends in fuel research programmes directed to achievement sufficient safety margins at high burnups with regard to normal and transient operational conditions. Majority of countries reported on the stable situation of the nuclear fuel industry, i.e. without significant additions/cuts in nuclear power plant and fuel fabrication plant (NPP) capacities. Refs, figs, tabs

  11. International Working Group on Water Reactor Fuel Performance and Technology. Summary report of the 14. plenary meeting. Working material

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-01

    The fourteenth Plenary Meeting of the International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) was held at IAEA Headquarters, Vienna, from 21 to 23 May 1997. Twenty-seven participants, from twenty two Member States and two international organizations, attended the meeting. These presentations generally gave: The general situation of the nuclear industry in the country; Fuel fabrication; Fuel performance, high burnup fuel (including MOX) operational experience; Status and trends in fuel research programmes directed to achievement sufficient safety margins at high burnups with regard to normal and transient operational conditions. Majority of countries reported on the stable situation of the nuclear fuel industry, i.e. without significant additions/cuts in nuclear power plant and fuel fabrication plant (NPP) capacities. Refs, figs, tabs.

  12. 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

  13. Advances in fuel pellet technology for improved performance at high burnup. Proceedings of a Technical Committee meeting

    International Nuclear Information System (INIS)

    1998-08-01

    The IAEA has recently completed two co-ordinated Research Programmes (CRPs) on The Development of Computer Models for Fuel Element Behaviour in Water Reactors, and on Fuel Modelling at Extended Burnup. Through these CRPs it became evident that there was a need to obtain data on fuel behaviour at high burnup. Data related o thermal behaviour, fission gas release and pellet to clad mechanical interaction were obtained and presented at the Technical Committee Meeting on Advances in Fuel Pellet Technology for Improved Performance at High Burnup which was recommended by the International Working Group on Fuel Performance and Technology (IWGFPT). The 34 papers from 10 countries are published in this proceedings and presented by a separate abstract. The papers were grouped in 6 sessions. First two sessions covered the fabrication of both UO 2 fuel and additives and MOX fuel. Sessions 3 and 4 covered the thermal behaviour of both types of fuel. The remaining two sessions dealt with fission gas release and the mechanical aspects of pellet to clad interaction

  14. Proceedings of fuel safety research specialists' meeting

    International Nuclear Information System (INIS)

    Suzuki, Motoe

    2002-08-01

    Fuel Safety Research Specialists' Meeting, which was organized by Japan Atomic Energy Research Institute, was held on March 4-5, 2002 at JAERI in Tokai Establishment. Purposes of the Meeting are to exchange information and views on LWR fuel safety topics among the specialist participants from domestic and foreign organizations, and to discuss the recent and future fuel research activities in JAERI. In the Meeting, presentations were given and discussions were made on general report of fuel safety research activities, fuel behaviors in normal operation and accident conditions, FP release behaviors in severe accident conditions, and JAERI's ''Advanced LWR Fuel Performance and Safety Research Program''. A poster exhibition was also carried out. The Meeting significantly contributed to planning future program and cooperation in fuel research. This proceeding integrates all the pictures and papers presented in the Meeting. The 10 of the presented papers are indexed individually. (J.P.N.)

  15. Sixth meeting of the International Working Group on Water Reactor Fuel Performance and Technology, Vienna, 14-15 May 1984

    International Nuclear Information System (INIS)

    1984-10-01

    The summary report of the 6th meeting of the IWGFPT had been approved. The review of IWGFPT activities since the last meeting are given in Appendix 3, which is a list of meetings and proceedings publications for that period. Co-ordinated research programmes were discussed in the fields of fuel element cladding interaction with water coolant in power reactors and examination and documentation methodology for water reactor fuel. The main activities for the next three years were settled. The next Working Group meeting will be held at the VIC, November 1985

  16. 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

  17. Ninth plenary meeting of the International Working Group on Water Reactor Fuel Performance and Technology Vienna, 2-4 October 1989

    International Nuclear Information System (INIS)

    1990-03-01

    The 9th Plenary Meeting of the International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) was held in Vienna from 2 to 4 October 1989. The meeting was attended by 24 participants from 18 Member States and one international organization. This publication is a summary report of this meeting. It contains review of the IWGFPT activities of the period since the 8th meeting. The representatives of the Member States presented information about National Programs and multilateral projects status with emphasis on new results. Summary of the discussion on schedule of IWGFPT activities for the period of 1990-91 is also published in this report. Refs, figs and tabs

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-12-01

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

  19. 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

  20. Summary of fuel safety research meeting 2004

    International Nuclear Information System (INIS)

    Fuketa, Toyoshi; Hidaka, Akihide; Nakamura, Jinichi; Suzuki, Motoe; Nagase, Fumihisa; Sasajima, Hideo; Fujita, Misao; Otomo, Takashi; Kudo, Tamotsu; Amaya, Masaki; Sugiyama, Tomoyuki; Ikehata, Hisashi; Iwasaki, Ryo; Ozawa, Masaaki; Kida, Mitsuko

    2004-10-01

    Fuel Safety Research Meeting 2004, which was organized by the Japan Atomic Energy Research Institute, was held on March 1-2, 2004 at Toranomon Pastoral, Tokyo. The purposes of the meeting are to present and discuss the results of experiments and analyses on reactor fuel safety and to exchange views and experiences among the participants. The technical topics of the meeting covered the status of fuel safety research activities, fuel behavior under RIA and LOCA conditions, high burnup fuel behavior, and radionuclides release under severe accident conditions. This summary contains all the abstracts and OHP sheets presented in the meeting. (author)

  1. 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

  2. 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

  3. 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

  4. 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)

  5. Summary of fuel safety research meeting 2005

    International Nuclear Information System (INIS)

    Fuketa, Toyoshi; Nakamura, Takehiko; Nagase, Fumihisa; Nakamura, Jinichi; Suzuki, Motoe; Sasajima, Hideo; Sugiyama, Tomoyuki; Amaya, Masaki; Kudo, Tamotsu; Chuto, Toshinori; Tomiyasu, Kunihiko; Udagawa, Yutaka; Ikehata, Hisashi; Kida, Mitsuko; Ikatsu, Nobuhiko; Hosoyamada, Ryuji; Hamanishi, Eizou; Iwasaki, Ryo; Ozawa, Masaaki

    2006-03-01

    Fuel Safety Research Meeting 2005, which was organized by the Japan Atomic Energy Agency (Establishment of the new organization in Oct. 1, 2005 integrated of JAERI and JNC) was held on March 2-3, 2005 at Toshi Center Hotel, Tokyo. The purposes of the meeting are to present and discuss the results of experiments and analyses on reactor fuel safety and to exchange views and experiences among the participants. The technical topics of the meting covered the status of fuel safety research activities, fuel behavior under Reactivity Initiated Accident (RIA) and Loss of coolant accident (LOCA) conditions, high fuel behavior, and radionuclide release under severe accident conditions. This summary contains all the abstracts and sheets of viewgraph presented in the meeting. (author)

  6. FAST FLUX TEST FACILITY DRIVER FUEL MEETING

    Energy Technology Data Exchange (ETDEWEB)

    None,

    1966-06-01

    The Pacific Northwest Laboratory has convened this meeting to enlist the best talents of our laboratories and industry in soliciting factual, technical information pertinent to the Pacific Northwest's Laboratory's evaluation of the potential fuel systems for the Fast Flux Test Facility. The particular factors emphasized for these fuel systems are those associated with safety, ability to meet testing objectives, and economics. The proceedings includes twenty-three presentations, along with a transcript of the discussion following each, as well as a summary discussion.

  7. 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

  8. Advancing PWR fuel to meet customer needs

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, F W

    1987-03-01

    Since the introduction of the Optimized Fuel Assembly (OFA) for PWRs in the late 1970s, Westinghouse has continued to work with the utility customers to identify the greatest needs for further advance in fuel performance and reliability. The major customer requirements include longer fuel cycle at lower costs, increased fuel discharge burn-up, enhanced operating flexibility, all accompanied by even greater reliability. In response to these needs, Westinghouse developed Vantage 5 PWR fuel. To optimize reactor operations, Vantage 5 fuel features distinct advantages: integral fuel burnable absorbers, axial and radial blankets, intermediate flow mixers, a removable top nozzle, and assembly modifications to accommodate increased discharge burn-up.

  9. 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)

  10. 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.)

  11. 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)

  12. Twenty-fifth water reactor safety information meeting: Proceedings. Volume 2: Human reliability analysis and human performance evaluation; Technical issues related to rulemakings; Risk-informed, performance-based initiatives; High burn-up fuel research

    International Nuclear Information System (INIS)

    Monteleone, S.

    1998-03-01

    This three-volume report contains papers presented at the conference. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Japan, Norway, and Russia. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. This volume contains the following: (1) human reliability analysis and human performance evaluation; (2) technical issues related to rulemakings; (3) risk-informed, performance-based initiatives; and (4) high burn-up fuel research

  13. Twenty-fifth water reactor safety information meeting: Proceedings. Volume 2: Human reliability analysis and human performance evaluation; Technical issues related to rulemakings; Risk-informed, performance-based initiatives; High burn-up fuel research

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1998-03-01

    This three-volume report contains papers presented at the conference. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Japan, Norway, and Russia. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. This volume contains the following: (1) human reliability analysis and human performance evaluation; (2) technical issues related to rulemakings; (3) risk-informed, performance-based initiatives; and (4) high burn-up fuel research. Selected papers have been indexed separately for inclusion in the Energy Science and Technology Database.

  14. 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)

  15. 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)

  16. Faraday Discussions meeting Catalysis for Fuels.

    Science.gov (United States)

    Fischer, Nico; Kondrat, Simon A; Shozi, Mzamo

    2017-05-02

    Welcome to Africa was the motto when after more than 100 years the flag ship conference series of the Royal Society of Chemistry, the Faraday Discussions was hosted for the first time on the African Continent. Under the fitting topic 'Catalysis for Fuels' over 120 delegates followed the invitation by the conference chair Prof. Graham Hutchings FRS (Cardiff Catalysis Institute), his organizing committee and the co-organizing DST-NRF Centre of Excellence in Catalysis c*change (). In the presentations of 21 invited speakers and 59 posters, cutting edge research in the field of catalysis for fuels, designing new catalysts for synthetic fuels, hydrocarbon conversion in the production of synthetic fuels and novel photocatalysis was presented over the two-day meeting. The scene was set by the opening lecture of Prof. Enrique Iglesias (UC Berkeley) and wrapped-up with the concluding remarks by Philip Gibson (SASOL).

  17. 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

  18. 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

  19. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1985-07-01

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

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

    International Nuclear Information System (INIS)

    1985-01-01

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

  2. 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.

  3. 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

  4. 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

  5. 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

  6. 77 FR 19278 - Informational Meeting on Nuclear Fuel Cycle Options

    Science.gov (United States)

    2012-03-30

    ... DEPARTMENT OF ENERGY Informational Meeting on Nuclear Fuel Cycle Options AGENCY: Office of Fuel... activities leading to a comprehensive evaluation and screening of nuclear fuel cycle options in 2013. At this... fuel cycle options developed for the evaluation and screening provides a comprehensive representation...

  7. Proceedings of the fuel cells `95 review meeting

    Energy Technology Data Exchange (ETDEWEB)

    George, T.J.

    1995-08-01

    This document contains papers presented at the Fuel Cells `95` Review Meeting. Topics included solid oxide fuel cells; DOE`s transportation program; ARPA advanced fuel cell development; molten carbonate fuel cells; and papers presented at a poster session. Individual papers have been processed separately for the U.S. DOE databases.

  8. 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

  9. Advanced Fuel Pellet Materials and Fuel Rod Design for Water Cooled Reactors. Proceedings of a Technical Committee Meeting

    International Nuclear Information System (INIS)

    2010-10-01

    The economics of current nuclear power plants have improved through increased fuel burnup and longer fuel cycles, i.e. increasing the effective time that fuel remains in the reactor core and the amount of energy it generates. Efficient consumption of fissile material in the fuel element before it is discharged from the reactor means that less fuel is required over the reactor's life cycle, which results in lower amounts of fresh fuel, lower spent fuel storage costs, and less waste for ultimate disposal. Better utilization of fissile nuclear materials, as well as more flexible power manoeuvring, place challenging operational demands on materials used in reactor components, and first of all, on fuel and cladding materials. It entails increased attention to measures ensuring desired in-pile fuel performance parameters that require adequate improvements in fuel material properties and fuel rod designs. These are the main reasons that motivated the IAEA Technical Working Group on Fuel Performance and Technology (TWG-FPT) to recommend the organization of a Technical Committee Meeting on Advanced Fuel Pellet Materials and Fuel Rod Designs for Power Reactors. The proposal was supported by the IAEA TWGs on Advanced Technologies for Light and Heavy Water-Cooled Reactors (TWG-LWR and TWG-HWR), and the meeting was held at the invitation of the Government of Switzerland at the Paul Scherrer Institute in Villigen, from 23 to 26 November 2009. This was the third IAEA meeting on these subjects (the first was held in 1996 in Tokyo, Japan, and the second in 2003 in Brussels, Belgium), which reflects the continuous interest in the above issues among Member States. The purpose of the meeting was to review the current status in the development of fuel pellet materials and to explore recent improvements in fuel rod designs for light and heavy water cooled power reactors. The meeting was attended by 45 specialists representing fuel vendors, nuclear utilities, research and development

  10. 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

  11. 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

  12. 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

  13. 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

  14. 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

  15. 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

  16. 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.

  17. Modelling of Water Cooled Fuel Including Design Basis and Severe Accidents. Proceedings of a Technical Meeting

    International Nuclear Information System (INIS)

    2015-11-01

    The demands on nuclear fuel have recently been increasing, and include transient regimes, higher discharge burnup and longer fuel cycles. This has resulted in an increase of loads on fuel and core internals. In order to satisfy these demands while ensuring compliance with safety criteria, new national and international programmes have been launched and advanced modelling codes are being developed. The Fukushima Daiichi accident has particularly demonstrated the need for adequate analysis of all aspects of fuel performance to prevent a failure and also to predict fuel behaviour were an accident to occur.This publication presents the Proceedings of the Technical Meeting on Modelling of Water Cooled Fuel Including Design Basis and Severe Accidents, which was hosted by the Nuclear Power Institute of China (NPIC) in Chengdu, China, following the recommendation made in 2013 at the IAEA Technical Working Group on Fuel Performance and Technology. This recommendation was in agreement with IAEA mid-term initiatives, linked to the post-Fukushima IAEA Nuclear Safety Action Plan, as well as the forthcoming Coordinated Research Project (CRP) on Fuel Modelling in Accident Conditions. At the technical meeting in Chengdu, major areas and physical phenomena, as well as types of code and experiment to be studied and used in the CRP, were discussed. The technical meeting provided a forum for international experts to review the state of the art of code development for modelling fuel performance of nuclear fuel for water cooled reactors with regard to steady state and transient conditions, and for design basis and early phases of severe accidents, including experimental support for code validation. A round table discussion focused on the needs and perspectives on fuel modelling in accident conditions. This meeting was the ninth in a series of IAEA meetings, which reflects Member States’ continuing interest in nuclear fuel issues. The previous meetings were held in 1980 (jointly with

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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.)

  3. Recycling of plutonium and uranium in water reactor fuel. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    1997-05-01

    The Technical Committee Meeting on Recycling of Plutonium and Uranium in Water Reactor Fuel was recommended by the International Working Group on Fuel Performance and Technology (IWGFPT). Its aim was to obtain an overall picture of MOX fabrication capacity and technology, actual performance of this kind of fuel, and ways explored to dispose of the weapons grade plutonium. The subject of this meeting had been reviewed by the International Atomic Energy Agency every 5 to 6 years and for the first time the problem of weapons grade plutonium disposal was included. The papers presented provide a summary of experience on MOX fuel and ongoing research in this field in the participating countries. The meeting was hosted by British Nuclear Fuels plc, at Newby Bridge, United Kingdom, from 3 to 7 July 1995. Fifty-six participants from twelve countries or international organizations took part. Refs, figs, tabs

  4. 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

  5. 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)

  6. 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)

  7. Proceedings of the fuels technology contractors review meeting

    Energy Technology Data Exchange (ETDEWEB)

    Malone, R.D. [ed.

    1993-11-01

    The Fuels Technology Contractors Review Meeting was held November 16-18, 1993, at the Morgantown Energy Technology Center (METC) in Morgantown, West Virginia. This meeting was sponsored and hosted by METC, the Office of Fossil Energy, U.S. Department of Energy (DOE). METC periodically provides an opportunity to bring together all of the R&D participants in a DOE-sponsored contractors review meeting to present key results of their research and to provide technology transfer to the active research community and to the interested public. This meeting was previously called the Natural Gas Technology Contractors Review Meeting. This year it was expanded to include DOE-sponsored research on oil shale and tar sands and so was retitled the Fuels Technology Contractors Review Meeting. Current research activities include efforts in both natural gas and liquid fuels. The natural gas portion of the meeting included discussions of results summarizing work being conducted in fracture systems, both natural and induced; drilling, completion, and stimulation research; resource characterization; delivery and storage; gas to liquids research; and environmental issues. The meeting also included project and technology summaries on research in oil shale, tar sands, and mild coal gasification, and summaries of work in natural-gas fuel cells and natural-gas turbines. The format included oral and poster session presentations. Individual papers have been processed separately for inclusion in the Energy Science and Technology database.

  8. 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

  9. Enlarged Halden programme group meeting on high burn-up fuel performance, safety and reliability and degradation of in-core materials and water chemistry effects and man-machine systems research. Volume II

    International Nuclear Information System (INIS)

    1999-01-01

    An Enlarged Halden Programme Group Meeting on Fuel Performance and Materials Testing and Man-Machine Research was held at Loen, Norway 24 to 29 May, 1999. The papers presented covered work performed at the OECD Halden Reactor Project during the years up to April 1999 as well as studies conducted within organisations participating in the Halden Project. The Halden Boiling Water Reactor, HBWR, was built in the years 1955 - 1958 by the Norwegian Institutt for Atomenergi, and has since 1958 been operated internationally as one of the joint undertakings of the OECD Nuclear Energy Agency. The experimental operation of the Halden Boiling Water Reactor and associated programmes are sponsored through an international agreement by Institute for energy technology, Norway, the Belgian Nuclear Research Centre SCK/CEN, acting also on behalf of other public or private organisations in Belgium, the Risoe National Laboratory in Denmark, the Finnish Ministry of Trade and Industry, the Electricite de France, the Gesellschaft fur Anlagen- und Reaktorsicherheit (GRS) mbH, representing a German group of companies working in agreement with the German Federal Ministry for Education, Science, Research and Technology, the Italian Ente per le Nouve Tecnologie, l'Energia e l'Ambiente, the Japan Atomic Energy Research Institute, the Korean Atomic Energy Research Institute, the Spanish Centre de Investigaciones Energeticas, Medioambientales y Tecnologias, representing a group of national and industry organisations in Spain, the Swedish Nuclear Power Inspectorate, the Swiss Federal Nuclear Safety Inspectorate, representing also the Swiss nuclear utilities and the Paul Scherrer Institute, the Nuclear Electric pic, representing a group of nuclear research and industry organisations in the United Kingdom, and the United States Nuclear Regulatory Commission, and as associated parties: the Brazilian Comissao Nacional de Energia Nuclear (CNEN), the Czech Nuclear Research Institute, the Hungarian

  10. Low-enriched fuel particle performance review

    International Nuclear Information System (INIS)

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

    1978-08-01

    The available data on low-enriched (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

  11. Poolside inspection, repair and reconstitution of LWR fuel elements. Proceedings of a Technical Committee meeting

    International Nuclear Information System (INIS)

    1998-11-01

    The Technical Committee Meeting on Poolside Inspection, repair and reconstruction of LWR Fuel Elements was organize by IAEA upon the recommendations of the International Working Group on Fuel performance Technology and held in Switzerland in October 1997. The purpose of the Meeting was to review the state of art in the area of poolside inspection, repair and reconstruction of light water fuel elements and to evaluate the progress achieved in this area since previous IAEA Meetings on the same topic in 1981 and 1984. The Meeting provided a forum on exchange of information between utilities, fuel designers and other authorities and specialists on a topic of current interest and real concern to industries in many Member States. The respective technologies are widely used or planned to be used in order to identify elementary major causes of fuel failure and to improve fuel utilization by repair and subsequent reuse of fuel elements. The Proceedings includes papers presented at the Meeting each described by a separate abstract

  12. 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

  13. 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

  14. 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.

  15. 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.

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

    International Nuclear Information System (INIS)

    1979-01-01

    recent meetings in related fields and included a cost-benefit assessment, indicated that the trends identified at this meeting were part of a general observation, namely that fuel fabrication technology is in a position to meet today's performance requirements, but that it is worth while pursuing efforts to broaden fuel utilization limits and economy. As the Chairman of the IWGFPT indicated in his concluding remarks, the fact that no novel discovery had to be announced at the meeting is the best indication that nuclear fuel has really reached maturity and has become an industrial, rather than a scientific, product. Technological improvements however are still possible and are being implemented. The symposium proved a good opportunity to evaluate the present status of fuel fabrication in relation to its performance. (author)

  17. 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.

  18. 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

  19. 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

  20. 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

  1. Dry process fuel performance technology development

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-06-15

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

  2. 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)

  3. 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

  4. 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

  5. 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.

  6. Proceedings of the Fuel Cells `97 Review Meeting

    Energy Technology Data Exchange (ETDEWEB)

    None

    1998-01-01

    The Federal Energy Technology Center (FETC) sponsored the Fuel Cells '97 Review Meeting on August 26-28, 1997, in Morgantown, West Virginia. The purpose of the meeting was to provide an annual forum for the exchange of ideas and discussion of results and plans related to the research on fuel cell power systems. The total of almost 250 conference participants included engineers and scientists representing utilities, academia, and government from the U.S. and eleven other countries: Canada, China, India, Iran, Italy, Japan, Korea, Netherlands, Russia, Taiwan, and the United Kingdom. On first day, the conference covered the perspectives of sponsors and end users, and the progress reports of fuel-cell developers. Papers covered phosphoric, carbonate, and solid oxide fuel cells for stationary power applications. On the second day, the conference covered advanced research in solid oxide and other fuel cell developments. On the third day, the conference sponsored a workshop on advanced research and technology development. A panel presentation was given on fuel cell opportunities. Breakout sessions with group discussions followed this with fuel cell developers, gas turbine vendors, and consultants.

  7. Technical Meeting on Design, Manufacturing and Irradiation Behaviour of Fast Reactors Fuels. Presentations

    International Nuclear Information System (INIS)

    2011-01-01

    The purpose of this meeting was to enable a rationalization and advancement of the design and manufacturing processes, a better selection of promising fuels, and a reduction of the time and costs currently required for R and D and testing, as well as to contribute to the improvement of the safety features of fuels under all operational states and accidental conditions. An overview of the status and perspective of the design, manufacturing and irradiation behaviour of fast reactors fuels were provided during this meeting. The main objectives are the following: Ensure sharing and dissemination of knowledge and expertise; Discuss specific features and issues of existing fuels; Improve knowledge and data for the design and engineering of fast reactor fuel and core structural materials; Discuss perspectives on advanced fuels; Consider modern technological, design and testing tools enabling reliable performance of fuels in current and planned operational environments; Establish international consensus in the developmental efforts on advanced fast reactor technologies, including collaborative programs and experiments. Contribute to the preparation and outline of the planned IAEA Coordinated Research Project on 'Examination of advanced fast reactor fuel and core structural materials. Each of the 24 presentations made at the meeting have been indexed separately

  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. 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)

  10. Design, Manufacturing and Irradiation Behaviour of Fast Reactor Fuel. Proceedings of a Technical Meeting

    International Nuclear Information System (INIS)

    2013-04-01

    Fast reactors are vital for ensuring the sustainability of nuclear energy in the long term. They offer vastly more efficient use of uranium resources and the ability to burn actinides, which are otherwise the long-lived component of high level nuclear waste. These reactors require development, qualification, testing and deployment of improved and innovative nuclear fuel and structural materials having very high radiation resistance, corrosion/erosion and other key operational properties. Several IAEA Member States have made efforts to advance the design and manufacture of technologies of fast reactor fuels, as well as to investigate their irradiation behaviour. Due to the acute shortage of fast neutron testing and post-irradiation examination facilities and the insufficient understanding of high dose radiation effects, there is a need for international exchange of knowledge and experience, generation of currently missing basic data, identification of relevant mechanisms of materials degradation and development of appropriate models. Considering the important role of nuclear fuels in fast reactor operation, the IAEA Technical Working Group on Fuel Performance and Technology (TWGFPT) proposed a Technical Meeting (TM) on 'Design, Manufacturing and Irradiation Behaviour of Fast Reactors Fuels', which was hosted by the Institute of Physics and Power Engineering (IPPE) in Obninsk, Russian Federation, from 30 May to 3 June 2011. The TM included a technical visit to the fuel production plant MSZ in Elektrostal. The purpose of the meeting was to provide a forum to share knowledge, practical experience and information on the improvement and innovation of fuels for fast reactors through scientific presentations and brainstorming discussions. The meeting brought together 34 specialists from national nuclear agencies, R and D and design institutes, fuel vendors and utilities from 10 countries. The presentations were structured into four sections: R and D Programmes on FR Fuel

  11. 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.

  12. 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

  13. Advanced fuel pellet materials and designs for water cooled reactors. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    2004-10-01

    This meeting was the second IAEA meeting on this subject. The first was held in 1996 in Tokyo, Japan. They are all part of a cooperative effort through the Technical Working Group on Water Reactor Fuel Performance and Technology (TWGFPT) of IAEA, with a series of three further meetings organized by CEA, France and co-sponsored by the IAEA and OECD/NEA. In the seven years since the first meeting took place, the demands on fuel duties have increased, with higher burnup, longer fuel cycles and higher temperatures. This places additional demands on fuel performance to comply with safety requirements. Criteria relative to fuel components, i.e. pellets and fuel rod column, require limiting of fission gas release and pellet-cladding interaction (PCI). This means that fuel components should maintain the composite of rather contradictory properties from the beginning until the end of its in-pile operation. Fabrication and design tools are available to influence, and to some extent, to ensure desirable in-pile fuel properties. Discussion of these tools was one of the objectives of the meeting. The second objective was the analysis of fuel characteristics at high burnup and the third and last objective was the discussion of specific feature of MOX and urania gadolinia fuels. Sixty specialists in the field of fuel fabrication technology attended the meeting from 18 countries. Twenty-five papers were presented in five sessions covering all relevant topics from the practices and modelling of fuel fabrication technology to its optimization. Eight papers were presented in session 'Optimization of fuel fabrication technology' which all were devoted to fuel fabrication technology. They mostly treated methods for optimizing fuel manufacturing processes, but gave also a good overview on nuclear fabrication needs and capabilities in different countries. During Session 'UO 2 , MOX and UO 2 -Gd 2 O 3 pellets with additives', six papers were presented in this session, which dealt mainly

  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. ENS RRFM 2005: 9th international topical meeting on research reactor fuel management. Transactions

    International Nuclear Information System (INIS)

    2005-01-01

    The ENS topical meeting on research reactor fuel management is an annual conference launched successfully in 1997. It has since then grown into well established international forum for the exchange and expertise on all significant aspects of the nuclear fuel cycle of research reactors. Oral presentations at this meeting were divided in the following four sessions: International Topics; Fuel Development, Qualification, Fabrication and Licensing; Reactor Operation, Fuel Safety and Core Conversion; Spent Fuel Management, Back-end Options, Transportation. The three poster sessions were devoted to fuel development, qualification, fabrication and licensing; reactor operation, fuel safety, core conversion, spent fuel; spent fuel management, fuel cycle back-end options, transportation

  17. Technical and economic limits to fuel burnup extension. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    2002-07-01

    For many years, the increase of efficiency in the production of nuclear electricity has been an economic challenge in many countries which have developed this kind of energy. The increase of fuel burnup leads to a reduction in the volume of spent fuel discharged to longer fuel cycles in the reactor, which means bigger availability and capacity factors. After having increased the authorized burnup in plants, developing new alloys capable of resisting high burnup, and having accumulated data on fuel evolution with burnup, it has become necessary to establish the limitations which could be imposed by the physical evolution of the fuel, influencing fuel management, neutron properties, reprocessing or, more generally, the management of waste and irradiated fuels. It is also necessary to verify whether the benefits of lower electricity costs would not be offset by an increase in fuel management costs. The main questions are: Are technical and economic limits to the increasing of fuel burnup in parallel? Can we envisage nowadays the hardest limitation in some of these areas? Which are the main points to be solved from the technical point of view? Is this effort worthwhile considering the economy of the cycle? To which extent? For these reasons, the IAEA, following a recommendation by the International Working Group on Fuel Performance and Technology, held a Technical Committee Meeting on Technical and Economic Limits to Fuel Burnup Extension. The purpose of this meeting was to provide an international forum to review the evolution of fuel properties at increased burnup in order to estimate the limitations both from a physical and an economic point of view. The meeting was therefore divided into two parts. The first part, focusing on technical limits, was devoted to the improvement of the fuel element, such as fission gas release (FGR), RIM effect, cladding, etc. and the fabrication, core management, spent fuel and reprocessing. Eighteen related papers were presented which

  18. IAEA specialists' meeting on power ramping and cycling behaviour of water reactor fuel. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1983-06-01

    At its fourth Annual Meeting, the IAEA International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) recommended that the Agency should hold a second Specialists' Meeting on 'Power Ramping and Cycling Behaviour of Water Reactor Fuel'. As research activities related to power ramping and cycling of water reactor fuel have been pursued vigorously, it was the objective of this meeting to review and discuss today's State of the Art and current understanding of water reactor fuel behaviour related to this these. Emphasis should be on practical experience and experimental investigations. The meeting was organised in five sessions: Power ramping and power cycling programs in power and and research reactors; Experimental methods; Power ramping and cycling results; Investigations and results of separate effects, especially related to PCI, defect mechanism, mechanical response, fuel design, and specially related to fission gas release; Operational strategies, recommendations and economic implications. The session chairmen, together with the speakers, prepared and presented reports with summary, conclusions and recommendations of the individual sessions. These reports are added to this summary report.

  19. IAEA specialists' meeting on power ramping and cycling behaviour of water reactor fuel. Summary report

    International Nuclear Information System (INIS)

    1983-06-01

    At its fourth Annual Meeting, the IAEA International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) recommended that the Agency should hold a second Specialists' Meeting on 'Power Ramping and Cycling Behaviour of Water Reactor Fuel'. As research activities related to power ramping and cycling of water reactor fuel have been pursued vigorously, it was the objective of this meeting to review and discuss today's State of the Art and current understanding of water reactor fuel behaviour related to this these. Emphasis should be on practical experience and experimental investigations. The meeting was organised in five sessions: Power ramping and power cycling programs in power and and research reactors; Experimental methods; Power ramping and cycling results; Investigations and results of separate effects, especially related to PCI, defect mechanism, mechanical response, fuel design, and specially related to fission gas release; Operational strategies, recommendations and economic implications. The session chairmen, together with the speakers, prepared and presented reports with summary, conclusions and recommendations of the individual sessions. These reports are added to this summary report

  20. 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

  1. Proceedings of the specialist meeting on the safety of water reactors fuel elements

    International Nuclear Information System (INIS)

    1973-01-01

    This specialist meeting on the safety of water reactors fuel elements was held in Saclay (France) in October 1973, and was organized by CSNI and CEA. It attracted specialists from 14 countries. Session I was devoted to normal operating conditions (coolant-cladding and fuel-cladding interactions, fission product release, effects of cladding deformation on fuel element performances and reactor operating limits); Session II was devoted to operating reactor accidents and failures, anomalous transients and handling accidents; Session III was devoted to modifications to be applied to fuel elements in order to enhance their safety and reliability; Session IV was devoted to Loss-of-Coolant Accidents (LOCA)(cladding behaviour during the accident, assembly behaviour during the accident, criteria to be considered for the study of fuel element behaviour during a LOCA)

  2. 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

  3. 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.

  4. 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)

  5. Structural behaviour of fuel assemblies for water cooled reactors. Proceedings of a technical meeting

    International Nuclear Information System (INIS)

    2005-07-01

    At the invitation of the Government of France and in response to a proposal of the IAEA Technical Working Group on Water Reactor Fuel Performance and Technology (TWGFPT), the IAEA convened a Technical Meeting on Fuel Assembly Structural Behaviour in Cadarache, France, from 22 to 26 November 2004. The meeting was hosted by the CEA Cadarache Centre, AREVA Framatome-ANP and Electricite de France. The meeting aimed to provide in depth technical exchanges on PWR and WWER operational experience in the field of fuel assembly mechanical behaviour and the potential impact of future high burnup fuel management on fuel reliability. It addressed in-service experience and remedial solutions, loop testing experience, qualification and damage assessment methods (analytic or experimental ones), mechanical behaviour of the fuel assembly including dynamic and fluid structure interaction aspects, modelling and numerical analysis methods, and impact of the in-service evolution of the structural materials. Sixty-seven participants from 17 countries presented 30 papers in the course of four sessions. The topics covered included the impact of hydraulic loadings on fuel assembly (FA)performance, FA bow and control rod (CR) drop kinetics, vibrations and rod-to-grid wear and fretting, and, finally, evaluation and modelling of accident conditions, mainly from seismic causes. FA bow, CR drop kinetics and hydraulics are of great importance under conditions of higher fuel duties including burnup increase, thermal uprates and longer fuel cycles. Vibrations and rod-to-grid wear and fretting have been identified as a key cause of fuel failure at PWRs during the past several years. The meeting demonstrated that full-scale hydraulic tests and modelling provide sufficient information to develop remedies to increase FA skeleton resistance to hydraulic loads, including seismic ones, vibrations and wear. These proceedings are presented as a book with an attached CD-ROM. The first part of the CD

  6. 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.

  7. 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.

  8. 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

  9. 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

  10. Water Chemistry and Clad Corrosion/Deposition Including Fuel Failures. Proceedings of a Technical Meeting

    International Nuclear Information System (INIS)

    2013-03-01

    Corrosion is a principal life limiting degradation mechanism in nuclear steam supply systems, particularly taking into account the trends in increasing fuel burnup, thermal ratings and cycle length. Further, many plants have been operating with varying water chemistry regimes for many years, and issues of crud (deposition of corrosion products on other surfaces in the primary coolant circuit) are of significant concern for operators. At the meeting of the Technical Working Group on Fuel Performance and Technology (TWGFPT) in 2007, it was recommended that a technical meeting be held on the subject of water chemistry and clad corrosion and deposition, including the potential consequences for fuel failures. This proposal was supported by both the Technical Working Group on Advanced Technologies for Light Water Reactors (TWG-LWR) and the Technical Working Group on Advanced Technologies for Heavy Water Reactors (TWG-HWR), with a recommendation to hold the meeting at the National Nuclear Energy Generating Company ENERGOATOM, Ukraine. This technical meeting was part of the IAEA activities on water chemistry, which have included a series of coordinated research projects, the most recent of which, Optimisation of Water Chemistry to Ensure Reliable Water Reactor Fuel Performance at High Burnup and in Ageing Plant (FUWAC) (IAEATECDOC-1666), concluded in 2010. Previous technical meetings were held in Cadarache, France (1985), Portland, Oregon, USA (1989), Rez, Czech Republic (1993), and Hluboka nad Vltavou, Czech Republic (1998). This meeting focused on issues associated with the corrosion of fuel cladding and the deposition of corrosion products from the primary circuit onto the fuel assembly, which can cause overheating and cladding failure or lead to unplanned power shifts due to boron deposition in the clad deposits. Crud deposition on other surfaces increases radiation fields and operator dose and the meeting considered ways to minimize the generation of crud to avoid

  11. CANDU fuel performance and development

    International Nuclear Information System (INIS)

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

    1978-12-01

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

  12. Status and prospects for gas cooled reactor fuels. Proceedings of two IAEA meetings held in June 2004 and June 2005

    International Nuclear Information System (INIS)

    2009-04-01

    following topics: overview of recent developments in nine countries; power and limitations of coated particle fuel modelling; Fuel performance technology; and novel ideas/applications/disposal questions. The meeting critically reviewed advanced fuel designs, including conventional ones, fabrication technology, quality assurance/quality control of fuel, fuel irradiation qualification, fuel performance, fuel modelling for transport and performance and overall fuel cycle issues

  13. Space reactor fuels performance and development issues

    International Nuclear Information System (INIS)

    Wewerka, E.M.

    1984-01-01

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

  14. 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.

  15. In-Pile Testing and Instrumentation for Development of Generation-IV Fuels and Materials. Proceedings of a Technical Meeting

    International Nuclear Information System (INIS)

    2013-12-01

    For many years, the increase in efficiency in the production of nuclear electricity has been an economic challenge in many countries which have developed this kind of energy. The increase in fuel burnup and fuel residence time leads to a reduction in the volume of fresh fuel loaded and spent fuel discharged, respectively. More demanding nuclear fuel cycle parameters are combined with a need to operate nuclear power plants with maximal availability and load factors, in load-follow mode and with longer fuel cycles. In meeting these requirements, fuel has to operate in a demanding environment of high radiation fields, high temperatures, high mechanical stresses and high coolant flow. Requirements of increased fuel reliability and minimal fuel failures also remain in force. Under such circumstances, continuous development of more radiation resistant fuel materials, especially advanced cladding, careful and incremental examinations, and improved understanding and modelling of high burnup fuel behaviour are required. Following a recommendation of the IAEA Technical Working Group on Fuel Performance and Technology, the Technical Meeting on In-pile Testing and Instrumentation for Development of Generation-IV Fuels and Materials was held in Halden, Norway, on 21-24 August 2012. The purpose of the meeting was to review the current status and the progress in methods and technologies used for the in-pile testing of nuclear fuel achieved since the previous IAEA meeting on In-core Instrumentation and Reactor Core Assessment, also held in Halden in 2007. Emphasis was placed on advanced techniques applied for the understanding of high burnup fuel behaviour of water cooled power reactors that represent the vast majority of the current nuclear reactor fleet. However, the meeting also included papers and discussion on testing techniques applied or developed specifically for new fuel and structural materials considered for Generation-IV systems. The meeting was attended by 43

  16. 77 FR 18243 - Hydrogen and Fuel Cell Technical Advisory Committee (HTAC); Notice of Open Meeting

    Science.gov (United States)

    2012-03-27

    ... DEPARTMENT OF ENERGY Hydrogen and Fuel Cell Technical Advisory Committee (HTAC); Notice of Open... open meeting. SUMMARY: This notice announces a meeting of the Hydrogen and Fuel Cell Technical Advisory... Committee: The Hydrogen and Fuel Cell Technical Advisory Committee (HTAC) was established under Section 807...

  17. 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.

  18. Proceedings of the specialist meeting on nuclear fuel and control rods: operating experience, design evolution and safety aspects

    International Nuclear Information System (INIS)

    1997-01-01

    Design and management of nuclear fuel has undergone a strong evolution process during past years. The increase of the operating cycle length and of the discharge burnup has led to the use of more advanced fuel designs, as well as to the adoption of fuel efficient operational strategies. The analysis of recent operational experience highlighted a number of issues related to nuclear fuel and control rod events raising concerns about the safety aspects of these new designs and operational strategies, which led to the organisation of this Specialists Meeting on fuel and control rod issues. The meeting was intended to provide a forum for the exchange of information on lessons learned and safety concern related to operating experience with fuel and control rods (degradation, reliability, experience with high burnup fuel, and others). After an opening session 6 papers), this meeting was subdivided into four sessions: Operating experience and safety concern (technical session I - 6 papers), Fuel performance and operational issues (technical session II - 7 papers), Control rod issues (technical session III - 9 papers), Improvement of fuel design (technical session IV.A - 4 papers), Improvement on fuel fabrication and core management (technical session IV.B - 6 papers)

  19. 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)

  20. 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

  1. 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.

  2. 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

  3. 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.

  4. 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

  5. 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.

  6. 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)

  7. 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)

  8. 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

  9. 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)

  10. 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

  11. 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)

  12. 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)

  13. 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)

  14. 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.

  15. Proton Conducting Fuel Cells where Electrochemistry Meets Material Science

    DEFF Research Database (Denmark)

    Li, Qingfeng

    Fuel cells are electrochemical devices which directly convert the chemical energy of fuels into electrical energy. They are featured of high energy conversion efficiency and minimized pollutant emission. Proton conducting electrolytes are primarily used as separator materials for low and intermed...... science point of view including novel proton conducting materials and non-precious metal catalysts. The discussion will be made with highlights of DTU´s recent research and of course addressing a diverse technical audience.......Fuel cells are electrochemical devices which directly convert the chemical energy of fuels into electrical energy. They are featured of high energy conversion efficiency and minimized pollutant emission. Proton conducting electrolytes are primarily used as separator materials for low...... followed by a review of the state-of-the-art in terms of performance, lifetime and cost. Technically faced challenges are then outlined on a system level and traced back to fundamental issues of the proton conducting mechanisms and materials. Perspectives and future research are sketched from a materials...

  16. DOE Centers of Excellence Performance Portability Meeting

    Energy Technology Data Exchange (ETDEWEB)

    Neely, J. R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-04-21

    Performance portability is a phrase often used, but not well understood. The DOE is deploying systems at all of the major facilities across ASCR and ASC that are forcing application developers to confront head-on the challenges of running applications across these diverse systems. With GPU-based systems at the OLCF and LLNL, and Phi-based systems landing at NERSC, ACES (LANL/SNL), and the ALCF – the issue of performance portability is confronting the DOE mission like never before. A new best practice in the DOE is to include “Centers of Excellence” with each major procurement, with a goal of focusing efforts on preparing key applications to be ready for the systems coming to each site, and engaging the vendors directly in a “shared fate” approach to ensuring success. While each COE is necessarily focused on a particular deployment, applications almost invariably must be able to run effectively across the entire DOE HPC ecosystem. This tension between optimizing performance for a particular platform, while still being able to run with acceptable performance wherever the resources are available, is the crux of the challenge we call “performance portability”. This meeting was an opportunity to bring application developers, software providers, and vendors together to discuss this challenge and begin to chart a path forward.

  17. 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.

  18. 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

  19. 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

  20. 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)

  1. 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

  2. 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)

  3. Advanced methods of process/quality control in nuclear reactor fuel manufacture. Proceedings of a technical committee meeting

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-07-01

    Nuclear fuel plays an essential role in ensuring the competitiveness of nuclear energy and its acceptance by the public. The economic and market situation is not favorable at present for nuclear fuel designers and suppliers. The reduction in fuel prices (mainly to compete with fossil fuels) and in the number of fuel assemblies to be delivered to customers (mainly due to burnup increase) has been offset by the rising number of safety and other requirements, e.g. the choice of fuel and structural materials and the qualification of equipment. In this respect, higher burnup and thermal rates, longer fuel cycles and the use of MOX fuels are the real means to improve the economics of the nuclear fuel cycle as a whole. Therefore, utilities and fuel vendors have recently initiated new research and development programmes aimed at improving fuel quality, design and materials to produce robust and reliable fuel for safe and reliable reactor operation more demanding conditions. In this connection, improvement of fuel quality occupies an important place and this requires continuous effort on the part of fuel researchers, designers and producers. In the early years of commercial fuel fabrication, emphasis was given to advancements in quality control/quality assurance related mainly to the product itself. Now, the emphasis is transferred to improvements in process control and to implementation of overall total quality management (TQM) programmes. In the area of fuel quality control, statistical methods are now widely implemented, replacing 100% inspection. The IAEA, recognizing the importance of obtaining and maintaining high standards in fuel fabrication, has paid particular attention to this subject. In response to the rapid progress in development and implementation of advanced methods of process/quality control in nuclear fuel manufacture and on the recommendation of the International Working Group on Water Reactor Fuel Performance and Technology, the IAEA conducted a

  4. Advanced methods of process/quality control in nuclear reactor fuel manufacture. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    2000-07-01

    Nuclear fuel plays an essential role in ensuring the competitiveness of nuclear energy and its acceptance by the public. The economic and market situation is not favorable at present for nuclear fuel designers and suppliers. The reduction in fuel prices (mainly to compete with fossil fuels) and in the number of fuel assemblies to be delivered to customers (mainly due to burnup increase) has been offset by the rising number of safety and other requirements, e.g. the choice of fuel and structural materials and the qualification of equipment. In this respect, higher burnup and thermal rates, longer fuel cycles and the use of MOX fuels are the real means to improve the economics of the nuclear fuel cycle as a whole. Therefore, utilities and fuel vendors have recently initiated new research and development programmes aimed at improving fuel quality, design and materials to produce robust and reliable fuel for safe and reliable reactor operation more demanding conditions. In this connection, improvement of fuel quality occupies an important place and this requires continuous effort on the part of fuel researchers, designers and producers. In the early years of commercial fuel fabrication, emphasis was given to advancements in quality control/quality assurance related mainly to the product itself. Now, the emphasis is transferred to improvements in process control and to implementation of overall total quality management (TQM) programmes. In the area of fuel quality control, statistical methods are now widely implemented, replacing 100% inspection. The IAEA, recognizing the importance of obtaining and maintaining high standards in fuel fabrication, has paid particular attention to this subject. In response to the rapid progress in development and implementation of advanced methods of process/quality control in nuclear fuel manufacture and on the recommendation of the International Working Group on Water Reactor Fuel Performance and Technology, the IAEA conducted a

  5. 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.)

  6. Water reactor fuel element modelling at high burnup and its experimental support. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    1997-08-01

    The Technical Committee Meeting on Fuel Element Modelling at High Burnup and its Experimental Support was recommended by the International Working Group on Fuel Performance and Technology (IWGFPT). Its subject had been touched on in many of the IAEA's activities; however for the first time modellers and experimentalists were brought together to have an exchange of views on the research under way and to identify areas where new knowledge is necessary to improve the safety, reliability and/or economics of nuclear fuel. The timely organization of this meeting in conjunction with the second meeting of the Co-ordinated Research Programme on Fuel Modelling at Extended Burnup, in short ''FUMEX'', allowed fruitful participation of representatives of developing countries which are only rarely exposed to such a scientific event. The thirty-nine papers presented covered the status of codes and experimental facilities and the main phenomena affecting the fuel during irradiation, namely: thermal fuel performance, clad corrosion and pellet-cladding interaction (PCI) and fission gas release (FGR). Refs, figs, tabs

  7. Water reactor fuel element modelling at high burnup and its experimental support. Proceedings of a technical committee meeting

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-08-01

    The Technical Committee Meeting on Fuel Element Modelling at High Burnup and its Experimental Support was recommended by the International Working Group on Fuel Performance and Technology (IWGFPT). Its subject had been touched on in many of the IAEA`s activities; however for the first time modellers and experimentalists were brought together to have an exchange of views on the research under way and to identify areas where new knowledge is necessary to improve the safety, reliability and/or economics of nuclear fuel. The timely organization of this meeting in conjunction with the second meeting of the Co-ordinated Research Programme on Fuel Modelling at Extended Burnup, in short ``FUMEX``, allowed fruitful participation of representatives of developing countries which are only rarely exposed to such a scientific event. The thirty-nine papers presented covered the status of codes and experimental facilities and the main phenomena affecting the fuel during irradiation, namely: thermal fuel performance, clad corrosion and pellet-cladding interaction (PCI) and fission gas release (FGR). Refs, figs, tabs.

  8. 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)

  9. 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

  10. 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

  11. 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

  12. 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

  13. Practices and developments in spent fuel burnup credit applications. Proceedings of a technical committee meeting

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-10-01

    The International Atomic Energy Agency convened a technical committee Meeting on Requirements, Practices and Developments in Burnup Credit (BUC) Applications in Madrid, Spain, from 22 to 26 April 2002. The purpose of this meeting was to explore the progress and status of international activities related to the BUC applications for spent nuclear fuel. This meeting was the third major meeting on the uses of BUC for spent fuel management systems held since the IAEA began to monitor the uses of BUC in spent fuel management systems in 1997. The first major meeting was an Advisory Group meeting (AGM), which was held in Vienna, in October 1997. The second major meeting was a technical committee meeting (TCM), which was held in Vienna, in July 2000. Several consultants meetings were held since 1997 to advise and assist the IAEA in planning and conducting its BUC activities. The proceedings of the 1997 AGM were published as IAEA-TECDOC-1013, and the proceedings of the 2000 TCM as IAEA-TECDOC-1241. BUC for wet and dry storage systems, spent fuel transport, reprocessing and final disposal is needed in many Member States to allow for increased enrichment, and to increase storage capacities, cask capacities and dissolver capacities avoiding the need for extensive modifications. The use of BUC is a necessity for spent fuel disposal.

  14. 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

  15. 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

  16. 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.

  17. 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)

  18. 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)

  19. Proceedings of the international meeting on research and test reactor core conversions from HEU to LEU fuels

    International Nuclear Information System (INIS)

    Travelli, A.

    1983-09-01

    Conversion of research and test reactor cores from the use of high enrichment uranium to the use of low enrichment uranium depends on the cooperation of many research organizations, reactor operators, and government agencies. At a technical level, it involves almost all aspects of the fuel cycle, including fuel development, testing, shipping and reprocessing; experiment performance; economics; and safety and licensing aspects. The reactors involved and the conversion activities are distributed among approximately 25 countries, making this a subject which is best dealt with on an international basis. To foster direct communication in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the fifth of a series which began in 1978. The previous meetings were held at Argonne (International Meeting of Research Reactor Fuel Designers, Developers, and Fabricators, Argonne National Laboratory, Argonne, U.S.A., November 910, 1978), at Saclay (IAEA Consultants' Meeting on Research Reactor Core Conversions from HEU to LEU, Centre d'etudes Nucleaires de Saclay, Saclay, France, December 12-14, 1979), at Argonne (International Meeting on Development, Fabrication and Application of Reduced Enrichment Fuels for Research and Test Reactors, Argonne National Laboratory, Argonne, U.S.A., November 12-14, 1980) and at Juelich (Seminar on Research Reactor Operation and Use, Juelich Nuclear Research Center, Juelich, F.R.G., September 48, 1981). Proceedings from the two most recent previous meetings were published as ANL/RERTR/TM-3 (CONF-801144) and IAEA-SR-77. The spirit of this meeting differs slightly from that of the previous meetings. The advances which have been made and the growing maturity of the effort have caused a gradual shift of emphasis away from those topics which dominated the floor during the first meetings, such as fuel and methods development, and towards topics which concern more

  20. Proceedings of the international meeting on research and test reactor core conversions from HEU to LEU fuels

    Energy Technology Data Exchange (ETDEWEB)

    Travelli, A [Argonne National Laboratory, Argonne, IL (United States)

    1983-09-01

    Conversion of research and test reactor cores from the use of high enrichment uranium to the use of low enrichment uranium depends on the cooperation of many research organizations, reactor operators, and government agencies. At a technical level, it involves almost all aspects of the fuel cycle, including fuel development, testing, shipping and reprocessing; experiment performance; economics; and safety and licensing aspects. The reactors involved and the conversion activities are distributed among approximately 25 countries, making this a subject which is best dealt with on an international basis. To foster direct communication in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the fifth of a series which began in 1978. The previous meetings were held at Argonne (International Meeting of Research Reactor Fuel Designers, Developers, and Fabricators, Argonne National Laboratory, Argonne, U.S.A., November 910, 1978), at Saclay (IAEA Consultants' Meeting on Research Reactor Core Conversions from HEU to LEU, Centre d'etudes Nucleaires de Saclay, Saclay, France, December 12-14, 1979), at Argonne (International Meeting on Development, Fabrication and Application of Reduced Enrichment Fuels for Research and Test Reactors, Argonne National Laboratory, Argonne, U.S.A., November 12-14, 1980) and at Juelich (Seminar on Research Reactor Operation and Use, Juelich Nuclear Research Center, Juelich, F.R.G., September 48, 1981). Proceedings from the two most recent previous meetings were published as ANL/RERTR/TM-3 (CONF-801144) and IAEA-SR-77. The spirit of this meeting differs slightly from that of the previous meetings. The advances which have been made and the growing maturity of the effort have caused a gradual shift of emphasis away from those topics which dominated the floor during the first meetings, such as fuel and methods development, and towards topics which concern more

  1. Fuel failure in water reactors: Causes and mitigation. Proceedings of a technical meeting

    International Nuclear Information System (INIS)

    2003-03-01

    The objective of this technical meeting (TM) was to review the present knowledge of the causes and mechanisms of fuel failure in water reactors during normal operational conditions. Emphasis has been given to analysis of failure causes and their mitigation by means of design as well as plant and core operation including strategies for operation with failed fuel. Some information on detection techniques (on-line monitoring and diagnostics, flux tilting, sipping techniques, etc) has also been presented. This TM presented also the progress on the above-mentioned subjects since the last meeting held in 1992 (Dimitrovgrad, Russian Federation). The topics covered in the papers were as follows: Experience feedback on fuel reliability (8 papers); Strategies to avoid or mitigate fuel failures (4 papers); Experimental studies on fuel failures and degradation mechanisms (4 papers); Modelling of fuel failure mechanisms (3 papers); Detection and monitoring during operation or outage (4 papers); Modelling and assessment of fuel failures (3 papers)

  2. 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

  3. Studies on production planning of IPEN fuel-element plant in order to meet RMB demand

    Energy Technology Data Exchange (ETDEWEB)

    Negro, Miguel L.M.; Saliba-Silva, Adonis M.; Durazzo, Michelangelo, E-mail: mlnegro@ipen.br, E-mail: saliba@ipen.br, E-mail: mdurazzo@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    The plant of the Nuclear Fuel Center (CCN) will have to change its current laboratorial production level to an industrial level in order to meet the fuel demand of RMB and of IEA-R1. CCN's production process is based on the hydrolysis of UF6, which is not a frequent production route for nuclear fuel. The optimization of the production capacity of such a production route is a new field of studies. Two different approaches from the area of Operations Research (OR) were used in this paper. The first one was the PERT/CPM technique and the second one was the creation of a mathematical linear model for minimization of the production time. PERT/CPM's results reflect the current situation and disclose which production activities may not be critical. The results of the second approach show a new average time of 3.57 days to produce one Fuel Element and set the need of inventory. The mathematical model is dynamic, so that it issues better results if performed monthly. CCN's management team will therefore have a clearer view of the process times and production and inventory levels. That may help to shape the decisions that need to be taken for the enlargement of the plant's production capacity. (author)

  4. Studies on production planning of IPEN fuel-element plant in order to meet RMB demand

    International Nuclear Information System (INIS)

    Negro, Miguel L.M.; Saliba-Silva, Adonis M.; Durazzo, Michelangelo

    2015-01-01

    The plant of the Nuclear Fuel Center (CCN) will have to change its current laboratorial production level to an industrial level in order to meet the fuel demand of RMB and of IEA-R1. CCN's production process is based on the hydrolysis of UF6, which is not a frequent production route for nuclear fuel. The optimization of the production capacity of such a production route is a new field of studies. Two different approaches from the area of Operations Research (OR) were used in this paper. The first one was the PERT/CPM technique and the second one was the creation of a mathematical linear model for minimization of the production time. PERT/CPM's results reflect the current situation and disclose which production activities may not be critical. The results of the second approach show a new average time of 3.57 days to produce one Fuel Element and set the need of inventory. The mathematical model is dynamic, so that it issues better results if performed monthly. CCN's management team will therefore have a clearer view of the process times and production and inventory levels. That may help to shape the decisions that need to be taken for the enlargement of the plant's production capacity. (author)

  5. 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.

  6. 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)

  7. 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

  8. Proceedings of the fuel cells 1994 contractors review meeting

    Science.gov (United States)

    Carpenter, C. P., II; Mayfield, M. J.

    1994-08-01

    METC annually sponsors this conference to provide a forum for energy executives, engineers, etc. to discuss advances in fuel cell research and development projects, to exchange ideas with private sector attendees, and to review relevant results in fuel cell technology programs. Two hundred and three people from industry, academia, and Government attended. The conference attempts to showcase the partnerships with the Government and with industry, by seeking activity participation and involvement from the Office of Energy Efficiency and Renewable Energy, EPRI, GRI, and APRA. In addition to sessions on fuel cells (solid oxide, molten carbonate, etc.) for stationary electric power generation, sessions on US DOE's Fuel Cell Transportation Program and on DOD/APRA's fuel cell logistic fuel program were presented. In addition to the 29 technical papers, an abstract of an overview of international fuel cell development and commercialization plans in Europe and Japan is included. Selected papers were indexed separately for inclusion in the Energy Science and Technology Database.

  9. Status and trends in spent fuel reprocessing. Proceedings of an advisory group meeting

    International Nuclear Information System (INIS)

    1999-08-01

    Spent fuel management has always been an important part of the nuclear fuel cycle and is still one of the most important activities in all countries exploiting the peaceful use of nuclear energy. Continuous attention is being given by the IAEA to the collection, analysis and exchange of information on spent fuel management. Its role in this area is to provide a forum for exchanging information and to coordinate and encourage closer co-operation among Member States in certain research and developing activities that are of common interest. As part of spent fuel management, reprocessing activities have been reviewed from time to time on a low profile level under the terminology 'spent fuel treatment'. However, spent fuel treatment covers, in broad terms, spent fuel storage (short, interim and long term), fuel rod consolidation, reprocessing and, in case the once-through cycle is selected, conditioning of the spent fuel for disposal. Hence the reprocessing activities under the heading 'spent fuel treatment' were somewhat misleading. Several meetings on spent fuel treatment have been organized during the fast decade: an Advisory Group meeting (AGM) in 1992, a Technical Committee meeting in 1995 and recently an Advisory Group meeting from 7 to 10 September 1998. The objectives of the meetings were to review the status and trends of spent fuel reprocessing, to discuss the environmental impact and safety aspects of reprocessing facilities and to define the most important issues in this field. Notwithstanding the fact that the Summary of the report does not include aspects of military reprocessing, some of the national presentations do refer to some relevant aspects (e.g. experience, fissile stockpiles)

  10. 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

  11. 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

  12. 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

  13. Implementation of burnup credit in spent fuel management systems. Proceedings of a technical committee meeting

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-08-01

    The purpose of this Technical Committee Meeting was to explore the status of international activities related to the use of burnup credit for spent fuel applications. This was the second major meeting on the issues of burnup credit for spent fuel management systems held since the IAEA began to monitor the uses of burnup credit in spent fuel management systems in 1997. Burnup credit for wet and dry storage systems is needed in many Member States to allow for increased initial fuel enrichment, and to increase the storage capacity and thus to avoid the need for extensive modifications of the spent fuel management systems involved. This document contains 31 individual papers presented at the Meeting; each of the papers was indexed separately.

  14. Implementation of burnup credit in spent fuel management systems. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    2001-08-01

    The purpose of this Technical Committee Meeting was to explore the status of international activities related to the use of burnup credit for spent fuel applications. This was the second major meeting on the issues of burnup credit for spent fuel management systems held since the IAEA began to monitor the uses of burnup credit in spent fuel management systems in 1997. Burnup credit for wet and dry storage systems is needed in many Member States to allow for increased initial fuel enrichment, and to increase the storage capacity and thus to avoid the need for extensive modifications of the spent fuel management systems involved. This document contains 31 individual papers presented at the Meeting; each of the papers was indexed separately

  15. Nuclear fuel behaviour modelling at high burnup and its experimental support. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    2001-07-01

    The Technical Committee Meeting (TCM) included separate sessions on the specific topics of fuel thermal performance and fission product retention. On thermal performance, it is apparent that the capability exists to measure conductivity in high burnup fuel either by out-of-pile measurement or by instrumentation of test reactor rods. State-of-the-art modelling codes contain models for the conductivity degradation process, and hence adequate predictions of fuel temperature are achievable. Concerning fission product release, it is clear that many groups around the world are actively investigating the subject, with experimental and modelling programmes being pursued. However, a general consensus on the exact mechanisms of gas release and related gas bubble swelling has yet to emerge, even at medium burnup levels. Fission gas phenomena, not only the release to open volumes, but the whole sequence of processes taking place prior to this, need to be modelled in any modern fuel performance code. The presence of gaseous fission products may generate rapid fuel swelling during power transients, and this can cause PCI and rod failure. At high burnups, the quantity of released gases could give rise to pressures exceeding the safe limits. Modelling of pellet-cladding interaction (PCI) effects during transient operation is also an active area of study for many groups. In some situations a purely empirical approach to failure modelling can be justified, while for other applications a more detailed mechanistic approach is required. Another aspect of cladding modelling which was featured at the TCM concerned corrosion and hydriding. Although this issue can be the main life-limiting factor on fuel duty, it is apparent that modelling methods, and the experimental measurement techniques that underpin them, are adequate. A session was included on MOX fuel modelling. Substantial programmes of work, especially by the MOX vendors, appear to be underway to bring the level of understanding

  16. International topical meeting. Research Reactor Fuel Management (RRFM) and meeting of the International Group on Reactor Research (IGORR)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2007-07-01

    Nuclear research and test reactors have been in operation for over 60 years, over 270 research reactors are currently operating in more than 50 countries. This meeting is dedicated to different aspects of research reactor fuels: new fuels for new reactors, the conversion to low enriched uranium fuels, spent fuel management and computational tools for core simulation. About 80 contributions are reported in this document, they are organized into 7 sessions: 1) international topics and overview on new projects and fuel, 2) new projects and upgrades, 3) fuel development, 4) optimisation and research reactor utilisation, 5) innovative methods in research reactors physics, 6) safety, operation and research reactor conversion, 7) fuel back-end management, and a poster session. Experience from Australian, Romanian, Libyan, Syrian, Vietnamese, South-African and Ghana research reactors are reported among other things. The Russian program for research reactor spent fuel management is described and the status of the American-driven program for the conversion to low enriched uranium fuels is presented. (A.C.)

  17. 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

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

    Science.gov (United States)

    Acosta, Waldo A.; Beckel, Stephen A.

    1989-01-01

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

  19. Technical committee meeting on fuel and cladding interaction. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1977-04-01

    Experiments and experiences concerning fuel-cladding interaction in thermal and fast neutron flux burnup are dealt with. A number of results from in-pile and out-of pile experiments with different fuel pins with cladding made of different stainless steels showed the importance of corrosion process, dependent on the burnup, core temperature, metal-oxide ratio, and other steady state parameters in the core of fast reactors (most frequently LMFBRs). This is of importance for fuel pins design and fabrication. Mixed oxide fuel is treated in many cases.

  20. Technical committee meeting on fuel and cladding interaction. Summary report

    International Nuclear Information System (INIS)

    1977-04-01

    Experiments and experiences concerning fuel-cladding interaction in thermal and fast neutron flux burnup are dealt with. A number of results from in-pile and out-of pile experiments with different fuel pins with cladding made of different stainless steels showed the importance of corrosion process, dependent on the burnup, core temperature, metal-oxide ratio, and other steady state parameters in the core of fast reactors (most frequently LMFBRs). This is of importance for fuel pins design and fabrication. Mixed oxide fuel is treated in many cases

  1. Meeting the special requirements of the fuel cycle industry

    International Nuclear Information System (INIS)

    McKenzie, N.C.

    1979-01-01

    The nuclear fuel cycle industry is generally thriving, despite the slow-down in nuclear power plant construction. Turnover is expected to grow from the present Pound4.7 billion to Pound12 billion in 1985 and nearly Pound16 billion in 1988. The emerging fuel cycle companies have large demands for investment and working capital but are not yet financially strong. Utilities also are finding the funding of fuel onerous. Special nuclear fuel finance companies have been formed to help both satisfy their financing needs. A particular problem is the security of loans and much creative thinking is being applied to the use of contracts as collaterals. (U.K.)

  2. The Impact of Corporate Board Meetings on Corporate Performance ...

    African Journals Online (AJOL)

    Our findings suggest a statistically significant and positive association between the frequency of corporate board meetings and corporate performance, implying that SA boards that meet more frequently tend to generate higher financial performance. A further investigation indicates a significant non-monotonic link between ...

  3. Proceedings of a topical meeting on safety of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    1996-01-01

    The topical meeting on the safety of the nuclear fuel cycle is composed of 17 papers grouped into four sessions which titles are: operational safety in nuclear fuel facilities; safety criteria and regulatory philosophy; plant hazard analysis and mitigation; plant experience and emergency planning

  4. Technical meeting on 'Review of solid and mobile fuels for partitioning and transmutation systems'. Working material

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-07-01

    The topics covered during the Meeting were divided into two Sessions. Session 1 - Qualification of Solid and Mobile Fuels delt with: Neutronic, fuel and material properties of a molten salt transmuter; and Preliminary analysis of transmutation fuels for KALIMER. Session 2 - Reactor Physics and Safety Characteristics of Transmutation Systems based on Solid and Mobile Fuel Types included the following: Activity in NEA for P and T area; IAEA activities in the area of partitioning and transmutation; The R and D activity in Brazil: A conceptual fast energy amplifier ADS cooled by helium double stata Th/U fuel cycle; Closed fuel cycle and contemporary tendencies of the nuclear facilities development; Current Russian activities in P and T area; Pyrochemical reprocessing and nuclear spent fuel disposal project; Fuel selection criteria specific for double stratum minor actinide burners.

  5. 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

  6. Hot Cell Post-Irradiation Examination and Poolside Inspection of Nuclear Fuel. Proceedings of the IAEA-HOTLAB Technical Meeting

    International Nuclear Information System (INIS)

    2013-04-01

    The growing operational requirements for nuclear fuel, such as longer fuel cycles, higher burnups and wider use of transient regimes, require more robust fuel designs and more radiation resistant materials. Development of such advanced fuels is only possible with testing and analysis of their performance and application of adequate post-irradiation examination (PIE) methods and techniques. In addition, operational feedback data from poolside and PIE facilities are absolutely necessary for verification of fuel modelling codes and analysis of fuel failure mechanisms. For these reasons, the International Atomic Energy Agency (IAEA) has supported the international exchange of knowledge and sharing of best practices in the application of modern destructive and non-destructive methods of investigation of highly radioactive materials through a series of technical meetings (TMs), the last of which was held in 2006 in Buenos Aires. Since 1963, similar meetings, initially at the European level, have been organized by the Hot Laboratories and Remote Handling Working Group (HOTLAB), a partner in the development of the IAEA's Post Irradiation Examination Facilities Database (PIEDB), part of the IAEA's Integrated Nuclear Fuel Cycle Information System. With this successful partnership in mind, in 2010 the IAEA Technical Working Group on Fuel Performance and Technology recommended that a joint IAEA-HOTLAB TM be held on 'Hot Cell Post-Irradiation Examination and Pool-Side Inspection of Nuclear Fuel', covering questions relevant to the IAEA sub-programmes on 'Nuclear Power Reactor Fuel Engineering' and 'Management of Spent Fuel from Nuclear Power Reactors'. The TM was held on 23-27 May 2011, in Smolenice, Slovakia, with the participation of a large number of interested organizations and comprehensive coverage of major PIE and poolside inspection issues relating to both operation and storage of fuel for nuclear power reactors. The proceedings, summaries and conclusions of that joint

  7. IAEA Technical committee meeting on methods used in design of spent fuel storage facilities

    International Nuclear Information System (INIS)

    Vitikainen, E.; Silfverberg, P.

    1985-01-01

    The meeting was held in Espoo, Finland and hosted by the Technical Research Centre of Finland (VTT), and was arranged to report and discuss design methods, licensing practise, operational experience as well as economic aspects connectied with spent fuel storage. This report contains session summaries by the session chairmen and the papers presented at the meeting

  8. An approach to meeting the spent fuel standard

    Energy Technology Data Exchange (ETDEWEB)

    Makhijani, A. [Institute for Energy and Environmental Research, Takoma Park, MD (United States)

    1996-05-01

    The idea of the spent fuel standard is that there should be a high surface gamma radiation to prevent theft. For purposes of preventing theft, containers should be massive, and the plutonium should be difficult to extract. This report discusses issues associated with the spent fuel standard.

  9. An approach to meeting the spent fuel standard

    International Nuclear Information System (INIS)

    Makhijani, A.

    1996-01-01

    The idea of the spent fuel standard is that there should be a high surface gamma radiation to prevent theft. For purposes of preventing theft, containers should be massive, and the plutonium should be difficult to extract. This report discusses issues associated with the spent fuel standard

  10. 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

  11. 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...

  12. Remote technology in spent fuel management. Proceedings of an advisory group meeting

    International Nuclear Information System (INIS)

    1999-01-01

    Spent fuel management has always been one of the important stages in the nuclear fuel cycle and it is still one of the most vital problems common to all countries with nuclear reactors. It begins with the discharge of spent fuel from a power or research reactor and ends with its ultimate disposition either by direct disposal or by reprocessing of the spent fuel. Continuous attention is being given by the IAEA to the collection, analysis and exchange of information on spent fuel management. Its role in this area is to provide a forum for exchanging information and development activities that are of common interest. Within its spent fuel management programme, the IAEA has monitored the progress, the benefits and the implementation of remote technologies such as remote tools, robotics, etc. An Advisory Group Meeting on Remote Technology in Spent Fuel Management was held in September 1997 in order to bring together specialists working in this field and to collect information on new technical and economic developments. The objective of the Advisory Group meeting was to review remote technologies in use for the complete range of spent fuel handling and spent fuel management covering wet and dry environments, to describe ongoing developments and to prepare a technical report. This document contains contributions presented at the Meeting. Each paper was indexed and provided with an abstract

  13. High Burnup Fuel: Implications and Operational Experience. Proceedings of a Technical Meeting

    International Nuclear Information System (INIS)

    2016-08-01

    This publication reports on the outcome of a technical meeting on high burnup fuel experience and economics, held in Buenos Aires, Argentina in 2013. The purpose of the meeting was to revisit and update the current operational experience and economic conditions associated with high burnup fuel. International experts with significant experience in experimental programmes on high burnup fuel discussed and evaluated physical limitations at pellet, cladding and structural component levels, with a wide focus including fabrication, core behaviour, transport and intermediate storage for most types of commercial nuclear power plants

  14. 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)

  15. Proceedings of the fuel cells `94 contractors review meeting

    Energy Technology Data Exchange (ETDEWEB)

    Carpenter, C.P. II; Mayfield, M.J. [eds.] [USDOE Morgantown Energy Technology Center, WV (United States)

    1994-08-01

    METC annually sponsors this conference to provide a forum for energy executives, engineers, etc. to discuss advances in fuel cell research and development projects, to exchange ideas with private sector attendees, and to review relevant results in fuel cell technology programs. Two hundred and three people from industry, academia, and Government attended. The conference attempts to showcase the partnerships with the Government and with industry, by seeking activity participation and involvement from the Office of Energy Efficiency and Renewable Energy, EPRI, GRI, and APRA. In addition to sessions on fuel cells (solid oxide, molten carbonate, etc.) for stationary electric power generation, sessions on US DOE`s Fuel Cell Transporation Program and on DOD/APRA`s fuel cell logistic fuel program were presented. In addition to the 29 technical papers, an abstract of an overview of international fuel cell development and commercialization plans in Europe and Japan is included. Selected papers were indexed separately for inclusion in the Energy Science and Technology Database.

  16. Proceedings of the Second Meeting of the OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs, 23-25 September 2014, OECD-NEA HQ

    International Nuclear Information System (INIS)

    Massara, S.; ); Bragg-Sitton, Shannon; Pasamehmetoglu, K.; Yang, Jae Ho; Dolley, Evan J.; Rebak, Raul B.; Sowder, Andrew; Cheng, Bo; Kurata, Masaki; Van Nieuwenhove, Rudi; Li, R.; McClellan, Ken; Nelson, Andy; Carmack, Jon; Harp, Jason; Finck, Phillip; ); Kakicuhi, K.

    2014-09-01

    Under the guidance of the OECD-NEA Nuclear Science Committee, the expert group acts as a forum for scientific and technical information exchange on advanced light water reactor (LWR) fuels with enhanced accident tolerance. The expert group focusses on the fundamental properties and behaviour under normal operations and accident conditions for advanced core materials and components (fuels, cladding, control rods, etc.). The materials considered are applicable to Gen II and Gen III Light Water Reactors, as well as Gen III+ reactors under construction. The objective of the expert group is to define and coordinate a programme of work to help advance the scientific knowledge needed to provide the technical underpinning for the development of advanced LWR fuels with enhanced accident tolerance compared to currently used zircaloy/UO 2 fuel systems, as well as other non-fuel core components with important roles in LWR performance under accident conditions. This document brings together the available presentations (slides) given at the Second Meeting of the OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs. Content: 1 - Proposed Agenda; 2 - Expert Group meeting - 23 September 2014: - Introduction and background (S. Massara, OECD-NEA) - Expected outcomes from the TFs meetings scheduled on 24-25 September (K. Pasamehmetoglu, EG Chair, INL); 3 - Task Force 1 (Systems assessment) meeting - 24 September 2014: - Metrics for the Evaluation of LWR Accident Tolerant Fuel (S. Bragg-Sitton, INL); 4 - Task Force 2 (Cladding/core materials) meeting - 24 September 2014: - Summary on SiC Task Force 2 (Clad) meeting (J.H. Yang, KAERI); - Accident Tolerant Advanced Steels Cladding for Commercial Light Water Reactors (E. Dolley, GE); - Molybdenum-Alloy Fuel Cladding Development and Testing - Update from April 2014 NEA ATF Meeting (A. Sowder, EPRI); - Accident Tolerant Control Rod Development in Japan (M. Kurata, JAEA); - IFA-774: The first in-pile test with coated fuel rods (R. Van

  17. Studies on fuels with low fission gas release. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    1997-10-01

    For more than a decade, the IAEA has organized various specialists meetings to discuss advances in nuclear fuel technology for non-water cooled reactors. In order to review progress in research and development of fuels with low fission gas release for light water reactors, fast reactors and research reactors, an IAEA Technical Committee meeting was organized in October 1996. At the invitation of the Government of the Russian Federation, the meeting was held in Moscow. Experts from seven Member States and one international organization participated. The objective of the meeting was to exchange topical information on such fuels, to evaluate their advantages and drawbacks, and to explore their commercial utilization. The present volume contains the full text of the sixteen papers presented at the meeting. The information compiled in these proceedings should be useful for engineers, scientists and managers from nuclear fuel development organizations, fuel fabrication plants, utilities and regulatory bodies who are involved in the analysis of fuel behaviour under normal and accident conditions. Refs, figs, tabs

  18. Accident Tolerant Fuel Concepts for Light Water Reactors. Proceedings of a Technical Meeting

    International Nuclear Information System (INIS)

    2016-06-01

    Nuclear fuel is a highly complex material that has been subject to continuous development over the past 40 years and has reached a stage where it can be safely and reliably irradiated up to 65 GWd/tU in commercial nuclear reactors. During this time, there have been many improvements to the original designs and materials used. However, the basic design of uranium oxide fuel pellets clad with zirconium alloy tubing has remained the fuel choice for the vast majority of commercial nuclear power plants. Severe accidents, such as those at the Three Mile Island and Fukushima Daiichi have shown that under such extreme conditions, nuclear fuel will fail and the high temperature reactions between zirconoi alloys and water will lead to the generation of hydrogen, with the potential for explosions to occur, daming the plant further. Recognizing that the current fuel designs are vulnerable to severe accident conditions, tehre is renewed interesst in alternative fuel designs that would be more resistant to fuel failure and hydrogen production. Such new fuel designs will need to be compatible with existing fuel and reactor systems if they are to be utilized in the current reactor fleet and in current new build designs, but there is also the possibility of new designs for new reactor systems. This publication provides a record of the Technical Meeting on Accident Tolerant Fuel Concepts for Light Water Reactors, held at Oak Ridge National Laboratories (ORNL), United States of America, 13-16 October 2014, to consider the early stages of research and development into accident tolerant fuel. There were 45 participants from 10 countries taking part in the meeting, with 32 papers organized into 7 sessions, of which 27 are included in this publication. This meeting is part of a wider investigation into such designs, and it is anticipated that further Technical Meetings and research programmes will be undertaken in this field

  19. 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

  20. Proceedings of the third specialist meeting on sodium/fuel interaction in fast reactors

    International Nuclear Information System (INIS)

    1976-01-01

    This specialist meeting, sponsored by the OECD-NEA and organized by the Power Reactor and Nuclear Fuel Development Corporation, was attended by 56 delegates from 6 countries and the CEC (Commission of the European Communities). The purpose of the meeting was to bring together and discuss in depth the Fuel-Sodium Interaction, a phenomenon of major importance in the assessment of the Hypothetical Core Disruptive Accident in the Liquid Metal Fast Breeder Reactor. The meeting was essentially a follow-up of an earlier meeting held at Ispra in December 1973. In all, 29 papers were presented, covering the following topics: 1. Current perspective on sodium-fuel interaction in LMFBR safety; 2. Basic experimental and theoretical studies including other materials; 3. In-pile and out-of-pile experimental studies on sodium-fuel interaction; 4. Theoretical models for the interpretation of experiments and for application to reactor situations. The meeting is considered useful in narrowing down the chain of events necessary to get energetic interaction, large work potential, but many points are being clarified on the gap between the basic vapor explosions and the real fuel sodium interactions in the HCDA scenario of LMFBR. Finally another meeting of the same nature as this one has been recommended

  1. 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.

  2. 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

  3. Assessment of technologies to meet a low carbon fuel standard.

    Science.gov (United States)

    Yeh, Sonia; Lutsey, Nicholas P; Parker, Nathan C

    2009-09-15

    California's low carbon fuel standard (LCFS) was designed to incentivize a diverse array of available strategies for reducing transportation greenhouse gas (GHG) emissions. It provides strong incentives for fuels with lower GHG emissions, while explicitly requiring a 10% reduction in California's transportation fuel GHG intensity by 2020. This paper investigates the potential for cost-effective GHG reductions from electrification and expanded use of biofuels. The analysis indicates that fuel providers could meetthe standard using a portfolio approach that employs both biofuels and electricity, which would reduce the risks and uncertainties associated with the progress of cellulosic and battery technologies, feedstock prices, land availability, and the sustainability of the various compliance approaches. Our analysis is based on the details of California's development of an LCFS; however, this research approach could be generalizable to a national U.S. standard and to similar programs in Europe and Canada.

  4. 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)

  5. 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

  6. 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

  7. 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

  8. 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)

  9. 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)

  10. 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

  11. Impact of nuclear fuel performance and reliability in a deregulated electricity market

    International Nuclear Information System (INIS)

    Barnoski, M.; Kaiser, B.

    1997-01-01

    US nuclear plants are now placing a great deal of importance on fuel performance and reliability because of their direct effect on a plant's capacity factor, availability, operating flexibility, outage duration, and worker exposure. This is reflected by utilities making fuel reliability a principal vendor evaluation criterion and making fuel performance an important bonus/penalty provision. Defect-free fuel, combined with ever improving design features, makes it possible for plants to operate at higher outputs while reducing operating and maintenance costs. Higher electrical output at lower cost is essential for US nuclear plants to meet the expected competition in a deregulated market. Through continued investment in manufacturing improvements and a vigilant, systematic design process, ABB's nuclear fuel has contributed to helping its nuclear utility customers reduce their cost of doing business. (author) 1 fig., 1 tab

  12. Are board meetings proactive or reactive to performance?

    Directory of Open Access Journals (Sweden)

    Yi Wang

    2008-11-01

    Full Text Available The purpose of this paper is to investigate the relationship between the intensity of board activity, as represented by board meeting frequency, and firm financial performance, using data from the top 500 Australian companies. Firm performance measures include return on assets, return on equity and shareholder return; several control variables are introduced in the analysis. The results indicate that board meeting frequency has a positive impact on subsequent shareholder return. Regarding the explanatory factors for the level of board activity, it is reported that firms with more board committees tend to have relatively more board and committee meetings; firms with larger boards have less board meetings. In addition, lower managerial ownership leads to more committee meetings.

  13. 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.

  14. 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)

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

    International Nuclear Information System (INIS)

    Anniel, Marc; Bordy, Michel-Aristide

    2009-01-01

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

  16. Development of an analytical model to assess fuel property effects on combustor performance

    Science.gov (United States)

    Sutton, R. D.; Troth, D. L.; Miles, G. A.; Riddlebaugh, S. M.

    1987-01-01

    A generalized first-order computer model has been developed in order to analytically evaluate the potential effect of alternative fuels' effects on gas turbine combustors. The model assesses the size, configuration, combustion reliability, and durability of the combustors required to meet performance and emission standards while operating on a broad range of fuels. Predictions predicated on combustor flow-field determinations by the model indicate that fuel chemistry, as defined by hydrogen content, exerts a significant influence on flame retardation, liner wall temperature, and smoke emission.

  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. 75 FR 11806 - Notice of Public Meeting; Tire Fuel Efficiency

    Science.gov (United States)

    2010-03-12

    .... After the NPRM was issued, NHTSA conducted an internet survey to further explore what influences... CONTACT: Ms. Mary Versailles, Telephone: 1-202- 366-2057, Office of International Vehicle, Fuel Economy... proposed in the NPRM. After the NPRM was issued, NHTSA conducted an internet survey to further explore what...

  19. 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)

  20. 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

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

    Science.gov (United States)

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

    1994-07-01

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

  2. Performance of 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)

  3. 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.

  4. Meeting Flooring for Schools Performance Needs

    Science.gov (United States)

    Simoneau, Mike

    2012-01-01

    Many school and college campuses are like small cities--they offer a little bit of everything. Although academic buildings and residence halls account for a sizable percentage of the facilities on a campus, the academic community on a school campus can include other buildings with specialized design and performance needs: laboratories,…

  5. Development of High Performance Hybrid Fuels

    Data.gov (United States)

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

  6. 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)

  7. 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)

  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. Safety of and regulations for nuclear fuel cycle facilities. Report of a technical committee meeting

    International Nuclear Information System (INIS)

    2001-05-01

    In order to compile information on the nature of the safety concerns and current status of the regulations concerning nuclear fuel cycle facilities in Member States, an IAEA Technical Committee meeting on this topic was convened from 8 to 12 May 2000 in Vienna. The present publication contains the results of this meeting. The contributions of the participants in Annex 3 exemplify the work done in some Member States to develop an adequate regulatory framework to oversee the safe operation of these facilities

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

    International Nuclear Information System (INIS)

    Jicha, J.J. Jr.

    1990-01-01

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

  11. 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

  12. 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)

  13. Meeting U.S. passenger vehicle fuel economy standards in 2016 and beyond

    International Nuclear Information System (INIS)

    Cheah, Lynette; Heywood, John

    2011-01-01

    New fuel economy standards require new U.S. passenger vehicles to achieve at least 34.1 miles per gallon (MPG) on average by model year 2016, up from 28.8 MPG today. In this paper, the magnitude, combinations and timings of the changes required in U.S. vehicles that are necessary in order to meet the new standards, as well as a target of doubling the fuel economy within the next two decades are explored. Scenarios of future vehicle characteristics and sales mix indicate that the 2016 mandate is aggressive, requiring significant changes starting from today. New vehicles must forgo horsepower improvements, become lighter, and a greater number will use advanced, more fuel-efficient powertrains, such as smaller turbocharged engines, hybrid-electric drives. Achieving a factor-of-two increase in fuel economy by 2030 is also challenging, but more feasible since the auto industry will have more lead time to respond. A discussion on the feasibility of meeting the new fuel economy mandate is included, considering vehicle production planning realities and challenges in deploying new vehicle technologies into the market. - Research Highlights: → The new 2016 fuel economy standards for passenger vehicles in the U.S. can be met, but are aggressive. → Future vehicles must forgo horsepower improvements, become lighter, and a greater number will use advanced, more fuel efficient powertrains. → The challenge of meeting the fuel economy targets is defined by both the magnitude and the timing of these requirements. → Doubling the fuel economy by 2030 is also challenging, but more feasible since the auto industry will have more lead time to respond.

  14. OECD/CSNI specialist meeting on fuel coolant interactions: summary and conclusions

    International Nuclear Information System (INIS)

    1997-01-01

    Research activities and interest on fuel-coolant interaction (FCI) have been increased and broadened since the last CSNI Specialist Meeting held in January 1993. Significant experimental and analytical research has been performed in many OECD countries and others. The growing international interest is, in large part, due to the emphasis on broader aspects of FCI ranging from melt quenching and coolability to energetic explosions (both in- and ex-vessel), and their relevance and applications to next-generation reactor design as well as accident management strategies. The objectives of the meeting are to review the knowledge and to obtain consensus on the phenomenology of FCI and in predicting FCI behavior in LWRs severe accidents; to identify those areas of FCI phenomena and prediction which are important for reactor safety but still poorly understood and require further study with clear methodologies; to inform the community and the regulatory agencies of the status of FCI issues, especially in the application to accident management and future reactor designs. The various sessions are: reactor applications, pre-mixing, propagation / trigger, experiments

  15. 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

  16. 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

  17. 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

  18. 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

  19. 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.)

  20. 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.

  1. Report of 5th new nuclear fuel research meeting, Yayoi Research Group. Trend of advanced basic research in nuclear fuel technical development

    International Nuclear Information System (INIS)

    1994-03-01

    Theme of this meeting is 'Trend of advanced basic research in nuclear fuel technical development', and it was attempted to balance both sides of the basic research and the development. At the meeting, lectures were given on the chemical form of FPs in oxide fuel pins, the absorption of hydrogen of fuel cladding tubes, the application of hydride fuel to thorium cycle, the thermal properties of fuel cladding tubes, the preparation of NpN and heat conductivity, the high temperature chemical reprocessing of nitride fuel, the research on the annihilation treatment of minor actinide in fast reactors, the separation of TRU by dry process and the annihilation using a metallic fuel FBR. In this report, the summaries of the lectures are collected, and also the program of the meeting and the list of attendants are shown. (K.I.)

  2. International Experts’ Meeting on Reactor and Spent Fuel Safety in the Light of the Accident at the Fukushima Daiichi Nuclear Power Plant. Presentations

    International Nuclear Information System (INIS)

    2012-01-01

    The primary objectives of this International Experts’ Meeting (IEM) were: to analyse relevant technical aspects of reactor and spent nuclear fuel management safety and performance related to severe accidents; to review what is known to date about the accident at the Fukushima Daiichi nuclear power plant in order to understand more fully its root causes; and to share the lessons learned from the accident. The meeting identified the necessary priorities for further actions in these areas in different power reactor types, focusing in particular on boiling water reactors (BWRs) and pressurized water reactors (PWRs). The meeting provided a forum for discussions and exchange of information among technical experts from Member States on reactor and spent nuclear fuel safety and performance under severe conditions. The meeting was of particular interest to technical experts from utilities, research and design organizations, regulatory bodies, manufacturing and service companies and other stakeholders. In particular, the objectives of the meeting was to: • Identify and analyse reactor and spent nuclear fuel safety and performance issues; • Consider the design, engineering and analysis of current and new systems for accident prevention and mitigation; • Exchange information on national assessments of reactor and spent nuclear fuel safety and performance; and • Identify potential priority areas for research and development, technology development and management

  3. 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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  5. Closed Nuclear Fuel Cycle Technologies to Meet Near-Term and Transition Period Requirements

    International Nuclear Information System (INIS)

    Collins, E.D.; Felker, L.K.; Benker, D.E.; Campbell, D.O.

    2008-01-01

    A scenario that very likely fits conditions in the U.S. nuclear power industry and can meet the goals of cost minimization, waste minimization, and provisions of engineered safeguards for proliferation resistance, including no separated plutonium, to close the fuel cycle with full actinide recycle is evaluated. Processing aged fuels, removed from the reactor for 30 years or more, can provide significant advantages in cost reduction and waste minimization. The UREX+3 separations process is being developed to separate used fuel components for reuse, thus minimizing waste generation and storage in geologic repositories. Near-term use of existing and new thermal spectrum reactors can be used initially for recycle actinide transmutation. A transition period will eventually occur, when economic conditions will allow commercial deployment of fast reactors; during this time, recycled plutonium can be diverted into fast reactor fuel and conversion of depleted uranium into additional fuel material can be considered. (authors)

  6. Closed Nuclear Fuel Cycle Technologies to Meet Near-Term and Transition Period Requirements

    Energy Technology Data Exchange (ETDEWEB)

    Collins, E.D.; Felker, L.K.; Benker, D.E.; Campbell, D.O. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, Tennessee, 37831-6152 (United States)

    2008-07-01

    A scenario that very likely fits conditions in the U.S. nuclear power industry and can meet the goals of cost minimization, waste minimization, and provisions of engineered safeguards for proliferation resistance, including no separated plutonium, to close the fuel cycle with full actinide recycle is evaluated. Processing aged fuels, removed from the reactor for 30 years or more, can provide significant advantages in cost reduction and waste minimization. The UREX+3 separations process is being developed to separate used fuel components for reuse, thus minimizing waste generation and storage in geologic repositories. Near-term use of existing and new thermal spectrum reactors can be used initially for recycle actinide transmutation. A transition period will eventually occur, when economic conditions will allow commercial deployment of fast reactors; during this time, recycled plutonium can be diverted into fast reactor fuel and conversion of depleted uranium into additional fuel material can be considered. (authors)

  7. Implications of using alternate fuel cycles to meet Ontario's nuclear power demand

    International Nuclear Information System (INIS)

    Lau, J.H.K.

    1978-08-01

    The use of alternate fuel cycles to meet an assumed nuclear capacity growth rate in Ontario is examined. Two criteria are used: the ability of the alternate fuel cycles to lessen the uranium demand; and the ease of commercialization. The nuclear strategies considered assume the use of the natural uranium cycle and, starting in the year 2000, the gradual introduction of an alternate fuel cycle. The alternate fuel cycles reviewed are enriched uranium, mixed oxides, and a variety of thorium cycles. The cumulative uranium requirement to the year 2070, and the growth and size of the reprocessing and fuel fabrication industries are discussed in detail. Sensitivity analyses on nuclear capacity growth rate, recycling loss and delay time are also described. (auth)

  8. Proceedings of the CSNI specialists meeting on fuel-coolant interactions

    Energy Technology Data Exchange (ETDEWEB)

    None

    1994-03-01

    A specialists meeting on fuel-coolant interactions was held in Santa Barbara, CA from January 5-7, 1993. The meeting was sponsored by the United States Nuclear Regulatory Commission in collaboration with the Committee on the Safety of Nuclear Installation (CSNI) of the OECD Nuclear Energy Agency (NEA) and the University of California at Santa Barbara. The objectives of the meeting are to cross-fertilize on-going work, provide opportunities for mutual check points, seek to focus the technical issues on matters of practical significance and re-evaluate both the objectives as well as path of future research. Individual papers have been cataloged separately.

  9. Proceedings of the CSNI specialists meeting on fuel-coolant interactions

    International Nuclear Information System (INIS)

    1994-03-01

    A specialists meeting on fuel-coolant interactions was held in Santa Barbara, CA from January 5--7, 1993. The meeting was sponsored by the United States Nuclear Regulatory Commission in collaboration with the Committee on the Safety of Nuclear Installation (CSNI) of the OECD Nuclear Energy Agency (NEA) and the University of California at Santa Barbara. The objectives of the meeting are to cross-fertilize on-going work, provide opportunities for mutual check points, seek to focus the technical issues on matters of practical significance and re-evaluate both the objectives as well as path of future research. Individual papers have been cataloged separately

  10. Advanced post-irradiation examination techniques for water reactor fuel. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    2002-03-01

    The purpose of the meeting was to provide and overview of the status of post-irradiation examination (PIE) techniques for water cooled reactor fuel assemblies and their components with emphasis given to advanced PIE techniques applied to high burnup fuel. Papers presented at the meeting described progress obtained in non-destructive (e.g. dimensional measurements, oxide layer thickness measurements, gamma scanning and tomography, neutron and X-ray radiography, etc.) and destructive PIE techniques (e.g. microstructural studies, elemental and isotopic analysis, measurement of physical and mechanical properties, etc.) used for investigation of water reactor fuel. Recent practice in high burnup fuel investigation revealed the importance of advanced PIE techniques, such as 3-D tomography, secondary ion mass spectrometry, laser flash, high resolution transmission and scanning electron microscopy, image analysis in microstructural studies, for understanding mechanisms of fuel behaviour under irradiation. Importance and needs for in-pile irradiation of samples and rodlets in instrumented rigs were also discussed. This TECDOC contains 20 individual papers presented at the meeting; each of the papers has been indexed separately

  11. Factors Contributing to SMEs Failure in Meeting Supplier Performance Standards

    Directory of Open Access Journals (Sweden)

    van Scheers Louise

    2016-01-01

    Full Text Available This study sought to determine the factors that contribute to the failure of small and medium enterprises (SMEs in meeting supplier performance standards. Suppliers are faced with the challenge of SMEs failing to meet supplier performance standards because they rely on these providers of products and services, mainly SMEs, for their operations and to ultimately achieve their mandate. The researchers were able to establish the main factors contributing to SME failure in meeting supplier performance standards. These are as follows: unclear specifications and terms of references from suppliers; insufficient information provided to SME suppliers to ensure understanding of requirements; insufficient feedback and support; no support of SME suppliers based on their Black Economic Empowerment (BEE status; limited or no use of business support programs by SME suppliers; and external factors such as access to finance, changes in the economy, and location.

  12. Feasibility of solid oxide fuel cell dynamic hydrogen coproduction to meet building demand

    Science.gov (United States)

    Shaffer, Brendan; Brouwer, Jacob

    2014-02-01

    A dynamic internal reforming-solid oxide fuel cell system model is developed and used to simulate the coproduction of electricity and hydrogen while meeting the measured dynamic load of a typical southern California commercial building. The simulated direct internal reforming-solid oxide fuel cell (DIR-SOFC) system is controlled to become an electrical load following device that well follows the measured building load data (3-s resolution). The feasibility of the DIR-SOFC system to meet the dynamic building demand while co-producing hydrogen is demonstrated. The resulting thermal responses of the system to the electrical load dynamics as well as those dynamics associated with the filling of a hydrogen collection tank are investigated. The DIR-SOFC system model also allows for resolution of the fuel cell species and temperature distributions during these dynamics since thermal gradients are a concern for DIR-SOFC.

  13. 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.

  14. 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

  15. 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.

  16. 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.

  17. 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

  18. 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.

  19. 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)

  20. 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

  1. 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.

  2. Recent developments in post-irradiation examination techniques for water reactor fuel. Proceedings of a technical committee meeting held in Cadarache, France, 17-21 October 1994

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    At the invitation of the Government of France, following a proposal of the International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) the IAEA convened a Technical Committee meeting from 14 to 21 October 1994 in Cadarache to discuss recent technical advances and improvements in the field of post-irradiation examination (PIE) of fuel used in nuclear power plants. Fifty participants representing 14 countries attended the meeting and 30 papers were presented and discussed during five technical sessions. Working Groups composed of the session chairmen and authors of papers prepared summaries of each session including conclusions and recommendations for future work. Refs, figs and tabs.

  3. Recent developments in post-irradiation examination techniques for water reactor fuel. Proceedings of a technical committee meeting held in Cadarache, France, 17-21 October 1994

    International Nuclear Information System (INIS)

    1995-09-01

    At the invitation of the Government of France, following a proposal of the International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) the IAEA convened a Technical Committee meeting from 14 to 21 October 1994 in Cadarache to discuss recent technical advances and improvements in the field of post-irradiation examination (PIE) of fuel used in nuclear power plants. Fifty participants representing 14 countries attended the meeting and 30 papers were presented and discussed during five technical sessions. Working Groups composed of the session chairmen and authors of papers prepared summaries of each session including conclusions and recommendations for future work. Refs, figs and tabs

  4. 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)

  5. 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)

  6. Spent fuel management: Current status and prospects 1997. Proceedings of a regular advisory group meeting

    International Nuclear Information System (INIS)

    1998-03-01

    Spent fuel management has always been one of the important stages in the nuclear fuel cycle and it is still most vital problems common to all countries with nuclear reactors. It begins with the discharge of spent fuel from a power or a research reactor and ends with its ultimate disposition. Two options exist - an open, once-through cycle with direct disposal of the spent fuel and a closed cycle with reprocessing of the spent fuel, recycling of reprocessed plutonium and uranium in new mixed oxide fuels and disposal of the radioactive waste. Continuous attention is being given by the IAEA to the collection, analysis and exchange of information on spent fuel management. Its role in this area is to provide a forum for exchanging information and to co-ordinate and to encourage closer co-operation among Member States in certain research and development activities that are of common interest. Spent fuel management is recognized as a high priority IAEA activity. The Regular Advisory Group on Spent Fuel Management was established in 1982. The objective of the Regular Advisory Group is to serve as a means of exchanging information on the current status and progress of national programmes on spent fuel management and to provide advice to the IAEA. The results of the last Regular Advisory Group meeting (9-12 September 1997) are reflected in this report. It gives an overview of the status of spent fuel management programmes in a number of countries, a description of the current status and prospects of activities in this field and recommendations of the participants

  7. 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.

  8. 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

  9. 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.

  10. Technical Meeting on Fast Reactors and Related Fuel Cycle Facilities with Improved Economic Characteristics. Working Material

    International Nuclear Information System (INIS)

    2013-01-01

    The objectives of the meeting were: - To identify the main issues and technical features that affect capital and energy production costs of fast reactors and related fuel cycle facilities; - To present fast reactor concepts and designs with enhanced economic characteristics, as well as innovative technical solutions (components, subsystems, etc.) that have the potential to reduce the capital costs of fast reactors and related fuel cycle facilities; - To present energy models and advanced tools for the cost assessment of innovative fast reactors and associated nuclear fuel cycles; - To discuss the results of studies and on-going R&D activities that address cost reduction and the future economic competitiveness of fast reactors; and - To identify research and technology development needs in the field, also in view of new IAEA initiatives to help and support Member States in improving the economic competitiveness of fast reactors and associated nuclear fuel cycles

  11. Technical Meeting on Fast Reactors and Related Fuel Cycle Facilities with Improved Economic Characteristics. Presentations

    International Nuclear Information System (INIS)

    2013-01-01

    The objectives of the meeting were: • To identify the main issues and technical features that affect capital and energy production costs of fast reactors and related fuel cycle facilities; • To present fast reactor concepts and designs with enhanced economic characteristics, as well as innovative technical solutions (components, subsystems, etc.) that have the potential to reduce the capital costs of fast reactors and related fuel cycle facilities; • To present energy models and advanced tools for the cost assessment of innovative fast reactors and associated nuclear fuel cycles; • To discuss the results of studies and ongoing R&D activities that address cost reduction and the future economic competitiveness of fast reactors; • To identify research and technology development needs in the field, also in view of new IAEA initiatives to help and support Member States in improving the economic competitiveness of fast reactors and associated nuclear fuel cycles

  12. 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)

  13. 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)

  14. 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

  15. 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.

  16. Specialists' meeting on theoretical modelling of LMFBR fuel pin behaviour. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1979-12-01

    The purpose of the meeting was to provide an opportunity for exchanging views of theoretical modelling of LMFBR fuel pin behaviour and to summarise the IWGFR member countries' knowledge in this field. The special emphasis was placed on normal operating conditions. The technical part of the meeting was divided into six sessions, as follows: An overview of fuel modelling studies; Key factors and basic phenomena relevant to fuel pin behaviour modelling; Application to steady state operation and normal transients; Experimental validation through pins in service and specific irradiation experiments; Advanced fuels; and Brief review of existing codes. During the meeting, papers were presented by the delegates on behalf of their countries or organization. The papers, which are included in this report, were either in the form of a general survey of the subject, or on specific technical subjects. In each subject area presentations appropriate to the subject were made from the submitted papers. The presentations were followed by discussions of the questions raised and summary is made.

  17. Specialists' meeting on theoretical modelling of LMFBR fuel pin behaviour. Summary report

    International Nuclear Information System (INIS)

    1979-12-01

    The purpose of the meeting was to provide an opportunity for exchanging views of theoretical modelling of LMFBR fuel pin behaviour and to summarise the IWGFR member countries' knowledge in this field. The special emphasis was placed on normal operating conditions. The technical part of the meeting was divided into six sessions, as follows: An overview of fuel modelling studies; Key factors and basic phenomena relevant to fuel pin behaviour modelling; Application to steady state operation and normal transients; Experimental validation through pins in service and specific irradiation experiments; Advanced fuels; and Brief review of existing codes. During the meeting, papers were presented by the delegates on behalf of their countries or organization. The papers, which are included in this report, were either in the form of a general survey of the subject, or on specific technical subjects. In each subject area presentations appropriate to the subject were made from the submitted papers. The presentations were followed by discussions of the questions raised and summary is made

  18. 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)

  19. 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.

  20. 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

  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. Proceedings of the Second OECD/NEA Organisation Meeting on Increased Accident Tolerance of Fuels for LWRs

    International Nuclear Information System (INIS)

    Massara, Simone; ); Bragg-Sitton, Shannon; Braase, Lori; Merrill, Brad; Teague, Melissa; Stanek, Chris R.; Montgomery, Robert H.; Ott, Larry J.; Robb, Kevin; Snead, Lance; Farmer, Mitch; Billone, Michael C.; Todosow, Michael; Brown, Nicholas; Brachet, J.C.; Le Flem, M.; Sauder, C.; Idarraga-Trujillo, I.; Michaux, A.; Lorrette, C.; Le Saux, M.; Blanpain, P.; Park, Jeong-Yong; Yang, Jae-Ho; Kim, Weon-Ju; Koo, Yang-Hyun; Liu, T.; Hallstadius, Lars; Lahoda, Ed; Waeckel, N.; Bonnet, J.M.; Vitanza, Carlo; Ohta, Hirokazu; Ogata, Takanari; Nakamura, Kinya; Dyck, Gary; Inozemtsev, Victor; )

    2013-01-01

    Under the guidance of the OECD-NEA Nuclear Science Committee, the expert group acts as a forum for scientific and technical information exchange on advanced light water reactor (LWR) fuels with enhanced accident tolerance. The expert group focusses on the fundamental properties and behaviour under normal operations and accident conditions for advanced core materials and components (fuels, cladding, control rods, etc.). The materials considered are applicable to Gen II and Gen III Light Water Reactors, as well as Gen III+ reactors under construction. The objective of the expert group is to define and coordinate a programme of work to help advance the scientific knowledge needed to provide the technical underpinning for the development of advanced LWR fuels with enhanced accident tolerance compared to currently used zircaloy/UO 2 fuel systems, as well as other non-fuel core components with important roles in LWR performance under accident conditions. This document brings together the available presentations (slides) given at the 2. Meeting on Increased Accident Tolerance of Fuels for LWRs. Content: 1 - Overview of the exchanges after the December-2012 Workshop through the discussion forum established at the OECD-NEA (S. Massara, NEA); 2 - Metrics Development for Enhanced Accident Tolerant LWR Fuels (S. Bragg-Sitton, INL); 3 - Candidate ATF Clad Technologies and Key Feasibility Issues (L. Snead, ORNL); 4 - CEA studies on nuclear fuel claddings for LWRs enhanced accident tolerant fuel. Some recent results, pending issues and prospects (J.C. Brachet, CEA); 5 - Current status on the accident tolerant fuel development in the Republic of Korea (J.Y. Park, J.H. Chang, KAERI); 6 - The current status of fuel R and D in the P.R. of China (T. Liu, CGN). Session 2: Key elements for a work programme on ATF: 7 - Beneficial characteristics of ATF (metrics) (L. Hallstadius, Westinghouse); 8 - Reactor types of interest (applicability) (L. Ott, ORNL); 9 - Impact on normal operations

  4. 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

  5. Advanced fuels with reduced actinide generation. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    1996-11-01

    Nuclear energy can play an important future role in supplying the world population with energy. However, this form of energy will be successful only under certain conditions: it must meet very strict safety requirements, it must be economically competitive, and it must be acceptable to the public. Nuclear power produces radioactive wastes and in several countries the public raises concern about safety. Much development work on advanced nuclear power systems is going on in several countries, with participation of both governmental and private industries to meet these conditions. In the framework of this IAEA activity the Technical Committee Meeting on Advanced Fuels with Reduced Actinide Generation was organized. The aim of the meeting was to highlight current research activities and to identify new research areas and fields of possible co-operation. The scope of the meeting included advanced fuels for all types of nuclear reactors: light water reactors, heavy water reactors, high temperature reactors, fast reactors, molten salt reactors and for accelerator driven systems. Other topics covered a wide range of investigations made, or to be made in the Member States. Refs, figs, tabs

  6. 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

  7. 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.

  8. 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

  9. 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)

  10. 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.

  11. Proceedings of the Topical Meeting on the safety of nuclear fuel cycle intermediate storage facilities

    International Nuclear Information System (INIS)

    1998-01-01

    The CSNI Working Group on Fuel Cycle Safety held an International Topical Meeting on safety aspects of Intermediate Storage Facilities in Newby Bridge, England, from 28 to 30 October 1997. The main purpose of the meeting was to provide a forum for the exchange of information on the technical issues on the safety of nuclear fuel cycle facilities (intermediate storage). Titles of the papers are: An international view on the safety challenges to interim storage of spent fuel. Interim storage of intermediate and high-level waste in Belgium: a description and safety aspects. Encapsulated intermediate level waste product stores at Sellafield. Safety of interim storage facilities of spent fuel: the international dimension and the IAEA's activities. Reprocessing of irradiated fuel and radwaste conditioning at Belgoprocess site: an overview. Retrieval of wastes from interim storage silos at Sellafield. Outline of the fire and explosion of the bituminization facility and the activities of the investigation committee (STAIJAERI). The fire and explosion incident of the bituminization facility and the lessons learned from the incident. Study on the scenario of the fire incident and related analysis. Study on the scenario of the explosion incident and related analysis. Accident investigation board report on the May 14, 1997 chemical explosion at the plutonium reclamation facility, Hanford site, Richland, Washington. Dry interim storage of spent nuclear fuel elements in Germany. Safe and effective system for the bulk receipt and storage of light water reactor fuel prior to reprocessing. Receiving and storage of glass canisters at vitrified waste storage center of Japan Nuclear Fuel Ltd. Design and operational experience of dry cask storage systems. Sellafield MOX plant; Plant safety design (BNFL). The assessment of fault studies for intermediate term waste storage facilities within the UK nuclear regulatory regime. Non-active and active commissioning of the thermal oxide

  12. 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

  13. 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

  14. 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

  15. 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)

  16. 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

  17. 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.

  18. 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)

  19. 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)

  20. CRP on Demonstrating Performance of Spent Fuel and Related Storage Systems beyond the Long Term

    International Nuclear Information System (INIS)

    Bevilacqua, Arturo

    2014-01-01

    At the initial Coordinated Research Project (CRP) planning meeting held in August 2011, international experts in spent fuel performance confirmed the value of further coordination and development of international efforts to demonstrate the performance of spent fuel and related storage system components as durations extend. Furthermore, in recognition that the Extended Storage Collaboration Program (ESCP) managed by the Electric Power Research Institute (EPRI) in the USA, from now on ESCP, provided a broad context for the research and development work to be performed in the frame of this CRP, it was agreed that its objectives should target specific ESCP needs in order to make a relevant contribution. Accordingly, the experts examined on-going gap analyses - gaps between anticipated technical needs and existing technical data - for identify the specific research objectives. Additionally, during the planning meeting it was pointed out the need to coordinate and cooperate with the OECD/NEA counterparts involved in the organization of the International Workshop planned in autumn 2013 and with the on-going third phase of the CRP on Spent Fuel Performance Assessment and Research (SPAR-III). Given the importance to assess the performance of spent fuel and related important storage system components in order to confirm the viability of very long term storage for supporting the need to extend or renew licenses for storage facilities the CRP was approved by the IAEA in November 2011. While a full range of spent fuel types and storage conditions are deployed around the world, this CRP is focused on existing systems and, more specifically, water reactor fuel in dry storage with the overall research objective to support the technical basis for water reactor spent fuel management as dry storage durations extend. In March 2012 the group of international experts who participated at the initial CRP planning meeting in August 2011 evaluated and recommended for approval 9 research

  1. 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.

  2. 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)

  3. 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.

  4. 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

  5. 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

  6. Report of the IAEA advisory group meeting on LMFBR fuel reprocessing

    International Nuclear Information System (INIS)

    1976-05-01

    A summary of the papers and discussions of the meeting is presented, reviewing the status of development in LMFBR fuel reprocessing and focusing attention on important problem areas. The following topics are discussed: Transport, storage and removal of sodium; decladding and shearing; dissolution; Purex process; fluoride volatility method; off-gas purification; waste disposal. Status reports of national programmes of Belgium, France, Federal Republic of Germany, Italy, Japan, United Kingdom, USSR and USA are included

  7. 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

  8. Performance Evaluation of the Neutron Coincidence Counter for the Advanced Spent Fuel Conditioning Process

    International Nuclear Information System (INIS)

    Lee, S.Y.; Li, T.K.; Menlove, Howard O.; Kim, H.D.; Ko, W.I.; Park, S.W.

    2005-01-01

    The Advanced Spent Fuel Conditioning Process (ACP) is a pyrochemical dry reprocessing technique to convert oxide-type spent nuclear fuel into a metallic form. The Korea Atomic Energy Research Institute (KAERI) has been developing this technology for the purpose of spent fuel management and is planning to perform a lab-scale demonstration in 2006. With this technology, a significant reduction of the volume and heat load of spent fuel is expected, which could decrease the burden of safety and economics. In this study, MCNPX code calculations were carried out to estimate the performance of a neutron coincidence counter designed for measruement of the process materials in the pilot-scale ACP facility. To verify the design requirement, the singles and doubles counting rates of the detectors were simulated with the latest coincidence capability of the MCNPX code. Then, the precision of the coincidence measurements were evaluated on various process materials from the ACP. It was verified that the performance of the neutron coincidence counter could meet the design criteria for all samples in the ACP, and the material accounting system for the pilot-scale ACP facility could meet the IAEA safeguards goals.

  9. 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)

  10. Fuel taxes, motor vehicle emission standards and patents related to the fuel-efficiency and emissions of motor vehicles. Joint Meetings of Tax and Environment Experts

    International Nuclear Information System (INIS)

    Vollebergh, H.

    2010-01-01

    Contribution to the project on Taxation, Innovation and the Environment of OECD's Joint Meetings of Tax and Environment Experts. It studies the impacts of motor vehicle fuel taxes and mandatory fuel efficiency standards on relevant car-related innovation activity in selected car-producing countries.

  11. 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

  12. 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

  13. 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)

  14. 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)

  15. 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)

  16. Effects of New Fossil Fuel Developments on the Possibilities of Meeting 2C Scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Meindertsma, W.; Blok, K.

    2012-12-15

    Recent years have seen an increasing activity in developing new fossil fuel production capacity. This includes unconventional fossil fuels, such as tar sands and shale gas, fossil fuels from remote locations, and fossil fuels with a very large increase in production in the near future. In this report, the impact of such developments on our ability to mitigate climate change is investigated. Our inventory shows that the new fossil fuel developments currently underway consist of 29,400 billion cubic meters of natural gas, 260,000 million barrels of oil and 49,600 million tonnes of coal. The development of these new fossil fuels would result in emissions of 300 billion tonnes of CO2 -equivalent (CO2e) from 2012 until 2050. Until 2050, a 'carbon budget' of 1550 billion tonnes CO2e is still available if we want to of keep global warming below 2C with a 50% probability. For a 75% probability to stay below 2C this budget is only 1050 billion tonnes CO2e. So, the new fossil fuel developments identified in this report consume 20-33% of the remaining carbon budget until 2050. In a scenario where the new fossil fuels are developed, we need to embark on a rapid emission reductions pathway at the latest in 2019 in order to meet the 50% probability carbon budget. Avoiding the development of new fossil fuels will give us until 2025 to start further rapid emission reductions. These calculations are based on the assumption that the maximum emission reduction rate is 4% per year and that the maximum change in emission trend is 0.5 percentage point per year. The starting year for rapid emission reductions depends on the choice of these parameters. A sensitivity analysis shows that, in all cases, refraining from new fossil fuel development allows for a delay of 5 to 8 years before we should embark on a rapid emission reduction pathway. The high investments required for developing new fossil fuels lead to a lock in effect; once developed, these fossil fuels need to be

  17. 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.

  18. 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

  19. 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.

  20. 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.

  1. 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

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

    Science.gov (United States)

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

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

  3. 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

  4. 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)

  5. 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

  6. 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)

  7. 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

  8. 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.

  9. Diesel fuel component contributions to engine emissions and performance: Clean fuel study

    Energy Technology Data Exchange (ETDEWEB)

    Erwin, J.; Ryan, T.W. III; Moulten, D.S. [Southwest Research Inst., San Antonio, TX (United States)

    1994-08-01

    The emissions characteristics of diesel engines are dominated by current engine design parameters as long as the fuels conform to the current industry-accepted specifications. The current and future emissions standard, are low enough that the fuel properties and compositions are starting to play a more significant role in meeting the emerging standards. The potential role of the fuel composition has been recognized by state and federal government agencies, and for the first time, fuel specifications have become part of the emissions control legislation. In this work, five different fuel feed and blend stocks were hydrotreated to two levels of sulfur and aromatic content. These materials were then each distilled to seven or eight fractions of congruent boiling points. After this, the raw materials and all of the fractions were characterized by a complement of tests from American Society for Testing and Materials and by hydrocarbon-type analyses. The sample matrix was subjected to a series of combustion bomb and engine tests to determine the ignition, combustion, and emissions characteristics of each of the 80 test materials.

  10. Technical Meeting on Fast Reactors and Related Fuel Cycle Facilities with Improved Economic Characteristics. Working Material

    International Nuclear Information System (INIS)

    2013-01-01

    In recent years, engineering oriented work, rather than basic research and development (R&D), has led to significant progress in improving the economics of innovative fast reactors and associated fuel cycle facilities, while maintaining and even enhancing the safety features of these systems. Optimization of plant size and layout, more compact designs, reduction of the amount of plant materials and the building volumes, higher operating temperatures to attain higher generating efficiencies, improvement of load factor, extended core lifetimes, high fuel burnup, etc. are good examples of achievements to date that have improved the economics of fast neutron systems. The IAEA, through its Technical Working Group on Fast Reactors (TWG-FR) and Technical Working Group on Nuclear Fuel Cycle Options and Spent Fuel Management (TWG-NFCO), devotes many of its initiatives to encouraging technical cooperation and promoting common research and technology development projects among Member States with fast reactor and advanced fuel cycle development programmes, with the general aim of catalysing and accelerating technology advances in these fields. In particular the theme of fast reactor deployment, scenarios and economics has been largely debated during the recent IAEA International Conference on Fast Reactors and Related Fuel Cycles: Safe Technologies and Sustainable Scenarios, held in Paris in March 2013. Several papers presented at this conference discussed the economics of fast reactors from different national and regional perspectives, including business cases, investment scenarios, funding mechanisms and design options that offer significant capital and energy production cost reductions. This Technical Meeting on Fast Reactors and Related Fuel Cycle Facilities with Improved Economic Characteristics addresses Member States’ expressed need for information exchange in the field, with the aim of identifying the main open issues and launching possible initiatives to help and

  11. 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)

  12. Proceedings of the OECD/CSNI specialists meeting on fuel-coolant interactions

    Energy Technology Data Exchange (ETDEWEB)

    Akiyama, Mamoru; Yamano, Norihiro; Sugimoto, Jun [eds.

    1998-01-01

    The OECD/CSNI Specialists Meeting on Fuel Coolant Interactions (FCI) was held at Tokai-mura in Japan on May 19 through 21, 1997, and attended by 80 participants from 14 countries and one international organizations. In the meeting 36 papers were presented followed by active discussions in six sessions on various aspects of FCI issues, such as reactor application, premixing, propagation/trigger, experiments and code/models. At the end of the Meeting, the participants have reached to the consensus on the summary and recommendations, which consists of the following items; (1) We find no new evidence that would change or violate the conclusion of SERG-2 (1996) that alpha-mode failure is not risk significant. (2) Significant progress has been made since the Santa Barbara meeting (1993). (3) Several areas have been identified, which need further investigations to understand the basic FCI phenomena, and to improve the modeling. (4) We recommend maximizing open communication between various research groups in order to accelerate the resolution of the remaining issues. (5) We recommend that the next specialist meeting be held within 3 to 5 years in order to synthesize the activities described above. (J.P.N.)

  13. 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)

  14. Can Hawaii Meet Its Renewable Fuel Target? Case Study of Banagrass-Based Cellulosic Ethanol

    Directory of Open Access Journals (Sweden)

    Chinh Tran

    2016-08-01

    Full Text Available Banagrass is a biomass crop candidate for ethanol production in the State of Hawaii. This study examines: (i whether enough banagrass can be produced to meet Hawaii’s renewable fuel target of 20% highway fuel demand produced with renewable sources by 2020 and (ii at what cost. This study proposes to locate suitable land areas for banagrass production and ethanol processing, focusing on the two largest islands in the state of Hawaii—Hawaii and Maui. The results suggest that the 20% target is not achievable by using all suitable land resources for banagrass production on both Hawaii and Maui. A total of about 74,224,160 gallons, accounting for 16.04% of the state’s highway fuel demand, can be potentially produced at a cost of $6.28/gallon. Lower ethanol cost is found when using a smaller production scale. The lowest cost of $3.31/gallon is found at a production processing capacity of about 9 million gallons per year (MGY, which meets about 2% of state demand. This cost is still higher than the average imported ethanol price of $3/gallon. Sensitivity analysis finds that it is possible to produce banagrass-based ethanol on Hawaii Island at a cost below the average imported ethanol price if banagrass yield increases of at least 35.56%.

  15. 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)

  16. 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)

  17. 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

  18. 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

  19. Fuel leak testing performance at NPP Jaslovske Bohunice

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  20. Fuel leak testing performance at NPP Jaslovske Bohunice

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-12-31

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

  1. 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)

  2. 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.

  3. 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

  4. 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

  5. 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.

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

    Science.gov (United States)

    Kannaiyan, Kumaran; Sadr, Reza

    2015-11-01

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

  7. 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)

  8. 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.)

  9. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-11-01

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

  11. 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

  12. NTL 11 spent fuel flask - meeting the challenge of regulatory and technological change

    International Nuclear Information System (INIS)

    Cory, A.R.

    2004-01-01

    By June 2005, when shipments of spent fuel for reprocessing from Germany are concluded, the NTL11 flask type will have been responsible for transporting a total of 1500 tonnes of heavy metal in the form of spent fuel. Excluding domestic transports in France and the UK, this represents 25% of the total European spent fuel transported for reprocessing since the flasks came into service in 1977. Approximately 40% of the total for the flask type will have been transported to BNFL's Sellafield facility, the remainder to Cogema at La Hague. The NTL11 flask can justifiably be described as being the workhorse of BNFL's European spent fuel transport business. The NTL11 flask started life under the ownership of Nuclear Transport Limited, an associate company of BNFL, and in recent years the original fleet of five flasks has been absorbed into the BNFL inventory. A recent build programme has seen a further four flasks added to the fleet, an expedient measure to cope with the additional transport requirements imposed by the need to meet the June 2005 deadline for the removal of contracted fuels from Germany. While there have been certain evolutionary changes affecting the package design, there have also been more significant changes in the Design Safety Case. These have sometimes been necessary to meet regulatory changes, or the challenges posed by the regulators. In other cases advantage has been taken of improvements in analytical techniques to demonstrate increased margins of operational safety. Where possible those margins have also been increased by other means, such as taking advantage of commercial trends to reduce package thermal loads. The NTL11 flask was designed around the reactor and fuel characteristics prevailing in the 1970's. Over the lifetime of the flask the responsible engineering teams have faced and met the successive challenges to develop the capability of the Package to face the changing requirements of the industry and the Transport Regulations. Both

  13. 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.

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

    International Nuclear Information System (INIS)

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

    1995-07-01

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

  15. Proceedings of the Third Meeting of the OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs, 3-5 March 2015, OECD-NEA HQ

    International Nuclear Information System (INIS)

    Bischoff, Jeremy; Gandrille, Pascal; Forgeron, Thierry; Brachet, Jean-Christophe; Lorrette, Christophe; Valot, C.; Freyss, M.; Braun, J.; Sauder, C.; Moatti, Marie; Waeckel, Nicolas; Ambard, Antoine; Pasamehmetoglu, Kemal; Johnston, Emma; Bragg-Sitton, Shannon M.; Kurata, M.; Hallstadius, Lars; Ohta, H.; Ogata, T.; Besmann, T.; Chauvin, Nathalie; Cornet, Stephanie; Massara, S.; Kohyama, Akira; Kishimoto, Hirotatsu; Park, Joon Soo; Nakazato, Naofumi; Hayasaka, Daisuke; Asakura, Yuuki; Kanda, Chisato; Kohyama, Akira; Terrani, Kurt; Katoh, Yutai; Yamamoto, Yukinori; Field, Kevin; Snead, Lance; Hu, Xunxiang; Dryepondt, Sebastien; Unocic, Kinga A.; Hoelzer, David T.; Pint, Bruce A.; Besmann, T.; Steinbrueck, M.; Grosse, M.; Jianu, A.; Weisenburger, A.; Avincola, V.; Ahmad, S.; Tang, C.; Heuser, Brent J.; Sickafus, Kurt; Kim, Hyun-Gil; Kim, Il-Hyun; Jung, Yang-Il; Park, Dong-Jun; Park, Jung-Hwan; Park, Jeong-Yong; Koo, Yang-Hyun; Lee, B.O.; Van Nieuwenhove, Rudi; Kim, Young; Rebak, Raul; Dolly, Evan; Dolley, E.J.; Rebak, R.B.; Maloy, Stu; Yang, Jae-Ho; Kim, Dong-Joo; Kim, Keon-Sik; Koo, Yang-Hyun; Lee, Won Jae; Tulenko, James S.; Puide, Mattias; Liu, T.; Gueneau, C.; Gosse, S.; Dupin, N.; Barber, D.; Corcoran, E.; Dumas, J.C.; Hania, R.; Kaye, M.; Turchi, P.

    2015-03-01

    Under the guidance of the OECD-NEA Nuclear Science Committee, the expert group acts as a forum for scientific and technical information exchange on advanced light water reactor (LWR) fuels with enhanced accident tolerance. The expert group focusses on the fundamental properties and behaviour under normal operations and accident conditions for advanced core materials and components (fuels, cladding, control rods, etc.). The materials considered are applicable to Gen II and Gen III Light Water Reactors, as well as Gen III+ reactors under construction. The objective of the expert group is to define and coordinate a programme of work to help advance the scientific knowledge needed to provide the technical underpinning for the development of advanced LWR fuels with enhanced accident tolerance compared to currently used zircaloy/UO 2 fuel systems, as well as other non-fuel core components with important roles in LWR performance under accident conditions. This document brings together the available presentations (slides) given at the Third Meeting of the OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs. Content: 1 - Task Force 1 (Systems assessment) meeting, 3-4 March 2015: - French evaluation of ATF Concepts (J. Bischoff, AREVA); - Technology Readiness Levels - TRL - for Fuels (K. Pasamehmetoglu, INL); - TRL-definition for advanced fuel concept applied for commercial LWRs in Japan (M. Kurata, JAEA); - Application of TRLs in NNL (E. Johnston, NNL); - Technology Readiness Levels for Advanced Nuclear Fuel and Materials (S. Bragg-Sitton, INL); 1a - Definition of the illustrative scenarios: - AREVA's proposal concerning scenario for Accident Tolerant Fuel studies (P. Gandrille); - A Simplified Accident Scenario (L. Hallstadius); - Accident Scenarios for ATF Performance Evaluation of BWR and PWR in Japan (H. Ohta, CRIEPI); 1b - Related NEA activities: - Working Party on Multi-scale Modelling of Fuels and Structural Materials for Nuclear Systems - WPMM, Expert

  16. 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

  17. 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

  18. 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.

  19. 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.

  20. 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...

  1. 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.

  2. Meeting performance goals by the use of experience data

    International Nuclear Information System (INIS)

    Salmon, M.W.; Kennedy, R.P.

    1993-01-01

    DOE Order 5480.28 requires that structures, systems and components (SSCs) be designed and constructed to withstand the effects of natural phenomena hazards. For SSCs to be acceptable, it must be demonstrated that there is a sufficiently low probability of failure of those SSCs consistent with established performance goals. For new design, NPH loads are taken from probabilistic hazard assessments and coupled with response and evaluation methods to control the levels of conservatism required to achieve performance goals. For components qualified by test, performance goals are achieved by specifying a test response spectrum that envelops a required response spectrum coupled with minimal acceptance standards. DOE Standard 1020-92 adapts both of these approaches to ensure that the required performance goals are met for new installations. For existing installations these approaches are generally not applicable. There is a need for a simple approach for use in verifying the performance of existing equipment subject to seismic hazards. The USNRC has adapted such an approach for the resolution of USI A-46 in the Generic Implementation Procedure (GIP). A simple set of screening rules, keyed to a generic bounding spectrum forms the basis of the USNRC approach. A similar approach is being adapted for use in the DOE. The DOE approach, however, must also ensure that appropriate performance goals are met when the general screens are met. This paper summarizes research to date on the topic of meeting performance goals by the use of experience data. The paper presents a review of the background material, a summary of the requirements for existing components, a summary of the approach used in establishing the performance goals associated with experience data approaches, and a summary of results to date. Simplified criteria are proposed

  3. 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)

  4. 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.

  5. 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.)

  6. 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)

  7. 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

  8. 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.

  9. 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

  10. 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.

  11. WFIRST: Managing Telescope Wavefront Stability to Meet Coronagraph Performance

    Science.gov (United States)

    Noecker, Martin; Poberezhskiy, Ilya; Kern, Brian; Krist, John; WFIRST System Engineering Team

    2018-01-01

    The WFIRST coronagraph instrument (CGI) needs a stable telescope and active wavefront control to perform coronagraph science with an expected sensitivity of 8x10-9 in the exoplanet-star flux ratio (SNR=10) at 200 milliarcseconds angular separation. With its subnanometer requirements on the stability of its input wavefront error (WFE), the CGI employs a combination of pointing and wavefront control loops and thermo-mechanical stability to meet budget allocations for beam-walk and low-order WFE, which enable stable starlight speckles on the science detector that can be removed by image subtraction. We describe the control strategy and the budget framework for estimating and budgeting the elements of wavefront stability, and the modeling strategy to evaluate it.

  12. 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.

  13. Performance of Fragema fuel in pressurized water reactors

    International Nuclear Information System (INIS)

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

    1986-06-01

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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  15. 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

  16. 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

  17. 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

  18. Procedures and techniques for the management of experimental fuels from research and test reactors. Proceedings of an advisory group meeting

    International Nuclear Information System (INIS)

    1999-04-01

    Almost all countries that have undertaken fuel development programs for power, research or military reactors have experimental and exotic fuels, either stored at the original research reactors where they have been tested or at some away-from-reactor storage facility. These spent fuel liabilities cannot follow the standard treatment recognized for modern power reactor fuels. They include experimental and exotic fuels ranging from liquids to coated spheres and in configurations ranging from full test assemblies to post irradiation examination specimens set in resin. This document contains an overview of the extent of the problem of managing experimental and exotic fuels from research and test reactors and an expert evaluation of the overall situation in countries which participated in the meeting

  19. Control assembly materials for water reactors: Experience, performance and perspectives. Proceedings of a technical committee meeting

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-02-01

    The safe, reliable and economic operation of water cooled nuclear power reactors depends to a large extent upon the reliable operation of control assemblies for the regulation and shutdown of the reactors. These consist of neutron absorbing materials clad in stainless steel or zirconium based alloys, guide tubes and guide cards, and other structural components. Current designs have worked extremely well in normal conditions, but less than ideal behaviour limits the lifetimes of control materials, imposing an economic penalty which acts as a strong incentive to produce improved materials and designs that are more reliable. Neutron absorbing materials currently in use include the ceramic boron carbide, the high melting point metal hafnium and the low melting point complex alloy Ag-In-Cd. Other promising neutron absorbing materials, such as dysprosium titanate, are being evaluated in the Russian Federation. These control materials exhibit widely differing mechanical, physical and chemical properties, which must be understood in order to be able to predict the behaviour of control rod assemblies. Identification of existing failure mechanisms, end of life criteria and the implications of the gradual introduction of extended burnup, mixed oxide (MOX) fuels and more complex fuel cycles constitutes the first step in a search for improved materials and designs. In the early part of this decade, it was recognized by the International Working Group on Fuel Performance and Technology (IWGFPT) that international conferences, symposia and published reviews on the materials science aspects of control assemblies were few and far between. Consequently, the IWGFPT recommended that the IAEA should rectify this situation with a series of Technical Committee meetings (TCMs) devoted entirely to the materials aspects of reactor control assemblies. The first was held in 1993 and in the intervening five years considerable progress has been made. In bringing together experts in the

  20. Control assembly materials for water reactors: Experience, performance and perspectives. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    2000-02-01

    The safe, reliable and economic operation of water cooled nuclear power reactors depends to a large extent upon the reliable operation of control assemblies for the regulation and shutdown of the reactors. These consist of neutron absorbing materials clad in stainless steel or zirconium based alloys, guide tubes and guide cards, and other structural components. Current designs have worked extremely well in normal conditions, but less than ideal behaviour limits the lifetimes of control materials, imposing an economic penalty which acts as a strong incentive to produce improved materials and designs that are more reliable. Neutron absorbing materials currently in use include the ceramic boron carbide, the high melting point metal hafnium and the low melting point complex alloy Ag-In-Cd. Other promising neutron absorbing materials, such as dysprosium titanate, are being evaluated in the Russian Federation. These control materials exhibit widely differing mechanical, physical and chemical properties, which must be understood in order to be able to predict the behaviour of control rod assemblies. Identification of existing failure mechanisms, end of life criteria and the implications of the gradual introduction of extended burnup, mixed oxide (MOX) fuels and more complex fuel cycles constitutes the first step in a search for improved materials and designs. In the early part of this decade, it was recognized by the International Working Group on Fuel Performance and Technology (IWGFPT) that international conferences, symposia and published reviews on the materials science aspects of control assemblies were few and far between. Consequently, the IWGFPT recommended that the IAEA should rectify this situation with a series of Technical Committee meetings (TCMs) devoted entirely to the materials aspects of reactor control assemblies. The first was held in 1993 and in the intervening five years considerable progress has been made. In bringing together experts in the

  1. Effect of Variable Compression Ratio on Performance of a Diesel Engine Fueled with Karanja Biodiesel and its Blends

    Science.gov (United States)

    Mishra, Rahul Kumar; soota, Tarun, Dr.; singh, Ranjeet

    2017-08-01

    Rapid exploration and lavish consumption of underground petroleum resources have led to the scarcity of underground fossil fuels moreover the toxic emissions from such fuels are pernicious which have increased the health hazards around the world. So the aim was to find an alternative fuel which would meet the requirements of petroleum or fossil fuels. Biodiesel is a clean, renewable and bio-degradable fuel having several advantages, one of the most important of which is being its eco-friendly and better knocking characteristics than diesel fuel. In this work the performance of Karanja oil was analyzed on a four stroke, single cylinder, water cooled, variable compression ratio diesel engine. The fuel used was 5% - 25% karanja oil methyl ester by volume in diesel. The results such obtained are compared with standard diesel fuel. Several properties i.e. Brake Thermal Efficiency, Brake Specific Fuel Consumptions, Exhaust Gas Temperature are determined at all operating conditions & at variable compression ratio 17 and 17.5.

  2. Study of Gasohol as Alternative Fuel for Gasoline Substitution: Characteristics and Performances

    Directory of Open Access Journals (Sweden)

    Bardi Murachman

    2014-12-01

    Full Text Available Gasohol is a mixture of premium petrol (gasoline with alcohol, in this case ethanol. The use of gasohol can reduce fuel consumption without having to modify the existing engine. Therefore, this research is conducted to study the characteristics and performance of gasohol in various mixing ratios, which includes analysis of physical properties and the use of gasohol in the machine. Results show that the addition of technical ethanol at 7.0169%v increases the value of gasohol vapor pressure on the value of 8.6682 psi (7.7 psi for regular gasoline. Gasohol with technical ethanol content above 30%v decreases vapor pressure, promotes phase separation, and causes a sharp drop in temperature from 40%v distillation. In term of corrosivity, gasohol with up to 50%v ethanol content has the same corrosion level with regular gasoline, which is corrosion level 1A. Based on gasohol characteristics test, it is known that gasohol with technical ethanol content below 20%v can be used as a fuel substitute for gasoline. Real-time performance test of gasohol in engines has shown that the addition of ethanol content in gasohol tend to increase the engine power at a certain compression ratio, but it also increases fuel consumption because the heat value of ethanol is lower than gasoline. Machine in gasohol with ethanol content below 20%v can operate smoothly without having to modify the engine. Based on the studies that have been done, gasohol in the range of 10%v ethanol content is well-functioned as a substitute for gasoline fuel and meets fuel specifications required by the General Director of Oil and Gas. The feasibility of using gasohol as an alternative fuel can be studied further.

  3. 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)

  4. World Biofuels Production Potential Understanding the Challenges to Meeting the U.S. Renewable Fuel Standard

    Energy Technology Data Exchange (ETDEWEB)

    Sastri, B.; Lee, A.

    2008-09-15

    . Within the mandate, amounts of advanced biofuels, including biomass-based diesel and cellulosic biofuels, are required beginning in 2009. Imported renewable fuels are also eligible for the RFS. Another key U.S. policy is the $1.01 per gal tax credit for producers of cellulosic biofuels enacted as part of the 2008 Farm Bill. This credit, along with the DOE's research, development and demonstration (RD&D) programs, are assumed to enable the rapid expansion of U.S. and global cellulosic biofuels production needed for the U.S. to approach the 2022 RFS goal. While the Environmental Protection Agency (EPA) has yet to issue RFS rules to determine which fuels would meet the greenhouse gas (GHG) reduction and land use restrictions specified in EISA, we assume that cellulosic ethanol, biomass-to-liquid fuels (BTL), sugar-derived ethanol, and fatty acid methyl ester biodiesel would all meet the EISA advanced biofuel requirements. We also assume that enough U.S. corn ethanol would meet EISA's biofuel requirements or otherwise be grandfathered under EISA to reach 15 B gal per year.

  5. 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

  6. 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

  7. Advances in applications of burnup credit to enhance spent fuel transportation, storage, reprocessing and disposition. Proceedings of a technical meeting

    International Nuclear Information System (INIS)

    2007-05-01

    Given a trend towards higher burnup power reactor fuel, the IAEA began an active programme in burnup credit (BUC) with major meetings in 1997 (IAEA-TECDOC-1013), 2000 (IAEA-TECDOC-1241) and 2002 (IAEA-TECDOC-1378) exploring worldwide interest in using BUC in spent fuel management systems. This publication contains the proceedings of the IAEA's 4th major BUC meeting, held in London. Sixty participants from 18 countries addressed calculation methodology, validation and criticality, safety criteria, procedural compliance with safety criteria, benefits of BUC applications, and regulatory aspects in BUC. This meeting encouraged the IAEA to continue its activities on burnup credit including dissemination of related information, given the number of Member States having to deal with increased spent fuel quantities and extended durations. A 5th major meeting on burnup credit is planned 2008. Burnup credit is a concept that takes credit for the reduced reactivity of fuel discharged from the reactor to improve loading density of irradiated fuel assemblies in storage, transportation, and disposal applications, relative to the assumption of fresh fuel nuclide inventories in loading calculations. This report has described a general four phase approach to be considered in burnup credit implementation. Much if not all of the background research and data acquisition necessary for successful burnup credit development in preparation for licensing has been completed. Many fuel types, facilities, and analysis methods are encompassed in the public knowledge base, such that in many cases this guidance will provide a means for rapid development of a burnup credit program. For newer assembly designs, higher enrichment fuels, and more extensive nuclide credit, additional research and development may be necessary, but even this work can build on the foundation that has been established to date. Those, it is hoped that this report will serve as a starting point with sufficient reference to

  8. Technologies for gas cooled reactor decommissioning, fuel storage and waste disposal. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    1998-09-01

    Gas cooled reactors (GCRs) and other graphite moderated reactors have been important part of the world's nuclear programme for the past four decades. The wide diversity in status of this very wide spectrum of plants from initial design to decommissioning was a major consideration of the International Working group on Gas Cooled Reactors which recommended IAEA to convene a Technical Committee Meeting dealing with GCR decommissioning, including spent fuel storage and radiological waste disposal. This Proceedings includes papers 25 papers presented at the Meeting in three sessions entitled: Status of Plant Decommissioning Programmes; Fuels Storage Status and Programmes; waste Disposal and decontamination Practices. Each paper is described here by a separate abstract

  9. 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.

  10. 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

  11. 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

  12. 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

  13. 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%

  14. 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.

  15. Fossil fuels, renewable and nuclear options to meet the energy and the environmental challenges

    International Nuclear Information System (INIS)

    Bacher, P.; Moncomble, J.E.

    1995-01-01

    In order to meet the world strongly growing energy demand, and especially electricity demand, there are a number of primary energy sources: hydro and new renewable, oil, natural gas, coal and nuclear. The energy mix adopted in each country will depend on a number of factors, such as geography, security of supply, financing capacity, environment, etc. Shares of the different fuels in electricity output are reviewed. Nuclear energy facts and issues are discussed from safety, environment and economics points of view, with a particular view on long-lived wastes that can be and are strictly controlled; properly managed, a nuclear program can be very cost efficient as demonstrated in France, Belgium or Canada, and it has many advantages: site selection, security of supply, no air pollution. 3 refs., 5 figs

  16. 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)

  17. 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.

  18. 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.

  19. 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

  20. Use of fuel cells to meet military requirements for mobile power

    International Nuclear Information System (INIS)

    Andrukaitis, E.

    2004-01-01

    'Full text:' The use of fuel cell technology in military applications will depend on safe, high energy density systems being developed. An important part of using this technology is also the development of alternative hydrogen producing fuels with high energy densities and are easy to transport. Fuel cells are now a very large R and D effort for several military applications around the world. The major reason is because of the high power demands needed requires electrical energy sources that far exceed the capabilities of batteries currently being fielded for portable applications. Fuel cells are regarded as highly efficient, tactical energy converters that can be adapted for wide range of power requirements. They are potentially the lowest weight power source when coupled with batteries or capacitors to form hybrid systems. Generally electrical power is needed to support a number of applications from ultra-high power for electrical pulses (radios, sensors) to reliable, conditioned power for command and control systems. In the future, sustained power for electric drive systems, will also be required. Some of the promising applications in the military and the R and D challenges that remain to reach performance and reliability targets suitable for military requirements will be discussed. (author)

  1. Management and storage of spent nuclear fuel at research and test reactors. Proceedings of an advisory group meeting

    International Nuclear Information System (INIS)

    1996-08-01

    Irradiated fuel from research and test reactors has been stored at various facilities for several decades. As these facilities age and approach or exceed their original design lifetimes, there is mounting concern about closure of the fuel cycle and about the integrity of ageing fuels from the materials point of view as well as some concern about the loss of self-protection of the fuels as their activity decays. It is clear that an international effort is necessary to give these problems sufficient exposure and to ensure that work continues on appropriate solutions. The future of nuclear research, with its many benefits to mankind, is in jeopardy in some countries, especially countries without nuclear power programmes, because effective solutions for extended interim storage and final disposition of spent research reactor fuels are not yet available. An advisory Group meeting was convened in Vienna to consider a Database on the Management and Storage of Spent Nuclear Fuel from Research and Test Reactors. Sixteen experts from sixteen different countries participated in the Advisory Group meeting and presented country reports, which together represent an overview of the technologies used in spent fuel management and storage at research and test reactors world-wide. The sixteen country reports together with the database summary are presented in this publication. Refs, figs, tabs

  2. Remote technology related to the handling, storage and disposal of spent fuel. Proceedings of a technical committee meeting

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-11-01

    Reduced radiation exposure, greater reliability and cost savings are all potential benefits of the application of remote technologies to the handling of spent nuclear fuel. Remote equipment and technologies are used to some extent in all facilities handling fuel and high-level wastes whether they are for interim storage, processing/repacking, reprocessing or disposal. In view of the use and benefits of remote technologies, as well as recent technical and economic developments in the area, the IAEA organized the Technical Committee Meeting (TCM) on Remote Technology Related to the Handling, Storage and/or Disposal of Spent Fuel. Twenty-one papers were presented at the TCM, divided into five general areas: 1. Choice of technologies; 2. Use of remote technologies in fuel handling; 3. Use of remote technologies for fuel inspection and characterization; 4. Remote maintenance of facilities; and 5. Current and future developments. Refs, figs and tabs.

  3. 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

  4. Integrating patient satisfaction into performance measurement to meet improvement challenges.

    Science.gov (United States)

    Smith, J E; Fisher, D L; Endorf-Olson, J J

    2000-05-01

    A Value Compass has been proposed to guide health care data collection. The "compass corners" represent the four types of data needed to meet health care customer expectations: appropriate clinical outcomes, improved functional status, patient satisfaction, and appropriate costs. Collection of all four types of data is necessary to select processes in need of improvement, guide improvement teams, and monitor the success of improvement efforts. INTEGRATED DATA AT BRYANLGH: BryanLGH Medical Center in Lincoln, Nebraska, has adopted multiple performance measurement systems to collect clinical outcome, financial, and patient satisfaction data into integrated databases. Data integration allows quality professionals at BryanLGH to identify quality issues from multiple perspectives and track the interrelated effects of improvement efforts. A CASE EXAMPLE: Data from the fourth quarter of 1997 indicated the need to improve processes related to cesarean section (C-section) deliveries. An interdisciplinary team was formed, which focused on educating nurses, physicians, and the community about labor support measures. Physicians were given their own rates of C-section deliveries. The C-section rate decreased from 27% to 19%, but per-case cost increased. PickerPLUS+ results indicated that BryanLGH obstetric patients reported fewer problems with receiving information than the Picker norm, but they reported more problems with the involvement of family members and friends. The data collected so far have indicated a decrease in the C-section rate and a need to continue to work on cost and psychosocial issues. A complete analysis of results was facilitated by integrated performance management systems. Successes have been easily tracked over time, and the need for further work on related processes has been clearly identified.

  5. 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)

  6. 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)

  7. 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)

  8. 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)

  9. Evaluated nuclear data for Th-U fuel cycle. Summary report of the second research coordination meeting

    International Nuclear Information System (INIS)

    Schillebeeckx, P.; Trkov, A.

    2004-12-01

    Highlights of the second research coordination meeting are given with respect to progress, and the agreed route forward to achieve the primary objectives of the coordinated research project on 'Evaluated Nuclear data for Th-U Fuel Cycle'. Participants debated their findings and requirements for such files, and formulated a list of assigned tasks. (author)

  10. 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

  11. 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)

  12. 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.

  13. 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.

  14. 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...

  15. CONVERSION OF DIESEL ENGINE INTO SPARK IGNITION ENGINE TO WORK WITH CNG AND LPG FUELS FOR MEETING NEW EMISSION NORMS

    Directory of Open Access Journals (Sweden)

    Syed Kaleemuddin

    2010-01-01

    Full Text Available Fluctuating fuel prices and associated pollution problems of largely exploited petroleum liquid fuel has stimulated the research on abundantly available gaseous fuels to keep the mobility industry intact. In the present work an air cooled diesel engine was modified suitably into a spark ignition engine incorporating electronic ignition and variable speed dependant spark timing to accommodate both LPG and CNG as fuels. Engine was optimized for stoichiometric operation on engine dynamometer. Materials of a few intricate engine components were replaced to suit LPG and CNG application. Ignition timing was mapped to work with gaseous fuels for different speeds. Compensation was done for recovering volumetric efficiency when operated with CNG by introducing more volume of air through resonator. Ignition timing was observed to be the pertinent parameter in achieving good performance with gaseous fuels under consideration. Performance and emission tests were carried out on engine dynamometer and chassis dynamometer. Under wide open throttle and at rated speed condition, it was observed that the peak pressure with LPG was lying between diesel fuel and CNG fuel operation due to slow burning nature of gaseous fuels. As compression ratio was maintained same for LPG and CNG fuel operation, low CO emissions were observed with LPG where as HC + NOx emissions were lower with CNG fuel operation. Chassis dynamometer based emission tests yielded lower CO2 levels with CNG operation.

  16. 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)

  17. 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.

  18. Validating Westinghouse atom 16 x 16 and 18 x 18 PWR fuel performance

    International Nuclear Information System (INIS)

    Andersson, S.; Gustafson, J.; Jourdain, P.; Lindstroem, L.; Hallstadius, L.; Hofling, C.G.

    2001-01-01

    Westinghouse Atom designs and fabricates PWR fuel for all major European fuel types: 17 x 17 standard (12 ft) and 17 x 17 XL (14 ft) for Westinghouse type PWRs, and 16 x 16 and 18 x 18 fuel for Siemens type PWRs. The W Atom PWR fuel designs are based on the extensive Westinghouse CE PWR fuel experience from combustion engineering type PWRs. The W atom designs utilise basic design features from the W CE fuel tradition, such as all-Zircaloy mid grids and the proven ( 6 rod years) Guardian TM debris catcher, which is integrated in the bottom Inconel grid. Several new features have been developed to meet with stringent European requirements originating from requirements on very high burnup, in combination with low-leakage core operating strategies and high coolant temperatures. The overall reliability of the Westinghouse Atom PWR fuel is very high; no fuel failure has been detected since 1997. (orig.)

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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.

  5. 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

  6. 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

  7. 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.

  8. 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.

  9. 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.

  10. 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)

  11. 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

  12. EDF nuclear fuel cycle: report for the HCTISN meeting on the 20 of November 2009

    International Nuclear Information System (INIS)

    2009-01-01

    The first part of this report describes the different steps of the EDF fuel cycle and the associated flows: overview, description of each step (natural uranium mines, enrichment, fabrication, production in reactors, processing, recycling). The second part deals with valuable materials and wastes (history of waste flows since 1994, MOX flows since 1994). It presents supply sources and flows (diversification to ensure supply safety, flows among different supply steps) and actions performed by EDF to promote information. Some propositions are given to improve information

  13. 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

  14. 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

  15. 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

  16. A high performance cathode for proton conducting solid oxide fuel cells

    KAUST Repository

    Wang, Zhiquan

    2015-01-01

    Intermediate temperature solid-oxide fuel cells (IT-SOFCs)), as one of the energy conversion devices, have attracted worldwide interest for their great fuel efficiency, low air pollution, much reduced cost and excellent longtime stability. In the intermediate temperature range (500-700°C), SOFCs based on proton conducting electrolytes (PSOFCs) display unique advantages over those based on oxygen ion conducting electrolytes. A key obstacle to the practical operation of past P-SOFCs is the poor stability of the traditionally used composite cathode materials in the steam-containing atmosphere and their low contribution to proton conduction. Here we report the identification of a new Ruddlesden-Popper-type oxide Sr3Fe2O7-δ that meets the requirements for much improved long-term stability and shows a superior single-cell performance. With a Sr3Fe2O7-δ-5 wt% BaZr0.3Ce0.5Y0.2O3-δ cathode, the P-SOFC exhibits high power densities (683 and 583 mW cm-2 at 700°C and 650°C, respectively) when operated with humidified hydrogen as the fuel and air as the cathode gas. More importantly, no decay in discharging was observed within a 100 hour test. © The Royal Society of Chemistry 2015.

  17. 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).

  18. 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...

  19. 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.

  20. 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.

  1. 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.

  2. 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

  3. Enhanced Emission Performance and Fuel Efficiency for HD Methane Engines. Literature Study. Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Broman, R.; Staalhammar, P.; Erlandsson, L.

    2010-05-15

    A literature survey has been conducted in order to define state-of-the-art for methane fuelled engines to be used in heavy duty vehicles. Use of methane can be favourable to increase security of supply and mitigate CO2 emissions, especially when the methane origins from biomass. Furthermore, methane used as a fuel in heavy duty engines has a potential to reduce toxic exhaust emissions. Historically, use of methane in heavy duty engines has often been hampered by poor efficiency, i.e. high fuel consumption when using the Otto-cycle. However, current generation technology engines might be within 5-10 % of the efficiency of Diesel engine technology. In this context it is worth mentioning that compliance-driven changes for meeting future emission regulations for Diesel engines may have a negative impact on fuel efficiency, thereby narrowing the gap. This may present an opportunity for heavy methane fuelled engines. The reliability and durability of the exhaust aftertreatment devices for methane fuelled engines has also given rise to some concerns. Some concepts are performing acceptable while others do not meet expectations. This is partly due to difficulties in handling methane in the aftertreatment device and partly to issues in the design of the ignition system. Methane is a fuel used worldwide and has a potential to be an important complement to Diesel oil. There are two categories of HD methane engines available to end-users: Retrofitted engines, which often include computer controlled retrofit systems developed as 'bolt-on' technologies that can be removed if necessary, to resell the vehicle with a normal diesel engine, and those developed specifically for and in conjunction with engine manufacturers and delivered to customers as factory-built engines or vehicles (OEM). Additionally, both these categories can include engines that use the Otto- or Diesel combustion cycles. When adapting a HD Diesel engine to run on methane there are two options, either

  4. 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)

  5. 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

  6. 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.

  7. Twenty-third water reactor safety information meeting: Volume 1, plenary session, high burnup fuel behavior, thermal hydraulic research. Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1996-03-01

    This three-volume report contains papers presented at the Twenty- Third Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 23-25, 1995. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Italy, Japan, Norway, Russia, Sweden, and Switzerland. This document, Volume 1, present topics on High Burnup Fuel Behavior, Thermal Hydraulic Research, and Plenary Session topics. Individual papers have been cataloged separately.

  8. Twenty-third water reactor safety information meeting: Volume 1, plenary session, high burnup fuel behavior, thermal hydraulic research. Proceedings

    International Nuclear Information System (INIS)

    Monteleone, S.

    1996-03-01

    This three-volume report contains papers presented at the Twenty- Third Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 23-25, 1995. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Italy, Japan, Norway, Russia, Sweden, and Switzerland. This document, Volume 1, present topics on High Burnup Fuel Behavior, Thermal Hydraulic Research, and Plenary Session topics. Individual papers have been cataloged separately

  9. Integrated fuel processor development

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  10. 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)

  11. 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

  12. 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)

  13. 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)

  14. 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.

  15. Results on Technical and Consultants Service Meetings on Lessons Learned from Operating Experience in Wet and Dry Spent Fuel Storage

    International Nuclear Information System (INIS)

    White, B.; Zou, X.

    2015-01-01

    Spent fuel storage has been and will continue to be a vital portion of the nuclear fuel cycle, regardless of whether a member state has an open or closed nuclear fuel cycle. After removal from the reactor core, spent fuel cools in the spent fuel pool, prior to placement in dry storage or offsite transport for disposal or reprocessing. Additionally, the inventory of spent fuel at many reactors worldwide has or will reach the storage capacity of the spent fuel pool; some facilities are alleviating their need for additional storage capacity by utilizing dry cask storage. While there are numerous differences between wet and dry storage; when done properly both are safe and secure. The nuclear community shares lessons learned worldwide to gain knowledge from one another’s good practices as well as events. Sharing these experiences should minimize the number of incidents worldwide and increase public confidence in the nuclear industry. Over the past 60 years, there have been numerous experiences storing spent fuel, in both wet and dry mediums, that when shared effectively would improve operations and minimize events. These lessons learned will also serve to inform countries, who are new entrants into the nuclear power community, on designs and operations to avoid and include as best practices. The International Atomic Energy Agency convened a technical and several consultants’ meetings to gather these experiences and produce a technical document (TECDOC) to share spent fuel storage lessons learned among member states. This paper will discuss the status of the TECDOC and briefly discuss some lessons learned contained therein. (author)

  16. 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)

  17. 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 ...

  18. 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)

  19. 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.

  20. 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.

  1. 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 .

  2. 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.

  3. 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

  4. 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.

  5. Design, fabrication and performance evaluation of an integrated reformed methanol fuel cell for portable use

    Science.gov (United States)

    Zhang, Shubin; Zhang, Yufeng; Chen, Junyu; Yin, Congwen; Liu, Xiaowei

    2018-06-01

    In this paper, an integrated reformed methanol fuel cell (RMFC) as a portable power source is designed, fabricated and tested. The RMFC consists of a methanol steam reformer (MSR), a high temperature proton exchange membrane fuel cell (HT-PEMFC) stack, a microcontroller unit (MCU) and other auxiliaries. First, a system model based on Matlab/Simulink is established to investigate the mass and energy transport characteristics within the whole system. The simulation results suggest a hydrogen flow rate of at least 670 sccm is needed for the system to output 30 W and simultaneously maintain thermal equilibrium. Second, a metallic MSR and an HT-PEMFC stack with 12 cells are fabricated and tested. The tests show that the RMFC system is able to function normally when the performances of all the components meet the minimum requirements. At last, in the experiment of successfully powering a laptop, the RMFC system exhibits a stable performance during the complete work flow of all the phases, namely start-up, output and shutdown. Moreover, with a conservative design of 20 W power rating, maximum energy conversion efficiency of the RMFC system can be achieved (36%), and good stability in long-term operation is shown.

  6. 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.

  7. Meeting

    Indian Academy of Sciences (India)

    July 1989 No.19 Newsletter of the Indian Academy of Sciences. 55th Annual. Meeting ... in the world, keeping alive atthe same time his research interests, abreast .... theory made a comeback with many new ideas and with the success of the ...

  8. 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

  9. 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.

  10. 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

  11. 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

  12. 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

  13. 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...

  14. 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.

  15. 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)

  16. Improving of diesel combustion-pollution-fuel economy and performance by gasoline fumigation

    International Nuclear Information System (INIS)

    Şahin, Zehra; Durgun, Orhan

    2013-01-01

    Highlights: • The effects of gasoline fumigation on the engine performance and NO x emission were investigated in Ford XLD 418 T automotive diesel engine. • Gasoline at approximately (2, 4, 6, 8 10, and 12)% (by vol.) ratios was injected into intake air by a carburetor. • GF enhances effective power and reduces brake specific fuel consumption, cost, and NO x emission. - Abstract: One of the most important objectives of the studies worldwide is to improve combustion of diesel engine to meet growing energy needs and to reduce increasing environmental pollution. To accomplish this goal, especially to reduce pollutant emissions, researchers have focused their interest on the field of alternative fuels and alternative solutions. Gasoline fumigation (GF) is one of these alternative solutions, by which diesel combustion, fuel economy, and engine performance are improved, and environmental pollution is decreased. In the fumigation method, gasoline is injected into intake air, either by a carburetor, which main nozzle section is adjustable or by a simple injection system. In the present experimental study, a simple carburetor was used, and the effects of gasoline fumigation at (2, 4, 6, 8, 10, 12)% (by vol.) gasoline ratios on the combustion, NO x emission, fuel economy, and engine performance sophisticatedly investigated for a fully instrumented, four-cylinder, water-cooled indirect injection (IDI), Ford XLD 418 T automotive diesel engine. Tests were conducted for each of the above gasoline fumigation ratios at three different speeds and for (1/1, 3/4, and 1/2) fuel delivery ratios (FDRs). GF test results showed that NO x emission is lower than that of neat diesel fuel (NDF). NO x emission decreases approximately 4.20%, 2.50%, and 9.65% for (1/1, 3/4, and 1/2) FDRs, respectively. Effective power increases approximately 2.38% for 1/1 FDR. At (2500 and 3000) rpms, effective power decreases at low gasoline ratios, but it increases at high gasoline ratios for 3/4 and 1

  17. 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)

  18. 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)

  19. 77 FR 5615 - First Meeting: RTCA Special Committee 227, Standards of Navigation Performance

    Science.gov (United States)

    2012-02-03

    ... Performance--Background/History Review of SC 181 Products and Intended Applications Walk through of DO-236B... Responsibilities Any Other Business Establish Agenda for Next Meeting Date and Place of Next Meeting Adjourn... Washington, DC, on January 30, 2012. John Raper, Manager, Business Operations Branch, Federal Aviation...

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  1. 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

  2. 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

  3. 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...

  4. 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

  5. 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.

  6. 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

  7. 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...

  8. 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

  9. Proceedings from Specialists Meeting on human performance in operational events

    International Nuclear Information System (INIS)

    1998-01-01

    This conference on human performance in operational events is composed of 34 papers, grouped in 11 sessions. After an invited contribution on the human factor in the nuclear industry, the sessions are: