WorldWideScience

Sample records for research scale fuel

  1. Fuel safety research 1999

    Energy Technology Data Exchange (ETDEWEB)

    Uetsuka, Hiroshi (ed.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-07-01

    In April 1999, the Fuel Safety Research Laboratory was newly established as a result of reorganization of the Nuclear Safety Research Center, JAERI. The laboratory was organized by combining three laboratories, the Reactivity Accident Laboratory, the Fuel Reliability Laboratory, and a part of the Sever Accident Research Laboratory. Consequently, the Fuel Safety Research Laboratory is now in charge of all the fuel safety research in JAERI. Various types of experimental and analytical researches are conducted in the laboratory by using the unique facilities such as the Nuclear Safety Research Reactor (NSRR), the Japan Material Testing Reactor (JMTR), the Japan Research Reactor 3 (JRR-3) and hot cells in JAERI. The laboratory consists of five research groups corresponding to each research fields. They are; (a) Research group of fuel behavior under the reactivity initiated accident conditions (RIA group). (b) Research group of fuel behavior under the loss-of-coolant accident conditions (LOCA group). (c) Research group of fuel behavior under the normal operation conditions (JMTR/BOCA group). (d) Research group of fuel behavior analysis (FEMAXI group). (e) Research group of FP release/transport behavior from irradiated fuel (VEGA group). This report summarizes the outline of research activities and major outcomes of the research executed in 1999 in the Fuel Safety Research Laboratory. (author)

  2. Fuel safety research 2001

    Energy Technology Data Exchange (ETDEWEB)

    Uetsuka, Hiroshi (ed.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-11-01

    The Fuel Safety Research Laboratory is in charge of research activity which covers almost research items related to fuel safety of water reactor in JAERI. Various types of experimental and analytical researches are being conducted by using some unique facilities such as the Nuclear Safety Research Reactor (NSRR), the Japan Material Testing Reactor (JMTR), the Japan Research Reactor 3 (JRR-3) and the Reactor Fuel Examination Facility (RFEF) of JAERI. The research to confirm the safety of high burn-up fuel and MOX fuel under accident conditions is the most important item among them. The laboratory consists of following five research groups corresponding to each research fields; Research group of fuel behavior under the reactivity initiated accident conditions (RIA group). Research group of fuel behavior under the loss-of-coolant accident conditions (LOCA group). Research group of fuel behavior under the normal operation conditions (JMTR/BOCA group). Research group of fuel behavior analysis (FEMAXI group). Research group of radionuclides release and transport behavior from irradiated fuel under severe accident conditions (VEGA group). The research conducted in the year 2001 produced many important data and information. They are, for example, the fuel behavior data under BWR power oscillation conditions in the NSRR, the data on failure-bearing capability of hydrided cladding under LOCA conditions and the FP release data at very high temperature in steam which simulate the reactor core condition during severe accidents. This report summarizes the outline of research activities and major outcomes of the research executed in 2001 in the Fuel Safety Research Laboratory. (author)

  3. Fuel safety research 2000

    Energy Technology Data Exchange (ETDEWEB)

    Uetsuka, Hiroshi (ed.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-03-01

    In April 1999, the Fuel Safety Research Laboratory was newly established as a part of reorganization of the Nuclear Safety Research Center, JAERI. The new laboratory was organized by combining three pre-existing laboratories, Reactivity Accident Laboratory, Fuel Reliability Laboratory, and a part of Severe Accident Research Laboratory. The Fuel Safety Research Laboratory becomes to be in charge of all fuel safety research in JAERI. Various experimental and analytical researches are conducted in the laboratory by using the unique facilities such as the Nuclear Safety Research Reactor (NSRR), the Japan Material Testing Reactor (JMTR), the Japan Research Reactor 3 (JRR-3) and hot cells in JAERI. The laboratory consists of following five research groups corresponding to each research fields; (a) Research group of fuel behavior under the reactivity initiated accident conditions (RIA group). (b) Research group of fuel behavior under the loss-of-coolant accident conditions (LOCA group). (c) Research group of fuel behavior under the normal operation conditions (JMTR/BOCA group). (d) Research group of fuel behavior analysis (FEMAXI group). (e) Research group of FP release/transport behavior from irradiated fuel (VEGA group). The research activities in year 2000 produced many important data and information. They are, for example, failure of high burnup BWR fuel rod under RIA conditions, data on the behavior of hydrided Zircaloy cladding under LOCA conditions and FP release data from VEGA experiments at very high temperature/pressure condition. This report summarizes the outline of research activities and major outcomes of the research executed in 2000 in the Fuel Safety Research Laboratory. (author)

  4. Fuel Chemistry Research | Transportation Research | NREL

    Science.gov (United States)

    Fuel Chemistry Research Fuel Chemistry Research Photo of a hand holding a beaker containing a clear oils. Photo by Dennis Schroeder, NREL NREL's fuel chemistry research explores how biofuels, advanced , emissions control catalysts, and infrastructure materials. Results from NREL's fuel chemistry studies feed

  5. Fuel Cell Research

    Energy Technology Data Exchange (ETDEWEB)

    Weber, Peter M. [Brown University

    2014-03-30

    Executive Summary In conjunction with the Brown Energy Initiative, research Projects selected for the fuel cell research grant were selected on the following criteria: They should be fundamental research that has the potential to significantly impact the nation’s energy infrastructure. They should be scientifically exciting and sound. They should synthesize new materials, lead to greater insights, explore new phenomena, or design new devices or processes that are of relevance to solving the energy problems. They involve top-caliper senior scientists with a record of accomplishment, or junior faculty with outstanding promise of achievement. They should promise to yield at least preliminary results within the given funding period, which would warrant further research development. They should fit into the overall mission of the Brown Energy Initiative, and the investigators should contribute as partners to an intellectually stimulating environment focused on energy science. Based on these criteria, fourteen faculty across three disciplines (Chemistry, Physics and Engineering) and the Charles Stark Draper Laboratory were selected to participate in this effort.1 In total, there were 30 people supported, at some level, on these projects. This report highlights the findings and research outcomes of the participating researchers.

  6. Advanced Research Reactor Fuel Development

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-04-15

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

  7. Fuel tank integrity research : fuel tank analyses and test plans

    Science.gov (United States)

    2013-04-15

    The Federal Railroad Administrations Office of Research : and Development is conducting research into fuel tank : crashworthiness. Fuel tank research is being performed to : determine strategies for increasing the fuel tank impact : resistance to ...

  8. Caramel fuel for research reactors

    International Nuclear Information System (INIS)

    Bussy, P.

    1979-11-01

    This fuel for research reactors is made of UO 2 pellets in a zircaloy cladding to replace 93% enriched uranium. It is a cold fuel, non contaminating and non proliferating, enrichment is only 7 to 8%. Irradiation tests were performed until burn-up of 50000 MWD/t [fr

  9. Large-scale fuel cycle centres

    International Nuclear Information System (INIS)

    Smiley, S.H.; Black, K.M.

    1977-01-01

    The US Nuclear Regulatory Commission (NRC) has considered the nuclear energy centre concept for fuel cycle plants in the Nuclear Energy Centre Site Survey 1975 (NECSS-75) Rep. No. NUREG-0001, an important study mandated by the US Congress in the Energy Reorganization Act of 1974 which created the NRC. For this study, the NRC defined fuel cycle centres as consisting of fuel reprocessing and mixed-oxide fuel fabrication plants, and optional high-level waste and transuranic waste management facilities. A range of fuel cycle centre sizes corresponded to the fuel throughput of power plants with a total capacity of 50,000-300,000MW(e). The types of fuel cycle facilities located at the fuel cycle centre permit the assessment of the role of fuel cycle centres in enhancing the safeguard of strategic special nuclear materials - plutonium and mixed oxides. Siting fuel cycle centres presents a smaller problem than siting reactors. A single reprocessing plant of the scale projected for use in the USA (1500-2000t/a) can reprocess fuel from reactors producing 50,000-65,000MW(e). Only two or three fuel cycle centres of the upper limit size considered in the NECSS-75 would be required in the USA by the year 2000. The NECSS-75 fuel cycle centre evaluation showed that large-scale fuel cycle centres present no real technical siting difficulties from a radiological effluent and safety standpoint. Some construction economies may be achievable with fuel cycle centres, which offer opportunities to improve waste-management systems. Combined centres consisting of reactors and fuel reprocessing and mixed-oxide fuel fabrication plants were also studied in the NECSS. Such centres can eliminate shipment not only of Pu but also mixed-oxide fuel. Increased fuel cycle costs result from implementation of combined centres unless the fuel reprocessing plants are commercial-sized. Development of Pu-burning reactors could reduce any economic penalties of combined centres. The need for effective fissile

  10. Fuel Combustion Laboratory | Transportation Research | NREL

    Science.gov (United States)

    Fuel Combustion Laboratory Fuel Combustion Laboratory NREL's Fuel Combustion Laboratory focuses on designs, using both today's technology and future advanced combustion concepts. This lab supports the combustion chamber platform for fuel ignition kinetics research, was acquired to expand the lab's

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

    Science.gov (United States)

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

  12. Research reactor fuel - an update

    International Nuclear Information System (INIS)

    Finlay, M.R.; Ripley, M.I.

    2003-01-01

    In the two years since the last ANA conference there have been marked changes in the research reactor fuel scene. A new low-enriched uranium (LEU) fuel, 'monolithic' uranium molybdenum, has shown such promise in initial trials that it may be suitable to meet the objectives of the Joint Declaration signed by Presidents Bush and Putin to commit to converting all US and Russian research reactors to LEU by 2012. Development of more conventional aluminium dispersion UMo LEU fuel has continued in the meantime and is entering the final qualification stage of multiple full sized element irradiations. Despite this progress, the original 2005 timetable for UMo fuel qualification has slipped and research reactors, including the RRR, may not convert from silicide to UMo fuel before 2007. The operators of the Swedish R2 reactor have been forced to pursue the direct route of qualifying a UMo lead test assembly (LTA) in order to meet spent fuel disposal requirements of the Swedish law. The LTA has recently been fabricated and is expected to be loaded shortly into the R2 reactor. We present an update of our previous ANA paper and details of the qualification process for UMo fuel

  13. Fuels for Canadian research reactors

    International Nuclear Information System (INIS)

    Feraday, M.A.

    1993-01-01

    This paper includes some statements and remarks concerning the uranium silicide fuels for which there is significant fabrication in AECL, irradiation and defect performance experience; description of two Canadian high flux research reactors which use high enrichment uranium (HEU) and the fuels currently used in these reactors; limited fabrication work done on Al-U alloys to uranium contents as high as 40 wt%. The latter concerns work aimed at AECL fast neutron program. This experience in general terms is applied to the NRX and NRU designs of fuel

  14. Navy Fuel Research Facility

    Data.gov (United States)

    Federal Laboratory Consortium — FUNCTION: Performs basic and applied research to understand the underlying chemistry that impacts the use, handling, and storage of current and future Navy mobility...

  15. Fuels Combustion Research: Supercritical Fuel Pyrolysis

    National Research Council Canada - National Science Library

    Glassman, Irvin

    2001-01-01

    .... The focus during the subject period was directed to understanding the pyrolysis and combustion of endothermic fuels under subcritical conditions and the pyrolysis of these fuels under supercritical conditions...

  16. Fuels Combustion Research: Supercritical Fuel Pyrolysis

    National Research Council Canada - National Science Library

    Glassman, Irvin

    2000-01-01

    .... The focus during the subject period was directed to understanding the pyrolysis and combustion of endothermic fuels under subcritical conditions and the pyrolysis of these fuels under supercritical conditions...

  17. Large-scale fuel cycle centers

    International Nuclear Information System (INIS)

    Smiley, S.H.; Black, K.M.

    1977-01-01

    The United States Nuclear Regulatory Commission (NRC) has considered the nuclear energy center concept for fuel cycle plants in the Nuclear Energy Center Site Survey - 1975 (NECSS-75) -- an important study mandated by the U.S. Congress in the Energy Reorganization Act of 1974 which created the NRC. For the study, NRC defined fuel cycle centers to consist of fuel reprocessing and mixed oxide fuel fabrication plants, and optional high-level waste and transuranic waste management facilities. A range of fuel cycle center sizes corresponded to the fuel throughput of power plants with a total capacity of 50,000 - 300,000 MWe. The types of fuel cycle facilities located at the fuel cycle center permit the assessment of the role of fuel cycle centers in enhancing safeguarding of strategic special nuclear materials -- plutonium and mixed oxides. Siting of fuel cycle centers presents a considerably smaller problem than the siting of reactors. A single reprocessing plant of the scale projected for use in the United States (1500-2000 MT/yr) can reprocess the fuel from reactors producing 50,000-65,000 MWe. Only two or three fuel cycle centers of the upper limit size considered in the NECSS-75 would be required in the United States by the year 2000 . The NECSS-75 fuel cycle center evaluations showed that large scale fuel cycle centers present no real technical difficulties in siting from a radiological effluent and safety standpoint. Some construction economies may be attainable with fuel cycle centers; such centers offer opportunities for improved waste management systems. Combined centers consisting of reactors and fuel reprocessing and mixed oxide fuel fabrication plants were also studied in the NECSS. Such centers can eliminate not only shipment of plutonium, but also mixed oxide fuel. Increased fuel cycle costs result from implementation of combined centers unless the fuel reprocessing plants are commercial-sized. Development of plutonium-burning reactors could reduce any

  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 cells: Trends in research and applications

    Science.gov (United States)

    Appleby, A. J.

    Various aspects of fuel cells are discussed. The subjects addressed include: fuel cells for electric power production; phosphoric acid fuel cells; long-term testing of an air-cooled 2.5 kW PAFC stack in Italy; status of fuel cell research and technology in the Netherlands, Bulgaria, PRC, UK, Sweden, India, Japan, and Brazil; fuel cells from the manufacturer's viewpoint; and fuel cells using biomass-derived fuels. Also examined are: solid oxide electrolye fuel cells; aluminum-air batteries with neutral chloride electrolyte; materials research for advanced solid-state fuel cells at the Energy Research Laboratory in Denmark; molten carbonate fuel cells; the impact of the Siemens program; fuel cells at Sorapec; impact of fuel cells on the electric power generation systems in industrial and developing countries; and application of fuel cells to large vehicles.

  20. Small-scale irradiated fuel electrorefining

    International Nuclear Information System (INIS)

    Benedict, R.W.; Krsul, J.R.; Mariani, R.D.; Park, K.; Teske, G.M.

    1993-01-01

    In support of the metallic fuel cycle development for the Integral Fast Reactor (IFR), a small scale electrorefiner was built and operated in the Hot Fuel Examination Facility (HFEF) at Argonne National Laboratory-West. The initial purpose of this apparatus was to test the single segment dissolution of irradiated metallic fuel via either direct dissolution in cadmium or anodic dissolution. These tests showed that 99.95% of the uranium and 99.99% of the plutonium was dissolved and separated from the fuel cladding material. The fate of various fission products was also measured. After the dissolution experiments, the apparatus was upgraded to stady fission product behavior during uranium electrotransport. Preliminary decontamination factors were estimated for different fission products under different processing conditions. Later modifications have added the following capabilities: Dissolution of multiple fuel segments simultaneously, electrotransport to a solid cathode or liquid cathode and actinide recovery with a chemical reduction crucible. These capabilities have been tested with unirradiated uranium-zirconium fuel and will support the Fuel Cycle Demonstration program

  1. World scale fuel methanol facility siting

    International Nuclear Information System (INIS)

    Stapor, M.C.; Hederman, W.F.

    1990-01-01

    Since the Administration announced a clean alternative fuels initiative, industry and government agencies' analyses of the economics of methanol as an alternative motor vehicle fuel have accelerated. In the short run, methanol appears attractive because excess production capacity currently has depressed methanol prices and marginal costs of production are lower than other fuels (current excess capacity). In the long run, however, full costs are the more relevant. To lower average production costs, U.S. policy interest has focused on production from a world-scale, 10,000 tons per day (tpd) methanol plant facility on a foreign site. This paper reviews several important site and financial considerations in a framework to evaluate large scale plant development. These considerations include: risks associated with a large process plant; supply economics of foreign sites; and investment climates and financial incentives for foreign investment at foreign sites

  2. Advanced research reactor fuel development

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-05-01

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

  3. Multiplexed electrospray scaling for liquid fuel injection

    International Nuclear Information System (INIS)

    Waits, C Mike; Hanrahan, Brendan; Lee, Ivan

    2010-01-01

    Evaporation and space-charge requirements are evaluated to understand the effect of device scaling and fuel preheating for a liquid fuel injector using a multiplexed electrospray (MES) configuration in compact combustion applications. This work reveals the influence of the droplet diameter, droplet velocity and droplet surface temperature as well as the surrounding gas temperature on the size and performance of microfabricated MES. Measurements from MES devices are used in the model to accurately account for the droplet diameter versus flow rate relationship, the minimum droplet diameter and the relevant droplet velocities. A maximum extractor electrode to ground electrode distance of 3.1 mm required to overcome space-charge forces is found to be independent of voltage or droplet velocity for large levels of multiplexing. This maximum distance also becomes the required evaporation length scale which imposes minimum fuel pre-heating requirements for large flow densities. Required fuel preheating is therefore evaluated for both ethanol and 1-butanol with combustor parameters relevant to fuel reformation, thermoelectric conversion, thermophotovoltaic conversion and thermionic conversion

  4. Research and development of thorium fuel cycle

    International Nuclear Information System (INIS)

    Oishi, Jun.

    1994-01-01

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

  5. Composite Solid Fuel: Research of Formation Parameters

    Directory of Open Access Journals (Sweden)

    Tabakaev Roman

    2016-01-01

    Full Text Available Involving of local low-grade fuels resources in fuel and energy balance is actual question of research in the present. In this paper the possibility of processing low-grade fuel in the solid fuel composite was considered. The aim of the work is to define the optimal parameters for formation of the solid composite fuel. A result of researches determined that dextrin content in the binder allows to obtain solid composite fuel having the highest strength. The drying temperature for the various fuels was determined: for pellets production was 20-80 °C, for briquettes – 20-40 °C.

  6. Unified fuel elements development for research reactors

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  7. Fuel elements of research reactors in China

    International Nuclear Information System (INIS)

    Zhou Yongmao; Chen Dianshan; Tan Jiaqiu

    1987-01-01

    This paper describes the current status of design, fabrication of fuel elements for research reactors in China, emphasis is placed on the technology of fuel elements for the High Flux Engineering Test Reactor (HFETR). (author)

  8. 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.)

  9. 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)

  10. 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

  11. Trends in fuel reprocessing safety research

    International Nuclear Information System (INIS)

    Tsujino, Takeshi

    1981-01-01

    With the operation of a fuel reprocessing plant in the Power Reactor and Nuclear Fuel Development Corporation (PNC) and the plan for a second fuel reprocessing plant, the research on fuel reprocessing safety, along with the reprocessing technology itself, has become increasingly important. As compared with the case of LWR power plants, the safety research in this field still lags behind. In the safety of fuel reprocessing, there are the aspects of keeping radiation exposure as low as possible in both personnel and local people, the high reliability of the plant operation and the securing of public safety in accidents. Safety research is then required to establish the safety standards and to raise the rate of plant operation associated with safety. The following matters are described: basic ideas for the safety design, safety features in fuel reprocessing, safety guideline and standards, and safety research for fuel reprocessing. (J.P.N.)

  12. Dispersion fuel for nuclear research facilities

    International Nuclear Information System (INIS)

    Kushtym, A.V.; Belash, M.M.; Zigunov, V.V.; Slabospitska, O.O.; Zuyok, V.A.

    2017-01-01

    Designs and process flow sheets for production of nuclear fuel rod elements and assemblies TVS-XD with dispersion composition UO_2+Al are presented. The results of fuel rod thermal calculation applied to Kharkiv subcritical assembly and Kyiv research reactor VVR-M, comparative characteristics of these fuel elements, the results of metallographic analyses and corrosion tests of fuel pellets are given in this paper

  13. Nuclear Fusion Fuel Cycle Research Perspectives

    International Nuclear Information System (INIS)

    Chung, Hongsuk; Koo, Daeseo; Park, Jongcheol; Kim, Yeanjin; Yun, Sei-Hun

    2015-01-01

    As a part of the International Thermonuclear Experimental Reactor (ITER) Project, we at the Korea Atomic Energy Research Institute (KAERI) and our National Fusion Research Institute (NFRI) colleagues are investigating nuclear fusion fuel cycle hardware including a nuclear fusion fuel Storage and Delivery System (SDS). To have a better knowledge of the nuclear fusion fuel cycle, we present our research efforts not only on SDS but also on the Fuel Supply System (FS), Tokamak Exhaust Processing System (TEP), Isotope Separation System (ISS), and Detritiation System (DS). To have better knowledge of the nuclear fusion fuel cycle, we presented our research efforts not only on SDS but also on the Fuel Supply System (FS), Tokamak Exhaust Processing System (TEP), Isotope Separation System (ISS), and Detritiation System (DS). Our efforts to enhance the tritium confinement will be continued for the development of cleaner nuclear fusion power plants

  14. The Canadian research reactor spent fuel situation

    International Nuclear Information System (INIS)

    Ernst, P.C.

    1996-01-01

    This paper summarizes the present research reactor spent fuel situation in Canada. The research reactors currently operating are listed along with the types of fuel that they utilize. Other shut down research reactors contributing to the storage volume are included for completeness. The spent fuel storage facilities associated with these reactors and the methods used to determine criticality safety are described. Finally the current inventory of spent fuel and where it is stored is presented along with concerns for future storage. (author). 3 figs

  15. Unification of fuel elements for research reactors

    International Nuclear Information System (INIS)

    Vatulyn, A.V.; Stetskyi, Y.A.; Dobrikova, I.V.

    1997-01-01

    To the purpose of fuel elements unification the possibility of rod fuel assembly (FA) using in the cores of research reactors have been considered in this paper. The calculation results of geometric, hydraulic and thermotechnical parameters of rod assembly are submitted. Several designs of finned square fuel element and fuel assembly are proposed on base of analysis of rod FA characteristics in compare of tube ones. The fuel elements specimens and the model assembly are manufactured. The developed designs are the basis for further optimization after neutron-physical calculations of cores. (author)

  16. Atomic scale modelling of materials of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Bertolus, M.

    2011-10-01

    This document written to obtain the French accreditation to supervise research presents the research I conducted at CEA Cadarache since 1999 on the atomic scale modelling of non-metallic materials involved in the nuclear fuel cycle: host materials for radionuclides from nuclear waste (apatites), fuel (in particular uranium dioxide) and ceramic cladding materials (silicon carbide). These are complex materials at the frontier of modelling capabilities since they contain heavy elements (rare earths or actinides), exhibit complex structures or chemical compositions and/or are subjected to irradiation effects: creation of point defects and fission products, amorphization. The objective of my studies is to bring further insight into the physics and chemistry of the elementary processes involved using atomic scale modelling and its coupling with higher scale models and experimental studies. This work is organised in two parts: on the one hand the development, adaptation and implementation of atomic scale modelling methods and validation of the approximations used; on the other hand the application of these methods to the investigation of nuclear materials under irradiation. This document contains a synthesis of the studies performed, orientations for future research, a detailed resume and a list of publications and communications. (author)

  17. Fuel Fabrication Capability Research and Development Plan

    Energy Technology Data Exchange (ETDEWEB)

    Senor, David J.; Burkes, Douglas

    2014-04-17

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

  18. Fuel cell research: Towards efficient energy

    CSIR Research Space (South Africa)

    Rohwer, MB

    2008-11-01

    Full Text Available fuel cells by optimising the loading of catalyst (being expensive noble metals) and ionomer; 2) Improving conventional acidic direct alcohol fuel cells by developing more efficient catalysts and by investigating other fuels than methanol; 3... these components add significantly to the overall cost of a PEMFC. 1 We focused our research activities on: 1) The effect of the loading of catalytic ink on cell performance; 2) The effect of the ionomer content in the catalytic ink; 3) Testing...

  19. Research on CDA for advanced fuel FBR

    International Nuclear Information System (INIS)

    Hirano, Go; Hirakawa, Naohiro; Kawada, Ken-ichi; Niwa, Hazime.

    1997-03-01

    For the purpose of evaluating possibility of the re-criticality of a metallic fueled reactor, Tohoku university and Power Reactor and Nuclear Fuel Development Corporation have made a joint research entitled 'Research on CDA for advanced fuel FBR'. The results of this year are the following. The accident initiator considered is a loss-of-flow accident with ATWS. The LOF analysis was performed for the metallic fueled 600 MWe homogeneous two region reactors, both for a metallic fuel only and for a metallic fuel core with ZrH pin. The SAS3D CDA initiation phase analysis code was used to investigate the re-criticality potential at the severe accident. The change mainly in the constants was necessary to apply the code for the analysis of a metallic fueled reactor. These changes were made by assuming appropriate models. LOF with flow decay half time of t 1/2 =0.5(s) (all blackout case) and 5.5(s) (ordinary LOF case) were analyzed. Independent of the conditions of the analysis, the results show all the cases could avoid to become prompt-critical. Depending on the analysis condition, it becomes necessary to transfer to the transient phase, it is also shown there is a possibility to avoid re-criticality due to the motion of molten fuel both for the metallic fuel and for the metallic fuel with ZrH moderator. However, because of the constants used for the material property the results might overestimate the fuel motion. It is shown that the moderator is effective to terminate the accident at an early stage. The behavior of metallic fueled reactors at CDA was analyzed with SAS3D code by modifying the constants of material properties to be applied to the reactor. It is shown that a metallic fueled reactor has a possibility to avoid re-criticality at CDA. (J.P.N.)

  20. Fuel Fabrication Capability Research and Development Plan

    Energy Technology Data Exchange (ETDEWEB)

    Senor, David J.; Burkes, Douglas

    2013-06-28

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

  1. Research reactor spent fuel in Ukraine

    International Nuclear Information System (INIS)

    Trofimenko, A.P.

    1996-01-01

    This paper describes the research reactors in Ukraine, their spent fuel facilities and spent fuel management problems. Nuclear sciences, technology and industry are highly developed in Ukraine. There are 5 NPPs in the country with 14 operating reactors which have total power capacity of 12,800 MW

  2. Advanced fuel in the Budapest research reactor

    International Nuclear Information System (INIS)

    Hargitai, T.; Vidovsky, I.

    1997-01-01

    The Budapest Research Reactor, the first nuclear facility of Hungary, started to operate in 1959. The main goal of the reactor is to serve neutron research, but applications as neutron radiography, radioisotope production, pressure vessel surveillance test, etc. are important as well. The Budapest Research Reactor is a tank type reactor, moderated and cooled by light water. After a reconstruction and upgrading in 1967 the VVR-SM type fuel elements were used in it. These fuel elements provided a thermal power of 5 MW in the period 1967-1986 and 10 MW after the reconstruction from 1992. In the late eighties the Russian vendor changed the fuel elements slightly, i.e. the main parameters of the fuel remained unchanged, however a higher uranium content was reached. This new fuel is called VVR-M2. The geometry of VVR-SM and VVR-M2 are identical, allowing the use to load old and new fuel assemblies together to the active core. The first new type fuel assemblies were loaded to the Budapest Research Reactor in 1996. The present paper describes the operational experience with the new type of fuel elements in Hungary. (author)

  3. Research reactor de-fueling and fuel shipment

    International Nuclear Information System (INIS)

    Ice, R.D.; Jawdeh, E.; Strydom, J.

    1998-01-01

    Planning for the Georgia Institute of Technology Research Reactor operations during the 1996 Summer Olympic Games began in early 1995. Before any details could be outlined, several preliminary administrative decisions had to be agreed upon by state, city, and university officials. The two major administrative decisions involving the reactor were (1) the security level and requirements and (2) the fuel status of the reactor. The Georgia Tech Research Reactor (GTRR) was a heavy-water moderated and cooled reactor, fueled with high-enriched uranium. The reactor was first licensed in 1964 with an engineered lifetime of thirty years. The reactor was intended for use in research applications and as a teaching facility for nuclear engineering students and reactor operators. Approximately one year prior to the olympics, the Georgia Tech administration decided that the GTRR fuel would be removed. In addition, a heightened, beyond regulatory requirements, security system was to be implemented. This report describes the scheduling, operations, and procedures

  4. Alternative Fuels DISI Engine Research ? Autoignition Metrics.

    Energy Technology Data Exchange (ETDEWEB)

    Sjoberg, Carl Magnus Goran [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Vuilleumier, David [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2018-02-01

    Improved engine efficiency is required to comply with future fuel economy standards. Alternative fuels have the potential to enable more efficient engines while addressing concerns about energy security. This project contributes to the science base needed by industry to develop highly efficient direct injection spark igniton (DISI) engines that also beneficially exploit the different properties of alternative fuels. Here, the emphasis is on quantifying autoignition behavior for a range of spark-ignited engine conditions, including directly injected boosted conditions. The efficiency of stoichiometrically operated spark ignition engines is often limited by fuel-oxidizer end-gas autoignition, which can result in engine knock. A fuel’s knock resistance is assessed empirically by the Research Octane Number (RON) and Motor Octane Number (MON) tests. By clarifying how these two tests relate to the autoignition behavior of conventional and alternative fuel formulations, fuel design guidelines for enhanced engine efficiency can be developed.

  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. Fuels for Canadian research reactors

    International Nuclear Information System (INIS)

    Feraday, M.A.

    1993-01-01

    For a period of about 10 years AECL had a significant program looking into the possibility of developing U 3 Si as a high density replacement for the UO 2 pellet fuel in use in CANDU power reactors. The element design consisted of a Zircaloy-clad U 3 Si rod containing suitable voidage to accommodate swelling. We found that the binary U 3 Si could not meet the defect criterion for our power reactors, i.e., one month in 300 degree C water with a defect in the sheath and no significant damage to the element. Since U 3 Si could not do the job, a new corrosion resistant ternary U-Si-Al alloy was developed and patented. Fuel elements containing this alloy came close to meeting the defect criterion and showed slightly better irradiation stability than U 3 Si. Shortly after this, the program was terminated for other reasons. We have made much of this experience available to the Low Enrichment Fuel Development Program and will be glad to supply further data to assist this program

  7. Radionuclide release from research reactor spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Curtius, H., E-mail: h.curtius@fz-juelich.de [Forschungszentrum Juelich, Institut fuer Energieforschung, IEF-6 Sicherheitsforschung und Reaktortechnik, Geb. 05.3, D-52425 Juelich (Germany); Kaiser, G.; Mueller, E.; Bosbach, D. [Forschungszentrum Juelich, Institut fuer Energieforschung, IEF-6 Sicherheitsforschung und Reaktortechnik, Geb. 05.3, D-52425 Juelich (Germany)

    2011-09-01

    Numerous investigations with respect to LWR fuel under non oxidizing repository relevant conditions were performed. The results obtained indicate slow corrosion rates for the UO{sub 2} fuel matrix. Special fuel-types (mostly dispersed fuels, high enriched in {sup 235}U, cladded with aluminium) are used in German research reactors, whereas in German nuclear power plants, UO{sub 2}-fuel (LWR fuel, enrichment in {sup 235}U up to 5%, zircaloy as cladding) is used. Irradiated research reactor fuels contribute less than 1% to the total waste volume. In Germany, the state is responsible for fuel operation and for fuel back-end options. The institute for energy research (IEF-6) at the Research Center Juelich performs investigation with irradiated research reactor spent fuels under repository relevant conditions. In the study, the corrosion of research reactor spent fuel has been investigated in MgCl{sub 2}-rich salt brine and the radionuclide release fractions have been determined. Leaching experiments in brine with two different research reactor fuel-types were performed in a hot cell facility in order to determine the corrosion behaviour and the radionuclide release fractions. The corrosion of two dispersed research reactor fuel-types (UAl{sub x}-Al and U{sub 3}Si{sub 2}-Al) was studied in 400 mL MgCl{sub 2}-rich salt brine in the presence of Fe{sup 2+} under static and initially anoxic conditions. Within these experimental parameters, both fuel types corroded in the experimental time period of 3.5 years completely, and secondary alteration phases were formed. After complete corrosion of the used research reactor fuel samples, the inventories of Cs and Sr were quantitatively detected in solution. Solution concentrations of Am and Eu were lower than the solubility of Am(OH){sub 3}(s) and Eu(OH){sub 3}(s) solid phases respectively, and may be controlled by sorption processes. Pu concentrations may be controlled by Pu(IV) polymer species, but the presence of Pu(V) and Pu

  8. Scaling up microbial fuel cells and other bioelectrochemical systems

    KAUST Repository

    Logan, Bruce E.

    2009-12-15

    Scientific research has advanced on different microbial fuel cell (MFC) technologies in the laboratory at an amazing pace, with power densities having reached over 1 kW/m3 (reactor volume) and to 6.9 W/m2 (anode area) under optimal conditions. The main challenge is to bring these technologies out of the laboratory and engineer practical systems for bioenergy production at larger scales. Recent advances in new types of electrodes, a better understanding of the impact of membranes and separators on performance of these systems, and results from several new pilot-scale tests are all good indicators that commercialization of the technology could be possible within a few years. Some of the newest advances and future challenges are reviewed here with respect to practical applications of these MFCs for renewable energy production and other applications. © 2009 Springer-Verlag.

  9. Scaling up microbial fuel cells and other bioelectrochemical systems

    KAUST Repository

    Logan, Bruce E.

    2009-01-01

    Scientific research has advanced on different microbial fuel cell (MFC) technologies in the laboratory at an amazing pace, with power densities having reached over 1 kW/m3 (reactor volume) and to 6.9 W/m2 (anode area) under optimal conditions. The main challenge is to bring these technologies out of the laboratory and engineer practical systems for bioenergy production at larger scales. Recent advances in new types of electrodes, a better understanding of the impact of membranes and separators on performance of these systems, and results from several new pilot-scale tests are all good indicators that commercialization of the technology could be possible within a few years. Some of the newest advances and future challenges are reviewed here with respect to practical applications of these MFCs for renewable energy production and other applications. © 2009 Springer-Verlag.

  10. About fuel assemblies optimization in research reactor

    International Nuclear Information System (INIS)

    Malers, Yu.P.

    1992-01-01

    Ealier was considered an algorithm for optimization of fuel assembly arrangement in a research reator. The alggorithm was based on an analytical relation between distributions of energy release and fuel concentration and on the method of succesive linearization and partially integral-number programming. In the paper are solved the problems, appeared as a result of realization of the used approach and required more correct formulation of the algorithm and introduction in it some variations

  11. Modeling of large-scale oxy-fuel combustion processes

    DEFF Research Database (Denmark)

    Yin, Chungen

    2012-01-01

    Quite some studies have been conducted in order to implement oxy-fuel combustion with flue gas recycle in conventional utility boilers as an effective effort of carbon capture and storage. However, combustion under oxy-fuel conditions is significantly different from conventional air-fuel firing......, among which radiative heat transfer under oxy-fuel conditions is one of the fundamental issues. This paper demonstrates the nongray-gas effects in modeling of large-scale oxy-fuel combustion processes. Oxy-fuel combustion of natural gas in a 609MW utility boiler is numerically studied, in which...... calculation of the oxy-fuel WSGGM remarkably over-predicts the radiative heat transfer to the furnace walls and under-predicts the gas temperature at the furnace exit plane, which also result in a higher incomplete combustion in the gray calculation. Moreover, the gray and non-gray calculations of the same...

  12. 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.

  13. PCR+ In Diesel Fuels and Emissions Research

    Energy Technology Data Exchange (ETDEWEB)

    McAdams, H.T.

    2002-04-15

    In past work for the U.S. Department of Energy (DOE) and Oak Ridge National Laboratory (ORNL), PCR+ was developed as an alternative methodology for building statistical models. PCR+ is an extension of Principal Components Regression (PCR), in which the eigenvectors resulting from Principal Components Analysis (PCA) are used as predictor variables in regression analysis. The work was motivated by the observation that most heavy-duty diesel (HDD) engine research was conducted with test fuels that had been ''concocted'' in the laboratory to vary selected fuel properties in isolation from each other. This approach departs markedly from the real world, where the reformulation of diesel fuels for almost any purpose leads to changes in a number of interrelated properties. In this work, we present new information regarding the problems encountered in the conventional approach to model-building and how the PCR+ method can be used to improve research on the relationship between fuel characteristics and engine emissions. We also discuss how PCR+ can be applied to a variety of other research problems related to diesel fuels.

  14. HANARO Neutron Radiography Facility and Fuel Cell Research

    International Nuclear Information System (INIS)

    Kim, Taejoo

    2013-01-01

    Fuel cell which generates electric energy from hydrogen and oxygen is one of noticed renewable energy system because this has high efficiency and free from CO 2 . Especially, PEMFC (Polymer Electrolyte Membrane Fuel Cell) is focused by automotive companies because PEMFC, which has high power rate per volume and low operating temperature (60∼80), is suited due to the compact design and short start-up time. The water management is one of the most critical issues for fuel cell commercialization. In order to make a proper scheme for water management, thein formation of water distribution and behavior is very important. Neutron imaging is the best method to visualize the water at fuel cell and has been applied worldwide with qualitative and quantitative results. Because the NRF has large beam size (350Χ450mm 2 ) and relatively high neutron flux (2Χ107 n/cm 2 sec), it is suitable for large scale fuel cell research. Neutron imaging technique was used to investigate the water distribution and behavior in PEMFC under different operating conditions. The NRF has contributed the improvement of fuel cell performance and is one of the best choices for fuel cell study

  15. French experience in research reactor fuel transportation

    International Nuclear Information System (INIS)

    Raisonnier, Daniele

    1996-01-01

    Since 1963 Transnucleaire has safely performed a large number of national and international transports of radioactive material. Transnucleaire has also designed and supplied suitable packaging for all types of nuclear fuel cycle radioactive material from front-end and back-end products and for power or for research reactors. Transportation of spent fuel from power reactors are made on a regular and industrial basis, but this is not yet the case for the transport of spent fuel coming from research reactors. Each shipment is a permanent challenge and requires a reactive organization dealing with all the transportation issues. This presentation will explain the choices made by Transnucleaire and its associates to provide and optimize the corresponding services while remaining in full compliance with the applicable regulations and customer requirements. (author)

  16. Testing of a transport cask for research reactor spent fuel

    International Nuclear Information System (INIS)

    Mourao, Rogerio P.; Silva, Luiz Leite da; Miranda, Carlos A.; Mattar Neto, Miguel; Quintana, Jose F.A.; Saliba, Roberto O.; Novara, Oscar E.

    2011-01-01

    Since the beginning of the last decade three Latin American countries which operate research reactors - Argentina, Brazil and Chile - have been joining efforts to improve the regional capability in the management of spent fuel elements from the reactors operated in the region. As a step in this direction, a packaging for the transport of irradiated fuel from research reactors was designed by a tri-national team and a half-scale model for MTR fuel constructed in Argentina and tested in Brazil. Two test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions. Although the specimen has not successfully performed the tests, its overall performance was considered very satisfactory, and improvements are being introduced to the design. A third test sequence is planned for 2011. (author)

  17. History of research reactor fuel fabrication at Babcock and Wilcox

    International Nuclear Information System (INIS)

    Freim, James B.

    1983-01-01

    B and W Research Reactor Fuel Element facility at Lynchburg, Virginia now produces national laboratory and university fuel assemblies. The Company's 201000 square foot facility is devoted entirely to supplying research fuel and related products. B and W re-entered the research reactor fuel market in 1981

  18. Reducing enrichment of fuel for research reactors

    International Nuclear Information System (INIS)

    Kanda, Keiji; Matsuura, Shojiro.

    1980-01-01

    In research reactors, highly enriched uranium (HEU) is used as fuel for their purposes of operation. However, the United States strongly required in 1977 that these HEU should be replaced by low enrichment uranium (LEU) of 20% or less, or even in unavoidable cases, it should be replaced by medium enrichment uranium (MEU). INFCE (International Nuclear Fuel Cycle Evaluation) which started its activity just at that time decided to discuss this problem in the research reactor group of No. 8 sectional committee. Japan has been able to forward the work, taking a leading part in the international opinion because she has taken the countermeasures quickly. INFCE investigated the problem along the lines of policy that the possibility of reducing the degree of enrichment should be limited to the degree in which the core structures and equipments of research reactors will be modified as little as possible, and the change of fuel element geometry will be done within the permissible thermohydrodynamic capacity, and concluded that it might be possible in near future to reduce the degree of enrichment to about 45% MEU, while the reduction to 20% LEU might require considerable research, development and verification. On the other hand, the joint researches by Kyoto University and ANL (Argonne National Laboratory) and by Japan Atomic Energy Research Institute and ANL are being continued. IAEA has edited the guidebook (IAEA-TECDOC-233) for reducing the degree of enrichment for developing countries. (Wakatsuki, Y.)

  19. Integrated multi-scale modelling and simulation of nuclear fuels

    International Nuclear Information System (INIS)

    Valot, C.; Bertolus, M.; Masson, R.; Malerba, L.; Rachid, J.; Besmann, T.; Phillpot, S.; Stan, M.

    2015-01-01

    This chapter aims at discussing the objectives, implementation and integration of multi-scale modelling approaches applied to nuclear fuel materials. We will first show why the multi-scale modelling approach is required, due to the nature of the materials and by the phenomena involved under irradiation. We will then present the multiple facets of multi-scale modelling approach, while giving some recommendations with regard to its application. We will also show that multi-scale modelling must be coupled with appropriate multi-scale experiments and characterisation. Finally, we will demonstrate how multi-scale modelling can contribute to solving technology issues. (authors)

  20. One-fifth-scale and full-scale fuel element rocking tests

    International Nuclear Information System (INIS)

    Nau, P.V.; Olsen, B.E.

    1978-06-01

    Using 1 / 5 -scale and 1 / 1 -scale (prototype H451) fuel elements, one, two, or three stacked elements on a clamped base element were rocked from an initial release position. Relative displacement, rock-down loads, and dowel pin shear forces were measured. A scaled comparison between 1 / 5 -scale and 1 / 1 -scale results was made to evaluate the model scaling laws, and an error analysis was performed to assess the accuracy and usefulness of the test data

  1. US spent fuel research and experience

    Energy Technology Data Exchange (ETDEWEB)

    Machiels, A [EPRI and USDOE (United States)

    2012-07-01

    The structural performance of high-burnup spent fuel cladding during dry storage and transportation has been the subject of research and evaluation at EPRI for several years. The major issues addressed in this research program have included the following: Characterization and development of predictive models for damage mechanisms perceived to be potentially active during dry storage; Modeling and analysis of deformation processes during long-term dry storage; Development of cladding failure models and failure criteria, considering cladding material and physical conditions during dry storage and transportation; Failure analysis, considering end-of-dry-storage conditions, of spent fuel systems subjected to normal and accident conditions of transport, prescribed in Part 71 of Title 10 of the Code of Federal Regulations (10CFR71) While issues related to dry storage have largely been resolved, transportation issues have not, at least for spent fuel with discharge burnups greater than 45 GWd/MTU. A research program was launched in late 2002 following two NRC-industry meetings held on September 6, 2002 and October 23, 2002. The aim of the research program was to assess the performance of high-burnup spent fuel cladding under normal and accident conditions of transportation, as prescribed by 10CFR71, considering the physical characteristics and mechanical properties of cladding at the end of dry storage. The objective is to present a synthesis of the information that collectively forms a part of a technical basis intended to facilitate resolution of regulatory issues associated with the transportation of spent nuclear fuel characterized by discharge burnups greater than 45 GWd/MTU.

  2. Research reactors fuel cycle problems and dilemma

    International Nuclear Information System (INIS)

    Romano, R.

    2004-01-01

    During last 10 years, some problems appeared in different steps of research reactors fuel cycle. Actually the majority of these reactors have been built in the 60s and these reactors were operated during all this long period in a cycle with steps which were dedicated to this activity. Progressively and for reasons often economical, certain steps of the cycle became more and more difficult to manage due to closing of some specialised workshops in the activities of scraps recycling, irradiated fuel reprocessing, even fuel fabrication. Other steps of the cycle meet or will meet difficulties, in particular supplying of fissile raw material LEU or HEU because this material was mostly produced in enrichment units existing mainly for military reason. Rarefaction of fissile material lead to use more and more enriched uraniums said 'of technical quality', that is to say which come from mixing of varied qualities of enriched material, containing products resulting from reprocessing. Actually, problems of end of fuel cycle are increased, either consisting of intermediary storage on the site of reactor or on specialised sites, or consisting of reprocessing. This brief summary shows most difficulties which are met today by a major part of industrials of the fuel cycle in the exercise of their activities

  3. 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

  4. Biorefinery and Hydrogen Fuel Cell Research

    Energy Technology Data Exchange (ETDEWEB)

    K.C. Das; Thomas T. Adams; Mark A. Eiteman; John Stickney; Joy Doran Peterson; James R. Kastner; Sudhagar Mani; Ryan Adolphson

    2012-06-12

    In this project we focused on several aspects of technology development that advances the formation of an integrated biorefinery. These focus areas include: [1] establishment of pyrolysis processing systems and characterization of the product oils for fuel applications, including engine testing of a preferred product and its pro forma economic analysis; [2] extraction of sugars through a novel hotwater extaction process, and the development of levoglucosan (a pyrolysis BioOil intermediate); [3] identification and testing of the use of biochar, the coproduct from pyrolysis, for soil applications; [4] developments in methods of atomic layer epitaxy (for efficient development of coatings as in fuel cells); [5] advancement in fermentation of lignocellulosics, [6] development of algal biomass as a potential substrate for the biorefinery, and [7] development of catalysts from coproducts. These advancements are intended to provide a diverse set of product choices within the biorefinery, thus improving the cost effectiveness of the system. Technical effectiveness was demonstrated in the pyrolysis biooil based diesel fuel supplement, sugar extraction from lignocelluose, use of biochar, production of algal biomass in wastewaters, and the development of catalysts. Economic feasibility of algal biomass production systems seems attractive, relative to the other options. However, further optimization in all paths, and testing/demonstration at larger scales are required to fully understand the economic viabilities. The various coproducts provide a clear picture that multiple streams of value can be generated within an integrated biorefinery, and these include fuels and products.

  5. Main results and status of the development of LEU fuel for Russian research reactors

    International Nuclear Information System (INIS)

    Vatulin, A.; Morozov, A.; Suprun, V.; Dobrikova, I.

    2005-01-01

    VNIINM develops low enrichment uranium (LEU) fuel on base U-Mo alloys and a novel design of pin-type fuel elements. The development is carried out both for existing reactors, and for new advanced designs of reactors. The work is carried on the following main directions: - irradiate LEU U-Mo dispersion fuel (the uranium density up to 6,0 g/cm 3 ) in two Russian research reactors: MIR (RIAR, Dimitrovgrad) as pin type fuel mini-elements and in WWR-M (PINP, Gatchina) within full-scaled fuel assembly (FA) with pin type fuel elements; - finalize development of design and fabrication process of IRT type FA with pin type fuel elements; - develop methods of reducing of U-Mo fuel --Al matrix interaction under irradiation; - develop fabricating methods of fuel elements on base of monolithic U-Mo fuel. The paper generally reviews the results of calculation, design and technology investigations accomplished by now. (author)

  6. Concepts for Small-Scale Testing of Used Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Marschman, Steven Craig [Idaho National Lab. (INL), Idaho Falls, ID (United States); Winston, Philip Lon [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    This report documents a concept for a small-scale test involving between one and three Boiling Water Rector (BWR) high burnup (HBU) fuel assemblies. This test would be similar to the DOE funded High Burn-Up (HBU) Confirmatory Data Project to confirm the behavior of used high burn-up fuel under prototypic conditions, only on a smaller scale. The test concept proposed would collect data from fuel stored under prototypic dry storage conditions to mimic, as closely as possible, the conditions HBU UNF experiences during all stages of dry storage: loading, cask drying, inert gas backfilling, and transfer to an Independent Spent Fuel Storage Installation (ISFSI) for multi-year storage.

  7. Fuel pin integrity assessment under large scale transients

    International Nuclear Information System (INIS)

    Dutta, B.K.

    2006-01-01

    The integrity of fuel rods under normal, abnormal and accident conditions is an important consideration during fuel design of advanced nuclear reactors. The fuel matrix and the sheath form the first barrier to prevent the release of radioactive materials into the primary coolant. An understanding of the fuel and clad behaviour under different reactor conditions, particularly under the beyond-design-basis accident scenario leading to large scale transients, is always desirable to assess the inherent safety margins in fuel pin design and to plan for the mitigation the consequences of accidents, if any. The severe accident conditions are typically characterized by the energy deposition rates far exceeding the heat removal capability of the reactor coolant system. This may lead to the clad failure due to fission gas pressure at high temperature, large- scale pellet-clad interaction and clad melting. The fuel rod performance is affected by many interdependent complex phenomena involving extremely complex material behaviour. The versatile experimental database available in this area has led to the development of powerful analytical tools to characterize fuel under extreme scenarios

  8. Research reactor spent fuel management in Argentina

    International Nuclear Information System (INIS)

    Audero, M.A.; Bevilacqua, A.M.; Mehlich, A.M.; Novara, O.

    2002-01-01

    The research reactor spent fuel (RRSF) management strategy will be presented as well as the interim storage experience. Currently, low-enriched uranium RRSF is in wet interim storage either at reactor site or away from reactor site in a centralized storage facility. High-enriched uranium RRSF from the centralized storage facility has been sent to the USA in the framework of the Foreign Research Reactor Spent Nuclear Fuel Acceptance Program. The strategy for the management of the RRSF could implement the encapsulation for interim dry storage. As an alternative to encapsulation for dry storage some conditioning processes are being studied which include decladding, isotopic dilution, oxidation and immobilization. The immobilized material will be suitable for final disposal. (author)

  9. Criticality evaluation of long term for spent fuel, using Scale

    International Nuclear Information System (INIS)

    Esquivel E, J.; Vargas E, S.; Ramirez S, J. R.

    2013-10-01

    Once carried out the spent fuel discharge, of the reactor core, this continues generating decay heat and diverse fission products, reason why is important to store this fuel inside containers able to dissipate the heat generated by the isotopes decay of the fuel and to maintain the fuels arrangement in subcritical condition. This means that: is necessary to assure the sub-criticality of those fuel assemblies in the time. This work, presents a criticality evaluation of fuel assemblies type PWR in a storage generic container. For this purpose have been used two codes: GeeWiz, to carry out the geometric model of the container with the fuel assemblies, and Keno, with which, the criticality of the full container with fuel is determined until a 10 6 years period. These codes are part of the package Scale. The specifications for each one of the analyzed components are based on a Benchmark document of the Nea/OECD, of where, the results that reports are compared with the obtained results by the realized analysis. (Author)

  10. Qualification of high-density fuel manufacturing for research reactors at CNEA

    Energy Technology Data Exchange (ETDEWEB)

    Adelfang, P.; Alvarez, L.; Boero, N.; Calabrese, R.; De La Fuente, M.; Echenique, P.; Markiewicz, M.; Pasqualini, E.; Ruggirello, G.; Taboada, H. [CNEA, Buenos Aires (Argentina)

    2001-07-01

    CNEA, the National Atomic Energy Commission of Argentina, is at the present a qualified supplier of uranium oxide fuel for research reactors. A new objective in this field is to develop and qualify the manufacturing of LEU high-density fuel for this type of reactors. According with the international trend Silicide fuel and U-xMo fuel are included in our program as the most suitable options. The facilities to complete the qualification of high-density MTR fuels, like the manufacturing plant installations, the reactor, the pool side fuel examination station and the hot cells are fully operational and equipped to perform all the activities required within the program. The programs for both type of fuels include similar activities: development and set up of the fuel material manufacturing technology, set up of fuel plate manufacturing, fabrication and irradiation of miniplates, fabrication and irradiation of full scale fuel elements, post-irradiation examination and feedback for manufacturing improvements. For silicide fuels most of these steps have already been completed. For U-xMo fuel the activities also include the development of alternative ways to obtain U-xMo powder, feasibility studies for large-scale manufacturing and the economical assessment. Set up of U-xMo fuel plate manufacturing is also well advanced and the fabrication of the first full scale prototype is foreseen during this year. (author)

  11. Administrative mechanics of research fuel transportation

    International Nuclear Information System (INIS)

    Harmon, Diane W.

    1983-01-01

    This presentation contains the discussion on the multitude of administrative mechanics that have to be meshed for the successful completion of a shipment of spent fuel, HEU or LEU in the research reactors fuel cycle. The costs associated with transportation may be the equivalent of 'a black hole', so an overview of cost factors is given. At the end one could find that this black hole factor in the budget is actually a bargain. The first step is the quotation phase. The cost variables in the quotation contain the cost of packaging i.e. containers; the complete routing of the packages and the materials. Factors that are of outmost importance are the routing restrictions and regulations, physical security regulations. All of this effort is just to provide a valid quotation not to accomplish the goal of completing a shipment. Public relations cannot be omitted either

  12. Administrative mechanics of research fuel transportation

    Energy Technology Data Exchange (ETDEWEB)

    Harmon, Diane W [Edlow International Company, Washington, DC (United States)

    1983-09-01

    This presentation contains the discussion on the multitude of administrative mechanics that have to be meshed for the successful completion of a shipment of spent fuel, HEU or LEU in the research reactors fuel cycle. The costs associated with transportation may be the equivalent of 'a black hole', so an overview of cost factors is given. At the end one could find that this black hole factor in the budget is actually a bargain. The first step is the quotation phase. The cost variables in the quotation contain the cost of packaging i.e. containers; the complete routing of the packages and the materials. Factors that are of outmost importance are the routing restrictions and regulations, physical security regulations. All of this effort is just to provide a valid quotation not to accomplish the goal of completing a shipment. Public relations cannot be omitted either.

  13. The scale analysis sequence for LWR fuel depletion

    International Nuclear Information System (INIS)

    Hermann, O.W.; Parks, C.V.

    1991-01-01

    The SCALE (Standardized Computer Analyses for Licensing Evaluation) code system is used extensively to perform away-from-reactor safety analysis (particularly criticality safety, shielding, heat transfer analyses) for spent light water reactor (LWR) fuel. Spent fuel characteristics such as radiation sources, heat generation sources, and isotopic concentrations can be computed within SCALE using the SAS2 control module. A significantly enhanced version of the SAS2 control module, which is denoted as SAS2H, has been made available with the release of SCALE-4. For each time-dependent fuel composition, SAS2H performs one-dimensional (1-D) neutron transport analyses (via XSDRNPM-S) of the reactor fuel assembly using a two-part procedure with two separate unit-cell-lattice models. The cross sections derived from a transport analysis at each time step are used in a point-depletion computation (via ORIGEN-S) that produces the burnup-dependent fuel composition to be used in the next spectral calculation. A final ORIGEN-S case is used to perform the complete depletion/decay analysis using the burnup-dependent cross sections. The techniques used by SAS2H and two recent applications of the code are reviewed in this paper. 17 refs., 5 figs., 5 tabs

  14. Dry Storage of Research Reactor Spent Nuclear Fuel - 13321

    Energy Technology Data Exchange (ETDEWEB)

    Adams, T.M.; Dunsmuir, M.D.; Leduc, D.R.; Severynse, T.F.; Sindelar, R.L. [Savannah River National Laboratory (United States); Moore, E.N. [Moore Nuclear Energy, LLC (United States)

    2013-07-01

    Spent fuel from domestic and foreign research reactors is received and stored at the Savannah River Site's L Area Material Storage (L Basin) Facility. This DOE-owned fuel consists primarily of highly enriched uranium in metal, oxide or silicide form with aluminum cladding. Upon receipt, the fuel is unloaded and transferred to basin storage awaiting final disposition. Disposition alternatives include processing via the site's H Canyon facility for uranium recovery, or packaging and shipment of the spent fuel to a waste repository. A program has been developed to provide a phased approach for dry storage of the L Basin fuel. The initial phase of the dry storage program will demonstrate loading, drying, and storage of fuel in twelve instrumented canisters to assess fuel performance. After closure, the loaded canisters are transferred to pad-mounted concrete overpacks, similar to those used for dry storage of commercial fuel. Unlike commercial spent fuel, however, the DOE fuel has high enrichment, very low to high burnup, and low decay heat. The aluminum cladding presents unique challenges due to the presence of an oxide layer that forms on the cladding surface, and corrosion degradation resulting from prolonged wet storage. The removal of free and bound water is essential to the prevention of fuel corrosion and radiolytic generation of hydrogen. The demonstration will validate models predicting pressure, temperature, gas generation, and corrosion performance, provide an engineering scale demonstration of fuel handling, drying, leak testing, and canister backfill operations, and establish 'road-ready' storage of fuel that is suitable for offsite repository shipment or retrievable for onsite processing. Implementation of the Phase I demonstration can be completed within three years. Phases II and III, leading to the de-inventory of L Basin, would require an additional 750 canisters and 6-12 years to complete. Transfer of the fuel from basin storage

  15. Fuel cycle industrialization program prepared by N-Fuel Research Committee, ANRE

    Energy Technology Data Exchange (ETDEWEB)

    1978-09-01

    To meet the new situation resulting from the scaling down of nuclear power development plan in Japan, and the changes due to the new U.S. nuclear non-proliferation policy, the Nuclear Fuel Research Committee of the Agency of Natural Resources and Energy of MITI has prepared the ''Interim Report on the Nuclear Fuel Cycle''. It sets out in precise terms the methods that should be followed for establishing the nuclear fuel cycle in Japan. Major items treated in this report are; uranium ore development, promotion of uranium stockpiling, construction of domestic uranium enrichment plant, promotion of the construction of a nuclear fuel park, Pu utilization and cooperation in international movement for nuclear non-proliferation, and the establishment of measures for radioactive waste management. Discussions are made from technological, economical, and political view points. Also attached are a table of the comprehensive industrialization plan up to the year 2000 and a table of estimated nuclear fuel demand and supply in Japan.

  16. Performance of a small-scale turbojet engine fed with traditional and alternative fuels

    International Nuclear Information System (INIS)

    Badami, M.; Nuccio, P.; Pastrone, D.; Signoretto, A.

    2014-01-01

    Highlights: • One synthetic fuel, one bio-fuel blend and one traditional kerosene were tested on a small-scale turbojet. • The three fuels were characterized and their chemical and physical specifications were compared. • The experimental results, in term of CO, UHC and NO x emissions, are discussed. • Emissions are compared with CFD analysis and semi-empirical equations found in literature. - Abstract: The growing use of aeronautical fuels, together with the increasing attention towards reduced environmental impact technologies, are driving governments and aeronautical companies to search for alternative aeronautical fuels to petroleum based kerosene. Analysis and measurements on real aeronautical turbines are complex and expensive; for this reason the use of parts of real engines or small size turbojets can be very interesting for research activities. The present paper describes the results of an experimental and numerical activity, regarding the influence of alternative fuels on the performance in term of emissions, on a small-scale turbojet engine with a nominal thrust of 80 N at 80,000 rpm. Three different fuels, a traditional Jet-A kerosene, a synthetic Gas To Liquid (GTL) fuel and a blend of 30% Jatropha Methyl Ester (JME) and 70% Jet-A, were tested. The experimental results, in term of CO, UHC and NO x emissions, are discussed and compared with results obtained from CFD analysis and from semi-empirical equations found in literature

  17. Fuel handling alternatives to prepare for large scale fuel channel replacement

    International Nuclear Information System (INIS)

    Martire, S.; Sandu, I.

    2007-01-01

    It is desirable to reduce the duration of defuelling the reactor in preparation for retube, as the cost of replacement power is $750K/day. Three fast defuelling concepts are presented. With the Through Flow Defuelling method, the fuel string is hydraulically pushed into the downstream Fuelling Machine (FM) by flow passing through the fuel channel. The Long Stroke C Ram method replaces the FM C Ram with a longer one capable of pushing all fuel bundles into the receiving FM. Defuelling Hardware uses enhanced design of ram extensions that interconnect mechanically to extend the Ram stroke to push fuel bundles into the receiving FM. This paper will present descriptions of each defuelling concept to prepare for Large Scale Fuel Channel Replacement. Advantages and disadvantages of each concept will be discussed and a recommendation will be made for future implementation. (author)

  18. From fuel cells to batteries: Synergies, scales and simulation methods

    OpenAIRE

    Bessler, Wolfgang G.

    2011-01-01

    The recent years have shown a dynamic growth of battery research and development activities both in academia and industry, supported by large governmental funding initiatives throughout the world. A particular focus is being put on lithium-based battery technologies. This situation provides a stimulating environment for the fuel cell modeling community, as there are considerable synergies in the modeling and simulation methods for fuel cells and batteries. At the same time, batter...

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  20. Status of research reactor spent fuel world-wide

    International Nuclear Information System (INIS)

    Ritchie, I.G.

    2004-01-01

    Results compiled in the research reactor spent fuel database are used to assess the status of research reactor spent fuel world-wide. Fuel assemblies, their types, enrichment, origin of enrichment and geological distribution among the industrialised and developed countries of the world are discussed. Fuel management practices in wet and dry storage facilities and the concerns of reactor operators about long-term storage of their spent fuel are presented and some of the activities carried out by the International Atomic Energy Agency to address the issues associated with research reactor spent fuel are outlined. (author)

  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. Fuels and Combustion | Transportation Research | NREL

    Science.gov (United States)

    Fuels and Combustion Fuels and Combustion This is the March 2015 issue of the Transportation and , combustion strategy, and engine design hold the potential to maximize vehicle energy efficiency and performance of low-carbon fuels in internal combustion engines with a whole-systems approach to fuel chemistry

  4. Overview of remote technologies applied to research reactor fuel

    International Nuclear Information System (INIS)

    Oerdoegh, M.; Takats, F.

    1999-01-01

    This paper gives a brief overview of the remote technologies applied to research reactor fuels. Due to many reasons, the remote technology utilization to research reactor fuel is not so widespread as it is for power reactor fuels, however, the advantages of the application of such techniques are obvious. (author)

  5. Social science to improve fuels management: a synthesis of research on aesthetics and fuels management

    Science.gov (United States)

    Robert L. Ryan

    2005-01-01

    A series of syntheses were commissioned by the USDA Forest Service to aid in fuels mitigation project planning. This synthesis focuses on research addressing aesthetic considerations of fuels management. A general finding is that fuels management activities can contribute to the visual quality of a landscape. Topics covered in the synthesis include research findings on...

  6. Management and storage of spent fuel from CEA research reactors

    International Nuclear Information System (INIS)

    Merchie, F.

    1996-01-01

    CEA research reactors and their interim spent fuel storage facilities are described. Long-term solutions for spent fuel storage problems, involving wet storage at PEGASE or dry storage at CASCAD, are outlined in some detail. (author)

  7. Thermal performance of a meso-scale liquid-fuel combustor

    International Nuclear Information System (INIS)

    Vijayan, V.; Gupta, A.K.

    2011-01-01

    Research highlights: → Demonstrated successful combustion of liquid fuel-air mixtures in a novel meso-scale combustor. → Flame quenching was eliminated using heat recirculation in a swiss roll type combustor that also extended the flammability limits. → Liquid fuel was rapidly vaporized with the use of hot narrow channel walls that eliminated the need of a fuel atomizer. → Maximum power density of the combustor was estimated to be about 8.5 GW/m3 and heat load in the range of 50-280W. → Overall efficiency of the combustor was estimated in the range of 12 to 20%. - Abstract: Combustion in small scale devices poses significant challenges due to the quenching of reactions from wall heat losses as well as the significantly reduced time available for mixing and combustion. In the case of liquid fuels there are additional challenges related to atomization, vaporization and mixing with the oxidant in the very short time-scale liquid-fuel combustor. The liquid fuel employed here is methanol with air as the oxidizer. The combustor was designed based on the heat recirculating concept wherein the incoming reactants are preheated by the combustion products through heat exchange occurring via combustor walls. The combustor was fabricated from Zirconium phosphate, a ceramic with very low thermal conductivity (0.8 W m -1 K -1 ). The combustor had rectangular shaped double spiral geometry with combustion chamber in the center of the spiral formed by inlet and exhaust channels. Methanol and air were introduced immediately upstream at inlet of the combustor. The preheated walls of the inlet channel also act as a pre-vaporizer for liquid fuel which vaporizes the liquid fuel and then mixes with air prior to the fuel-air mixture reaching the combustion chamber. Rapid pre-vaporization of the liquid fuel by the hot narrow channel walls eliminated the necessity for a fuel atomizer. Self-sustained combustion of methanol-air was achieved in a chamber volume as small as 32.6 mm 3

  8. Using fine-scale fuel measurements to assess wildland fuels, potential fire behavior and hazard mitigation treatments in the southeastern USA

    International Nuclear Information System (INIS)

    Ottmar, Roger D.; Blake, John I.; Crolly, William T.

    2012-01-01

    The inherent spatial and temporal heterogeneity of fuelbeds in forests of the southeastern United States may require fine scale fuel measurements for providing reliable fire hazard and fuel treatment effectiveness estimates. In a series of five papers, an intensive, fine scale fuel inventory from the Savanna River Site in the southeastern United States is used for building fuelbeds and mapping fire behavior potential, evaluating fuel treatment options for effectiveness, and providing a comparative analysis of landscape modeled fire behavior using three different data sources including the Fuel Characteristic Classification System, LANDFIRE, and the Southern Wildfire Risk Assessment. The research demonstrates that fine scale fuel measurements associated with fuel inventories repeated over time can be used to assess broad scale wildland fire potential and hazard mitigation treatment effectiveness in the southeastern USA and similar fire prone regions. Additional investigations will be needed to modify and improve these processes and capture the true potential of these fine scale data sets for fire and fuel management planning.

  9. Research of power fuel low-temperature vortex combustion in industrial boiler based on numerical modelling

    Directory of Open Access Journals (Sweden)

    Orlova K.Y.

    2017-01-01

    Full Text Available The goal of the presented research is to perform numerical modelling of fuel low-temperature vortex combustion in once-through industrial steam boiler. Full size and scaled-down furnace model created with FIRE 3D software and was used for the research. All geometrical features were observed. The baseline information for the low-temperature vortex furnace process are velocity and temperature of low, upper and burner blast, air-fuel ratio, fuel consumption, coal dust size range. The obtained results are: temperature and velocity three dimensional fields, furnace gases and solid fuel ash particles concentration.

  10. Multi-scale modelling of the physicochemical-mechanical coupling of fuel behaviour at high temperature in pressurized water reactors

    International Nuclear Information System (INIS)

    Julien, Jerome

    2008-01-01

    Within the frame of the problematic of pellet-sheath interaction in a nuclear fuel rod, a good description of the fuel thermo-mechanical behaviour is required. This research thesis reports the coupling of physics-chemistry (simulation of gas transfers between different cavities) and mechanics (assessment of fuel viscoplastic strains). A new micromechanical model is developed which uses a multi-scale approach to describe the evolution of the double population of cavities (cavities with two different scales) while taking internal pressures as well as the fuel macroscopic viscoplastic behaviour into account. The author finally describes how to couple this micromechanical mode to physics-chemistry models [fr

  11. The use of medium enriched uranium fuel for research reactors

    International Nuclear Information System (INIS)

    1979-01-01

    The evaluation described in the present paper concerns the use of medium enriched uranium fuel for our research reactors. The underlying assumptions set up for the evaluation are as follows: (1) At first, the use of alternative fuel should not affect, even to a small extent, research and development programs in nuclear energy utilization, which were described in the previous paper. Hence the use of lower enrichment fuel should not cause any reduction in reactor performances. (2) The fuel cycle cost for operating research reactors with alternative fuel, excepting R and D cost for such fuel, should not increase beyond an acceptable limit. (3) The use of alternative fuel should be satisfactory with respect to non-proliferation purposes, to the almost same degree as the use of 20% enriched uranium fuel

  12. Comparative Study of Laboratory-Scale and Prototypic Production-Scale Fuel Fabrication Processes and Product Characteristics

    International Nuclear Information System (INIS)

    Marshall, Douglas W.

    2014-01-01

    An objective of the High Temperature Gas Reactor fuel development and qualification program for the United States Department of Energy has been to qualify fuel fabricated in prototypic production-scale equipment. The quality and characteristics of the tristructural isotropic (TRISO) coatings on fuel kernels are influenced by the equipment scale and processing parameters. The standard deviations of some TRISO layer characteristics were diminished while others have become more significant in the larger processing equipment. The impact on statistical variability of the processes and the products, as equipment was scaled, are discussed. The prototypic production-scale processes produce test fuels meeting all fuel quality specifications. (author)

  13. 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.

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  15. Emissions from small scale biomass combustion - Research needs

    International Nuclear Information System (INIS)

    Gustavsson, L.; Karlsson, M.L.; Larfeldt, J.; Leckner, B.

    1994-01-01

    Earlier investigations have shown that small scale biomass combustion leads to unacceptable emissions in the air. The most important problem is high levels of unburnt hydrocarbons. This report analyzes which are the most important reasons to these emissions and which research efforts that are necessary to increase the knowledge about the combustion processes, thereby promoting the development of environmentally feasible equipment. The following factors are defined as most crucial to emission levels: size of combustion chamber, air excess ratio, means of combustion air supply, mixing between air and fuel, transient events, and fuel quality. It is concluded that both basic and research within the area is needed. More specific, research in the form of systematic analysis of best available technology, reactor experiments, compilation of knowledge about relevant basic combustion processes, mathematical modelling as well as development of measurement techniques are called for. 15 refs, 11 figs, 1 tab

  16. The risks of the Taiwan research reactor spent fuel project

    International Nuclear Information System (INIS)

    1991-06-01

    The proposed action is to transport up to 118 spent fuel rods, to include canned spent fuel rod particulates immobilized on filters, from a research reactor in Taiwan by sea to Hampton Roads, Virginia, and then overland by truck to the Receiving Basin for Offsite Fuels at the Savannah River Site (SRS). At SRS, the spent fuel will be reprocessed to recover uranium and plutonium. 55 refs., 8 tabs

  17. Research on bioorganic fuels as power sources

    Energy Technology Data Exchange (ETDEWEB)

    Lampinen, Markku J.; Spets, Jukka-Pekka [Aalto University, Department of Energy Technology-TKK, Applied Thermodynamics, PO Box 4400, FI-02201 TKK (Finland); Kiros, Yohannes [Royal Institute of Technology-KTH, Department of Chemical Engineering and Technology, S100-44 Stockholm (Sweden); Anttila, Tomi [Oy Hydrocell Ltd, Minkkikatu 1-3, 04430 Jaervenpaeae (Finland)

    2010-11-15

    This paper deals with the kind of the bioorganic fuel cells that are equipped with or without ion exchange membranes. The bioorganic materials of interest are alcohols (methanol, ethanol) and glucose, which are obtained from renewable energy sources such as biomass. The operation temperatures of the direct fuel cells cover from room temperature up to 150 C. The direct bioorganic fuel cells belong to the subject area of 'Advanced fuel cells' of the Working group 4 in the EU COST Action 543 among the collaborating Universities and Institutes. Bioorganic fuel cells are suitable for application in small portable power sources, such as backups, battery chargers and in electronic devices. A number of current and earlier works are summarised and advances are highlighted in this area with special emphasis on glucose as a fuel. (author)

  18. Scaling Research Results: Design and Evaluation | IDRC ...

    International Development Research Centre (IDRC) Digital Library (Canada)

    Design and evaluation The project will provide helpful guidance to IDRC management and ... scaling and programming for scalable research Offer the monograph in multiple forms, ... Asian outlook: New growth dependent on new productivity.

  19. Complete Flow Blockage of a Fuel Channel for Research Reactor

    International Nuclear Information System (INIS)

    Lee, Byeonghee; Park, Suki

    2015-01-01

    The CHF correlation suitable for narrow rectangular channels are implemented in RELAP5/MOD3.3 code for the analyses, and the behavior of fuel temperatures and MCHFR(minimum critical heat flux ratio) are compared between the original and modified codes. The complete flow blockage of fuel channel for research reactor is analyzed using original and modified RELAP5/MOD3.3 and the results are compared each other. The Sudo-Kaminaga CHF correlation is implemented into RELAP5/MOD3.3 for analyzing the behavior of fuel adjacent to the blocked channel. A flow blockage of fuel channels can be postulated by a foreign object blocking cooling channels of fuels. Since a research reactor with plate type fuel has isolated fuel channels, a complete flow blockage of one fuel channel can cause a failure of adjacent fuel plates by the loss of cooling capability. Although research reactor systems are designed to prevent foreign materials from entering into the core, partial flow blockage accidents and following fuel failures are reported in some old research reactors. In this report, an analysis of complete flow blockage accident is presented for a 15MW pool-type research reactor with plate type fuels. The fuel surface experience different heat transfer regime in the results from original and modified RELAP5/MOD3.3. By the discrepancy in heat transfer mode of two cases, a fuel melting is expected by the modified RELAP5/MOD3.3, whereas the fuel integrity is ensured by the original code

  20. Political consultation and large-scale research

    International Nuclear Information System (INIS)

    Bechmann, G.; Folkers, H.

    1977-01-01

    Large-scale research and policy consulting have an intermediary position between sociological sub-systems. While large-scale research coordinates science, policy, and production, policy consulting coordinates science, policy and political spheres. In this very position, large-scale research and policy consulting lack of institutional guarantees and rational back-ground guarantee which are characteristic for their sociological environment. This large-scale research can neither deal with the production of innovative goods under consideration of rentability, nor can it hope for full recognition by the basis-oriented scientific community. Policy consulting knows neither the competence assignment of the political system to make decisions nor can it judge succesfully by the critical standards of the established social science, at least as far as the present situation is concerned. This intermediary position of large-scale research and policy consulting has, in three points, a consequence supporting the thesis which states that this is a new form of institutionalization of science: These are: 1) external control, 2) the organization form, 3) the theoretical conception of large-scale research and policy consulting. (orig.) [de

  1. Advances and highlights of the CNEA qualification program as high density fuel manufacturer for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Adelfang, P.; Alvarez, L.; Boero, N.; Calabrese, R.; Echenique, P.; Markiewicz, M.; Pasqualini, E.; Ruggirello, G.; Taboada, H. [Unidad de Actividad Combustibles Nucleares Comision Nacional de Energia Atomica (CNE4), Avda. del Libertador, 8250 C1429BNO Buenos Aires (Argentina)

    2002-07-01

    One of the main objectives of CNEA regarding the fuel for research reactors is the development and qualification of the manufacturing of LEU high-density fuels. The qualification programs for both types of fuels, Silicide fuel and U- x Mo fuel, are similar. They include the following activities: development and set up of the fissile compound manufacturing technology, set up of fuel plate manufacturing, fabrication and irradiation of mini plates and plates, design and fabrication of fuel assembly prototypes for irradiation, post-irradiation examination and feedback for manufacturing improvements. This paper describes the different activities performed within each program during the last year and the main advances and achievements of the programs within this period. The main achievements may be summarized in the following activities: Continuation of the irradiation of the first silicide fuel element in the R A3. Completion of the manufacturing of the second silicide fuel element, licensing and beginning of its irradiation in the R A3. Development of the HMD Process to manufacture U-Mo powder (pUMA project). Set up of fuel plates manufacturing at industrial level using U-Mo powder. Preliminary studies and the design for the irradiation of mini plates, plates and full scale fuel elements with U-Mo and 7 g U/cm{sup 3}. PIE destructive studies for the P-04 silicide fuel prototype (accurate burnup determination through chemical analysis, metallography and SEM of samples from the irradiated fuel plates). Improvement and development of new characterization techniques for high density fuel plates quality control including US testing and densitometric analysis of X-ray examinations. The results obtained in this period are encouraging and also allow to foresee a wider participation of CNEA in the international effort to qualify U-Mo as a new material for the manufacturing of research reactor fuels. (author)

  2. Advances and highlights of the CNEA qualification program as high density fuel manufacturer for research reactors

    International Nuclear Information System (INIS)

    Adelfang, P.; Alvarez, L.; Boero, N.; Calabrese, R.; Echenique, P.; Markiewicz, M.; Pasqualini, E.; Ruggirello, G.; Taboada, H.

    2002-01-01

    One of the main objectives of CNEA regarding the fuel for research reactors is the development and qualification of the manufacturing of LEU high-density fuels. The qualification programs for both types of fuels, Silicide fuel and U- x Mo fuel, are similar. They include the following activities: development and set up of the fissile compound manufacturing technology, set up of fuel plate manufacturing, fabrication and irradiation of mini plates and plates, design and fabrication of fuel assembly prototypes for irradiation, post-irradiation examination and feedback for manufacturing improvements. This paper describes the different activities performed within each program during the last year and the main advances and achievements of the programs within this period. The main achievements may be summarized in the following activities: Continuation of the irradiation of the first silicide fuel element in the R A3. Completion of the manufacturing of the second silicide fuel element, licensing and beginning of its irradiation in the R A3. Development of the HMD Process to manufacture U-Mo powder (pUMA project). Set up of fuel plates manufacturing at industrial level using U-Mo powder. Preliminary studies and the design for the irradiation of mini plates, plates and full scale fuel elements with U-Mo and 7 g U/cm 3 . PIE destructive studies for the P-04 silicide fuel prototype (accurate burnup determination through chemical analysis, metallography and SEM of samples from the irradiated fuel plates). Improvement and development of new characterization techniques for high density fuel plates quality control including US testing and densitometric analysis of X-ray examinations. The results obtained in this period are encouraging and also allow to foresee a wider participation of CNEA in the international effort to qualify U-Mo as a new material for the manufacturing of research reactor fuels. (author)

  3. Fuels for research and test reactors, status review: July 1982

    International Nuclear Information System (INIS)

    Stahl, D.

    1982-12-01

    A thorough review is provided on nuclear fuels for steady-state thermal research and test reactors. The review was conducted to provide a documented data base in support of recent advances in research and test reactor fuel development, manufacture, and demonstration in response to current US policy on availability of enriched uranium. The review covers current fabrication practice, fabrication development efforts, irradiation performance, and properties affecting fuel utilization, including thermal conductivity, specific heat, density, thermal expansion, corrosion, phase stability, mechanical properties, and fission-product release. The emphasis is on US activities, but major work in Europe and elsewhere is included. The standard fuel types discussed are the U-Al alloy, UZrH/sub x/, and UO 2 rod fuels. Among new fuels, those given major emphasis include H 3 Si-Al dispersion and UO 2 caramel plate fuels

  4. Bench-scale and full-scale studies of nitric oxides reduction by gaseous fuel reburning

    International Nuclear Information System (INIS)

    Su, S.; Xiang, J.; Sun, L.S.; Hu, S.; Zhu, J.M.

    2008-01-01

    Nitrogen oxides (NOx) emissions from coal-fired boilers are significant contributors to atmospheric pollution. China has specified more rigorous legal limits for NOx emissions from power plants. As a result of the need to reduce NOx emissions, cost-effective NOx reduction strategies must be explored. This paper presented detailed experimental studies on a gaseous fuel reburning process that was performed in a 36 kilowatt bench-scale down-fired furnace to define the optimal reburning operating conditions when different Chinese coals were fired in the furnace. In addition, the combustion system of a 350 megawatt full-scale boiler was retrofitted according to the experimental results. Finally, the gaseous fuel reburning was applied to the retrofitted full-scale boiler. The purpose of the study was to obtain a better understanding of the influence of the key parameters on nitric oxide (NO) reduction efficiency of the reburning process and demonstrate the gaseous fuel reburning on a 350 MWe coal-fired boiler in China. The paper described the experimental procedure with particular reference to the experimental facility and measurement; a schematic diagram of the experimental system; experimental fuels; and characteristics of coals for the reburning experiments. Results that were presented included influence of reburn zone residence time; influence of gaseous reburn fuel per cent; influence of excess air coefficient; and unburned carbon in fly ash. It was concluded that both an above 50 per cent NO reduction efficiency and low carbon loss can be obtained by the gaseous fuel reburning process under the optimal operating conditions. 20 refs., 5 tabs., 10 figs

  5. Back-end of the research reactor fuel cycle

    International Nuclear Information System (INIS)

    Gruber, Gehard J.

    1996-01-01

    This paper outlines the status of topics and issues related to: (1) Research Reactor Spent Nuclear Fuel Return to the U.S., including policy, shipments and ports of entry, management sites, fees, storage technologies, contracts, actual shipment, and legal process, (2) UKAEA: MTR Spent Nuclear Fuel Reprocessing, (3) COGEMA: MTR Spent Nuclear Fuel Reprocessing, and (4) Intermediate Storage + Direct Disposal for Research Reactors. (author)

  6. New fuels for proliferation-safe research reactors. SCK-CEN a kingpin in international research

    International Nuclear Information System (INIS)

    Van den Berghe, S.

    2011-01-01

    A number of research reactors in the world, including BR2, are still working on highly enriched proliferation sensitive nuclear fuel. Extremely high-density fuel is required to make possible the switch to low-enriched fuel without loss of effectiveness. In the past few years, SCK-CEN has developed into the global test station for new, low-enriched fuels.

  7. Storage experience in Hungary with fuel from research reactors

    International Nuclear Information System (INIS)

    Gado, J.; Hargitai, T.

    1996-01-01

    In Hungary several critical assemblies, a training reactor and a research reactor have been in operation. The fuel used in the research and training reactors are of Soviet origin. Though spent fuel storage experience is fairly good, medium and long term storage solutions are needed. (author)

  8. Fire, Fuel, and Smoke Science Program 2015 Research Accomplishments

    Science.gov (United States)

    Faith Ann Heinsch; Charles W. McHugh; Colin C. Hardy

    2016-01-01

    The Fire, Fuel, and Smoke Science Program (FFS) of the U.S. Forest Service, Rocky Mountain Research Station focuses on fundamental and applied research in wildland fire, from fire physics and fire ecology to fuels management and smoke emissions. Located at the Missoula Fire Sciences Laboratory in Montana, the scientists, engineers, technicians, and support...

  9. Fire, Fuel, and Smoke Program: 2014 Research Accomplishments

    Science.gov (United States)

    Faith Ann Heinsch; Robin J. Innes; Colin C. Hardy; Kristine M. Lee

    2015-01-01

    The Fire, Fuel, and Smoke Science Program (FFS) of the U.S. Forest Service, Rocky Mountain Research Station focuses on fundamental and applied research in wildland fire, from fire physics and fire ecology to fuels management and smoke emissions. Located at the Missoula Fire Sciences Laboratory in Montana, the scientists, engineers, technicians, and support staff in FFS...

  10. Status of biomass fuels technologies research in the US

    Energy Technology Data Exchange (ETDEWEB)

    Koontz, R.P.; Parker, S.; Glenn, B.

    1984-07-01

    Biomass is a tremendous potential source of fuel and chemical feedstocks. The US Department of Energy has sponsored a broad spectrum of research on biomass at various US government laboratories, private installations, and universities. The status of biomass fuels technologies research in the US is discussed.

  11. Fire, Fuel, and Smoke Science Program: 2013 Research accomplishments

    Science.gov (United States)

    Faith Ann Heinsch; Robin J. Innes; Colin C. Hardy; Kristine M. Lee

    2014-01-01

    The Fire, Fuel, and Smoke Science Program (FFS) of the U.S. Forest Service, Rocky Mountain Research Station, focuses on fundamental and applied research in wildland fire, from fire physics and fire ecology to fuels management and smoke emissions. Located at the Missoula Fire Sciences Laboratory in Montana, the scientists, engineers, technicians, and support staff in...

  12. Kinetic parameters of a material test research reactor fueled with various low enriched uranium dispersion fuels

    International Nuclear Information System (INIS)

    Muhammad, Farhan; Majid, Asad

    2009-01-01

    The effects of using different low enriched uranium fuels, having same uranium density, on the kinetic parameters of a material test research reactor were studied. For this purpose, the original aluminide fuel (UAl x -Al) containing 4.40 gU/cm 3 of an MTR was replaced with silicide (U 3 Si-Al and U 3 Si 2 -Al) and oxide (U 3 O 8 -Al) dispersion fuels having the same uranium density as of the original fuel. Simulations were carried out to calculate prompt neutron generation time, effective delayed-neutron fraction, core excess reactivity and neutron flux spectrum. Nuclear reactor analysis codes including WIMS-D4 and CITATION were used to carry out these calculations. It was observed that both the silicide fuels had the same prompt neutron generation time 0.02% more than that of the original aluminide fuel, while the oxide fuel had a prompt neutron generation time 0.05% less than that of the original aluminide fuel. The effective delayed-neutron fraction decreased for all the fuels; the decrease was maximum at 0.06% for U 3 Si 2 -Al followed by 0.03% for U 3 Si-Al, and 0.01% for U 3 O 8 -Al fuel. The U 3 O 8 -Al fueled reactor gave the maximum ρ excess at BOL which was 21.67% more than the original fuel followed by U 3 Si-Al which was 2.55% more, while that of U 3 Si 2 -Al was 2.50% more than the original UAl x -Al fuel. The neutron flux of all the fuels was more thermalized, than in the original fuel, in the active fuel region of the core. The thermalization was maximum for U 3 O 8 -Al followed by U 3 Si-Al and then U 3 Si 2 -Al fuel.

  13. Researches of WWER fuel rods behaviour under RIA accident conditions

    International Nuclear Information System (INIS)

    Nechaeva, O.; Medvedev, A.; Novikov, V.; Salatov, A.

    2003-01-01

    Unirradiated fuel rod and refabricated fuel rod tests in the BIGR as well as acceptance criteria proving absence of fragmentation and the settlement modeling of refabricated fuel rods thermomechanical behavior in the BIGR-tests using RAPTA-5 code are discussed in this paper. The behaviour of WWER type simulators with E110 and E635 cladding was researched at the BIGR reactor under power pulse conditions simulating reactivity initiated accident. The results of the tests in four variants of experimental conditions are submitted. The behaviour of 12 WWER type refabricated fuel rods was researched in the BIGR reactor under power pulse conditions simulating reactivity initiated accident: burnup 48 and 60 MWd/kgU, pulse width 3 ms, peak fuel enthalpy 115-190 cal/g. The program of future tests in the research reactor MIR with high burnup fuel rod (up to 70 MWd/kgU) under conditions simulating design RIA in WWER-1000 is presented

  14. A Fuel-Sensitive Reduced-Order Model (ROM) for Piston Engine Scaling Analysis

    Science.gov (United States)

    2017-09-29

    of high Reynolds number nonreacting and reacting JP-8 sprays in a constant pressure flow vessel with a detailed chemistry approach . J Energy Resour...for rapid grid generation applied to in-cylinder diesel engine simulations. Society of Automotive Engineers ; 2007 Apr. SAE Technical Paper No.: 2007...ARL-TR-8172 ● Sep 2017 US Army Research Laboratory A Fuel-Sensitive Reduced-Order Model (ROM) for Piston Engine Scaling Analysis

  15. FTA fuel cell bus program : research accomplishments through 2011.

    Science.gov (United States)

    2012-03-01

    Prepared by the Federal Transit Administration (FTA) Office of Research, Demonstration, and Innovation (TRI), this report summarizes the accomplishments of fuel-cell-transit-bus-related research and demonstrations projects supported by FTA through 20...

  16. Electrocatalysis research for fuel cells and hydrogen production

    CSIR Research Space (South Africa)

    Mathe, MK

    2012-01-01

    Full Text Available The CSIR undertakes research in the Electrocatalysis of fuel cells and for hydrogen production. The Hydrogen South Africa (HySA) strategy supports research on electrocatalysts due to their importance to the national beneficiation strategy. The work...

  17. Research and Test Reactor Fuel Elements (RTRFE)

    International Nuclear Information System (INIS)

    Pace, Brett W.; Marinak, Edward A.

    1999-01-01

    BWX Technologies Inc. (BWXT) has experienced several production improvements over the past year. The homogeneity yields in 4.8 gU/cc U 3 Si 2 plates have increased over last year's already high yields. Through teamwork and innovative manufacturing techniques, maintaining high quality surface finishes on plates and elements is becoming easier and less expensive. Currently, BWXT is designing a fabrication development plan to reach a fuel loading of 9 gU/cc within 2 - 4 years. This development will involve a step approach requested by ANL to produce plates using U-8Mo at a loading of 6 gU/cc first and qualify the fuel at those levels. In achieving the goal of a very high-density fuel loading of 9 gU/cc, BWXT is considering employing several new, state of the art, ultrasonic testing techniques for fuel core evaluation. (author)

  18. Experimental research of fuel element reliability

    International Nuclear Information System (INIS)

    Cech, B.; Novak, J.; Chamrad, B.

    1980-01-01

    The rate and extent of the damage of the can integrity for fission products is the basic criterion of reliability. The extent of damage is measurable by the fission product leakage into the reactor coolant circuit. An analysis is made of the causes of the fuel element can damage and a model is proposed for testing fuel element reliability. Special experiments should be carried out to assess partial processes, such as heat transfer and fuel element surface temperature, fission gas liberation and pressure changes inside the element, corrosion weakening of the can wall, can deformation as a result of mechanical interactions. The irradiation probe for reliability testing of fuel elements is described. (M.S.)

  19. Pressure Scalings and Influence Region Research

    Science.gov (United States)

    2018-01-01

    AFRL-RQ-WP-TP-2018-0015 PRESSURE SCALINGS AND INFLUENCE REGION RESEARCH James H. Miller High Speed Systems Division...inside pages STINFO COPY AIR FORCE RESEARCH LABORATORY AEROSPACE SYSTEMS DIRECTORATE WRIGHT-PATTERSON AIR FORCE BASE, OH... RESEARCH 5a. CONTRACT NUMBER In-house 5b. GRANT NUMBER 5c. PROGRAM ELEMENT NUMBER 62201F 6. AUTHOR(S) James H. Miller 5d. PROJECT NUMBER

  20. Logistics of the research reactor fuel cycle: AREVA solutions

    International Nuclear Information System (INIS)

    Ohayon, David; Halle, Laurent; Naigeon, Philippe; Falgoux, Jean-Louis; Franck Obadia, Franck; Auziere, Philippe

    2005-01-01

    The AREVA Group Companies offer comprehensive solutions for the entire fuel cycle of Research Reactors comply with IAEA standards. CERCA and Cogema Logistics have developed a full partnership in the front end cycle. In the field of uranium CERCA and Cogema Logistics have the long term experience of the shipment from Russia, USA to the CERCA plant.. Since 1960, CERCA has manufactured over 300,000 fuel plates and 15,000 fuel elements of more than 70 designs. These fuel elements have been delivered to 40 research reactors in 20 countries. For the Back-End stage, Cogema and Cogema Logistics propose customised solutions and services for international shipments. Cogema Logistics has developed a new generation of packaging to meet the various needs and requirements of the Laboratories and Research Reactors all over the world, and complex regulatory framework. Comprehensive assistance dedicated, services, technical studies, packaging and transport systems are provided by AREVA for every step of research reactor fuel cycle. (author)

  1. Cryogenic hydrogen fuel for controlled inertial confinement fusion (formation of reactor-scale cryogenic targets)

    Energy Technology Data Exchange (ETDEWEB)

    Aleksandrova, I. V.; Koresheva, E. R., E-mail: elena.koresheva@gmail.com; Krokhin, O. N. [Russian Academy of Sciences, Lebedev Physical Institute (Russian Federation); Osipov, I. E. [Power Efficiency Centre, Inter RAO UES (Russian Federation)

    2016-12-15

    In inertial fusion energy research, considerable attention has recently been focused on low-cost fabrication of a large number of targets by developing a specialized layering module of repeatable operation. The targets must be free-standing, or unmounted. Therefore, the development of a target factory for inertial confinement fusion (ICF) is based on methods that can ensure a cost-effective target production with high repeatability. Minimization of the amount of tritium (i.e., minimization of time and space at all production stages) is a necessary condition as well. Additionally, the cryogenic hydrogen fuel inside the targets must have a structure (ultrafine layers—the grain size should be scaled back to the nanometer range) that supports the fuel layer survivability under target injection and transport through the reactor chamber. To meet the above requirements, significant progress has been made at the Lebedev Physical Institute (LPI) in the technology developed on the basis of rapid fuel layering inside moving free-standing targets (FST), also referred to as the FST layering method. Owing to the research carried out at LPI, unique experience has been gained in the development of the FST-layering module for target fabrication with an ultrafine fuel layer, including a reactor- scale target design. This experience can be used for the development of the next-generation FST-layering module for construction of a prototype of a target factory for power laser facilities and inertial fusion power plants.

  2. Enablers and Barriers to Large-Scale Uptake of Improved Solid Fuel Stoves: A Systematic Review

    Science.gov (United States)

    Puzzolo, Elisa; Stanistreet, Debbi; Pope, Daniel; Bruce, Nigel G.

    2013-01-01

    Background: Globally, 2.8 billion people rely on household solid fuels. Reducing the resulting adverse health, environmental, and development consequences will involve transitioning through a mix of clean fuels and improved solid fuel stoves (IS) of demonstrable effectiveness. To date, achieving uptake of IS has presented significant challenges. Objectives: We performed a systematic review of factors that enable or limit large-scale uptake of IS in low- and middle-income countries. Methods: We conducted systematic searches through multidisciplinary databases, specialist websites, and consulting experts. The review drew on qualitative, quantitative, and case studies and used standardized methods for screening, data extraction, critical appraisal, and synthesis. We summarized our findings as “factors” relating to one of seven domains—fuel and technology characteristics; household and setting characteristics; knowledge and perceptions; finance, tax, and subsidy aspects; market development; regulation, legislation, and standards; programmatic and policy mechanisms—and also recorded issues that impacted equity. Results: We identified 31 factors influencing uptake from 57 studies conducted in Asia, Africa, and Latin America. All domains matter. Although factors such as offering technologies that meet household needs and save fuel, user training and support, effective financing, and facilitative government action appear to be critical, none guarantee success: All factors can be influential, depending on context. The nature of available evidence did not permit further prioritization. Conclusions: Achieving adoption and sustained use of IS at a large scale requires that all factors, spanning household/community and program/societal levels, be assessed and supported by policy. We propose a planning tool that would aid this process and suggest further research to incorporate an evaluation of effectiveness. Citation: Rehfuess EA, Puzzolo E, Stanistreet D, Pope D, Bruce

  3. Material test reactor fuel research at the BR2 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dyck, Steven Van; Koonen, Edgar; Berghe, Sven van den [Institute for Nuclear Materials Science, SCK-CEN, Boeretang, Mol (Belgium)

    2012-03-15

    The construction of new, high performance material test reactor or the conversion of such reactors' core from high enriched uranium (HEU) to low enriched uranium (LEU) based fuel requires several fuel qualification steps. For the conversion of high performance reactors, high density dispersion or monolithic fuel types are being developed. The Uranium-Molybdenum fuel system has been selected as reference system for the qualification of LEU fuels. For reactors with lower performance characteristics, or as medium enriched fuel for high performance reactors, uranium silicide dispersion fuel is applied. However, on the longer term, the U-Mo based fuel types may offer a more efficient fuel alternative and-or an easier back-end solution with respect to the silicide based fuels. At the BR2 reactor of the Belgian nuclear research center, SCK-CEN in Mol, several types of fuel testing opportunities are present to contribute to such qualification process. A generic validation test for a selected fuel system is the irradiation of flat plates with representative dimensions for a fuel element. By flexible positioning and core loading, bounding irradiation conditions for fuel elements can be performed in a standard device in the BR2. For fuel element designs with curved plates, the element fabrication method compatibility of the fuel type can be addressed by incorporating a set of prototype fuel plates in a mixed driver fuel element of the BR2 reactor. These generic types of tests are performed directly in the primary coolant flow conditions of the BR2 reactor. The experiment control and interpretation is supported by detailed neutronic and thermal-hydraulic modeling of the experiments. Finally, the BR2 reactor offers the flexibility for irradiation of full size prototype fuel elements, as 200mm diameter irradiation channels are available. These channels allow the accommodation of various types of prototype fuel elements, eventually using a dedicated cooling loop to provide the

  4. Development of a transport cask for spent fuel elements of research reactors

    International Nuclear Information System (INIS)

    Quintana, F.; Saliba, R.O.; Furnari, J.C.; Mourao, R.P; Leite da Silva, L.; Novara, O.; Alexandre Miranda, C.; Mattar Neto, M.

    2012-01-01

    This article presents an overview of the development of a research reactor spent fuel transport cask. Through a project funded by the IAEA, Argentina, Brazil and Chile have collaborated to enhance regional capacity in the management of spent fuel elements from research reactors operated in the region. A packaging for the transport of research reactors spent fuel was developed. It was designed by a team of researchers from the countries mentioned and a 1:2 scale model for MTR type fuel was constructed in Argentina and subsequently tested in CDTN facilities in Belo Horizonte, Brazil. There were three test sequences to test the cask for normal transport and hypothetical accident conditions. It has successfully passed the tests and the overall performance was considered satisfactory. As part of the licensing process, a test sequence with the presence of regulatory authorities is scheduled for December, 2012 (author)

  5. Cost aspects of the research reactor fuel cycle

    International Nuclear Information System (INIS)

    2010-01-01

    Research reactors have made valuable contributions to the development of nuclear power, basic science, materials development, radioisotope production for medicine and industry, and education and training. In doing so, they have provided an invaluable service to humanity. Research reactors are expected to make important contributions in the coming decades to further development of the peaceful uses of nuclear technology, in particular for advanced nuclear fission reactors and fuel cycles, fusion, high energy physics, basic research, materials science, nuclear medicine, and biological sciences. However, in the context of decreased public sector support, research reactors are increasingly faced with financial constraints. It is therefore of great importance that their operations are based on a sound understanding of the costs of the complete research reactor fuel cycle, and that they are managed according to sound financial and economic principles. This publication is targeted at individuals and organizations involved with research reactor operations, with the aim of providing both information and an analytical framework for assessing and determining the cost structure of fuel cycle related activities. Efficient management of fuel cycle expenditures is an important component in developing strategies for sustainable future operation of a research reactor. The elements of the fuel cycle are presented with a description of how they can affect the cost efficient operation of a research reactor. A systematic review of fuel cycle choices is particularly important when a new reactor is being planned or when an existing reactor is facing major changes in its fuel cycle structure, for example because of conversion of the core from high enriched uranium (HEU) to low enriched uranium (LEU) fuel, or the changes in spent fuel management provision. Review and optimization of fuel cycle issues is also recommended for existing research reactors, even in cases where research reactor

  6. Research on burning of biomass fuels, KTH

    Energy Technology Data Exchange (ETDEWEB)

    Hagstroem, U.; Zoukatas, N.; Kutscher, E.; Megas, L.

    1983-05-01

    The three main principles of combustion, namely burning over the fuel bed, under the bed, and the inverted flame have been investigated. Combustion under the fuel bed rendered the lowest emission of carbon monoxide, hydrocarbons, benzopyrene, particulates and tar. Emission is also reduced by preheating the primary incoming air. Burning of pine gives variable emissions whereas birch tree and lying log gives satisfactory combustion. High flame intensity and Reynolds number of the flame zone in the interval 5 to 8 x 10/sup 3/ also give low emission. A conventional wood burner with its flame over the fuel bed and with a water cooled combustion chamber produces 100 times more carbon monoxide than an advanced construction.

  7. Growing dimensions. Spent fuel management at research reactors

    International Nuclear Information System (INIS)

    Ritchie, I.G.

    1998-01-01

    More than 550 nuclear research reactors are operating or shout down around the world. At many of these reactors, spent fuel from their operations is stored, pending decisions on its final disposition. In recent years, problems associated with this spent fuel storage have loomed larger in the international nuclear community. In efforts to determine the overall scope of problems and to develop a database on the subject, the IAEA has surveyed research reactor operators in its Member States. Information for the Research Reactor Spent Fuel Database (RRSFDB) so far has been obtained from a limited but representative number of research reactors. It supplements data already on hand in the Agency's more established Research Reactor Database (RRDB). Drawing upon these database resources, this article presents an overall picture of spent fuel management and storage at the world's research reactors, in the context of associated national and international programmes in the field

  8. Results of fuel elements fabrication on the basis of increased concentration dioxide fuel for research reactors

    International Nuclear Information System (INIS)

    Alexandrov, A.B.; Afanasiev, V.L.; Enin, A.A.; Suprun, V.B.

    1996-01-01

    According to the Russian Reduced Enrichment for Research and Test Reactors (RERTR) program, that were constructed under the Russian projects, at the Novosibirsk Chemical Concentrates Plant the pilot series of different configuration (WR-M2, MR, IRT-4M) fuel elements, based on increased concentration uranium dioxide fuel, have been fabricated for reactor tests. Comprehensive fabricated fuel elements quality estimation has been carried out. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-06-15

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

  10. Research and development of FBR fuel reprocessing in PNC

    International Nuclear Information System (INIS)

    Hoshino, T.

    1976-05-01

    The research program of the PNC for FBR fuel reprocessing in Japan is discussed. The general characteristics of FBR fuel reprocessing are pointed out and a comparison with LWR fuel is made. The R and D program is based on reprocessing using the aqueous Purex process. So far, some preliminary steps of the research program have been carried out, these include solvent extraction test, off-gas treatment test, voloxidation process study, solidification test of high-level liquid waste, and study of the dissolution behaviour of irradiated mixed oxide fuel. By the end of the 1980s, a pilot plant for FBR fuel reprocessing will be completed. For the design of the pilot plant, further research will be carried out in the following fields: head-end techniques; voloxidation process; dissolution and extraction techniques; waste treatment techniques. A time schedule for the different steps of the program is included

  11. United States Domestic Research Reactor Infrastructure TRIGA Reactor Fuel Support

    International Nuclear Information System (INIS)

    Morrell, Douglas

    2011-01-01

    The United State Domestic Research Reactor Infrastructure Program at the Idaho National Laboratory manages and provides project management, technical, quality engineering, quality inspection and nuclear material support for the United States Department of Energy sponsored University Reactor Fuels Program. This program provides fresh, unirradiated nuclear fuel to Domestic University Research Reactor Facilities and is responsible for the return of the DOE-owned, irradiated nuclear fuel over the life of the program. This presentation will introduce the program management team, the universities supported by the program, the status of the program and focus on the return process of irradiated nuclear fuel for long term storage at DOE managed receipt facilities. It will include lessons learned from research reactor facilities that have successfully shipped spent fuel elements to DOE receipt facilities.

  12. Conversion of research reactors to low-enrichment uranium fuels

    International Nuclear Information System (INIS)

    Muranaka, R.G.

    1983-01-01

    There are at present approximately 350 research reactors in 52 countries ranging in power from less than 1 watt to 100 Megawatt and over. In the 1970's, many people became concerned about the possibility that some fuels and fuel cycles could provide an easy route to the acquisition of nuclear weapons. Since enrichment to less than 20% is internationally recognized as a fully adequate barrier to weapons usability, certain Member States have moved to minimize the international trade in highly enriched uranium and have established programmes to develop the technical means to help convert research reactors to the use of low-enrichment fuels with minimum penalties. This could involve modifications in the design of the reactor and development of new fuels. As a result of these programmes, it is expected that most research reactors can be converted to the use of low-enriched fuel

  13. Ultra-Scale Visualization: Research and Education

    International Nuclear Information System (INIS)

    Ma, K-L; Ross, Robert; Huang Jian; Humphreys, Greg; Max, Nelson; Moreland, Kenneth; Owens, John D; Shen, H-W

    2007-01-01

    Understanding the science behind large-scale simulations and high-throughput experiments requires extracting meaning from data sets of hundreds of terabytes or more. Visualization is the most intuitive means for scientists to understand data at this scale, and the most effective way to communicate their findings with others. Even though visualization technology has matured over the past twenty years, it is still limited by the extent and scale of the data that it can be applied to, and also by the functionalities that were mostly designed for single-user, single-variable, and single-space investigation. The Institute for Ultra-Scale Visualization (IUSV), funded by the DOE SciDAC-2 program, has the mission to advance visualization technologies to enable knowledge discovery and dissemination for peta-scale applications. By working with the SciDAC application projects, Centers for Enabling Technology, and other Institutes, IUSV aims to lead the research innovation that can create new visualization capabilities needed for gleaning insights from data at petascale and beyond to solve forefront scientific problems. This paper outlines what we see as some of the biggest research challenges facing the visualization community, and how we can approach education and outreach to put successful research in the hands of scientists

  14. Management and storage of nuclear fuel from Belgian research reactors

    International Nuclear Information System (INIS)

    Gubel, P.

    1996-01-01

    Experiences and problems with the storage of irradiated fuel at research reactors in Belgium are described. In particular, interim storage problems exist for spent fuel elements at the BR2 and the shut down BR3 reactors in Mol. (author). 1 ref

  15. Research reactor fuel transport in the U.K

    Energy Technology Data Exchange (ETDEWEB)

    Panter, R [U.K. Atomic Energy Authority, Harwell (United Kingdom)

    1983-09-01

    This paper describes the containers currently used for transport of fresh or spent fuel elements for Research and Materials Test Reactors in the U.K., their status, operating procedures and some of the practical difficulties. In the U.K., MTR fuel cycle work is almost entirely the responsibility of the U.K. Atomic Energy Authority.

  16. CNG INJECTOR RESEARCH FOR DUAL FUEL ENGINE

    Directory of Open Access Journals (Sweden)

    Adam Majczak

    2017-03-01

    Full Text Available The article presents the tests results of the prototype design of hydraulically assisted injector, that is designed for gas supply into diesel engines. The construction of the injector allows for it positioning in the glow plug socket, so that the gas is injected directly into the combustion chamber. The cycle analysis of the four-cylinder Andoria ADCR engine with a capacity of 2.6 dm3 for different crankshaft rotational speeds allowed to determine the necessary time for fuel injection. Because of that, it was possible to determine the required mass flow rate of the injector, for replacing as much of the original fuel by gaseous fuel. To ensure a high value of flow inside the injector, supply pressure equal to 1 MPa was applied. High gas supply pressure requires high value of valve opening forces. For this purpose a injector with hydraulic control system, using a liquid under pressure for the opening process was designed. On the basis of air pressure measurements in the flow line after the injector, the analysis of opening and closing of the valve was made. Measurements of outflow mass of the injector were also carried out. The results showed that the designed injector meets the requirements necessary to supply ADCR engine by the CNG fuel.

  17. Alternative Fuel Research in Fischer-Tropsch Synthesis

    Science.gov (United States)

    Surgenor, Angela D.; Klettlinger, Jennifer L.; Yen, Chia H.; Nakley, Leah M.

    2011-01-01

    NASA Glenn Research Center has recently constructed an Alternative Fuels Laboratory which is solely being used to perform Fischer-Tropsch (F-T) reactor studies, novel catalyst development and thermal stability experiments. Facility systems have demonstrated reliability and consistency for continuous and safe operations in Fischer-Tropsch synthesis. The purpose of this test facility is to conduct bench scale Fischer-Tropsch (F-T) catalyst screening experiments while focusing on reducing energy inputs, reducing CO2 emissions and increasing product yields within the F-T process. Fischer-Tropsch synthesis is considered a gas to liquid process which reacts syn-gas (a gaseous mixture of hydrogen and carbon monoxide), over the surface of a catalyst material which is then converted into liquids of various hydrocarbon chain length and product distributions1. These hydrocarbons can then be further processed into higher quality liquid fuels such as gasoline and diesel. The experiments performed in this laboratory will enable the investigation of F-T reaction kinetics to focus on newly formulated catalysts, improved process conditions and enhanced catalyst activation methods. Currently the facility has the capability of performing three simultaneous reactor screening tests, along with a fourth fixed-bed reactor used solely for cobalt catalyst activation.

  18. Plant-scale anodic dissolution of unirradiated IFR fuel pins

    International Nuclear Information System (INIS)

    Gay, E.C.; Tomczuk, Z.; Miller, W.E.

    1993-01-01

    This report discusses anodic dissolution which is a major operation in the pyrometallurgical process for recycling spent metal fuels from the Integral Fast Reactor (IFR), an advanced reactor design developed at Argonne National Laboratory. This process involves electrorefining the heavy metals (uranium and plutonium) from chopped, steel-clad fuel segments. The heavy metals are electrotransported from anodic dissolution baskets to solid and liquid cathodes in a molten salt electrolyte (LiCl-KCI) at 500 degrees C. Uranium is recovered on a solid cathode mandrel, while a uranium-plutonium mixture is recovered in a liquid cadmium cathode. The anode configuration consists of four baskets mounted on an anode shaft. These baskets provide parallel circuits in the electrolyte and salt flow through the chopped fuelbed as the baskets are rotated. The baskets for the engineering-scale tests were sized to contain up to 2.5 kg of heavy metal. Anodic dissolution of 10 kg batches of chopped, steel-clad simulated tuel (U-10% Zr and U-Zr-Fs alloy) was demonstrated

  19. Application of ac impedance in fuel cell research and development

    Energy Technology Data Exchange (ETDEWEB)

    Selman, J R; Lin, Y P [Illinois Inst. of Tech., Chicago, IL (United States). Dept. of Chemical Engineering

    1993-10-01

    In applying ac impedance to fuel cells and their porous (gas diffusion) electrodes the emphasis lies on different fuel cell components, and their properties, according to the fuel cell type. The focus has been directed at the electrode/electrolyte interface in MCFC and PAFC, whereas in SOFC and PEMFC the ionic/electronic conductivity of the electrolyte or the characteristics of its composite with the electrocatalyst is of primary interest. The limitations of ac impedance in fuel cell application are in part due to difficulties of interpretation and in part due to experimental difficulties because of the generally fast electrode reaction kinetics. Further research directions are indicated. (author)

  20. Thermal fuel research and development facilities in BNFL

    International Nuclear Information System (INIS)

    Roberts, V.A.; Vickers, J.

    1996-01-01

    BNFL is committed to providing high quality, cost effective nuclear fuel cycle services to customers on a National and International level. BNFL's services, products and expertise span the complete fuel cycle; from fuel manufacture through to fuel reprocessing, transport, waste management and decommissioning and the Company maintains its technical and commercial lead by investment in continued research and development (R and D). This paper discusses BNFL's involvement in R and D and gives an account of the current facilities available together with a description of the advanced R and D facilities constructed or planned at Springfields and Sellafield. It outlines the work being carried out to support the company fuel technology business, to (1) develop more cost effective routes to existing fuel products; (2) maximize the use of recycled uranium, plutonium and tails uranium and (3) support a successful MOX business

  1. Emissions from small scale combustion of pelletized wood fuels

    International Nuclear Information System (INIS)

    Bachs, A.

    1998-01-01

    Combustion of wood pellets in small scale heating systems with an effect below 20 kW has increased. During the winter season 1995/96 1500 small plants for heating houses are estimated to be in operation. Stack emissions from three pellet burners and two pellet stoves have been studied at laboratory. Different pellet qualities were tested. When the fraction of fines increased also the NO x emissions increased with about 10 %. As reference fuel 8 mm pellets was used. Tests with 6 mm pellets gave, in most cases, significant lower emissions of CO and THC. Eleven stoves, burners and boilers were studied in a field test. The results show that all the plants generally have higher emissions in the field than during conditions when the plants are adjusted with a stack gas monitoring instrument. A conclusion is that it is difficult for the operator to adjust the plant without a monitoring instrument. The emissions from the tested plants give an estimation of stack gas emissions from small scale pellet plants. The difference between the 'best' and 'worst' technologies is big. The span of emissions with the best technology to the worst is given below. The interval is concerning normal combustion . During abnormal conditions the emissions are on a significant higher level: * CO 80-1 000 mg/MJ; * Tar 0,3-19 mg/MJ; * THC (as methane equivalents) 2-100 mg/MJ; * NO x 50-70 mg/W;, and * Dust emissions 20-40 mg/MJ. Emissions from pellets heating are lower than from wood combustion and the best technology is close to the emission from oil burners. Wood and pellets have the same origin but the conditions to burn them in an environmental friendly way differ. Combustion of pellets could be improved through improved control of the air and fuel ratio that will create more stable conditions for the combustion

  2. Corrosion of research reactor aluminium clad spent fuel in water

    International Nuclear Information System (INIS)

    2009-12-01

    A large variety of research reactor spent fuel with different fuel meats, different geometries and different enrichments in 235 U are presently stored underwater in basins located around the world. More than 90% of these fuels are clad in aluminium or aluminium based alloys that are notoriously susceptible to corrosion in water of less than optimum quality. Some fuel is stored in the reactor pools themselves, some in auxiliary pools (or basins) close to the reactor and some stored at away-from-reactor pools. Since the early 1990s, when corrosion induced degradation of the fuel cladding was observed in many of the pools, corrosion of research reactor aluminium clad spent nuclear fuel stored in light water filled basins has become a major concern, and programmes were implemented at the sites to improve fuel storage conditions. The IAEA has since then established a number of programmatic activities to address corrosion of research reactor aluminium clad spent nuclear fuel in water. Of special relevance was the Coordinated Research Project (CRP) on Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water (Phase I) initiated in 1996, whose results were published in IAEA Technical Reports Series No. 418. At the end of this CRP it was considered necessary that a continuation of the CRP should concentrate on fuel storage basins that had demonstrated significant corrosion problems and would therefore provide additional insight into the fundamentals of localized corrosion of aluminium. As a consequence, the IAEA started a new CRP entitled Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water (Phase II), to carry out more comprehensive research in some specific areas of corrosion of aluminium clad spent nuclear fuel in water. In addition to this CRP, one of the activities under IAEA's Technical Cooperation Regional Project for Latin America Management of Spent Fuel from Research Reactors (2001-2006) was corrosion monitoring and surveillance of research

  3. Reactivity feedbacks of a material test research reactor fueled with various low enriched uranium dispersion fuels

    International Nuclear Information System (INIS)

    Muhammad, Farhan; Majid, Asad

    2009-01-01

    The reactivity feedbacks of a material test research reactor using various low enriched uranium fuels, having same uranium density were calculated. For this purpose, the original aluminide fuel (UAl x -Al) containing 4.40 gU/cm 3 of an MTR was replaced with silicide (U 3 Si-Al and U 3 Si 2 -Al) and oxide (U 3 O 8 -Al) dispersion fuels having the same uranium density as of the original fuel. Calculations were carried out to find the fuel temperature reactivity feedback, moderator temperature reactivity feedback, moderator density reactivity feedback and moderator void reactivity feedback. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It was observed that the magnitudes all the respective reactivity feedbacks from 38 deg. C to 50 deg. C and 100 deg. C, at the beginning of life, of all the fuels were very close to each other. The fuel temperature reactivity feedback of the U 3 O 8 -Al was about 2% more than the original UAl x -Al fuel. The magnitudes of the moderator temperature, moderator density and moderator void reactivity feedbacks of all the fuels, showed very minor variations from the original aluminide fuel.

  4. Situation of test and research reactors' spent fuels

    International Nuclear Information System (INIS)

    Shimizu, Kenichi; Uchiyama, Junzo; Sato, Hiroshi

    1996-01-01

    The U.S. DOE decided a renewal Off-Site Fuel Policy for stopping to spread a highly enriched uranium which was originally enriched at the U.S., the policy declared that to receive all HEU spent fuels from Test and Research reactors in all the world. In Japan, under bilateral agreement of cooperation between the government of the United States and the government of Japan concerning peaceful uses of nuclear energy, the highly enriched uranium of Test and Research Reactors' fuels was purchased from the U.S. and the fuels had been manufactured in Japan, America, Germany and France. On the other hand, a former president of the U.S. J. Carter proposed that to convert the fuels from HEU to LEU concerning a nonproliferation of nuclear materials in 1978, and Japan absolutely supported this policy. Under this condition, the U.S. stopped to receive the spent fuels from the other countries concerning legal action to the Off-Site Fuels Policy. As a result, the spent fuels are increasing, and to cross to each reactor's storage capacity, and if this policy start, a faced crisis of Test and Research Reactors will be avoided. (author)

  5. Allometric scaling of microbial fuel cells and stacks: The lifeform case for scale-up

    Science.gov (United States)

    Greenman, John; Ieropoulos, Ioannis A.

    2017-07-01

    This case study reports for the first time on the comparison between allometric scaling of lifeforms and scale-up of microbial fuel cell entities; enlarging individual units in volume, footprint and electrode surface area but also multiplying a static size/footprint and electrode surface area to scale-up by stacking. A study published in 2010 by DeLong et al. showed for the first time that Kleiber's law does not apply uniformly to all lifeforms, and that in fact growth rate for prokaryotes is superlinear, for protists is linear and for metazoa is sublinear. The current study, which is utilising data from previous experiments, is showing for the first time that for individual MFC units, which are enlarged, growth rate/power is sublinear, whereas for stacks this is superlinear.

  6. Research reactor fuel development at AECL

    International Nuclear Information System (INIS)

    Sears, D.F.; Wang, N.

    2000-09-01

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

  7. Research reactor fuel development at AECL

    International Nuclear Information System (INIS)

    Sears, D.F.; Wang, N.

    2000-01-01

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

  8. Research Establishment progress report 1978 - uranium fuel cycle

    International Nuclear Information System (INIS)

    1978-12-01

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

  9. Bioenergy Research Programme. Yearbook 1994. Production of wood fuels

    International Nuclear Information System (INIS)

    Alakangas, E.

    1995-01-01

    BIOENERGIA Research Programme is one of energy technology programmes of the Finnish Ministry of Trade and Industry (in 1995 TEKES, Technology Development Center). The aim of Bioenergy Research Programme is to increase the use of economically profitable and environmentally sound bioenergy by improving the competitiveness of present peat and wood fuels. Research and development projects will also develop new economically competitive biofuels and new equipment and methods for production, handling and using of biofuels. The funding for 1994 was nearly 50 million FIM and projects numbered 60. The main goal of the production of wood fuels research area is to develop new production methods in order to decrease the production costs to the level of imported fuels. The total potential of the wood fuel use should be at least 1.0 million toe/a (5.5 million m 3 ). There were 27 projects in 1994 for research on wood fuel production. This part of the yearbook 1994 presents the main results of these projects. The wood reserves do not limit the obtainability of the target. Research and development work has, however, directed to development of equipment and research on wood fuels production chains. Many devices, designed for both separate and integrated production of wood fuels became ready or were becoming ready for prototyping, to be used for production tests. Results of the biomass harvesting and properties research were obtained for utilization in 1994. According to the results it is possible to obtain the desired targets both in integrated and separated production of wood fuels. (author)

  10. Safeguardability assessment on pilot-scale advanced spent fuel conditioning facility

    International Nuclear Information System (INIS)

    Lee, S.Y.; Li, T.K.; Pickett, S.E.; Miller, M.C.; Ko, W.I.; Kim, H.D.

    2006-01-01

    Full text: In South Korea, approximately 6,000 metric tons of spent nuclear fuel from commercial reactor operation has been accumulated with the expectation of more than 30,000 metric tons, three times the present storage capacity, by the end of 2040. To resolve these challenges in spent fuel management, the Korea Atomic Energy Research Institute (KAERI) has been developing a dry reprocessing technology called Advanced Spent Fuel Conditioning Process (ACP). This is an electrometallurgical treatment technique to convert oxide-type spent fuel into a metallic form, and the electrolytic reduction (ER) technology developed recently is known as a more efficient concept for spent fuel conditioning. The goal of the ACP study is to recover more than 99% of the actinide elements into a metallic form with minimizing the volume and heat load of spent fuel. The significant reduction of the volume and heat load of spent fuel is expected to lighten the burden of final disposal in terms of disposal size, safety, and economics. In the framework of R and D collaboration for the ACP safeguards, a joint study on the safeguardability of the ACP technology has been performed by the Los Alamos National Laboratory (LANL) and KAERI. The purpose of this study is to address the safeguardability of the ACP technology, through analysis of material flow and development of a proper safeguards system that meet IAEA's comprehensive safeguards objective. The sub-processes and material flow of the pilot-scale ACP facility were analyzed, and subsequently the relevant material balance area (MBA) and key measurement point (KMP) were designed for material accounting. The uncertainties in material accounting were also estimated with international target values, and design requirements for the material accounting systems were derived

  11. Internationalization Measures in Large Scale Research Projects

    Science.gov (United States)

    Soeding, Emanuel; Smith, Nancy

    2017-04-01

    Internationalization measures in Large Scale Research Projects Large scale research projects (LSRP) often serve as flagships used by universities or research institutions to demonstrate their performance and capability to stakeholders and other interested parties. As the global competition among universities for the recruitment of the brightest brains has increased, effective internationalization measures have become hot topics for universities and LSRP alike. Nevertheless, most projects and universities are challenged with little experience on how to conduct these measures and make internationalization an cost efficient and useful activity. Furthermore, those undertakings permanently have to be justified with the Project PIs as important, valuable tools to improve the capacity of the project and the research location. There are a variety of measures, suited to support universities in international recruitment. These include e.g. institutional partnerships, research marketing, a welcome culture, support for science mobility and an effective alumni strategy. These activities, although often conducted by different university entities, are interlocked and can be very powerful measures if interfaced in an effective way. On this poster we display a number of internationalization measures for various target groups, identify interfaces between project management, university administration, researchers and international partners to work together, exchange information and improve processes in order to be able to recruit, support and keep the brightest heads to your project.

  12. Russian research reactor fuel return program starts shipping fuel to Russia

    International Nuclear Information System (INIS)

    Dedik, T.; Bolshinsky, I.; Krass, A.

    2003-01-01

    For almost four years the United States (U.S), the Russian Federation (R.F.), and the International Atomic Energy Agency (IAEA) have been discussing an initiative to return Soviet/Russian-origin research reactor fuel to the Russian Federation. In a series of bilateral and trilateral meetings in Vienna and Moscow, considerable progress has been made toward defining the Russian Research Reactor Fuel Return Program as well as obtaining the necessary technical data to facilitate the return. More than 20 research reactors in 17 countries that have Soviet- or Russian-supplied fuel have identified. Most of these reactors have stocks of both fresh and irradiated HEU fuel that must be carefully stored and managed for many years to come. On September 21, 2003 the Russian Research Reactor Fuel Return program shipped 14 kg of fresh Russian-origin HEU fuel from Romania to the nuclear fuel fabrication facility in Russia, which represented the beginning of the practical implementation of the program. (author)

  13. Using fine-scale fuel measurements to assess wildland fuels, potential fire behavior and hazard mitigation treatments in the southeastern USA

    Science.gov (United States)

    Roger D. Ottmar; John I. Blake; William T. Crolly

    2012-01-01

    The inherent spatial and temporal heterogeneity of fuel beds in forests of the southeastern United States may require fine scale fuel measurements for providing reliable fire hazard and fuel treatment effectiveness estimates. In a series of five papers, an intensive, fine scale fuel inventory from the Savanna River Site in the southeastern United States is used for...

  14. IAEA activities related to research reactor fuel conversion and spent fuel return programs

    International Nuclear Information System (INIS)

    Goldman, Ira N.; Adelfang, Pablo; Ritchie, Iain G.

    2005-01-01

    The IAEA has been involved for more than twenty years in supporting international nuclear non-proliferation efforts associated with reducing the amount of highly enriched uranium (HEU) in international commerce. IAEA projects and activities have directly supported the Reduced Enrichment for Research and Test Reactors (RERTR) programme, as well as directly associated efforts to return research reactor fuel to the country where it was originally enriched. IAEA efforts have included the development and maintenance of several data bases with information related to research reactors and research reactor spent fuel inventories that have been essential in planning and managing both RERTR and spent fuel return programmes. Other IAEA regular budget programs have supported research reactor fuel conversion from HEU to low enriched uranium (LEU), and in addressing issues common to many member states with spent fuel management problems and concerns. The paper briefly describes IAEA involvement since the early 1980's in these areas, including regular budget and Technical Co-operation programme activities, and focuses on efforts in the past five years to continue to support and accelerate U.S. and Russian research reactor spent fuel return programmes. (author)

  15. IAEA activities related to research reactor fuel conversion and spent fuel return programmes

    International Nuclear Information System (INIS)

    Ritchie, I.G.; Adelfang, P.; Goldman, I.N.

    2004-01-01

    Full text: The IAEA has been involved for more than twenty years in supporting international nuclear non-proliferation efforts associated with reducing the amount of highly enriched uranium (HEU) in international commerce. IAEA projects and activities have directly supported the Reduced Enrichment for Research and Test Reactors (RERTR) programme, as well as directly associated efforts to return research reactor fuel to the country of origin where it was originally enriched. IAEA efforts have included the development and maintenance of several data bases with information related to research reactors and research reactor spent fuel inventories that have been essential in planning and managing both RERTR and spent fuel return programmes. Other IAEA regular budget programmes have supported research reactor fuel conversion from HEU to low enriched uranium, and in addressing issues common to many member states with spent fuel management problems and concerns. The paper briefly describes IAEA involvement since the early 1980's in these areas, including regular budget and Technical Co-operation programme activities, and focuses on efforts in the past five years to continue to support and accelerate U.S. and Russian research reactor spent fuel return programmes. It is hoped that an announcement of the extension of the U.S. Acceptance Programme, which is expected in the very near future, will facilitate the life extensions of many productive TRIGA reactors around the world. (author)

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

    International Nuclear Information System (INIS)

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

    1978-09-01

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

  17. Assessment of Research Needs for Advanced Fuel Cells

    Energy Technology Data Exchange (ETDEWEB)

    Penner, S.S.

    1985-11-01

    The DOE Advanced Fuel Cell Working Group (AFCWG) was formed and asked to perform a scientific evaluation of the current status of fuel cells, with emphasis on identification of long-range research that may have a significant impact on the practical utilization of fuel cells in a variety of applications. The AFCWG held six meetings at locations throughout the country where fuel cell research and development are in progress, for presentations by experts on the status of fuel cell research and development efforts, as well as for inputs on research needs. Subsequent discussions by the AFCWG have resulted in the identification of priority research areas that should be explored over the long term in order to advance the design and performance of fuel cells of all types. Surveys describing the salient features of individual fuel cell types are presented in Chapters 2 to 6 and include elaborations of long-term research needs relating to the expeditious introduction of improved fuel cells. The Introduction and the Summary (Chapter 1) were prepared by AFCWG. They were repeatedly revised in response to comments and criticism. The present version represents the closest approach to a consensus that we were able to reach, which should not be interpreted to mean that each member of AFCWG endorses every statement and every unexpressed deletion. The Introduction and Summary always represent a majority view and, occasionally, a unanimous judgment. Chapters 2 to 6 provide background information and carry the names of identified authors. The identified authors of Chapters 2 to 6, rather than AFCWG as a whole, bear full responsibility for the scientific and technical contents of these chapters.

  18. How Large-Scale Research Facilities Connect to Global Research

    DEFF Research Database (Denmark)

    Lauto, Giancarlo; Valentin, Finn

    2013-01-01

    Policies for large-scale research facilities (LSRFs) often highlight their spillovers to industrial innovation and their contribution to the external connectivity of the regional innovation system hosting them. Arguably, the particular institutional features of LSRFs are conducive for collaborative...... research. However, based on data on publications produced in 2006–2009 at the Neutron Science Directorate of Oak Ridge National Laboratory in Tennessee (United States), we find that internationalization of its collaborative research is restrained by coordination costs similar to those characterizing other...

  19. Basic research on cermet nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ohashi, Hiroshi; Sto, Seichi [Hokkaido Univ., Sapporo (Japan). Faculty of Engineering; Takano, Masahide; Minato, Kazuo; Fukuda, Kosaku

    1998-01-01

    Production of cermet nuclear fuel having fine uranium dioxide (UO{sub 2}) particles dispersed in matrix metal requires basic property data on the compatibility of matrix metal with fission product compounds. It is thermodynamically suggested that, as burnup increases, cesium in oxide fuel reacts with the fuel, other fission products or cladding pipe and produces cesium uranates, cesium molybdate, or cesium chromate in stainless steel cladding pipe. Attempt was made to measure the thermal expansion coefficient and thermal conductivity of cesium uranates (Cs{sub 2}UO{sub 4} and Cs{sub 2}U{sub 2}O{sub 7}), cesium molybdate (Cs{sub 2}MoO{sub 4}) and cesium chromate (Cs{sub 2}CrO{sub 4}). Thermal expansion was measured by X-ray diffraction and determined by Cohen`s method. Thermal conductivity was obtained by measuring thermal diffusion by laser flash method. The thermal expansion of Cs{sub 2}UO{sub 4} and Cs{sub 2}U{sub 2}O{sub 7} is as low as 1.2% for the former and 1.0% for the latter, up to 1000K. The thermal expansion of Cs{sub 2}MoO{sub 4} is as high as that of Cs{sub 2}CrO{sub 4}, 2.1% for the former and 2.5% for the latter at temperatures from room temperature to 873K. Average thermal expansion in this temperature range is 4.4 x 10{sup -5} K{sup -1} for Cs{sub 2}MoO{sub 4} and 4.2 x 10{sup -5} K{sup -1}. The thermal expansion of Cs{sub 2}CrO{sub 4} is four times higher than that of UO{sub 2} and five times higher than that of Cr{sub 2}O{sub 3}. The thermal conductivity of Cs{sub 2}UO{sub 4} is nearly equal to that of Cs{sub 2}U{sub 2}O{sub 7} in absolute value and temperature dependency. Cs{sub 2}U{sub 2}O{sub 7}, having different thermal conductivity between {alpha} and {beta} phases, shows higher conductivity with {beta} than with {alpha}, about 1/4 of that of UO{sub 2} at 1000K. The thermal conductivity of Cs{sub 2}CrO{sub 4} is nearly equal to that of Cs{sub 2}MoO{sub 4} in absolute value and temperature dependency. (N.H.)

  20. Handling of spent fuel from research reactors in Japan

    International Nuclear Information System (INIS)

    Kanda, K.

    1997-01-01

    In Japan eleven research reactors are in operation. After the 19th International Meeting on Reduced Enrichment for Research Reactors and Test Reactors (RERTR) on October 6-10, 1996, Seoul, Korea, the Five Agency Committee on Highly Enriched Uranium, which consists of Science and Technology Agency, the Ministry of Education, Science and Culture, the Ministry of Foreign Affairs, Japan Atomic Energy Research Institute (JAERI) and Kyoto University Research Reactor Institute (KURRI) met on November 7,1996, to discuss the handling of spent fuel from research reactors in Japan. Advantages and disadvantages to return spent fuel to the USA in comparison to Europe were discussed. So far, a number of spent fuel elements in JAERI and KURRI are to be returned to the US. The first shipment to the US is planned for 60 HEU elements from JMTR in 1997. The shipment from KURRI is planned to start in 1999. (author)

  1. Calculation of fuel burn-up and fuel reloading for the Dalat Nuclear Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lan, Nguyen Phuoc; Huy, Ngo Quang [Centre for Nuclear Technique Application, Ho Chi Minh City (Viet Nam); Thong, Ha Van; Binh, Do Quang [Nuclear Research Inst., Da Lat (Viet Nam)

    1994-10-01

    Calculation of fuel burnup and fuel reloading for the Dalat Nuclear Research Reactor was carried out by using a new programme named HEXA-BURNUP, realized in a PC. The programme is used to calculate the following parameters of the Dalat reactor: a/Critical configurations of the core loaded with 69, 72, 74, 86, 88, 89 and 92 fuel elements. The effective multiplication coefficients equal 1 within the error ranges of less than 0.38%. b/ The thermal neutron flux distribution in the reactor. The calculated results agree with the experimental data measured at 11 typical positions. c/The average fuel burn-up for the period from Feb. 1984 to Sep. 1992. The difference between calculation and experiment is only about 1.9%. 10 fuel reloading versions are calculated, from which an optimal version is proposed. (author). 9 refs., 4 figs., 5 tabs.

  2. United States Domestic Research Reactor Infrastructure - TRIGA Reactor Fuel Support

    International Nuclear Information System (INIS)

    Morrell, Douglas

    2008-01-01

    The purpose of the United State Domestic Research Reactor Infrastructure Program is to provide fresh nuclear reactor fuel to United States universities at no, or low, cost to the university. The title of the fuel remains with the United States government and when universities are finished with the fuel, the fuel is returned to the United States government. The program is funded by the United States Department of Energy - Nuclear Energy division, managed by Department of Energy - Idaho Field Office, and contracted to the Idaho National Laboratory's Management and Operations Contractor - Battelle Energy Alliance. Program has been at Idaho since 1977 and INL subcontracts with 26 United States domestic reactor facilities (13 TRIGA facilities, 9 plate fuel facilities, 2 AGN facilities, 1 Pulstar fuel facility, 1 Critical facility). University has not shipped fuel since 1968 and as such, we have no present procedures for shipping spent fuel. In addition: floor loading rate is unknown, many interferences must be removed to allow direct access to the reactor tank, floor space in the reactor cell is very limited, pavement ends inside our fence; some of the surface is not finished. The whole approach is narrow, curving and downhill. A truck large enough to transport the cask cannot pull into the lot and then back out (nearly impossible / refused by drivers); a large capacity (100 ton), long boom crane would have to be used due to loading dock obstructions. Access to the entrance door is on a sidewalk. The campus uses it as a road for construction equipment, deliveries and security response. Large trees are on both sides of sidewalk. Spent fuel shipments have never been done, no procedures approved or in place, no approved casks, no accident or safety analysis for spent fuel loading. Any cask assembly used in this facility will have to be removed from one crane, moved on the floor and then attached to another crane to get from the staging area to the reactor room. Reactor

  3. Full-scale tests of spent-nuclear-fuel shipping systems

    International Nuclear Information System (INIS)

    Yoshimura, H.R.; Huerta, M.

    1976-01-01

    Sandia Laboratories will be conducting, for the U.S. Energy Research and Development Administration, a series of tests involving spent-nuclear-fuel shipping systems. Large shipping casks in the 20500 to 70000-kg range will be included in the following full-scale tests: (1) Runaway tractor-trailer crash into a solid concrete barrier while carrying a shipping cask. (2) High-speed locomotive grade-crossing impact with a truck carrying a shipping cask. (3) High-speed derailment, collision, and fire involving a special railcar and shipping cask. The hardware and testing procedures for each of the tests are described. The analysis conducted in advance of the tests addresses the modelling technique used and a description of the scale-model tests. Analytical modelling being done before running the full-scale tests is also described. (author)

  4. 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.)

  5. The University of Missouri Research Reactor, its fuel and productivity

    International Nuclear Information System (INIS)

    Brugger, R.M.; Schlapper, G.A.; Alger, D.M.

    1993-01-01

    This paper describes the University of Missouri Research Reactor (MURR) and presents a summary of contributions to education, research, and service. These efforts have helped offset the impact on the U.S. economy of research from other countries. Special emphasis is placed on fuel design developments that have allowed MuRR to keep the cost megawatt day (MWD) of fuel essentially constant. Also noted is the fact that the United States has missed some research opportunities because of a hold-the-line attitude. The slipping position of U.S. research reactors is compared with the rest of the world. As will be further outlined in the text, the MURR cannot (with available technology) decrease its U-235 enrichment level and maintain present research capabilities and fuel cycle costs. Data is presented to show how MURR, if permitted to use advanced fuel technology, could reduce fuel fabrication costs and onsite U-235 inventory. In addition it is shown that MURR could increase its capabilities provided that arbitrary institutional limits are removed

  6. The University of Missouri Research Reactor, its fuel and productivity

    International Nuclear Information System (INIS)

    Brugger, Robert M.; Schlapper, Gerald A.; Alger, Don M.

    1993-01-01

    This paper describes the University of Missouri Research Reactor (MURR) and presents a summary of contributions to education, research, and service. These efforts have helped offset the impact on the U. S. economy of research from other countries. Special emphasis is placed on fuel design developments that have allowed MURR to keep the cost megawatt day (MWD) of fuel essentially constant. Also noted is the fact that the United States has missed some research opportunities because of a hold-the-line attitude. The slipping position of U.S. research reactors is compared with the rest of the world. As will be further outlined in the text, the MURR cannot (with available technology) decrease its U-235 enrichment level and maintain present research capabilities and fuel cycle costs. Data is presented to show how MURR if permitted to use advanced fuel technology, could reduce fuel fabrication costs and onsite U-235 inventory. In addition it is shown that MURR could increase its capabilities provided that arbitrary institutional limits are removed

  7. Status of USNRC research on fuel behavior under accident conditions

    International Nuclear Information System (INIS)

    Johnston, W.V.

    1976-01-01

    The program of the Fuel Behaviour Research is directed at providing a detailed understanding of the response of nuclear fuel assemblies to off-normal or accident conditions. This understanding is expressed in physical and analytical correlations which are incorporated into computer codes. The results of these experiments and the resulting codes are available to the licensing authorities for use in evaluating utility submissions. (orig.) [de

  8. Hydrogen and fuel cell research networking in Ontario

    Energy Technology Data Exchange (ETDEWEB)

    Peppley, B.A. [Queen' s-RMC Fuel Cell Research Centre, Kingston, ON (Canada)

    2009-07-01

    This presentation reviewed the activities of the Ontario Fuel Cell Research and Innovation Network since its launch in 2006. Funded by the Ontario Ministry of Research and Innovation, the project involves 17 academic researchers from 8 universities and is supported by 8 industrial partners. The group of researchers has made progress in supporting the developing fuel cell industry in Ontario and in Canada. Their work has the potential to help deploy the province's automotive-oriented manufacturing sector in directions that address the issues of clean air and climate change. New initiatives in the development of hydrogen and fuel cell technologies are instrumental in expanding this network to leverage new business activities in the post financial crisis period. These activities are expected to result in economic benefits for job and economic growth.

  9. Criticality safety research on nuclear fuel cycle facility

    Energy Technology Data Exchange (ETDEWEB)

    Miyoshi, Yoshinori [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2004-07-01

    This paper present d s current status and future program of the criticality safety research on nuclear fuel cycle made by Japan Atomic Energy Research Institute. Experimental research on solution fuel treated in reprocessing plant has been performed using two critical facilities, STACY and TRACY. Fundamental data of static and transient characteristics are accumulated for validation of criticality safety codes. Subcritical measurements are also made for developing a monitoring system for criticality safety. Criticality safety codes system for solution and power system, and evaluation method related to burnup credit are developed. (author)

  10. Fuel enrichment reduction for heavy water moderated research reactors

    International Nuclear Information System (INIS)

    McCulloch, D.B.

    1984-01-01

    Twelve heavy-water-moderated research reactors of significant power level (5 MW to 125 MW) currently operate in a number of countries, and use highly enriched uranium (HEU) fuel. Most of these reactors could in principle be converted to use uranium of lower enrichment, subject in some cases to the successful development and demonstration of new fuel materials and/or fuel element designs. It is, however, generally accepted as desirable that existing fuel element geometry be retained unaltered to minimise the capital costs and licensing difficulties associated with enrichment conversion. The high flux Australian reactor, HIFAR, at Lucas Heights, Sydney is one of 5 Dido-class reactors in the above group. It operates at 10 MW using 80% 235 U HEU fuel. Theoretical studies of neutronic, thermohydraulic and operational aspects of converting HIFAR to use fuels of reduced enrichment have been made over a period. It is concluded that with no change of fuel element geometry and no penalty in the present HEU fuel cycle burn-up performance, conversion to MEU (nominally 45% 235 U) would be feasible within the limits of current fully qualified U-Al fuel materials technology. There would be no significant, adverse effects on safety-related parameters (e.g. reactivity coefficients) and only small penalties in reactor flux. Conversion to LEU (nominally 20% 235 U) a similar basis would require that fuel materials of about 2.3 g U cm -3 be fully qualified, and would depress the in-core thermal neutron flux by about 15 per cent relative to HEU fuelling. In qualitative terms, similar conclusions would be expected to hold for a majority of the above heavy water moderated reactors. (author)

  11. Small-scale production and use of wood fuels. Report of the year 2005

    International Nuclear Information System (INIS)

    Alakangas, E.

    2005-01-01

    The target areas of the research programme are: Small-scale production and handling of wood fuels, Pellet production, distribution and use, Heating technology and Business and service concepts. Production and processing technology focuses on cost-effectiveness, fuel quality, logistics of production chains and storage, transport and feeding solutions. The quality of pellets in the view of the whole chain: production, storage, distribution and feeding, is under scrutiny. In addition, storage and distribution systems are being developed. The aim is to create functional and comprehensive heat production systems based on the use of wood pellets. Emissions from small-scale use are reduced and efficiency of combustion improved to meet the Central European standard. Modern control, automation and data management systems are applied cost-effectively. The aim is to create comprehensive systems and modular solutions. Business and service concepts relate to all target areas such as heat entrepreneurship and energy service companies (ESCO). The aim is to promote the networking of companies and develop new solutions for fuel and heat production services

  12. Basic research and industrialization of CANDU advanced fuel

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Suk Ho; Park, Joo Hwan; Jun, Ji Su [and others

    2000-04-01

    handling, the preliminary evaluation of the CANFLEX-RU economics and some specification and description for the irradiation test of the CANFLEX-RU fuel NRU research reactor.

  13. Basic research and industrialization of CANDU advanced fuel

    International Nuclear Information System (INIS)

    Chun, Suk Ho; Park, Joo Hwan; Jun, Ji Su

    2000-04-01

    handling, the preliminary evaluation of the CANFLEX-RU economics and some specification and description for the irradiation test of the CANFLEX-RU fuel NRU research reactor

  14. Underlying chemistry research for the nuclear fuel waste management program

    International Nuclear Information System (INIS)

    Torgerson, D.F.; Sagert, N.H.; Shoesmith, D.W.; Taylor, P.

    1984-04-01

    This document reviews the underlying chemistry research part of the Canadian Nuclear Fuel Waste Management Program, carried out in the Research Chemistry Branch. This research is concerned with developing the basic chemical knowledge and under-standing required in other parts of the Program. There are four areas of underlying research: Waste Form Chemistry, Solute and Solution Chemistry, Rock-Water-Waste Interactions, and Abatement and Monitoring of Gas-Phase Radionuclides

  15. Safety research in nuclear fuel cycle at PNC

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-09-01

    This report collects the results of safety research in nuclear fuel cycle at Power Reactor and Nuclear Fuel Development Corporation, in order to answer to the Questionnaire of OECD/NEA. The Questionnaire request to include information concerning to research topic, description, main results (if available), reference documents, research institutes involved, sponsoring organization and other pertinent information about followings: a) Recently completed research projects. b) Ongoing (current) research projects. Achievements on following items are omitted by the request of OECD/NEA, uranium mining and milling, uranium refining and conversion to UF{sub 6}, uranium enrichment, fuel manufacturers, spent fuel storage, radioactive waste management, transport of radioactive materials, decommissioning. We select topics from the fields of a) nuclear installation, b) seismic, and c) PSA, in projects from frame of annual safety research plan for nuclear installations established by Nuclear Safety Commission. We apply for the above a) and b) projects as follows: a) Achievements in Safety Research, fiscal 1991-1995, b) fiscal 1996 Safety Research Achievements: Progress. (author)

  16. Safety research in nuclear fuel cycle at PNC

    International Nuclear Information System (INIS)

    1998-09-01

    This report collects the results of safety research in nuclear fuel cycle at Power Reactor and Nuclear Fuel Development Corporation, in order to answer to the Questionnaire of OECD/NEA. The Questionnaire request to include information concerning to research topic, description, main results (if available), reference documents, research institutes involved, sponsoring organization and other pertinent information about followings: a) Recently completed research projects. b) Ongoing (current) research projects. Achievements on following items are omitted by the request of OECD/NEA, uranium mining and milling, uranium refining and conversion to UF 6 , uranium enrichment, fuel manufacturers, spent fuel storage, radioactive waste management, transport of radioactive materials, decommissioning. We select topics from the fields of a) nuclear installation, b) seismic, and c) PSA, in projects from frame of annual safety research plan for nuclear installations established by Nuclear Safety Commission. We apply for the above a) and b) projects as follows: a) Achievements in Safety Research, fiscal 1991-1995, b) fiscal 1996 Safety Research Achievements: Progress. (author)

  17. Small scale wood combustion in Germany. Recent research and trends

    Energy Technology Data Exchange (ETDEWEB)

    Maier, H.; Unterberger, S.; Hein, K.R.G. [Institute of Process Engineering and Power Plant Technology, University of Stuttgart (Germany)

    1998-12-31

    To reduce Europe`s greenhouse gas emission CO{sub 2} it is a challenging task utilising biomass fuels as there are wood or wood residues from the forest industry. The utilisation can be done either in commercially operated medium (> 50 kWth) or full scale (> 1 MWth) decentralised heat and power stations or in small scale (< 50 kWth) domestic heating systems. In small scale heating systems untreated wood logs, wood briquette or wood pellets and in few cases wood chips are used. The present market in Germany is focused on the use of wood logs. Presently, the use of wood pellets in small scale automatically operated boilers < 15 kW especially for low energy houses is discussed more and more. Since 1980 the installation of new wood fired small scale domestic heating systems reached a significant size due to the interest of the customers to have a alternative inhouse heating system and to increase the living comfort. In 1994 the amount of sold small scale heaters in Germany were in total about 133.258 units. The thermal power of in 1994 sold units is estimated of about 1350 MW which is a significant size in total with regard to domestic heating purposes. Since few years there is a clear market trend in Germany towards the installation of open fire stoves. Due to this trend in Germany and the design characteristic of open fire stoves using huge glass doors of glass windows it is very difficult to achieve a further reduction of emissions like CO and unburned volatile hydrocarbons (VOC). In the text the requirements for modern small scale wood fired stoves in Germany as well as the actual stage and trend of research and development (R and D) are discussed 4 refs.

  18. Research on assurance system of nuclear fuel supply (Contract research)

    International Nuclear Information System (INIS)

    Kobayashi, Naoki; Naoi, Yosuke; Wakabayashi, Shuji; Tazaki, Makiko; Senzaki, Masao

    2010-03-01

    Assurance of supply (AOS) of nuclear fuel is a special arrangement in case of nuclear fuel supply disruption caused by political reasons other than nonproliferation. It aims to support a stable supply of nuclear fuel while avoiding unnecessary spread of sensitive enrichment technology. Current discussions on AOS have been initiated by the IAEA Director-General's article published in The Economist entitled 'Towards a Safer World' Oct. 2003. Since then, various proposals on AOS have been presented. In order to facilitate international discussions on AOS, authors have conducted studies of AOS system based on Japanese Government's proposal 'IAEA Standby Arrangement System (INFCIRC/683)'. In this paper, we gave an overview of discussions on AOS since World War II, and elaborated on some of current proposals. We have been able to discuss feasibility of AOS system more specifically by including additional costs and period required for AOS, and to present a system which could work as a practical system. Issues we have tried to tackle here include definitions of AOS, and roles of consumer states, supplier states, IAEA and nuclear industries. We present some solutions including broadening coverage of AOS, declaration by supplier states on AOS, establishing advisory committee in the IAEA on the actual application of AOS, and setting up an IAEA fund for AOS. (author)

  19. Research on assurance system of nuclear fuel supply (Contract research)

    International Nuclear Information System (INIS)

    Kobayashi, Naoki; Naoi, Yosuke; Wakabayashi, Shuji; Tazaki, Makiko; Senzaki, Masao

    2010-08-01

    Assurance of supply (AOS) of nuclear fuel is a special arrangement in case of nuclear fuel supply disruption caused by political reasons other than nonproliferation. It aims to support a stable supply of nuclear fuel while avoiding spread of sensitive enrichment technology. Current discussions on AOS have been initiated by the IAEA Director-General's article published in The Economist entitled 'Towards a Safer World' Oct. 2003. Since then, various proposals on AOS have been presented. In order to facilitate international discussions on AOS, authors have conducted studies of AOS system based on Japanese Government's proposal 'IAEA Standby Arrangement System (INFCIRC/683)'. In this paper, we have been able to discuss feasibility of AOS system more specifically by including additional costs and period required for AOS, and to present a system which could work as a practical system. Issues we have tried to tackle here include definitions of AOS, and roles of consumer States, supplier States, IAEA and nuclear industries. We present some solutions including broadening coverage of AOS, declaration by supplier States on AOS, establishing advisory committee in the IAEA on the actual application of AOS, and setting up an IAEA fund for AOS. (author)

  20. A basic research on the transient behavior for a metallic fuel FBR

    International Nuclear Information System (INIS)

    Baba, Mamoru; Hirano, Go; Kawada, Ken-ichi; Niwa, Hajime

    1999-03-01

    A metallic fuel with novel design has received great deal of interest recently as an option of advanced fuel to be substituted MOX fuel, however, the behavior at the transient has not been studied in many aspects. Therefore, for the purpose to show the basic tendency of the behavior and released energy at CDA (core disruptive accident) for a metallic fuel FBR and to prepare the basic knowledge for consideration of the adoption of the advanced fuel, Tohoku university and Power Reactor and Nuclear Fuel Development Corporation have made a joint research entitled 'A basic research on the transient behavior for a metallic fuel FBR'. The results are the following. (1) Target and Results of analysis: The accident initiator considered is a LOF accident without scram. The LOF analysis was performed for a metallic fuel 600 MWe homogeneous two region core at the beginning of cycle, both for an ordinary metallic fuel core and for a metallic fuel core with ZrH pins. It was necessary mainly to change the constants of input parameters to apply the code for the analysis of a metallic fueled reactor. These changes were made by assuming appropriate models. Basic LOF cases and all blackout case that assumed using electromagnetic pumps were analyzed. The results show that the basic LOF cases for a metallic fuel core and all the cases for a metallic fuel core with ZrH pins could be avoided to become prompt-critical, and mildly transfer to the transition phase. It is shown that the moderator is quite elective to mitigate the accident at the initiation phase. However, it is necessary to analyze the transition phase to know if the re-criticality is totally avoided after the initiation phase. (2) Improvement of CDA initiation phase analysis code: At present, it is difficult for the code to adapt to the large scale material movement in the core at the transient. Therefore, the nuclear calculation model in the code was improved by using the adiabatic space dependent kinetics, and examined

  1. Green Routing Fuel Saving Opportunity Assessment: A Case Study on California Large-Scale Real-World Travel Data

    Energy Technology Data Exchange (ETDEWEB)

    Zhu, Lei [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Holden, Jacob [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Gonder, Jeffrey D [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Wood, Eric W [National Renewable Energy Laboratory (NREL), Golden, CO (United States)

    2017-07-31

    New technologies, such as connected and automated vehicles, have attracted more and more researchers for improving the energy efficiency and environmental impact of current transportation systems. The green routing strategy instructs a vehicle to select the most fuel-efficient route before the vehicle departs. It benefits the current transportation system with fuel saving opportunity through identifying the greenest route. This paper introduces an evaluation framework for estimating benefits of green routing based on large-scale, real-world travel data. The framework has the capability to quantify fuel savings by estimating the fuel consumption of actual routes and comparing to routes procured by navigation systems. A route-based fuel consumption estimation model, considering road traffic conditions, functional class, and road grade is proposed and used in the framework. An experiment using a large-scale data set from the California Household Travel Survey global positioning system trajectory data base indicates that 31% of actual routes have fuel savings potential with a cumulative estimated fuel savings of 12%.

  2. Green Routing Fuel Saving Opportunity Assessment: A Case Study on California Large-Scale Real-World Travel Data: Preprint

    Energy Technology Data Exchange (ETDEWEB)

    Zhu, Lei; Holden, Jacob; Gonder, Jeff; Wood, Eric

    2017-07-13

    New technologies, such as connected and automated vehicles, have attracted more and more researchers for improving the energy efficiency and environmental impact of current transportation systems. The green routing strategy instructs a vehicle to select the most fuel-efficient route before the vehicle departs. It benefits the current transportation system with fuel saving opportunity through identifying the greenest route. This paper introduces an evaluation framework for estimating benefits of green routing based on large-scale, real-world travel data. The framework has the capability to quantify fuel savings by estimating the fuel consumption of actual routes and comparing to routes procured by navigation systems. A route-based fuel consumption estimation model, considering road traffic conditions, functional class, and road grade is proposed and used in the framework. An experiment using a large-scale data set from the California Household Travel Survey global positioning system trajectory data base indicates that 31% of actual routes have fuel savings potential with a cumulative estimated fuel savings of 12%.

  3. Research on a Scania 11 liter ethanol fueled bus engine

    Energy Technology Data Exchange (ETDEWEB)

    Egebaeck, K E; Pettersson, E

    1996-05-01

    This report presents research carried out on an alcohol fueled bus engine. The engine used was a six cylinder 11 liter compression ignition turbo-charged engine with an inter cooler. The research program included studies of the impact on the emissions when changing different engine components and different settings of the engine. A study of the impact on the engine performance and emissions when using different fuel compositions was also carried out. During the course of the work, the engine was equipped with oxidation catalysts of two different types, one of which was more efficient than the other concerning the oxidation of unburnt fuel related components. One of the main purposes of the research was to improve the emission characteristics of the engine by an optimization of the engine and its setting. The exhaust emissions were thoroughly characterized with respect to both regulated and unregulated emissions. 13 refs, figs, tabs

  4. Research and development of nitride fuel cycle technology in Europe

    International Nuclear Information System (INIS)

    Wallenius, Janne

    2004-01-01

    Research and development on nitride fuels for minor actinide burning in accelerator driven systems is performed in Europe in context of the CONFIRM project. Dry and wet methods for fabrication of uranium free nitride fuels have been developed with the assistance of thermo-chemical modelling. Four (Pu, Zr) pins have been fabricated by PSI and will be irradiated in Studsvik at a rating of 40-50 kW/m. The thermal conductivity of (Pu, Zr)N has been measured and was found to be in agreement with earlier theoretical assessments. Safety modeling indicates that americium bearing nitride fuels, in spite of their relatively poor high temperature stability under atmospheric pressure, can survive power transients as long as the fuel cladding remains intact. (author)

  5. Nuclear criticality assessment of Oak Ridge research fuel element storage

    International Nuclear Information System (INIS)

    Thomas, J.T.

    1978-06-01

    Spent and partially spent Oak Ridge Research Reactor (ORR) fuel elements are retained in the storage section of the ORR pool facility. Determination of a maximum expected neutron multiplication factor for the storage area is accomplished by a validated calculational method. The KENO Monte Carlo code and the Hansen-Roach 16-group neutron cross section sets were validated by calculations of critical experiments performed with early ORR fuel elements and with SPERT-D fuel elements. Calculations of various fuel element arrangements are presented which confirm the subcriticality previously inferred from critical experiments and indicate the k/sub eff/ would not exceed 0.85, were the storage area to be filled to capacity with storage racks containing elements with the fissionable material loading increased to 350 g of 235 U

  6. Annual report 1984. Nuclear fuel research department

    International Nuclear Information System (INIS)

    Nebel, D.

    1985-12-01

    Research activities have been reviewed in the field of powder metallurgy and the production of new ceramic materials in particular silicon nitrides. Communications on the chemistry of solutions and the analytical chemistry have been compiled specially under the aspect of technical solutions and the development of new devices, equipments and methods. In a final chapter publications and lectures are assorted

  7. Nano-watt fueling from a micro-scale microbial fuel cell using black tea waste

    KAUST Repository

    Rojas, Jhonathan Prieto; Alqarni, Wejdan Mohammed Mofleh; Kalantan, Kalthom Kamil Saleh; Hussain, Muhammad Mustafa; Mink, Justine E.

    2016-01-01

    In this report, we show the rapid assessment of black tea as potential fuel to power up nanopower systems using a microsized, simplistic and sustainable air-cathode microbial fuel cell. It was found that tea produced more power compared

  8. 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

  9. Full-Scale Cask Testing and Public Acceptance of Spent Nuclear Fuel Shipments - 12254

    Energy Technology Data Exchange (ETDEWEB)

    Dilger, Fred [Black Mountain Research, Henderson, NV 81012 (United States); Halstead, Robert J. [State of Nevada Agency for Nuclear Projects Carson City, NV 80906 (United States); Ballard, James D. [Department of Sociology, California State University, Northridge Northridge, CA 91330 (United States)

    2012-07-01

    Full-scale physical testing of spent fuel shipping casks has been proposed by the National Academy of Sciences (NAS) 2006 report on spent nuclear fuel transportation, and by the Presidential Blue Ribbon Commission (BRC) on America's Nuclear Future 2011 draft report. The U.S. Nuclear Regulatory Commission (NRC) in 2005 proposed full-scale testing of a rail cask, and considered 'regulatory limits' testing of both rail and truck casks (SRM SECY-05-0051). The recent U.S. Department of Energy (DOE) cancellation of the Yucca Mountain project, NRC evaluation of extended spent fuel storage (possibly beyond 60-120 years) before transportation, nuclear industry adoption of very large dual-purpose canisters for spent fuel storage and transport, and the deliberations of the BRC, will fundamentally change assumptions about the future spent fuel transportation system, and reopen the debate over shipping cask performance in severe accidents and acts of sabotage. This paper examines possible approaches to full-scale testing for enhancing public confidence in risk analyses, perception of risk, and acceptance of spent fuel shipments. The paper reviews the literature on public perception of spent nuclear fuel and nuclear waste transportation risks. We review and summarize opinion surveys sponsored by the State of Nevada over the past two decades, which show consistent patterns of concern among Nevada residents about health and safety impacts, and socioeconomic impacts such as reduced property values along likely transportation routes. We also review and summarize the large body of public opinion survey research on transportation concerns at regional and national levels. The paper reviews three past cask testing programs, the way in which these cask testing program results were portrayed in films and videos, and examines public and official responses to these three programs: the 1970's impact and fire testing of spent fuel truck casks at Sandia National

  10. Fuel isolation research for the Canadian nuclear fuel waste management program

    International Nuclear Information System (INIS)

    1982-06-01

    This document is intended to give a broad outline of the Fuel Isolatikn program and to indicate how this program fits into the overall framework of the Canadian Nuclear Fuel Waste Management Program. Similar activities in other countries are described, and the differences in philosophy behind these and the Canadian program are highlighted. A program plan is presented that outlines the development of research programs that contribute to the safety assessment of the disposal concept and the development of technology required for selection and optimization of a feasible fuel isolation system. Some indication of the work that might take place beyond concept assessment, at the end of the decade, is also given. The current program is described in some detail, with emphasis on what the prkgram has achieved to date and hopes to achieve in the future for the concept assessment phase of the waste management program. Finally, some major capital facilities associated with the fuel isolation program are described

  11. Spent fuel performance assessment and research. Final report of a co-ordinated research project on Spent Fuel Performance Assessment and Research (SPAR) 1997-2001

    International Nuclear Information System (INIS)

    2003-03-01

    The report provides an overview of technical issues related to spent fuel wet and dry storage and summarizes the objectives and major findings of research, carried out within the framework of the Coordinated Research Program. Included are the fuel integrity aspects, fuel degradation mechanisms in dry and wet storage, behaviour of storage facility components (metallic components, reinforced concrete). Also included are issues related to long-term storage and monitoring technologies and techniques. Country reports on research projects within the SPAR Coordinated Research Program is presented. A brief history is given on the history of the BEFAST and SPAR Coordinated Research Projects

  12. Spent fuel performance assessment and research. Final report of a co-ordinated research project on Spent Fuel Performance Assessment and Research (SPAR) 1997-2001

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-03-01

    The report provides an overview of technical issues related to spent fuel wet and dry storage and summarizes the objectives and major findings of research, carried out within the framework of the Coordinated Research Program. Included are the fuel integrity aspects, fuel degradation mechanisms in dry and wet storage, behaviour of storage facility components (metallic components, reinforced concrete). Also included are issues related to long-term storage and monitoring technologies and techniques. Country reports on research projects within the SPAR Coordinated Research Program is presented. A brief history is given on the history of the BEFAST and SPAR Coordinated Research Projects.

  13. Spent Fuel Performance Assessment and Research. Final Report of a Coordinated Research Project (SPAR-II)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-07-01

    As storage of spent fuel has become a key technology in spent fuel management, wet and dry storage have become mature technologies and continue to demonstrate good performance. Increased spent fuel storage capacity in combination with longer storage durations will be needed over the foreseeable future as many countries have delayed their decision on spent fuel disposal or reprocessing. Extended spent fuel storage is, and will remain, an important activity for all countries with nuclear power programmes. A number of countries are planning or have already initiated research programmes on spent fuel storage performance, and there is a continuing benefit in exchanging spent fuel storage experience of the Member States in order to build a comprehensive technology knowledge base. Potential degradation mechanisms that may affect cladding integrity during wet storage are uniform corrosion, pitting, galvanic, and microbiologically-influenced corrosion. Potential degradation mechanisms that may affect cladding integrity during dry storage and subsequent handling and transportation operations are air oxidation, thermal creep, stress corrosion cracking (SCC), delayed hydride cracking (DHC), hydride re-orientation, hydrogen migration and re-distribution. Investigations carried out so far indicate that from the degradation mechanisms that may affect the integrity of spent fuel assembly/bundle structure during interim storage, hydride re-orientation has the potential to impair the ability of the cladding to effectively withstand potentially adverse mechanical challenges resulting from handling or transportation accidents. Fuel integrity issues are related to the definition and criteria of fuel integrity, failure classification, packaging and retrieval of damaged fuel and transport of damaged fuel assemblies. Various monitoring technologies have been developed and used to confirm the continued spent fuel integrity during storage or to provide an early indication of developing

  14. Spent Fuel Performance Assessment and Research. Final Report of a Coordinated Research Project (SPAR-II)

    International Nuclear Information System (INIS)

    2012-01-01

    As storage of spent fuel has become a key technology in spent fuel management, wet and dry storage have become mature technologies and continue to demonstrate good performance. Increased spent fuel storage capacity in combination with longer storage durations will be needed over the foreseeable future as many countries have delayed their decision on spent fuel disposal or reprocessing. Extended spent fuel storage is, and will remain, an important activity for all countries with nuclear power programmes. A number of countries are planning or have already initiated research programmes on spent fuel storage performance, and there is a continuing benefit in exchanging spent fuel storage experience of the Member States in order to build a comprehensive technology knowledge base. Potential degradation mechanisms that may affect cladding integrity during wet storage are uniform corrosion, pitting, galvanic, and microbiologically-influenced corrosion. Potential degradation mechanisms that may affect cladding integrity during dry storage and subsequent handling and transportation operations are air oxidation, thermal creep, stress corrosion cracking (SCC), delayed hydride cracking (DHC), hydride re-orientation, hydrogen migration and re-distribution. Investigations carried out so far indicate that from the degradation mechanisms that may affect the integrity of spent fuel assembly/bundle structure during interim storage, hydride re-orientation has the potential to impair the ability of the cladding to effectively withstand potentially adverse mechanical challenges resulting from handling or transportation accidents. Fuel integrity issues are related to the definition and criteria of fuel integrity, failure classification, packaging and retrieval of damaged fuel and transport of damaged fuel assemblies. Various monitoring technologies have been developed and used to confirm the continued spent fuel integrity during storage or to provide an early indication of developing

  15. Comparison of scale/triton and helios burnup calculations for high burnup LWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Tittelbach, S.; Mispagel, T.; Phlippen, P.W. [WTI Wissenschaftlich-Technische Ingenieurberatung GmbH, Juelich (Germany)

    2009-07-01

    The presented analyses provide information about the suitability of the lattice burnup code HELIOS and the recently developed code SCALE/TRITON for the prediction of isotopic compositions of high burnup LWR fuel. The accurate prediction of the isotopic inventory of high burnt spent fuel is a prerequisite for safety analyses in and outside of the reactor core, safe loading of spent fuel into storage casks, design of next generation spent fuel casks and for any consideration of burnup credit. Depletion analyses are performed with both burnup codes for PWR and BWR fuel samples which were irradiated far beyond 50 GWd/t within the LWR-PROTEUS Phase II project. (orig.)

  16. Comparative Study of Laboratory-Scale and Prototypic Production-Scale Fuel Fabrication Processes and Product Characteristics

    International Nuclear Information System (INIS)

    2014-01-01

    An objective of the High Temperature Gas Reactor fuel development and qualification program for the United States Department of Energy has been to qualify fuel fabricated in prototypic production-scale equipment. The quality and characteristics of the tristructural isotropic coatings on fuel kernels are influenced by the equipment scale and processing parameters. Some characteristics affecting product quality were suppressed while others have become more significant in the larger equipment. Changes to the composition and method of producing resinated graphite matrix material has eliminated the use of hazardous, flammable liquids and enabled it to be procured as a vendor-supplied feed stock. A new method of overcoating TRISO particles with the resinated graphite matrix eliminates the use of hazardous, flammable liquids, produces highly spherical particles with a narrow size distribution, and attains product yields in excess of 99%. Compact fabrication processes have been scaled-up and automated with relatively minor changes to compact quality to manual laboratory-scale processes. The impact on statistical variability of the processes and the products as equipment was scaled are discussed. The prototypic production-scale processes produce test fuels that meet fuel quality specifications.

  17. Flop casting of nuclear materials for advanced fuel cycle research - 5247

    International Nuclear Information System (INIS)

    Swift, A.J.; Koury, D.J.; Czerwinski, K.R.; Vollmer, J.M.

    2015-01-01

    Full text of publication follows. Next generation fast reactor designs of nuclear reactors utilizing metallic fuel are being developed as an alternative fuel cycle option in an effort to reduce carbon emissions. Metallic fuel systems are attractive because of their high thermal conductivity, fissile atom density, and inherent safety. Metallic fuel systems are also being investigated because of their potential to reach high burnups. The increased targeted burnups for metallic fuels lead to higher concentrations of actinides, lanthanides, and other fission products, which alter the fuel properties and impact the performance of the fuel. Before designs can be implemented, the fission product concentrations must be studied at variable fuel geometries and stages of fuel burnup. Arc flop casting serves as a viable option for casting alloys as the molds can be tailored to fit design specific requirements while cutting costs in time-consuming machining. Arc casting is done as the final preparation step in a small arc furnace with an argon or 5% hydrogen-argon atmosphere after the sample has been subsequently melted, overturned, and re-melted. The flop casting mold is then fitted to the chamber as needed and the previously prepared sample is quickly hit with a high current arc causing the molten metal to fill the copper mold. The U-Zr-Pu system will serve as the basis for this research as it has been extensively studied since the 1950 years, although flop casting can be adapted to any metallic fuel system. Multiple U-Zr-Pu with varying fission product concentrations alloys, Technetium metal, and Plutonium alloys have been flop cast based on burnup calculations. Prepared samples were cast using different molds and dimensions, then characterized by Scanning Electron Microscopy, X-ray diffraction, and Thermogravimetric Analysis. The goal of this research is to test and develop flop casting techniques for the production of metallic fuel alloys applicable for various stages and

  18. Appendix - A small scale research review

    DEFF Research Database (Denmark)

    Larsen, Lea Lund

    over two questions asked in the abstract concerning workshop C - Effects of teacher training in lifelong learning - at the current ASEM conference. At first it is asked what do we already know about teacher training effects in relation to adult learners? This will be identified by a search...... for systematic reviews in the international This will be identified by a search for systematic reviews in the international clearinghouses and afterwards by a small scale mapping done in the pedagogic databaseERIC. The second question sounds: which methods can be used to measure the effects of teacher...... training? This will be illustrated by describing a research design in progress at the National Centre of Competence Development, DK, regarding a program where teachers are taught Cooperative Learning as a pedagogical and didactical method. This appendix concerns the first question. In search for empiric results concerning: What do...

  19. Fuel element gamma scanning at the Oak Ridge Research Reactor

    International Nuclear Information System (INIS)

    Hobbs, R.W.

    1987-01-01

    In January 1986, a demonstration program was begun at the Oak Ridge Research Reactor (ORR) to convert operations from high-enrichment uranium fuel to the newly developed U 3 Si 2 low-enrichment fuel. A primary program objective is to validate neutronics calculations conducted by the Reduced Enrichment in Research and Test Reactors Program at Argonne National Laboratory. Accordingly, a new method for determining core-power distribution has been developed. The method is based on gamma-ray spectroscopy measurements to determine the relative levels of 140 La in the fuel elements after each operating cycle. The measurement and data analyses are described and a comparison of measured and diffusion theory calculated values of the core-power distribution is presented in this paper

  20. Research on nondestructive examination methods for CANDU fuel channel inspection

    International Nuclear Information System (INIS)

    Soare, M.; Petriu, F.; Toma, V.; Revenco, V.; Calinescu, A.; Ciocan, R.; Iordache, C.; Popescu, L.; Mihalache, M.; Murgescu, C.

    1995-01-01

    The requirements of the 1994 edition of CAN/CSA-N285.4 Periodic Inspection Standard, which address all known and postulated degradation mechanisms and introduce material surveillance demands, involve a growing need for improved nondestructive examination (NDE) methods and technologies. In order to have a proper technical support in its decisions concerning fuel channel inspections at Cernavoda NPP, the Romanian Power Authority (RENEL) initiated a Research Program regarding the nondestructive characterization of the fuel channels structural integrity. The paper presents the most significant results obtained on this Research Program: the ENDUS experimental system for Laboratory simulation of the fuel channel inspection, ultrasonic Rayleigh-Lamb waves technique for pressure tubes examination, phase analysis technique for near-surface flaws, influence of the metallurgical state of the pressure tube material on the eddy current defectoscopic signals, characterization of plastic deformation and fracture of zirconium alloys by acoustic emission. (author)

  1. Conjugate heat transfer simulations of advanced research reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Piro, M.H.A., E-mail: pirom@aecl.ca; Leitch, B.W.

    2014-07-01

    Highlights: • Temperature predictions are enhanced by coupling heat transfer in solid and fluid zones. • Seven different cases are considered to observe trends in predicted temperature and pressure. • The seven cases consider high/medium/low power, flow, burnup, fuel material and geometry. • Simulations provide temperature predictions for performance/safety. Boiling is unlikely. • Simulations demonstrate that a candidate geometry can enhance performance/safety. - Abstract: The current work presents numerical simulations of coupled fluid flow and heat transfer of advanced U–Mo/Al and U–Mo/Mg research reactor fuels in support of performance and safety analyses. The objective of this study is to enhance predictions of the flow regime and fuel temperatures through high fidelity simulations that better capture various heat transfer pathways and with a more realistic geometric representation of the fuel assembly in comparison to previous efforts. Specifically, thermal conduction, convection and radiation mechanisms are conjugated between the solid and fluid regions. Also, a complete fuel element assembly is represented in three dimensional space, permitting fluid flow and heat transfer to be simulated across the entire domain. Seven case studies are examined that vary the coolant inlet conditions, specific power, and burnup to investigate the predicted changes in the pressure drop in the coolant and the fuel, clad and coolant temperatures. In addition, an alternate fuel geometry is considered with helical fins (replacing straight fins in the existing design) to investigate the relative changes in predicted fluid and solid temperatures. Numerical simulations predict that the clad temperature is sensitive to changes in the thermal boundary layer in the coolant, particularly in simultaneously developing flow regions, while the temperature in the fuel is anticipated to be unaffected. Finally, heat transfer between fluid and solid regions is enhanced with

  2. State-of-the-Art Report on Multi-scale Modelling of Nuclear Fuels

    International Nuclear Information System (INIS)

    Bartel, T.J.; Dingreville, R.; Littlewood, D.; Tikare, V.; Bertolus, M.; Blanc, V.; Bouineau, V.; Carlot, G.; Desgranges, C.; Dorado, B.; Dumas, J.C.; Freyss, M.; Garcia, P.; Gatt, J.M.; Gueneau, C.; Julien, J.; Maillard, S.; Martin, G.; Masson, R.; Michel, B.; Piron, J.P.; Sabathier, C.; Skorek, R.; Toffolon, C.; Valot, C.; Van Brutzel, L.; Besmann, Theodore M.; Chernatynskiy, A.; Clarno, K.; Gorti, S.B.; Radhakrishnan, B.; Devanathan, R.; Dumont, M.; Maugis, P.; El-Azab, A.; Iglesias, F.C.; Lewis, B.J.; Krack, M.; Yun, Y.; Kurata, M.; Kurosaki, K.; Largenton, R.; Lebensohn, R.A.; Malerba, L.; Oh, J.Y.; Phillpot, S.R.; Tulenko, J. S.; Rachid, J.; Stan, M.; Sundman, B.; Tonks, M.R.; Williamson, R.; Van Uffelen, P.; Welland, M.J.; Valot, Carole; Stan, Marius; Massara, Simone; Tarsi, Reka

    2015-10-01

    The Nuclear Science Committee (NSC) of the Nuclear Energy Agency (NEA) has undertaken an ambitious programme to document state-of-the-art of modelling for nuclear fuels and structural materials. The project is being performed under the Working Party on Multi-Scale Modelling of Fuels and Structural Material for Nuclear Systems (WPMM), which has been established to assess the scientific and engineering aspects of fuels and structural materials, describing multi-scale models and simulations as validated predictive tools for the design of nuclear systems, fuel fabrication and performance. The WPMM's objective is to promote the exchange of information on models and simulations of nuclear materials, theoretical and computational methods, experimental validation and related topics. It also provides member countries with up-to-date information, shared data, models, and expertise. The goal is also to assess needs for improvement and address them by initiating joint efforts. The WPMM reviews and evaluates multi-scale modelling and simulation techniques currently employed in the selection of materials used in nuclear systems. It serves to provide advice to the nuclear community on the developments needed to meet the requirements of modelling for the design of different nuclear systems. The original WPMM mandate had three components (Figure 1), with the first component currently completed, delivering a report on the state-of-the-art of modelling of structural materials. The work on modelling was performed by three expert groups, one each on Multi-Scale Modelling Methods (M3), Multi-Scale Modelling of Fuels (M2F) and Structural Materials Modelling (SMM). WPMM is now composed of three expert groups and two task forces providing contributions on multi-scale methods, modelling of fuels and modelling of structural materials. This structure will be retained, with the addition of task forces as new topics are developed. The mandate of the Expert Group on Multi-Scale Modelling of

  3. An Expert System to Analyze Homogeneity in Fuel Element Plates for Research Reactors

    International Nuclear Information System (INIS)

    Tolosa, S.C.; Marajofsky, A.

    2004-01-01

    In the manufacturing control of Fuel Element Plates for Research Reactors, one of the problems to be addressed is how to determine the U-density homogeneity in a fuel plate and how to obtain qualitative and quantitative information in order to establish acceptance or rejection criteria for such, as well as carrying out the quality follow-up. This paper is aimed at developing computing software which implements an Unsupervised Competitive Learning Neural Network for the acknowledgment of regions belonging to a digitalized gray scale image. This program is applied to x-ray images. These images are generated when the x-ray beams go through a fuel plate of approximately 60 cm x 8 cm x 0.1 cm thick. A Nuclear Fuel Element for Research Reactors usually consists of 18 to 22 of these plates, positioned in parallel, in an arrangement of 8 x 7 cm. Carrying out the inspection of the digitalized x-ray image, the neural network detects regions with different luminous densities corresponding to U-densities in the fuel plate. This is used in quality control to detect failures and verify acceptance criteria depending on the homogeneity of the plate. This modality of inspection is important as it allows the performance of non-destructive measurements and the automatic generation of the map of U-relative densities of the fuel plate

  4. An expert system to analyze homogeneity in fuel element plates for research reactors

    International Nuclear Information System (INIS)

    Cativa Tolosa, Sebastian; Marajofsky, Adolfo

    2004-01-01

    In the manufacturing control of Fuel Element Plates for Research Reactors, one of the problems to be addressed is how to determine the U-density homogeneity in a fuel plate and how to obtain qualitative and quantitative information in order to establish acceptance or rejection criteria for such, as well as carrying out the quality follow-up.This paper is aimed at developing computing software which implements an Unsupervised Competitive Learning Neural Network for the acknowledgment of regions belonging to a digitalized gray scale image. This program is applied to X-ray images. These images are generated when the X-ray beams go through a fuel plate of approximately 60 cm x 8 cm x 0.1 cm thick. A Nuclear Fuel Element for Research Reactors usually consists of 18 to 22 of these plates, positioned in parallel, in an arrangement of 8 x 7 cm. Carrying out the inspection of the digitalized X-ray image, the neural network detects regions with different luminous densities corresponding to U-densities in the fuel plate. This is used in quality control to detect failures and verify acceptance criteria depending on the homogeneity of the plate. This modality of inspection is important as it allows the performance of non-destructive measurements and the automatic generation of the map of U-relative densities of the fuel plate. (author)

  5. Fuel Management at the Dalat Nuclear Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pham, V.L.; Nguyen, N.D.; Luong, B.V.; Le, V.V.; Huynh, T.N.; Nguyen, K.C. [Nuclear Research Institute, 01 Nguyen Tu Luc Street, Dalat City (Viet Nam)

    2011-07-01

    The Dalat Nuclear Research Reactor (DNRR) is a pool type research reactor which was reconstructed in 1982 from the old 250 kW TRIGA-MARK II reactor. The spent fuel storage was newly designed and installed in the place of the old thermalizing column for biological irradiation. The core was loaded by Russian WWR-M2 fuel assemblies (FAs) with 36% enrichment. The reconstructed reactor reached its initial criticality in November 1983 and attained it nominal power of 500 kW in February 1984. The first fuel reloading was executed in April 1994 after more than 10 years of operation with 89 highly enriched uranium (HEU) FAs. The third fuel reloading by shuffling of HEU FAs was executed in June 2004. After the shuffling the working configuration of reactor core kept unchanged of 104 HEU FAs. The fourth fuel reloading was executed in November 2006. The 2 new HEU FAs were loaded in the core periphery, at previous locations of wet irradiation channel and dry irradiation channel. After reloading the working configuration of reactor core consisted of 106 HEU FAs. Contracts for reactor core conversion between USA, Russia, Vietnam and the International Atomic Energy Agency for Nuclear fuel manufacture and supply for DNRR and Return of Russian-origin non-irradiated highly enriched uranium fuel to the Russian Federation have been realized in 2007. According to the results of design and safety analyses performed by the joint study between RERTR Program at Argonne National Laboratory and Vietnam Atomic Energy Institute the mixed core configurations of irradiated HEU and new low enriched uranium (LEU) FAs has been created on 12 September, 2007 and on 20 July, 2009. After reloading in 2009, the 14 HEU FAs with highest burnup were removed from the core and put in the interim storage in reactor pool. The works on full core conversion for the DNRR are being realized in cooperation with the organizations, DOE and IAEA. Contract for Nuclear fuel manufacture and supply of 66 LEU FAs for DNRR

  6. Pilot-scale equipment development for pyrochemical treatment of spent oxide fuel

    International Nuclear Information System (INIS)

    Herrmann, S. D.

    1999-01-01

    Fundamental objectives regarding spent nuclear fuel treatment technologies include, first, the effective distribution of spent fuel constituents among product and stable waste forms and, second, the minimization and standardization of waste form types and volumes. Argonne National Laboratory (ANL) has developed and is presently demonstrating the electrometallurgical treatment of sodium-bonded metal fuel from Experimental Breeder Reactor II, resulting in an uranium product and two stable waste forms, i.e. ceramic and metallic. Engineering efforts are underway at ANL to develop pilot-scale equipment which would precondition irradiated oxide fuel via pyrochemical processing and subsequently allow for electrometallurgical treatment of such non-metallic fuels into standard product and waste forms. This paper highlights the integration of proposed spent oxide fuel treatment with existing electrometallurgical processes. System designs and technical bases for development of pilot-scale oxide reduction equipment are also described

  7. A feasible approach to implement a commercial scale CANDU fuel manufacturing plant in Egypt

    International Nuclear Information System (INIS)

    El-Shehawy, I.; El-Sharaky, M.; Yasso, K.; Selim, I.; Graham, N.; Newington, D.

    1995-01-01

    Many planning scenarios have been examined to assess and evaluate the economic estimates for implementing a commercial scale CANDU fuel manufacturing plant in Egypt. The cost estimates indicated strong influence of the annual capital costs on total fuel manufacturing cost; this is particularly evident in a small initial plant where the proposed design output is only sufficient to supply reload fuel for a single CANDU-6 reactor. A modular approach is investigated as a possible way, to reduce the capital costs for a small initial fuel plant. In this approach the plant would do fuel assembly operations only and the remainder of a plant would be constructed and equipped in the stages when high production volumes can justify the capital expenses. Such approach seems economically feasible for implementing a small scale CANDU fuel manufacturing plant in developing countries such as Egypt and further improvement could be achieved over the years of operation. (author)

  8. Fossil fuel power generation within the European Research Area

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-12-10

    The report is the first in a series of three produced by the PowerClean Thematic Network that looks at and defines future requirements for research and development of fossil fuel power generation in the European Union. It makes the case for fossil fuel R & D with emphasis on the need for clean coal technologies (to increased efficiency and other CO{sub 2} capture and storage) For satisfying future energy demands of the enlarged European Union between now and 2030. The report concludes that affirmative R, D and D action is needed to support the EU power industry, working together on a Europe-wide basis, to establish the use of coal and other fossil fuels in near-zero emissions power plant. The role model would be the European Research Area, as in the Sixth Framework Programme (FP6), but with a more comprehensive range of technical objectives recognising the importance of fossil fuels. Section headings are: introduction; current energy use; future needs and requirements; the future for clean fossil fuel energy in Europe; comparison with approaches adopted elsewhere (USA Vision 21 and FutureGen programmes, Japan); and responsibilities for EU coal R, D & D. 14 refs., 9 figs., 4 tabs.

  9. Commercial Aspect of Research Reactor Fuel Element Production

    International Nuclear Information System (INIS)

    Susanto, B.G; Suripto, A

    1998-01-01

    Several aspects affecting the commercialization of the Research Reactor Fuel Element Production Installation (RR FEPI) under a BUMN (state-owned company)have been studied. The break event point (BEP) value based on total production cost used is greatly depending upon the unit selling price of the fuel element. At a selling price of USD 43,500/fuel element, the results of analysis shows that the BEP will be reached at 51% of minimum available capacity. At a selling price of US$ 43.500/fuel element the total income (after tax) for 7 years ahead is US $ 4.620.191,- The net present value in this study has a positive value is equal to US $ 2.827.527,- the internal rate of return will be 18% which is higher than normal the bank interest rare (in US dollar) at this time. It is concluded therefore that the nuclear research reactor fuel element produced by state-owned company BUMN has a good prospect to be sold commercially

  10. History of Research Reactor Fuel Fabrication at Babcock and Wilcox

    International Nuclear Information System (INIS)

    Freim, J.B.

    1983-01-01

    1982 was a year of tremendous growth for Babcock and Wilcox and its Research Reactor Fuel Facility. The Division has progressed from essentially being a non-competitor to a position where we are growing in strength. This paper describes some of the general aspects of past history and where B and W is now

  11. Small-scale production and utilization of wood fuels; Puupolttoaineen pientuotanto ja -kaeyttoe - katsaus tutkimus- projekteihin

    Energy Technology Data Exchange (ETDEWEB)

    Tuomi, S. [Work Efficiency Inst., Rajamaeki (Finland)

    1996-12-31

    The objective of the research on small-scale production of wood fuels was to promote the forest owners` own utilization and procurement of firewood. The profitability of firewood was improved by developing new farm-tractor mountable equipment and methods for forest owners and small-entrepreneurs for harvesting of first-thinning wood and other small-dimeter wood. Totally new solution for machine felling of small trees and chopwood production were developed to serial production level. Recyclable processing and delivery units were developed for delivery of chopwood. A calculation model for analysing the costs of small-scale production of firewood became ready. A guide on the development of heating-entrepreneur activities, serving the entrepreneurs, was published. The objective of the firewood utilization research was to reduce the technical barriers of the utilization of firewood in small-house and real-estate scales. The main aim was to reduce the flue-gas emissions. The emissions of the fireplaces were reduced by developing the construction of fireplaces, catalytic combustion and heating methods. An automatic stoker-burner was developed for real-estate scale and a boiler series was designed for biofuels

  12. Small-scale production and utilization of wood fuels; Puupolttoaineen pientuotanto ja -kaeyttoe - katsaus tutkimus- projekteihin

    Energy Technology Data Exchange (ETDEWEB)

    Tuomi, S [Work Efficiency Inst., Rajamaeki (Finland)

    1997-12-31

    The objective of the research on small-scale production of wood fuels was to promote the forest owners` own utilization and procurement of firewood. The profitability of firewood was improved by developing new farm-tractor mountable equipment and methods for forest owners and small-entrepreneurs for harvesting of first-thinning wood and other small-dimeter wood. Totally new solution for machine felling of small trees and chopwood production were developed to serial production level. Recyclable processing and delivery units were developed for delivery of chopwood. A calculation model for analysing the costs of small-scale production of firewood became ready. A guide on the development of heating-entrepreneur activities, serving the entrepreneurs, was published. The objective of the firewood utilization research was to reduce the technical barriers of the utilization of firewood in small-house and real-estate scales. The main aim was to reduce the flue-gas emissions. The emissions of the fireplaces were reduced by developing the construction of fireplaces, catalytic combustion and heating methods. An automatic stoker-burner was developed for real-estate scale and a boiler series was designed for biofuels

  13. Uncertainty and sensitivity analysis of fission gas behavior in engineering-scale fuel modeling

    Energy Technology Data Exchange (ETDEWEB)

    Pastore, Giovanni, E-mail: Giovanni.Pastore@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Swiler, L.P., E-mail: LPSwile@sandia.gov [Optimization and Uncertainty Quantification, Sandia National Laboratories, P.O. Box 5800, Albuquerque, NM 87185-1318 (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); 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); 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); Spencer, B.W., E-mail: Benjamin.Spencer@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Luzzi, L., E-mail: Lelio.Luzzi@polimi.it [Politecnico di Milano, Department of Energy, Nuclear Engineering Division, via La Masa 34, I-20156 Milano (Italy); Van Uffelen, P., E-mail: Paul.Van-Uffelen@ec.europa.eu [European Commission, Joint Research Centre, Institute for Transuranium Elements, Hermann-von-Helmholtz-Platz 1, D-76344 Karlsruhe (Germany); 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)

    2015-01-15

    The role of uncertainties in fission gas behavior calculations as part of engineering-scale nuclear fuel modeling is investigated using the BISON fuel performance code with a recently implemented physics-based model for fission gas release and swelling. Through the integration of BISON with the DAKOTA software, a sensitivity analysis of the results to selected model parameters is carried out based on UO{sub 2} single-pellet simulations covering different power regimes. The parameters are varied within ranges representative of the relative uncertainties and consistent with the information in the open literature. The study leads to an initial quantitative assessment of the uncertainty in fission gas behavior predictions with the parameter characterization presently available. Also, the relative importance of the single parameters is evaluated. Moreover, a sensitivity analysis is carried out based on simulations of a fuel rod irradiation experiment, pointing out a significant impact of the considered uncertainties on the calculated fission gas release and cladding diametral strain. The results of the study indicate that the commonly accepted deviation between calculated and measured fission gas release by a factor of 2 approximately corresponds to the inherent modeling uncertainty at high fission gas release. Nevertheless, significantly higher deviations may be expected for values around 10% and lower. Implications are discussed in terms of directions of research for the improved modeling of fission gas behavior for engineering purposes.

  14. Research efforts on fuels, fuel models, and fire behavior in eastern hardwood forests

    Science.gov (United States)

    Thomas A. Waldrop; Lucy Brudnak; Ross J. Phillips; Patrick H. Brose

    2006-01-01

    Although fire was historically important to most eastern hardwood systems, its reintroduction by prescribed burning programs has been slow. As a result, less information is available on these systems to fire managers. Recent research and nationwide programs are beginning to produce usable products to predict fuel accumulation and fire behavior. We introduce some of...

  15. Basic research on high-uranium density fuels for research and test reactors

    International Nuclear Information System (INIS)

    Ugajin, M.; Itoh, A.; Akabori, M.

    1992-01-01

    High-uranium density fuels, uranium silicides (U 3 Si 2 , U 3 Si) and U 6 Me-type uranium alloys (Me = Fe, Mn, Ni), were prepared and examined metallurgically as low-enriched uranium (LEU) fuels for research and test reactors. Miniature aluminum-dispersion plate-type fuel (miniplate) and aluminum-clad disk-type fuel specimens were fabricated and subjected to the neutron irradiation in JMTR (Japan Materials Testing Reactor). Fuel-aluminum compatibility tests were conducted to elucidate the extent of reaction and to identify reaction products. The relative stability of the fuels in an aluminum matrix was established at 350degC or above. Experiments were also performed to predict the chemical form of the solid fission-products in the uranium silicide (U 3 Si 2 ) simulating a high burnup anticipated for reactor service. (author)

  16. BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

    Energy Technology Data Exchange (ETDEWEB)

    Bevard, Bruce Balkcom [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mertyurek, Ugur [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Scaglione, John M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-10-01

    The objective of this report is to identify issues and support documentation and identify and detail existing research on spent fuel dry storage; provide information to support potential R&D for the UKABWR (United Kingdom Advanced Boiling Water Reactor) Spent Fuel Interim Storage (SFIS) Pre-Construction Safety Report; and support development of answers to questions developed by the regulator. Where there are gaps or insufficient data, Oak Ridge National Laboratory (ORNL) has summarized the research planned to provide the necessary data along with the schedule for the research, if known. Spent nuclear fuel (SNF) from nuclear power plants has historically been stored on site (wet) in spent fuel pools pending ultimate disposition. Nuclear power users (countries, utilities, vendors) are developing a suite of options and set of supporting analyses that will enable future informed choices about how best to manage these materials. As part of that effort, they are beginning to lay the groundwork for implementing longer-term interim storage of the SNF and the Greater Than Class C (CTCC) waste (dry). Deploying dry storage will require a number of technical issues to be addressed. For the past 4-5 years, ORNL has been supporting the U.S. Department of Energy (DOE) in identifying these key technical issues, managing the collection of data to be used in issue resolution, and identifying gaps in the needed data. During this effort, ORNL subject matter experts (SMEs) have become expert in understanding what information is publicly available and what gaps in data remain. To ensure the safety of the spent fuel under normal and frequent conditions of wet and subsequent dry storage, intact fuel must be shown to: 1.Maintain fuel cladding integrity; 2.Maintain its geometry for cooling, shielding, and subcriticality; 3.Maintain retrievability, and damaged fuel with pinhole or hairline cracks must be shown not to degrade further. Where PWR (pressurized water reactor) information is

  17. 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

  18. Fuel shuffling optimization for the Delft research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Geemert, R. van; Hoogenboom, J.E.; Gibcus, H.P.M. [Delft Univ. of Technology, Interfaculty Reactor Inst., Delft (Netherlands); Quist, A.J. [Delft Univ., Fac. of Applied Mathematics and Informatics, Delft (Netherlands)

    1997-07-01

    A fuel shuffling optimization procedure is proposed for the Hoger Onderwijs Reactor (HOR) in Delft, the Netherlands, a 2 MWth swimming-pool type research reactor. In order to cope with the fluctuatory behaviour of objective functions in loading pattern optimization, the proposed cyclic permutation optimization procedure features a gradual transition from global to local search behaviour via the introduction of stochastic tests for the number of fuel assemblies involved in a cyclic permutation. The possible objectives and the safety and operation constraints, as well as the optimization procedure, are discussed, followed by some optimization results for the HOR. (author) 5 figs., 4 refs.

  19. Fuel shuffling optimization for the Delft research reactor

    International Nuclear Information System (INIS)

    Geemert, R. van; Hoogenboom, J.E.; Gibcus, H.P.M.; Quist, A.J.

    1997-01-01

    A fuel shuffling optimization procedure is proposed for the Hoger Onderwijs Reactor (HOR) in Delft, the Netherlands, a 2 MWth swimming-pool type research reactor. In order to cope with the fluctuatory behaviour of objective functions in loading pattern optimization, the proposed cyclic permutation optimization procedure features a gradual transition from global to local search behaviour via the introduction of stochastic tests for the number of fuel assemblies involved in a cyclic permutation. The possible objectives and the safety and operation constraints, as well as the optimization procedure, are discussed, followed by some optimization results for the HOR. (author)

  20. Fuel shuffling optimization for the Delft research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Geemert, R. van; Hoogenboom, J.E.; Gibcus, H.P.M. [Delft Univ. of Technology, Interfaculty Reactor Inst., Delft (Netherlands); Quist, A.J. [Delft Univ., Fac. of Applied Mathematics and Informatics, Delft (Netherlands)

    1997-07-01

    A fuel shuffling optimization procedure is proposed for the Hoger Onderwijs Reactor (HOR) in Delft, the Netherlands, a 2 MWth swimming-pool type research reactor. In order to cope with the fluctuatory behaviour of objective functions in loading pattern optimization, the proposed cyclic permutation optimization procedure features a gradual transition from global to local search behaviour via the introduction of stochastic tests for the number of fuel assemblies involved in a cyclic permutation. The possible objectives and the safety and operation constraints, as well as the optimization procedure, are discussed, followed by some optimization results for the HOR. (author)

  1. Design and research of fuel element for pulsed reactor

    International Nuclear Information System (INIS)

    Tian Sheng

    1994-05-01

    The fuel element is the key component for pulsed reactor and its design is one of kernel techniques for pulsed reactor. Following the GA Company of US the NPIC (Nuclear Power Institute of China) has mastered this technique. Up to now, the first pulsed reactor in China (PRC-1) has been safely operated for about 3 years. The design and research of fuel element undertaken by NPIC is summarized. The verification and evaluation of this design has been carried out by using the results of measured parameters during operation and test of PRC-1 as well as comparing the design parameters published by others

  2. Some tooling for manufacturing research reactor fuel plates

    International Nuclear Information System (INIS)

    Knight, R.W.

    1999-01-01

    This paper will discuss some of the tooling necessary to manufacture aluminum-based research reactor fuel plates. Most of this tooling is intended for use in a high-production facility. Some of the tools shown have manufactured more than 150,000 pieces. The only maintenance has been sharpening. With careful design, tools can be made to accommodate the manufacture of several different fuel elements, thus, reducing tooling costs and maintaining tools that the operators are trained to use. An important feature is to design the tools using materials with good lasting quality. Good tools can increase return on investment. (author)

  3. Some Tooling for Manufacturing Research Reactor Fuel Plates

    International Nuclear Information System (INIS)

    Knight, R.W.

    1999-01-01

    This paper will discuss some of the tooling necessary to manufacture aluminum-based research reactor fuel plates. Most of this tooling is intended for use in a high-production facility. Some of the tools shown have manufactured more than 150,000 pieces. The only maintenance has been sharpening. With careful design, tools can be made to accommodate the manufacture of several different fuel elements, thus, reducing tooling costs and maintaining tools that the operators are trained to use. An important feature is to design the tools using materials with good lasting quality. Good tools can increase return on investment

  4. Final Technical Report for the MIT Annular Fuel Research Project

    International Nuclear Information System (INIS)

    Mujid S. Kazimi; Pavel Hejzlar

    2008-01-01

    MIT-NFC-PR-082 (January 2006) Abstract This summary provides an overview of the results of the U.S. DOE funded NERI (Nuclear Research Energy Initiative) program on development of the internally and externally cooled annular fuel for high power density PWRs. This new fuel was proposed by MIT to allow a substantial increase in power density (on the order of 30% or higher) while maintaining or improving safety margins. A comprehensive study was performed by a team consisting of MIT (lead organization), Westinghouse Electric Corporation, Gamma Engineering Corporation, Framatome ANP(formerly Duke Engineering) and Atomic Energy of Canada Limited

  5. Research-scale melter test report

    Energy Technology Data Exchange (ETDEWEB)

    Cooper, M.F.; Elliott, M.L.; Eyler, L.L.; Freeman, C.J.; Higginson, J.J.; Mahoney, L.A.; Powell, M.R.

    1994-05-01

    The Melter Performance Assessment (MPA) activity in the Pacific Northwest Laboratory`s (PNL) Hanford Waste Vitrification Plant (HWVP) Technology Development (PHTD) effort is intended to determine the impact of noble metals on the operational life of the reference HWVP melter. As a part of this activity, a parametric melter test was completed using a Research-Scale Melter (RSM). The RSM is a small, approximately 1/100-scale melter, 6-in.-diameter, that allows rapid changing of process conditions and subsequent re-establishment of a steady-state condition. The test matrix contained nine different segments that varied the melter operating parameters (glass and plenum temperatures) and feed properties (oxide concentration, redox potential, and noble metal concentrations) so that the effects of these parameters on noble metal agglomeration on the melter floor could be evaluated. The RSM operated for 48 days and consumed 1,300 L of feed, equating to 153 tank turnovers. The run produced 531 kg of glass. During the latter portion of the run, the resistance between the electrodes decreased. Upon destructive examination of the melter, a layer of noble metals was found on the bottom. This was surprising because the glass residence time in the RSM is only 10% of the HWVP plant melter. The noble metals layer impacted the melter significantly. Approximately 1/3 of one paddle electrode was melted or corroded off. The cause is assumed to be localized heating from short circuiting of the electrode to the noble metal layer. The metal layer also removed approximately 1/2 in. of the refractory on the bottom of the melter. The mechanism for this damage is not presently known.

  6. Research-scale melter test report

    International Nuclear Information System (INIS)

    Cooper, M.F.; Elliott, M.L.; Eyler, L.L.; Freeman, C.J.; Higginson, J.J.; Mahoney, L.A.; Powell, M.R.

    1994-05-01

    The Melter Performance Assessment (MPA) activity in the Pacific Northwest Laboratory's (PNL) Hanford Waste Vitrification Plant (HWVP) Technology Development (PHTD) effort is intended to determine the impact of noble metals on the operational life of the reference HWVP melter. As a part of this activity, a parametric melter test was completed using a Research-Scale Melter (RSM). The RSM is a small, approximately 1/100-scale melter, 6-in.-diameter, that allows rapid changing of process conditions and subsequent re-establishment of a steady-state condition. The test matrix contained nine different segments that varied the melter operating parameters (glass and plenum temperatures) and feed properties (oxide concentration, redox potential, and noble metal concentrations) so that the effects of these parameters on noble metal agglomeration on the melter floor could be evaluated. The RSM operated for 48 days and consumed 1,300 L of feed, equating to 153 tank turnovers. The run produced 531 kg of glass. During the latter portion of the run, the resistance between the electrodes decreased. Upon destructive examination of the melter, a layer of noble metals was found on the bottom. This was surprising because the glass residence time in the RSM is only 10% of the HWVP plant melter. The noble metals layer impacted the melter significantly. Approximately 1/3 of one paddle electrode was melted or corroded off. The cause is assumed to be localized heating from short circuiting of the electrode to the noble metal layer. The metal layer also removed approximately 1/2 in. of the refractory on the bottom of the melter. The mechanism for this damage is not presently known

  7. Bioenergy research programme. Yearbook 1996. Production of wood fuels; Bioenergian tutkimusohjelma. Vuosikirja 1996. Puupolttoaineiden tuotantotekniikka

    Energy Technology Data Exchange (ETDEWEB)

    Nikku, P [ed.

    1997-12-01

    The aim of the programme is to increase the use of economically profitable and environmentally sound bioenergy by improving the competitiveness of present peat and wood fuels. Research and development projects will also develop new economically competitive biofuels, new equipment and methods for production, handling and utilisation of biofuels. The total funding for 1996 was 27.3 million FIM and the number of projects 63. The number of projects concerning wood fuels production was 36. The main goals of the research are to develop new production methods for wood fuels in order to decrease the production costs to the level of imported fuels (100 km distance). The second goal is to decrease the small scale production costs by 20 % as compared with the 1992 technology level. Also, new harvesting technology and new work methods will be developed for forest owners and small-entrepreneurs in the course of the programme. Results of the projects carried out in 1996 in this programme are presented in this publication. The integrated harvesting methods, which supply both raw material to wood products industry and wood fuel for energy production, have been chosen the main research areas because they seem to be most promising. Most of the projects are focused in the wood fuel production from first thinnings and from final fellings. The projects broadly covered the research area focusing from material flows, productivity studies, basic wood properties to several case studies. The follow up project of Evaluation-drum chipper was completed with good fuel quality and productivity results. Also the large Forest Energy Project of Central Finland was completed. The project was a significant technology transfer and information dissemination project. (orig.)

  8. Atomization of U3Si2 for research reactor fuel

    International Nuclear Information System (INIS)

    Kim, C.K.; Kim, K.H.; Lee, C.T.; Kuk, I.H.

    1995-01-01

    Rotating disk atomization technique is applied to KMRR (Korea Multi-purpose Research Reactor) fuel fabrication. A rotating disk atomizer is designed and manufactured locally and U-4.0 wt. % Si alloy powders are produced. The atomized powders are heat-treated to transform into U 3 Si and the mixture of U 3 Si and Al are extruded to fuel meat. Most of the atomized powders are spherical in shape. The microstructure of the powder is fine due to the rapid solidification. The time required for peritectoid reaction is reduced due to the fine microstructures and the resultant U 3 Si grain size is finer than ever obtained from ingot process. The mechanical properties of the fuel meat are improved: yield strength about 30 %, tensile strength 10% and elongation 250 % increased. (author)

  9. Convective parameters in fuel elements for research nuclear reactors

    International Nuclear Information System (INIS)

    Lopez Martinez, C.D.

    1992-01-01

    The study of a prototype for the simulation of fuel elements for research nuclear reactors by natural convection in water is presented in this paper. This project is carry out in the thermofluids laboratory of National Institute of Nuclear Research. The fuel prototype has already been test for natural convection in air, and the first results in water are presented in this work. In chapter I, a general description of Triga Mark III is made, paying special atention to fuel-moderator components. In chapter II and III an approach to convection subject in its global aspects is made, since the intention is to give a general idea of the events occuring around fuel elements in a nuclear reactor. In chapter II, where an emphasis on forced convection is made, some basic concepts for forced convection as well as for natural convection are included. The subject of flow through cylinders is annotated only as a comparative reference with natural convection in vertical cylinders, noting the difference between used correlations and the involved variables. In chapter III a compilation of correlation found in the bibliography about natural convection in vertical cylinders is presented, since its geometry is the more suitable in the analysis of a fuel rod. Finally, in chapter IV performed experiments in the test bench are detailed, and the results are presented in form of tables and graphs, showing the used equations for the calculations and the restrictions used in each case. For the analysis of the prototypes used in the test bench, a constant and uniform flow of heat in the whole length of the fuel rod is considered. At the end of this chapter, the work conclusions and a brief explanation of the results are presented (Author)

  10. Application of IAEA's International Nuclear Event Scale to events at testing/research reactors in Japan

    International Nuclear Information System (INIS)

    Nozawa, Masao; Watanabe, Norio

    1999-01-01

    The International Nuclear Event Scale (INES) is a means for providing prompt, clear and consistent information related to nuclear events and facilitating communication between the nuclear community, the media and the public on such events. This paper describes the INES rating process for events at testing/research reactors and nuclear fuel processing facilities and experience on the application of the INES scale in Japan. (author)

  11. Feed process studies: Research-Scale Melter

    Energy Technology Data Exchange (ETDEWEB)

    Whittington, K.F.; Seiler, D.K.; Luey, J.; Vienna, J.D.; Sliger, W.A.

    1996-09-01

    In support of a two-phase approach to privatizing the processing of hazardous and radioactive waste at Hanford, research-scale melter (RSM) experiments were conducted to determine feed processing characteristics of two potential privatization Phase 1 high-level waste glass formulations and to determine if increased Ag, Te, and noble metal amounts would have bad effects. Effects of feed compositions and process conditions were examined for processing rate, cold cap behavior, off-gas, and glass properties. The 2 glass formulations used were: NOM-2 with adjusted waste loading (all components except silica and soda) of 25 wt%, and NOM-3 (max waste loaded glass) with adjusted waste loading of 30 wt%. The 25 wt% figure is the minimum required in the privatization Request for Proposal. RSM operated for 19 days (5 runs). 1010 kg feed was processed, producing 362 kg glass. Parts of runs 2 and 3 were run at 10 to 30 degrees above the nominal temperature 1150 C, with the most significant processing rate increase in run 3. Processing observations led to the choice of NOM-3 for noble metal testing in runs 4 and 5. During noble metal testing, processing rates fell 50% from baseline. Destructive analysis showed that a layer of noble metals and noble metal oxides settled on the floor of the melter, leading to current ``channeling`` which allowed the top section to cool, reducing production rates.

  12. Feed process studies: Research-Scale Melter

    International Nuclear Information System (INIS)

    Whittington, K.F.; Seiler, D.K.; Luey, J.; Vienna, J.D.; Sliger, W.A.

    1996-09-01

    In support of a two-phase approach to privatizing the processing of hazardous and radioactive waste at Hanford, research-scale melter (RSM) experiments were conducted to determine feed processing characteristics of two potential privatization Phase 1 high-level waste glass formulations and to determine if increased Ag, Te, and noble metal amounts would have bad effects. Effects of feed compositions and process conditions were examined for processing rate, cold cap behavior, off-gas, and glass properties. The 2 glass formulations used were: NOM-2 with adjusted waste loading (all components except silica and soda) of 25 wt%, and NOM-3 (max waste loaded glass) with adjusted waste loading of 30 wt%. The 25 wt% figure is the minimum required in the privatization Request for Proposal. RSM operated for 19 days (5 runs). 1010 kg feed was processed, producing 362 kg glass. Parts of runs 2 and 3 were run at 10 to 30 degrees above the nominal temperature 1150 C, with the most significant processing rate increase in run 3. Processing observations led to the choice of NOM-3 for noble metal testing in runs 4 and 5. During noble metal testing, processing rates fell 50% from baseline. Destructive analysis showed that a layer of noble metals and noble metal oxides settled on the floor of the melter, leading to current ''channeling'' which allowed the top section to cool, reducing production rates

  13. Geoscience research for the Canadian nuclear fuel waste management program

    International Nuclear Information System (INIS)

    Whitaker, S.H.

    1987-01-01

    The Canadian Nuclear Fuel Waste Management Program is assessing the concept of deep disposal of nuclear fuel waste in plutonic rock. As part of that assessment, a broad program of geoscience and geotechnical work has been undertaken to develop methods for characterizing sites, incorporating geotechnical data into disposal facility design, and incorporating geotechnical data into environmental and safety assessment of the disposal system. General field investigations are conducted throughout the Precambrian Shield, subsurface investigations are conducted at designated field research areas, and in situ rock mass experiments are being conducted in an Underground Research Laboratory. Samples from the field research areas and elsewhere are subjected to a wide range of tests and experiments in the laboratory to develop an understanding of the physical and chemical processes involved in ground-water-rock-waste interactions. Mathematical models to simulate these processes are developed, verified and validated. 114 refs.; 13 figs

  14. 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

  15. Proposed nuclear weapons nonproliferation policy concerning foreign research reactor spent nuclear fuel: Appendix B, foreign research reactor spent nuclear fuel characteristics and transportation casks. Volume 2

    International Nuclear Information System (INIS)

    1995-03-01

    This is Appendix B of a draft Environmental Impact Statement (EIS) on a Proposed Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel. It discusses relevant characterization and other information of foreign research reactor spent nuclear fuel that could be managed under the proposed action. It also discusses regulations for the transport of radioactive materials and the design of spent fuel casks

  16. Social science to improve fuels management: a synthesis of research on assessing social acceptability of fuels treatments

    Science.gov (United States)

    Terry C. Daniel; Michael Valdiserri; Carrie R. Daniel; Pamela Jakes; Pamela Jakes; Susan Barro

    2005-01-01

    A series of syntheses were commissioned by the USDA Forest Service to aid in fuels mitigation project planning. This synthesis focuses on research for assessing the social acceptability of fuels treatments. The synthesis is structured around six important considerations for any social acceptability assessment: defining the fuels treatments being assessed; representing...

  17. Summary engineering description of underwater fuel storage facility for foreign research reactor spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Dahlke, H.J.; Johnson, D.A.; Rawlins, J.K.; Searle, D.K.; Wachs, G.W.

    1994-10-01

    This document is a summary description for an Underwater Fuel Storage Facility (UFSF) for foreign research reactor (FRR) spent nuclear fuel (SNF). A FRR SNF environmental Impact Statement (EIS) is being prepared and will include both wet and dry storage facilities as storage alternatives. For the UFSF presented in this document, a specific site is not chosen. This facility can be sited at any one of the five locations under consideration in the EIS. These locations are the Idaho National Engineering Laboratory, Savannah River Site, Hanford, Oak Ridge National Laboratory, and Nevada Test Site. Generic facility environmental impacts and emissions are provided in this report. A baseline fuel element is defined in Section 2.2, and the results of a fission product analysis are presented. Requirements for a storage facility have been researched and are summarized in Section 3. Section 4 describes three facility options: (1) the Centralized-UFSF, which would store the entire fuel element quantity in a single facility at a single location, (2) the Regionalized Large-UFSF, which would store 75% of the fuel element quantity in some region of the country, and (3) the Regionalized Small-UFSF, which would store 25% of the fuel element quantity, with the possibility of a number of these facilities in various regions throughout the country. The operational philosophy is presented in Section 5, and Section 6 contains a description of the equipment. Section 7 defines the utilities required for the facility. Cost estimates are discussed in Section 8, and detailed cost estimates are included. Impacts to worker safety, public safety, and the environment are discussed in Section 9. Accidental releases are presented in Section 10. Standard Environmental Impact Forms are included in Section 11.

  18. Mononitride fuel and large scale nuclear power industry

    International Nuclear Information System (INIS)

    Orlov, V.V.; Sila-Novitsky, A.G.; Smirnov, V.S.

    1997-01-01

    Work to study nitride fuel, conducted in Russia a long time, has indicated interrelation between operational properties, including the data of gas fission release, the content of impurities and synthesis processes. Recommended method production fuel from initial metal, as more profitable that one from initial oxide, needs development of electro-chemical reprocessing with metal as end product and contributory to nonproliferation and cost efficient fuel cycle. Irradiation tests UN have shown no fission gaseous bubble within the lead sublayer. Taking as example the core design of BREST reactor of 300 MW(e) power there have been shown a possibility to put into reality a principle of nature safety in fast reactor by using inherent properties of U-Pu mononitride fuel (high density and thermal conductivity, low fission gas release and swelling) and lead coolant (low neutron absorption and moderation, high density and boiling temperature, chemical passive) as well as original technical solution on core major component (passive self-regulatory). (author). 19 refs, 8 figs, 1 tab

  19. Plant-scale anodic dissolution of unirradiated N-Reactor fuel

    International Nuclear Information System (INIS)

    Gay, E.C.; Miller, W.E.; Laidler, J.J.

    1995-01-01

    Anodic dissolution tests were made with unirradiated N-Reactor fuel to determine the fuel segment length, diameter, and shape required for high throughput electrorefiner treatment for ultimate disposal in a geologic repository. Based on these tests, a conceptual design was produced of an electrorefiner for a full-scale plant to process N-Reactor spent fuel. In this design, the diameter of an electrode assembly is about 0.6 m (25 in.). Eight of these assemblies in an electrorefiner would accommodate a 1.333-metric-ton batch of N-Reactor fuel. Electrorefining would proceed at a rate of 40 kg uranium per hour

  20. 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

  1. Laboratory Scale Coal And Biomass To Drop-In Fuels (CBDF) Production And Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Lux, Kenneth [Altex Technologies Corporation, Sunnyvale, CA (United States); Imam, Tahmina [Altex Technologies Corporation, Sunnyvale, CA (United States); Chevanan, Nehru [Altex Technologies Corporation, Sunnyvale, CA (United States); Namazian, Mehdi [Altex Technologies Corporation, Sunnyvale, CA (United States); Wang, Xiaoxing [Pennsylvania State Univ., University Park, PA (United States); Song, Chunshan [Pennsylvania State Univ., University Park, PA (United States)

    2016-06-29

    This Final Technical Report describes the work and accomplishments of the project entitled, “Laboratory Scale Coal and Biomass to Drop-In Fuels (CBDF) Production and Assessment.” The main objective of the project was to fabricate and test a lab-scale liquid-fuel production system using coal containing different percentages of biomass such as corn stover and switchgrass at a rate of 2 liters per day. The system utilizes the patented Altex fuel-production technology, which incorporates advanced catalysts developed by Pennsylvania State University. The system was designed, fabricated, tested, and assessed for economic and environmental feasibility relative to competing technologies.

  2. RANS based CFD methodology for a real scale 217-pin wire-wrapped fuel assembly of KAERI PGSFR

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Jae-Ho, E-mail: jhjeong@kaeri.re.kr [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseoung-gu, Daejeon (Korea, Republic of); Song, Min-Seop [Department of Nuclear Engineering, Seoul National University, 559 Gwanak-ro, Gwanak-gu, Seoul (Korea, Republic of); Lee, Kwi-Lim [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseoung-gu, Daejeon (Korea, Republic of)

    2017-03-15

    Highlights: • This paper presents a suitable way for a practical RANS based CFD methodology which is applicable to real scale 217-pin wire-wrapped fuel assembly of KAERI PGSFR. • A key point of differentiation of the RANS based CFD methodology in this study is adapting an innovative grid generation method using a fortran based in-house code with a GGI function in a general-purpose commercial CFD code, CFX. • The RANS based CFD methodology is implemented with high resolution scheme and SST turbulence model in the 7-pin 37-pin, and 127-pin wire-wrapped fuel assembly of PNC and JNC. Furthermore, the RANS based CFD methodology can be successfully extended to the real scale 217-pin wire-wrapped fuel bundles of KAERI PGSFR. • Three-dimensional thermal-hydraulic characteristics have been also investigated briefly. - Abstract: This paper presents a suitable way for a practical RANS (Reynolds Averaged Navier-Stokes simulation) based CFD (Computational Fluid Dynamics) methodology which is applicable to real scale 217-pin wire-wrapped fuel assembly of KAERI (Korea Atomic Energy Research Institute) PGSFR (Prototype Gen-IV Sodium-cooled Fast Reactor). The main purpose of the current study is to support license issue for the KAERI PGSFR core safety and to elucidate thermal-hydraulic characteristics in a 217-pin wire-wrapped fuel assembly of KAERI PGSFR. A key point of differentiation of the RANS based CFD methodology in this study is adapting an innovative grid generation method using a fortran based in-house code with a GGI (General Grid Interface) function in a general-purpose commercial CFD code, CFX. The innovative grid generation method with GGI function can achieve to simulate a real wire shape with minimizing cell skewness. The RANS based CFD methodology is implemented with high resolution scheme in convection term and SST (Shear Stress Transport) turbulence model in the 7-pin 37-pin, and 127-pin wire-wrapped fuel assembly of PNC (Power reactor and Nuclear fuel

  3. Effect of reduced enrichment on the fuel cycle for research reactors

    International Nuclear Information System (INIS)

    Travelli, A.

    1982-01-01

    The new fuels developed by the RERTR Program and by other international programs for application in research reactors with reduced uranium enrichment (<20% EU) are discussed. It is shown that these fuels, combined with proper fuel-element design and fuel-management strategies, can provide at least the same core residence time as high-enrichment fuels in current use, and can frequently significantly extend it. The effect of enrichment reduction on other components of the research reactor fuel cycle, such as uranium and enrichment requirements, fuel fabrication, fuel shipment, and reprocessing are also briefly discussed with their economic implications. From a systematic comparison of HEU and LEU cores for the same reference research reactor, it is concluded that the new fuels have a potential for reducing the research reactor fuel cycle costs while reducing, at the same time, the uranium enrichment of the fuel

  4. Pilot-scale equipment development for lithium-based reduction of spent oxide fuel

    International Nuclear Information System (INIS)

    Herrmann, S. D.

    1998-01-01

    An integral function of the electrometallurgical conditioning of DOE spent nuclear fuel is the standardization of waste forms. Argonne National Laboratory (ANL) has developed and is presently demonstrating the electrometallurgical conditioning of sodium-bonded metal fuel from Experimental Breeder Reactor II, resulting in uranium, ceramic waste, and metal waste forms. Engineering studies are underway at ANL in support of pilot-scale equipment development, which would precondition irradiated oxide fuel and likewise demonstrate the application of electrometallurgical conditioning to such non-metallic fuels. This paper highlights the integration of proposed spent oxide fuel conditioning with existing electrometallurgical processes. Additionally, technical bases for engineering activities to support a scale up of an oxide reduction process are described

  5. Can we combine symptom scales for collaborative research projects?

    LENUS (Irish Health Repository)

    Lyne, John P

    2012-02-01

    Collaborative research projects have the potential to answer important research questions, which may otherwise require huge resources, funding, and time to complete. There are several scales for measuring psychotic symptoms in schizophrenia and other psychotic disorders, with the Scale for Assessment of Positive Symptoms (SAPS), Scale for Assessment of Negative Symptoms (SANS), Positive and Negative Symptom Scale (PANSS), and the Brief Psychiatric Rating Scale (BPRS) being among the most commonly used. High quality research efforts have used these three scales in different projects, and in order to merge study efforts, some means of combining data from these scales may be necessary. We reviewed correlations in published studies for these three scales, finding them to be highly correlated, however on comparison of the three scales there were considerable clinical differences between them. The paper discusses potential methods for combining the scales in collaborative research, including use of the recently developed standardised remission criteria for schizophrenia.

  6. Thermal simulations and tests in the development of a helmet transport spent fuel elements Research Reactor

    International Nuclear Information System (INIS)

    Saliba, R.; Quintana, F.; Márquez Turiello, R.; Furnari, J.C.; Pimenta Mourão, R.

    2013-01-01

    A packaging for the transport of irradiated fuel from research reactors was designed by a group of researchers to improve the capability in the management of spent fuel elements from the reactors operated in the region. Two half-scale models for MTR fuel were constructed and tested so far and a third one for both MTR and TRIGA fuels will be constructed and tested next. Four test campaigns have been carried out, covering both normal and hypothetical accident conditions of transportation. The thermal test is part of the requirements for the qualification of transportation packages for nuclear reactors spent fuel elements. In this paper both the numerical modelling and experimental thermal tests performed are presented and discussed. The cask is briefly described as well as the finite element model developed and the main adopted hypotheses for the thermal phenomena. The results of both numerical runs and experimental tests are discussed as a tool to validate the thermal modelling. The impact limiters, attached to the cask for protection, were not modelled. (author) [es

  7. 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

  8. Gamma scanning of full scale HTR fuel elements

    International Nuclear Information System (INIS)

    Harrison, T.A.; Simpson, J.A.H.; Nabielek, H.

    1983-04-01

    Gamma scanning for the determination of burn-up and fission product inventory has been developed at the Dragon Project, suitable for measurements on fuel elements and segments from full-sized integral block elements. This involved the design and construction of a new lead flask with sophisticated collimator design. State-of-the art gamma spectrometric equipment was set up to cope with strong variations of count-rate and high data throughput. Software efforts concentrated on the calculation of the self absorption and absorption corrections in the complicated geometry of multi-hole graphite block segments with a corrugated circumference. The techniques described here are applicable to the non-destructive examination of a wide range of fuel element designs. (author)

  9. Development of large scale internal reforming molten carbonate fuel cell

    Energy Technology Data Exchange (ETDEWEB)

    Sasaki, A.; Shinoki, T.; Matsumura, M. [Mitsubishi Electric Corp., Hyogo (Japan)

    1996-12-31

    Internal Reforming (IR) is a prominent scheme for Molten Carbonate Fuel Cell (MCFC) power generating systems in order to get high efficiency i.e. 55-60% as based on the Higher Heating Value (HHV) and compact configuration. The Advanced Internal Reforming (AIR) technology has been developed based on two types of the IR-MCFC technology i.e. Direct Internal Reforming (DIR) and Indirect Internal Reforming (DIR).

  10. Supply of wood fuel from small-scale woodlands for small-scale heating

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-07-01

    This report summarises the findings of a study aimed at stimulating a market for wood fuels. A desk study of harvesting in existing small woodland was conducted, and thirteen case studies covering early broadleaved thinnings, mixed broadleaved coppice, and crownwood, scrub and residues were examined to obtain information on woodland types, wood fuel supply, and combustion equipment. Details are given of the measurement of moisture content of woodchips and stacked roundwood, wood volume and green density, harvesting options, crop and site variables, and production and costs of wood fuels. Usage of wood fuels, and the drying of small roundwood was considered. (UK)

  11. Spent fuel management options for research reactors in Latin America

    International Nuclear Information System (INIS)

    2006-06-01

    Research reactors (RRs) have been operated in Latin America since the late 1950s, and a total of 23 RRs have been built in the region. At the time of writing (November 2005), 18 RRs are in operation, 4 have been shut down and 1 has been decommissioned. The number of operating RRs in Latin America represents around 6% of the existing operational RRs worldwide and around 21% of the RRs operating in developing countries. Common to all RRs in the region is a consistent record of safe and successful operation. With the purpose of carrying out a collaborative study of different aspects of the management of spent fuel from RRs, some countries from the region proposed to the IAEA in 2000 the organization of a Regional Project. The project (IAEA TC Regional Project RLA/4/018) that was approved for the biennium 2001-2002 and extended for 2003-2004 included the participation of Argentina, Brazil, Chile, Mexico and Peru. The main objectives of this project were: (a) to define the basic conditions for a regional strategy for managing spent fuel that will provide solutions compatible with the economic and technological realities of the countries involved; and (b) to determine what is needed for the temporary wet and dry storage of spent fuel from the research reactors in the countries of the Latin American region that participated in the project. This TECDOC is based on the results of TC Regional Project RLA/4/018. This project was successful in identifying and assessing a number of viable alternatives for RRSF management in the Latin American region. Options for operational and interim storage, spent fuel conditioning and final disposal have been carefully considered. This report presents the views of Latin American experts on RR spent fuel management and will be useful as reference material for the Latin American RR community, decision making authorities in the region and the public in general

  12. Experimental research on safety assurance of advanced WWER fuel cycles

    International Nuclear Information System (INIS)

    Krainov, Ju.; Kukushkin, Ju.

    2002-01-01

    The paper presents the results of experimental investigations on substantiation of implementation of a modernized butt joint for the WWER-440 reactor, carried out in the critical test facility 'P' in the RRC 'Kurchatov Institute'. The comparison results of the calculation and experimental data obtained in the physical startup of Volgodonsk NPP-1 with the WWER-1000 are also given. In the implementation of four-year fuel cycle in the WWER-440 with the average enrichment of fuel makeup 3.82% it was solved to conduct experimental research of power distribution in the vicinity of control rod butt junction. Moreover, it was assumed that adequate actions should be applied to eliminate inadmissible power jumps, if necessary. It is not available to measure their values in NPP conditions. Therefore, the power distribution near the butt joint was studied in a 19-rod bank installed in the critical test facility 'P' first for the normal design of the joint when surrounding fuel assemblies enrichment goes up. Then a set of calculation and tests was fulfilled to optimize a butt junction design. On the base of this research the composition of a butt junction was advanced by placing Hf plates into the junction. The effectiveness of modernized butt joint design was experimentally confirmed. In Volgodonsk NPP-1 with WWER-1000 the four-year fuel cycle is being implemented. During the physical startup of the reactor the measurements of the reactivity effects and coefficients were measured at the minimum controlled flux level, and the parameters of a number of critical states were recorded. The data obtained were compared with the calculation. The validity of the certified code package for forecasting the neutronic characteristics of WWER-1000 cores in the implementation of a four year fuel cycle has been supported (Authors)

  13. Cost evaluation of a commercial-scale DUPIC fuel fabrication facility (Part I) -Summary

    Energy Technology Data Exchange (ETDEWEB)

    Ko, Won Il; Choi, Hang Bok; Yang, Myung Seung [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-08-01

    A conceptual design of a commercial scale DUPIC fuel fabrication facility was initiated to provide some insights into the costs associated with construction, operation, and decommissioning. The primary conclusion of this report is that it is feasible to design, license, construct, test, and operate a facility that will process 400 MTHE/yr of spent PWR fuel and reconfigure the fuel into CANDU fuel bundles at a reasonable unit cost of the fuel material. Although DUPIC fuel fabrication by vibropacking method is clearly cheaper than that of the pellet method, the feasibility of vibropac technology for DUPIC fuel fabrication and use of vibroac fuel in CANDU reactors may has to be studied in depth in order to use as an alternative to the conventional pellet fuel method. Especially, there are some questions on meeting the CANDU requirements in thermal and mechanical terms as well as density of fuel. Wherever possible, this report used representative costs of currently available technologies as the bases for cost estimation. It should also be noted that the conceptual design and cost information contained in this report was extracted from the public domain and general open literature. Later studies have to focus on other important areas of concern such as safety, security, safeguards, process optimization etc. 7 figs., 6 tabs. (Author)

  14. Scaling of the burning efficiency for multicomponent fuel pool fires

    DEFF Research Database (Denmark)

    van Gelderen, Laurens; Farahani, Hamed Farmahini; Rangwala, Ali S.

    In order to improve the validity of small scale crude oil burning experiments, which seem to underestimate the burning efficiency obtained in larger scales, the gasification mechanism of crude oil was studied. Gasification models obtained from literature were used to make a set of predictions for...... an external heat source to simulate the larger fire size are currently in process....

  15. Fuel burnup analysis for the Moroccan TRIGA research reactor

    International Nuclear Information System (INIS)

    El Bakkari, B.; El Bardouni, T.; Nacir, B.; El Younoussi, C.; Boulaich, Y.; Boukhal, H.; Zoubair, M.

    2013-01-01

    Highlights: ► A fuel burnup analysis of the 2 MW TRIGA MARK II Moroccan research reactor was established. ► Burnup calculations were done by means of the in-house developed burnup code BUCAL1. ► BUCAL1 uses the MCNP tallies directly in the calculation of the isotopic inventories. ► The reactor life time was found to be 3360 MW h considering full power operating conditions. ► Power factors and fluxes of the in-core irradiation positions are strongly affected by burnup. -- Abstract: The fundamental advantage and main reason to use Monte Carlo methods for burnup calculations is the possibility to generate extremely accurate burnup dependent one group cross-sections and neutron fluxes for arbitrary core and fuel geometries. Yet, a set of values determined for a material at a given position and time remains accurate only in a local region, in which neutron spectrum and flux vary weakly — and only for a limited period of time, during which changes of the local isotopic composition are minor. This paper presents the approach of fuel burnup evaluation used at the Moroccan TRIGA MARK II research reactor. The approach is essentially based upon the utilization of BUCAL1, an in-house developed burnup code. BUCAL1 is a FORTRAN computer code designed to aid in analysis, prediction, and optimization of fuel burnup performance in nuclear reactors. The code was developed to incorporate the neutron absorption reaction tally information generated directly by MCNP5 code in the calculation of fissioned or neutron-transmuted isotopes for multi-fueled regions. The fuel cycle length and changes in several core parameters such as: core excess reactivity, control rods position, fluxes at the irradiation positions, axial and radial power factors and other parameters are estimated. Besides, this study gives valuable insight into the behavior of the reactor and will ensure better utilization and operation of the reactor during its life-time and it will allow the establishment of

  16. Fuels from microalgae: Technology status, potential, and research requirements

    Energy Technology Data Exchange (ETDEWEB)

    Neenan, B.; Feinberg, D.; Hill, A.; McIntosh, R.; Terry, K.

    1986-08-01

    Although numerous options for the production of fuels from microalgae have been proposed, our analysis indicates that only two qualify for extensive development - gasoline and ester fuel. In developing the comparisons that support this conclusion, we have identified the major areas of microalgae production and processing that require extensive development. Technology success requires developing and testing processes that fully utilize the polar and nonpolar lipids produced by microalgae. Process designs used in these analyses were derived from fragmented, preliminary laboratory data. These results must be substantiated and integrated processes proposed, tested, and refined to be able to evaluate the commercial feasibility from microalgae. The production of algal feedstocks for processing to gasoline or ester fuel requires algae of high productivity and high lipid content that efficiently utilize saline waters. Species screening and development suggest that algae can achieve required standards taken individually, but algae that can meet the integrated requirements still elude researchers. Effective development of fuels from microalgae technology requires that R and D be directed toward meeting the integrated standards set out in the analysis. As technology analysts, it is inappropriate for us to dictate how the R and D effort should proceed to meet these standards. We end our role by noting that alternative approaches to meeting the feasibility targets have been identified, and it is now the task of program managers and scientists to choose the appropriate approach to assure the greatest likelihood of realizing a commercially viable technology. 70 refs., 39 figs., 35 tabs.

  17. Reprocessing of research reactor fuel the Dounreay option

    Energy Technology Data Exchange (ETDEWEB)

    Cartwright, P.

    1997-08-01

    Reprocessing is a proven process for the treatment of spent U/Al Research Reactor fuel. At Dounreay 12679 elements have been reprocessed during the past 30 years. For reactors converting to LEU fuel the uranium recovered in reprocessing can be blended down to less than 20% U{sub 235}, enrichment and be fabricated into new elements. For reactors already converted to LEU it is technically possible to reprocess spent silicide fuel to reduce the U{sub 235} burden and present to a repository only stable conditioned waste. The main waste stream from reprocessing which contains the Fission products is collected in underground storage tanks where it is kept for a period of at least five years before being converted to a stable solid form for return to the country of origin for subsequent storage/disposal. Discharges to the environment from reprocessing are low and are limited to the radioactive gases contained in the spent fuel and a low level liquid waste steam. Both of these discharges are independently monitored, and controlled within strict discharge limits set by the UK Government`s Scottish Office. Transportation of spent fuel to Dounreay has been undertaken using many routes from mainland Europe and has utilised over the past few years both chartered and scheduled vessel services. Several different transport containers have been handled and are currently licensed in the UK. This paper provides a short history of MTR reprocessing at Dounreay, and provides information to show reprocessing can satisfy the needs of MTR operators, showing that reprocessing is a valuable asset in non-proliferation terms, offers a complete solution and is environmentally acceptable.

  18. Irradiated fuel by-product separation research in Canada

    International Nuclear Information System (INIS)

    Burston, M.

    1984-01-01

    Although no decision has been made to reprocess irradiated CANDU fuel, by-product separation research has recently been initiated in Canada because of its potential importance to Canadian research programs in advanced fuel cycles (especially U/Pu cycle development in the near term) and nuclear waste management. In addition, separated by-products could have a significant commercial potential. Demonstrated applications include: heat sources, gamma radiation sources, light sources, new materials for productions of other useful isotopes, etc. For illustrative purposes the calculated market value of by-products currently stored in irradiated CANDU fuel is approximately $210/kgU. Ontario Hydro has initiated a program to study the application of new separation technolgies, such as laser-based techniques and the plasma ion cyclotron resonance separation technique, to either augment and/or supplant the chemical extraction methods. The main goal is to develop new, more economical extraction methods in order to increase the magnitude of the advantages resulting from this approach to reprocessing. (author)

  19. Irradiation experience of IPEN fuel at IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Perrotta, Jose A.; Neto, Adolfo; Durazzo, Michelangelo; Souza, Jose A.B. de; Frajndlich, Roberto

    1998-01-01

    IPEN/CNEN-SP produces, for its IEA-R1 Research Reactor, MTR fuel assemblies based on U 3 O 8 -Al dispersion fuel type. Since 1985 a qualification program on these fuel assemblies has been performed. Average 235 U burnup of 30% and peak burnup of 50% was already achieved by these fuel assemblies. This paper presents some results acquire, by these fuel assemblies, under irradiation at IEA-R1 Research Reactor. (author)

  20. Research in JAERI on the backend of nuclear fuel cycle

    International Nuclear Information System (INIS)

    Maeda, Mitsuru; Takeshita, Isao

    1999-01-01

    Japan's policy of the backend of nuclear fuel cycle is to reprocess spent fuels and recycle recovered plutonium and uranium, under the principle of no surplus plutonium. High-level radioactive waste separated during reprocessing will be disposed of after solidification in vitrified form, followed by the storage for 30 to 50 years and finally by ultimate disposal in a deep geological formation. The role of JAERI and the effective utilization of NUCEF would become more important. The current status of JAERI's research on backend cycle is reviewed together with the future research direction with emphases on NUCEF utilization. (1) Major objectives of safety research is to develop safety criteria and establish technical bases for licensing, to improve the safety of current or near future technology and to clarify the safety margin of licensed technology. (2) The present goal of fundamental research is to show or clarify the chemical or scientific feasibility of advanced system such as for recycling minor actinides or for incinerating long-lived nuclides. (3) Supporting research for nuclear material control is also conducted mainly for international contribution to strengthened safeguards by IAEA and to frame working of international monitoring system for CTBT. (J.P.N.)

  1. Nano-watt fueling from a micro-scale microbial fuel cell using black tea waste

    KAUST Repository

    Rojas, Jhonathan Prieto

    2016-02-03

    In this report, we show the rapid assessment of black tea as potential fuel to power up nanopower systems using a microsized, simplistic and sustainable air-cathode microbial fuel cell. It was found that tea produced more power compared with traditional sodium acetate media due in part to its acidophilic pH and its higher organics content. Although high internal resistance remains a big concern, this simple, curiosity-driven experiment gave us the preliminary results to say that energy could be extracted from the reuse of waste resources such the collection of our afternoon-tea\\'s leftovers.

  2. Multi-scale sustainability assessments for biomass-based and coal-based fuels in China.

    Science.gov (United States)

    Man, Yi; Xiao, Honghua; Cai, Wei; Yang, Siyu

    2017-12-01

    Transportation liquid fuels production is heavily depend on oil. In recent years, developing biomass based and coal based fuels are regarded as promising alternatives for non-petroleum based fuels in China. With the rapid growth of constructing and planning b biomass based and coal based fuels production projects, sustainability assessments are needed to simultaneously consider the resource, the economic, and the environmental factors. This paper performs multi-scale analyses on the biomass based and coal based fuels in China. The production cost, life cycle cost, and ecological life cycle cost (ELCC) of these synfuels are investigated to compare their pros to cons and reveal the sustainability. The results show that BTL fuels has high production cost. It lacks of economic attractiveness. However, insignificant resource cost and environmental cost lead to a substantially lower ELCC, which may indicate better ecological sustainability. CTL fuels, on the contrary, is lower in production cost and reliable for economic benefit. But its coal consumption and pollutant emissions are both serious, leading to overwhelming resource cost and environmental cost. A shifting from petroleum to CTL fuels could double the ELCC, posing great threat to the sustainability of the entire fuels industry. Copyright © 2017 Elsevier B.V. All rights reserved.

  3. Conceptual Biorefinery Design and Research Targeted for 2022: Hydrothermal Liquefacation Processing of Wet Waste to Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Snowden-Swan, Lesley J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Zhu, Yunhua [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Bearden, Mark D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Seiple, Timothy E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Jones, Susanne B. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Schmidt, Andrew J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Billing, Justin M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hallen, Richard T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hart, Todd R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Liu, Jian [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Albrecht, Karl O. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Fox, Samuel P. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Maupin, Gary D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Elliott, Douglas C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2017-12-28

    The Department of Energy Bioenergy Technologies Office (BETO) invests in research and development of new pathways for commercially viable conversion of biomass into drop-in ready transportation fuels, fuel blendstocks and products. The primary emphasis has been on terrestrial and algae feedstocks, but more recently BETO has begun to explore the potential of wet wastes for biofuel production, with focus on wastewater residuals, manure, food waste, and fats, oils and grease. A recent resource analysis estimates that 77 million dry tons per year of these wastes are generated annually, 65% of which are underutilized for any beneficial purpose. Approximately 14 million dry tons of the total resource is wastewater residuals (sludge and biosolids) generated at the nation’s wastewater treatment plants (WWTPs). Conversion of this resource into transportation fuels could significantly contribute to the creation of a new domestic bioenergy and bioproduct industry, while providing an economically and environmentally sustainable alternative for current waste disposal practices. Hydrothermal liquefaction (HTL) is a process that uses hot, pressurized water in the condensed phase to convert biomass to a thermally stable oil product, also known as “biocrude”, which can then be thermo-catalytically upgraded to hydrocarbon fuel blendstocks. HTL is conceptually simple, has a high carbon efficiency, and can be applied to a wide range of wet feedstocks at similar processing conditions. The purpose of this report is to document the conceptual design, economics and supporting data for a sludge-to-fuel pathway via HTL and biocrude upgrading. The configuration includes a HTL plant that is co-located with a WWTP and a larger scale biocrude upgrading plant for production of hydrocarbon fuel blendstocks. Experimental data from bench scale testing of a 1:1 mixture of primary:secondary sludges are used to establish the economic and technical assumptions for the analysis. The design

  4. Shifting to non-explosive fuels for research reactors

    International Nuclear Information System (INIS)

    Gilinsky, Victor

    1984-01-01

    The RERTR program is not just an American program, it is an international program and it can succeed only with wide support and participation. Excellent work underway at research laboratories in several countries is making a vital contribution. The cooperative spirit shown by all participants is particularly gratifying. Some practical difficulties may be encountered with the safety regulatory agencies in different countries when modifying the reactors. The US NRC intends to demonstrate that conversion to low enriched fuel is not a difficult process by taking steps to enable domestic reactors operating in the United States to convert to low enriched fuels. A proposed regulation is being prepared limiting the use of highly enriched uranium in domestic reactors. In this connection, The US NRC will be prepared tp cooperate with reactor operators from other countries in the safety area as it relates to the conversion process

  5. Basic research in support of innovative fuels design for the Generation IV systems (F-BRIDGE project)

    International Nuclear Information System (INIS)

    Valot, Carole; Bertolus, Marjorie; Konings, Rudy; Somers, Joe; Groot, Sander de

    2010-01-01

    F-BRIDGE (Basic Research in support of Innovative Fuels Design for the GEN IV systems) is a 4-year project which started in 2008. It seeks to bridge the gap between basic research and technological applications for generation IV nuclear reactor systems. One of the challenges for the next generation of reactors is to significantly increase the efficiency in designing innovative fuels. The object of the F-BRIDGE project is to complement the empirical approach by a physically-based description of fuel and cladding materials to enable a rationalization of the design process and a better selection of promising fuel systems. Advanced modelling and separate effects experiments are carried out in order to obtain more exact physical descriptions of ceramic fuels and cladding, at relevant scales from the atomic to the macroscopic scale. Research is also focused on assessing and improving 'sphere-pac' fuel, a composite-ceramics concept which has shown promise. The project activities can be broken down into four main areas: (i) Basic research investigations using a multi-scale approach in both experimentation and modelling to enable the generation of missing basic data, the identification of relevant mechanisms and the development of appropriate models; (ii) Transfer between technological issues and basic research by bringing together within the same project materials scientists, engineers and end-users; (iii) Assessment of the drawbacks and benefits of the sphere-pac fuel application to various Generation IV systems; (iv) Education and training to promote research in the field of fuel materials, to ensure the exchange of results and ideas among the participants and to link the project with other related European or international initiatives. The project relies on the complementary expertise of 19 partners: nuclear and non nuclear research organisations, universities, a nuclear engineering company, as well as technology and project management consultancy small and medium

  6. Technologies for small scale wood-fueled combined heat and power systems

    Energy Technology Data Exchange (ETDEWEB)

    Houmann Jakobsen, H.; Houmoeller, S.; Thaaning Pedersen, L.

    1998-01-01

    The aim of this study is to describe and compare different technologies for small cogeneration systems (up to 2-3 MW{sub e}), based on wood as fuel. For decentralized cogeneration, i.e. for recovering energy from saw mill wood wastes or heat supply for small villages, it is vital to know the advantages and disadvantages of the different technologies. Also, for the decision-makers it is of importance to know the price levels of the different technologies. A typical obstacle for small wood cogeneration systems is the installation costs. The specific price (per kW) is usually higher than for larger plants or plants using fossil fuels. For a saw mill choosing between cogeneration and simple heat production, however, the larger installation costs are counter weighed by the sale of electricity, while the fuel consumption is the same. Whether it is profitable or not to invest in cogeneration is often hard to decide. For many years small wood cogeneration systems have been too expensive, leading to the construction of only heat producing systems due to too high price levels of small steam turbines. In recent years a great deal of effort has been put into research and developing of new technologies to replace this traditional steam turbine. Among these are: Steam engines; Stirling engines; Indirectly fired gas turbines; Pressurized down draft combustion. Along with the small scale traditional steam turbines, these technologies will be evaluated in this study. When some or all these technologies are fully developed and commercial, a strong means of reducing the strain on the environment and the greenhouse effect will be available, as the total efficiency is high (up to 90%) and wood is an energy source in balance with nature. (au) EFP-95. 19 refs.

  7. ROSA-IV Large Scale Test Facility (LSTF) system description for second simulated fuel assembly

    International Nuclear Information System (INIS)

    1990-10-01

    The ROSA-IV Program's Large Scale Test Facility (LSTF) is a test facility for integral simulation of thermal-hydraulic response of a pressurized water reactor (PWR) during small break loss-of-coolant accidents (LOCAs) and transients. In this facility, the PWR core nuclear fuel rods are simulated using electric heater rods. The simulated fuel assembly which was installed during the facility construction was replaced with a new one in 1988. The first test with this second simulated fuel assembly was conducted in December 1988. This report describes the facility configuration and characteristics as of this date (December 1988) including the new simulated fuel assembly design and the facility changes which were made during the testing with the first assembly as well as during the renewal of the simulated fuel assembly. (author)

  8. Hydrogen and fuel cell research: Institute for Integrated Energy Systems (IESVic)

    International Nuclear Information System (INIS)

    Pitt, L.

    2006-01-01

    Vision: IESVic's mission is to chart feasible paths to sustainable energy. Current research areas of investigation: 1. Energy system analysis 2. Computational fuel cell engineering; Fuel cell parameter measurement; Microscale fuel cells 3. Hydrogen dispersion studies for safety codes 4. Active magnetic refrigeration for hydrogen liquifaction and heat transfer in metal hydrides 5. Hydrogen and fuel cell system integration (author)

  9. Scientific research on the back-end of the fuel cycle for the 21. century

    International Nuclear Information System (INIS)

    2000-01-01

    The aim of the Atalante-2000 conference is to present the major research axis concerning the nuclear fuel cycle back-end. The different topics are: - Present options concerning fuel cycle back-end; - Reprocessing of spent fuel; - Advanced separation for transmutation; - Processing and packaging of radioactive wastes; - Design and fabrication of targets for transmutation; and - Conversion of military plutonium into MOX fuels

  10. Optimal design and operation of solid oxide fuel cell systems for small-scale stationary applications

    Science.gov (United States)

    Braun, Robert Joseph

    The advent of maturing fuel cell technologies presents an opportunity to achieve significant improvements in energy conversion efficiencies at many scales; thereby, simultaneously extending our finite resources and reducing "harmful" energy-related emissions to levels well below that of near-future regulatory standards. However, before realization of the advantages of fuel cells can take place, systems-level design issues regarding their application must be addressed. Using modeling and simulation, the present work offers optimal system design and operation strategies for stationary solid oxide fuel cell systems applied to single-family detached dwellings. A one-dimensional, steady-state finite-difference model of a solid oxide fuel cell (SOFC) is generated and verified against other mathematical SOFC models in the literature. Fuel cell system balance-of-plant components and costs are also modeled and used to provide an estimate of system capital and life cycle costs. The models are used to evaluate optimal cell-stack power output, the impact of cell operating and design parameters, fuel type, thermal energy recovery, system process design, and operating strategy on overall system energetic and economic performance. Optimal cell design voltage, fuel utilization, and operating temperature parameters are found using minimization of the life cycle costs. System design evaluations reveal that hydrogen-fueled SOFC systems demonstrate lower system efficiencies than methane-fueled systems. The use of recycled cell exhaust gases in process design in the stack periphery are found to produce the highest system electric and cogeneration efficiencies while achieving the lowest capital costs. Annual simulations reveal that efficiencies of 45% electric (LHV basis), 85% cogenerative, and simple economic paybacks of 5--8 years are feasible for 1--2 kW SOFC systems in residential-scale applications. Design guidelines that offer additional suggestions related to fuel cell

  11. Full scale tests on remote handled FFTF fuel assembly waste handling and packaging

    International Nuclear Information System (INIS)

    Allen, C.R.; Cash, R.J.; Dawson, S.A.; Strode, J.N.

    1986-01-01

    Handling and packaging of remote handled, high activity solid waste fuel assembly hardware components from spent FFTF reactor fuel assemblies have been evaluated using full scale components. The demonstration was performed using FFTF fuel assembly components and simulated components which were handled remotely using electromechanical manipulators, shielding walls, master slave manipulators, specially designed grapples, and remote TV viewing. The testing and evaluation included handling, packaging for current and conceptual shipping containers, and the effects of volume reduction on packing efficiency and shielding requirements. Effects of waste segregation into transuranic (TRU) and non-transuranic fractions also are discussed

  12. An assessment of the commercial cost of farm scale wood fuel procurement and processing

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-11-01

    The capital cost of small scale biomass fired systems is currently significantly higher than equivalent rated fossil fuel fired systems. If the cost of producing willow or poplar coppice derived fuel can be shown to offer significant savings over the cost of conventional fossil fuel alternatives, the resultant total cost of energy production could in fact be less. The production of ''home grown'' fuel should always, in theory, be cheaper than bought in supplies due to the removal of cost components such as profit and risk and the possible use of labour during traditionally quiet periods on the farm. However, it has not been shown to date that small scale coppice plantations can successfully produce cost effective wood fuel to displace fuel that would otherwise be 'bought in'. It is likely that fuel from coppice will be harvested on a semi-manual basis using brush cutters and farm loaders etc. This report identifies appropriate systems and provides estimates of the key costs to a grower. Particular emphasis is given to operations surrounding cut-back, harvesting and comminution. The report provides an outline of the statutory requirements of employers engaged in coppice management. Key costs have been presented in their most useful form for a potential grower to compile enterprise gross margins. (Author)

  13. Generation of SCALE 6 Input Data File for Cross Section Library of PWR Spent Fuel

    International Nuclear Information System (INIS)

    Jeong, Chang Joon; Cho, Dong Keun

    2010-11-01

    In order to obtain the cross section libraries of the Korean Pressurized water reactor (PWR) spent fuel (SF), SCALE 6 code input files have been generated. The PWR fuel data were obtained from the nuclear design report (NDR) of the current operating PWRs. The input file were prepared for 16 fuel types such as 4 types of Westinghouse 14x14, 3 types of OPR-1000 16x16, 4 types of Westinghouse 16x16, and 6 types of Westinghouse 17x17. For each fuel type, 5 kinds of fuel enrichments have been considered such as 1.5, 2.0 ,3.0, 4.0 and 5.0 wt%. In the SCALE 6 calculation, a ENDF-V 44 group was used. The 25 burnup step until 72000 MWD/T was used. A 1/4 symmetry model was used for 16x16 and 17x17 fuel assembly, and 1/2 symmetry model was used for 14x14 fuel assembly The generated cross section libraries will be used for the source-term analysis of the PWR SF

  14. Air pollution and economics: Alternate use of fuels in small scale industries

    International Nuclear Information System (INIS)

    Rao, B.P.S.; Pandit, V.I.

    1999-01-01

    In developing countries the problem of air pollution was recognized earlier, however, it has acquired a greater dimension due to the conventional use of low grade fuels like coal, baggase, rice husk, etc. having high sulphur and ash content. The industrial sources contribute about 30--40% of the total emissions. In India, the small scale industries (low investment group) contribute about 60--80% of the total industrial emissions. These industries are characterized with various environmental pollution problems due to cluster of small scale industries located in sensitive area; use of low grade fuel, primitive processing techniques without emission abatement facilities etc., thus leading to enormous pollution in an confined region. Acute need was felt to reduce the pollution problem associated with small scale industries by use of cleaner fuel so as to reduce the localized problem. The paper presents the emissions associated with use of coal/coke, natural gas, LPG, and propane along with the fuel cost for small scale industrial sector of Agra, Firozabad and Mathura region. The studies carried out would find applicability to meet the air pollution standards based on shift in fuel and associated cost

  15. Status of research reactor spent fuel world-wide: Database summary

    International Nuclear Information System (INIS)

    Ritchie, I.G.

    1996-01-01

    Results complied in the research reactor spent fuel database are used to assess the status of research reactor spent fuel world-wide. Fuel assemblies, their types, enrichment, origin of enrichment and geological distribution among the industrialized and developed countries of the world are discussed. Fuel management practices in wet and dry storage facilities and the concerns of reactor operators about long-term storage of their spent fuel are presented and some of the activities carried out by the International Atomic Energy Agency to address the issues associated with research reactor spent fuel are outlined. (author). 4 refs, 17 figs, 4 tabs

  16. Investigation of small scale sphere-pac fuel fabrication plant with external gelation process

    International Nuclear Information System (INIS)

    Maekawa, Kazuhiko; Yoshimura, Tadahiro; Kikuchi, Toshiaki; Hoshino, Yasushi; Munekata, Hideki; Shimizu, Makoto

    2005-02-01

    In feasibility studies on commercialized FBR cycle system, comprehensive system investigation and properties evaluation for candidate FBR cycle systems have been implemented through view point of safety, economics, environmental burden reduction, non-proliferation resistivity, etc. As part of these studies, an investigation of small scale sphere-pac fuel fabrication plant with external gelation process was conducted. Until last fiscal year, equipment layout in cells and overall layout design of the 200t-HM/y scale fuel fabrication plant were conducted as well as schematical design studies on main equipments in gelation and reagent recovery processes of the plant. System property data concerning economics and environmental burden reduction of fuel fabrication plant was also acquired. In this fiscal year, the processes from vibropacking to fuel assemblies storage were added to the investigation range, and a conceptual design of whole fuel fabrication plant was studied as well as deepening the design study on main equipments. The conceptual design study was mainly conducted for small 50t-HM/y scale plant and a revising investigation was done for 200t-HM/y scale plant. Taking the planed comparative evaluation with pellet fuel fabrication system into account, design of equipments which should be equivalent with pellet system, especially in post-vibropacking processes, were standardized in each system. Based on these design studies, system properties data concerning economics and environmental burden reduction of the plant was also acquired. In comparison with existing design, the cell height was lowered on condition that plug type pneumatic system was adopted and fuel fabrication building was downsized by applying rationalized layout design of pellet system to post-vibropacking processes. Reduction of reagent usage at gelation process and rationalization of sintering and O/M controlling processes etc., are foremost tasks. (author)

  17. Scaling Science | IDRC - International Development Research Centre

    International Development Research Centre (IDRC) Digital Library (Canada)

    2018-02-23

    Feb 23, 2018 ... A new model for optimizing the impact of research and innovation. ... “Bigger is better” is the standard mantra of the business world where industry seeks ... Methods of measurement and analysis can be adapted as necessary ...

  18. NREL Fuels and Engines Research: Maximizing Vehicle Efficiency and

    Science.gov (United States)

    chemistry, conversion, and combustion to the evaluation of advanced fuels in actual engines and vehicles . With fuel chemistry, our scientific discoveries start out small. We use quantum mechanical modeling to explore how fuels with varying chemistry interact with engine and vehicle design. At our Fuel Combustion

  19. Corrosion of research reactor aluminium clad spent fuel in water

    International Nuclear Information System (INIS)

    2003-01-01

    This report describes research performed in ten laboratories within the framework of the IAEA Co-ordinated Research Project on Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water. The project consisted of exposure of standard racks of corrosion coupons in the spent fuel pools of the participating research reactor laboratories and the evaluation of the coupons after predetermined exposure times, along with periodic monitoring of the storage water. A group of experts in the field contributed a state of the art review and provided technical supervision of the project. Localized corrosion mechanisms are notoriously difficult to understand, and it was clear from the outset that obtaining consistency in the results and their interpretation from laboratory to laboratory would depend on the development of an excellent set of experimental protocols. These experimental protocols are described in the report together with guidelines for the maintenance of optimum water chemistry to minimize the corrosion of aluminium clad research reactor fuel in wet storage. A large database on corrosion of aluminium clad materials has been generated from the CRP and the SRS corrosion surveillance programme. An evaluation of these data indicates that the most important factors contributing to the corrosion of the aluminium are: (1) High water conductivity (100-200 μS/cm); (2) Aggressive impurity ion concentrations (Cl - ); (3) Deposition of cathodic particles on aluminium (Fe, etc.); (4) Sludge (containing Fe, Cl - and other ions in concentrations greater than ten times the concentrations in the water); (5) Galvanic couples between dissimilar metals (stainless steel-aluminium, aluminium-uranium, etc); (6) Scratches and imperfections (in protective oxide coating on cladding); (7) Poor water circulation. These factors operating both independently and synergistically may cause corrosion of the aluminium. The single most important key to preventing corrosion is maintaining good

  20. Japanese perspectives and research on packaging, transport and storage of spent fuel

    International Nuclear Information System (INIS)

    Saegusa, T.; Ito, C.; Yamakawa, H.; Shirai, K.

    2004-01-01

    The Japanese policy on spent fuel is reprocessing. Until, reprocessed, spent fuel shall be stored properly. This paper overviews current status of transport and storage of spent fuel with related research in Japan. The research was partly carried out under a contract of Ministry of Economy, Trade and Industry of the Japanese government

  1. EXPERIMENTAL RESEARCHES OF THERMO-PHYSICAL AND PHYSICOCHEMICAL INTERNALS OF BIO-DIESEL FUEL

    OpenAIRE

    V. N. Goryachkin; A. V. Ivaschenko

    2010-01-01

    The conducted researches are related to transfer of diesel engines to biodiesel fuel. The technique and results of an experimental research of thermo-physical and physical-and-chemical properties of biodiesel fuel as well as mixes of biodiesel fuel with the petroleum one are presented.

  2. EXPERIMENTAL RESEARCHES OF THERMO-PHYSICAL AND PHYSICOCHEMICAL INTERNALS OF BIO-DIESEL FUEL

    Directory of Open Access Journals (Sweden)

    V. N. Goryachkin

    2010-11-01

    Full Text Available The conducted researches are related to transfer of diesel engines to biodiesel fuel. The technique and results of an experimental research of thermo-physical and physical-and-chemical properties of biodiesel fuel as well as mixes of biodiesel fuel with the petroleum one are presented.

  3. The use of radiochemical techniques in fuel cell research

    International Nuclear Information System (INIS)

    Meier, H.

    1975-01-01

    The utilization of metal chelates as catalysts in fuel cell research gives rise to special problems which cannot be solved by the usual methods, but may be well clarified by isotope technical methods. The electrocatalytic efficiency of polymer iron phthalocyanine (on carbon carriers) can be proved by the plotting of potential-current density curves. Two questions, however, remain unanswered: a) What is the solubility behaviour of the catalyst, and b) is there an additional stabilizing interaction between the metal chelate catalyst and the carbon electrode. To answer the first question, the iron phthalocyanine was labelled with Fe-59 and the dissolving time of the complexed Fe ions measured; the results were compared with the potential time behaviour of the oxygen cathodes. To check the interactions between phthalocyamine catalysts and carbon carrier, Moessbauer spectroscopy was used. The evidence obtained suggest the application of isotope technical methods to an ever greater extent than up to now in fuel cell and battery research. (RB/LH) [de

  4. Used fuel disposition research and development roadmap - FY10 status.

    Energy Technology Data Exchange (ETDEWEB)

    Nutt, W. M. (Nuclear Engineering Division)

    2010-10-01

    Since 1987 the U.S. has focused research and development activities relevant to the disposal of commercial used nuclear fuel and U.S. Department of Energy (DOE) owned spent nuclear fuel and high level waste on the proposed repository at Yucca Mountain, Nevada. At the same time, the U.S. successfully deployed a deep geologic disposal facility for defense-related transuranic waste in bedded salt at the Waste Isolation Pilot Plant. In 2009 the DOE established the Used Fuel Disposition Campaign (UFDC) within the Office of Nuclear Energy. The Mission of the UFDC is to identify alternatives and conduct scientific research and technology development to enable storage, transportation and disposal of used nuclear fuel and wastes generated by existing and future nuclear fuel cycles. The U.S. national laboratories have participated on these programs and has conducted research and development related to these issues to a limited extent. However, a comprehensive research and development (R&D) program investigating a variety of geologic media has not been a part of the U.S. waste management program since the mid 1980s. Such a comprehensive R&D program is being developed in the UFDC with a goal of meeting the UFDC Grand Challenge to provide a sound technical basis for absolute confidence in the safety and security of long-term storage, transportation, and disposal of used nuclear fuel and wastes from the nuclear energy enterprise. The DOE has decided to no longer pursue the development of a repository at Yucca Mountain, Nevada. Since a repository site will ultimately have to be selected, sited, characterized, designed, and licensed, other disposal options must now be considered. In addition to the unsaturated volcanic tuff evaluated at Yucca Mountain, several different geologic media are under investigation internationally and preliminary assessments indicate that disposal of used nuclear fuel and high level waste in these media is feasible. Considerable progress has been made in

  5. Used fuel disposition research and development roadmap - FY10 status

    International Nuclear Information System (INIS)

    Nutt, W.M.

    2010-01-01

    Since 1987 the U.S. has focused research and development activities relevant to the disposal of commercial used nuclear fuel and U.S. Department of Energy (DOE) owned spent nuclear fuel and high level waste on the proposed repository at Yucca Mountain, Nevada. At the same time, the U.S. successfully deployed a deep geologic disposal facility for defense-related transuranic waste in bedded salt at the Waste Isolation Pilot Plant. In 2009 the DOE established the Used Fuel Disposition Campaign (UFDC) within the Office of Nuclear Energy. The Mission of the UFDC is to identify alternatives and conduct scientific research and technology development to enable storage, transportation and disposal of used nuclear fuel and wastes generated by existing and future nuclear fuel cycles. The U.S. national laboratories have participated on these programs and has conducted research and development related to these issues to a limited extent. However, a comprehensive research and development (R and D) program investigating a variety of geologic media has not been a part of the U.S. waste management program since the mid 1980s. Such a comprehensive R and D program is being developed in the UFDC with a goal of meeting the UFDC Grand Challenge to provide a sound technical basis for absolute confidence in the safety and security of long-term storage, transportation, and disposal of used nuclear fuel and wastes from the nuclear energy enterprise. The DOE has decided to no longer pursue the development of a repository at Yucca Mountain, Nevada. Since a repository site will ultimately have to be selected, sited, characterized, designed, and licensed, other disposal options must now be considered. In addition to the unsaturated volcanic tuff evaluated at Yucca Mountain, several different geologic media are under investigation internationally and preliminary assessments indicate that disposal of used nuclear fuel and high level waste in these media is feasible. Considerable progress has been

  6. Selection of fuel design for conversion and upgradation of Pakistan Research Reactor (PARR-1)

    International Nuclear Information System (INIS)

    Arshad, M.

    1991-01-01

    The Pakistan Research Reactor (PARR-1) is being converted from the use of Highly Enriched Uranium (HEU) to Low Enriched Uranium (LEU) fuel and its power is also being upgraded. In order to select new fuel for the converted and upgraded core ten different fuel element designs were analyzed and their relative performance was compared. Results of this study were later used to select appropriate design of the new fuel for PARR-1. This paper describes the computational methodology utilized for the analysis of various fuel element designs. Criteria for selecting the new fuel element are discussed and guidelines forming the selection basis of the new fuel design are given. (author)

  7. Application of wavelet scaling function expansion continuous-energy resonance calculation method to MOX fuel problem

    International Nuclear Information System (INIS)

    Yang, W.; Wu, H.; Cao, L.

    2012-01-01

    More and more MOX fuels are used in all over the world in the past several decades. Compared with UO 2 fuel, it contains some new features. For example, the neutron spectrum is harder and more resonance interference effects within the resonance energy range are introduced because of more resonant nuclides contained in the MOX fuel. In this paper, the wavelets scaling function expansion method is applied to study the resonance behavior of plutonium isotopes within MOX fuel. Wavelets scaling function expansion continuous-energy self-shielding method is developed recently. It has been validated and verified by comparison to Monte Carlo calculations. In this method, the continuous-energy cross-sections are utilized within resonance energy, which means that it's capable to solve problems with serious resonance interference effects without iteration calculations. Therefore, this method adapts to treat the MOX fuel resonance calculation problem natively. Furthermore, plutonium isotopes have fierce oscillations of total cross-section within thermal energy range, especially for 240 Pu and 242 Pu. To take thermal resonance effect of plutonium isotopes into consideration the wavelet scaling function expansion continuous-energy resonance calculation code WAVERESON is enhanced by applying the free gas scattering kernel to obtain the continuous-energy scattering source within thermal energy range (2.1 eV to 4.0 eV) contrasting against the resonance energy range in which the elastic scattering kernel is utilized. Finally, all of the calculation results of WAVERESON are compared with MCNP calculation. (authors)

  8. Chemistry research for the Canadian nuclear fuel waste management program

    International Nuclear Information System (INIS)

    Vikis, A.C.; Garisto, F.; Lemire, R.J.; Paquette, J.; Sagert, N.H.; Saluja, P.P.S.; Sunder, S.; Taylor, P.

    1988-01-01

    This publication reviews chemical research in support of the Canadian Nuclear Fuel Waste Management Program. The overall objective of this research is to develop the fundamental understanding required to demonstrate the suitability of waste immobilization media and processes, and to develop the chemical information required to predict the long-term behaviour of radionuclides in the geosphere after the waste form and the various engineered barriers containing it have failed. Key studies towards the above objective include experimental and theoretical studies of uranium dioxide oxidation/dissolution; compilation of thermodynamic databases and an experimental program to determine unavailable thermodynamic data; studies of hydrothermal alteration of minerals and radionuclide interactions with such minerals; and a study examining actinide colloid formation, as well as sorption of actinides on groundwater colloids

  9. Management of spent fuel from research and prototype power reactors and residues from post-irradiation examination of fuel

    International Nuclear Information System (INIS)

    1989-09-01

    The safe and economic management of spent fuel is important for all countries which have nuclear research or power reactors. It involves all aspects of the handling, transportation, storage, conditioning and reprocessing or final disposal of the spent fuel. In the case of spent fuel management from power reactors the shortage of available reprocessing capacity and the rising economic interest in the direct disposal of spent fuel have led to an increasing interest in the long term storage and management of spent fuel. The IAEA has played a major role in coordinating the national activities of the Member States in this area. It was against this background that the Technical Committee Meeting on ''Safe Management of Spent Fuel From Research Reactors, Prototype Power Reactors and Fuel From Commercial Power Reactors That Has Been Subjected to PIE (Post Irradiated Examination)'' (28th November - 1st December 1988) was organised. The aims of the current meeting have been to: 1. Review the state-of-the-art in the field of management of spent fuel from research and prototype power reactors, as well as the residues from post irradiation examination of commercial power reactor fuel. The emphasis was to be on the safe handling, conditioning, transportation, storage and/or disposal of the spent fuel during operation and final decommissioning of the reactors. Information was sought on design details, including shielding, criticality and radionuclide release prevention, heat removal, automation and remote control, planning and staff training; licensing and operational practices during each of the phases of spent fuel management. 2. Identify areas where additional research and development are needed. 3. Recommend areas for future international cooperation in this field. Refs, figs and tabs

  10. Plant-scale anodic dissolution of unirradiated N-Reactor fuel

    International Nuclear Information System (INIS)

    Gay, E.C.; Miller, W.E.; Laidler, J.J.

    1995-01-01

    Anodic dissolution tests were made with unirradiated N-Reactor fuel to determine the fuel segment length, diameter, and shape required for high throughput electro-refiner treatment for ultimate disposal in a geologic repository. Based on these tests, a conceptual design was produced of an electro-refiner for a full-scale plant to process N-Reactor spent fuel. In this design, the diameter of an electrode assembly is about 0.6 m (25 in.). Eight of these assemblies in an electro-refiner would accommodate a 1.333-metric-ton batch of N-Reactor fuel. Electrorefining would proceed at a rate of 40 kg uranium per hour. (author)

  11. Three-peat NREL Intern Pushes Boundaries of Early-Stage Fuels Research on

    Science.gov (United States)

    -Stage Fuels Research on Way to Master's Degree January 4, 2018 Woman preparing a fuel evaluation in a . Mission & Programs Awards History Leadership Community Education Center Economic Impact Environment

  12. Basic Research Needs for Clean and Efficient Combustion of 21st Century Transportation Fuels

    Energy Technology Data Exchange (ETDEWEB)

    McIlroy, A.; McRae, G.; Sick, V.; Siebers, D. L.; Westbrook, C. K.; Smith, P. J.; Taatjes, C.; Trouve, A.; Wagner, A. F.; Rohlfing, E.; Manley, D.; Tully, F.; Hilderbrandt, R.; Green, W.; Marceau, D.; O' Neal, J.; Lyday, M.; Cebulski, F.; Garcia, T. R.; Strong, D.

    2006-11-01

    To identify basic research needs and opportunities underlying utilization of evolving transportation fuels, with a focus on new or emerging science challenges that have the potential for significant long-term impact on fuel efficiency and emissions.

  13. Reprocessing ability of high density fuels for research and test reactors

    International Nuclear Information System (INIS)

    Gay, A.; Belieres, M.

    1997-01-01

    The development of a new high density fuel is becoming a key issue for Research Reactors operators. Such a new fuel should be a Low Enrichment Uranium (LEU) fuel with a high density, to improve present in core performances. It must be compatible with the reprocessing in an industrial plant to provide a steady back-end solution. Within the framework of a work group CEA/CERCA/COGEMA on new fuel development for Research Reactors, COGEMA has performed an evaluation of the reprocessing ability of some fuel dispersants selected as good candidates. The results will allow US to classify these fuel dispersants from a reprocessing ability point of view. (author)

  14. Small-scale fuel cell cogen: application potentials and market strategies

    International Nuclear Information System (INIS)

    Vogel, Bernd

    2000-01-01

    Small (less than 5 kW) fuel-cell cogeneration systems are now being developed for use in residential buildings. The devices are expected to be on the market in five years. The article discusses the potential for their large-scale introduction, the impact of this new technology on the natural gas business, potential applications and marketing strategies

  15. Conditioning spent fuels from research nuclear reactor in ceramic dies

    International Nuclear Information System (INIS)

    Russo, D.O; Rodriguez, D.S; Mateos, P; Heredia, A; Sangilippo, M; Sterba, M

    2002-01-01

    The problem of immobilizing nuclear wastes is a complex one and is vitally important in the nuclear fuels cycle. In the case of spent elements from research reactors, the presence of large amounts of aluminum makes the procedure more complex and, therefore, onerous. There are various alternatives proposed for processing these materials. Two methods were studied in the Nuclear Materials Division for obtaining, as a final product, a vitreous block that could be place definitively in a geological repository. The processes are briefly, as follows: 1.By mechanical and chemical processes eliminating all the exterior aluminum from the fuel plates and then placing the product which we will call 'meat' (with some additional treatment and mixing with the amount needed to produce a natural uranium compound or weakened by decreasing the isotope enrichment in U-235) in a vitreous matrix. 2.Mechanically eliminate the aluminum from the exterior frame (as shown below) by shearing and cutting off the sectors containing only the Al, but leaving the rest of the aluminum, a big part of which is still present (4511.03), then doing the same procedure as in the case above: mixing with a natural uranium compound or weakening and vitrifying this mixture. In both cases, the vitrification can be carried out by fusion as well as by sintering. Given that these methods imply a big increase in volume together with a big mass of uranium and an even bigger amount of glass we decided to study an alternative. The proposed process involves synthesizing the mixtures obtained from the pre-treatment of the fuel plates (as described later) with natural isotope uranium oxide in order to obtain a block with the appropriate properties for its final disposal in a deep geological repository (CW)

  16. Research and development of spent fuel shipping casks and the criteria for seagoing vessel carrying casks

    International Nuclear Information System (INIS)

    Aoki, S.; Ando, Y.

    1977-01-01

    Considering that the transportation of spent fuel will increase rapidly and extensively in the near future, Japanese Atomic Energy Committee enacted ''Technical Standard for Transportation of Radioactive Materials'' based on ''IAEA Regulation for the Safe Transport of Radioactive Materials 1973 Revised Edition''. Coping with the recommendation of AEC, Atomic Energy Bureau in Science and Technology Agency and other authorities concerned started to review the former ordinances for transportation of radioactive materials and to consolidate a unified system of relevant laws and standards. On the other hand, Atomic Energy Bureau has invested in research and development since ten years ago in order to obtain the data for design and licensing work of spent fuel shipping casks. In those studies some different scale models of a prototype of 80 t in weight have been used to make clear the scale effect at the drop, pucture and fire tests, which are one of the features of Japanese research and development. And also the immersion test in high pressure water up to about 500 bars is now carried out to investigate the integrity of cask body and sealing structure to prevent leakage of radioactive contents to the ambient when the cask falls into deep sea. In Japan, depending on the site conditions of nuclear plants, almost all transportations of unirradiated and spent fuels are done on the sea. Therefore, in order to secure the safety of transportation, the design criteria of the seagoing vessels for exclusive transportation of spent fuel shipping casks, namely full load shipping, has been enacted, which aims to make minimum the probability of sinking at collison, stranding and other unforeseen accidents at sea and also to restrain radiation exposure of the crew as low as possible

  17. Needs, opportunities, and options for large scale systems research

    Energy Technology Data Exchange (ETDEWEB)

    Thompson, G.L.

    1984-10-01

    The Office of Energy Research was recently asked to perform a study of Large Scale Systems in order to facilitate the development of a true large systems theory. It was decided to ask experts in the fields of electrical engineering, chemical engineering and manufacturing/operations research for their ideas concerning large scale systems research. The author was asked to distribute a questionnaire among these experts to find out their opinions concerning recent accomplishments and future research directions in large scale systems research. He was also requested to convene a conference which included three experts in each area as panel members to discuss the general area of large scale systems research. The conference was held on March 26--27, 1984 in Pittsburgh with nine panel members, and 15 other attendees. The present report is a summary of the ideas presented and the recommendations proposed by the attendees.

  18. Thermal characteristics of various biomass fuels in a small-scale biomass combustor

    International Nuclear Information System (INIS)

    Al-Shemmeri, T.T.; Yedla, R.; Wardle, D.

    2015-01-01

    Biomass combustion is a mature and reliable technology, which has been used for heating and cooking. In the UK, biomass currently qualifies for financial incentives such as the Renewable Heat Incentive (RHI). Therefore, it is vital to select the right type of fuel for a small-scale combustor to address different types of heat energy needs. In this paper, the authors attempt to investigate the performance of a small-scale biomass combustor for heating, and the impact of burning different biomass fuels on useful output energy from the combustor. The test results of moisture content, calorific value and combustion products of various biomass samples were presented. Results from this study are in general agreement with published data as far as the calorific values and moisture contents are concerned. Six commonly available biomass fuels were tested in a small-scale combustion system, and the factors that affect the performance of the system were analysed. In addition, the study has extended to examine the magnitude and proportion of useful heat, dissipated by convection and radiation while burning different biomass fuels in the small-scale combustor. It is concluded that some crucial factors have to be carefully considered before selecting biomass fuels for any particular heating application. - Highlights: • Six biomass materials combustion performance in a small combustor was examined. • Fuel combustion rate and amount of heat release has varied between materials. • Heat release by radiation, convection and flue gasses varied between materials. • Study helps engineers and users of biomass systems to select right materials

  19. Status of research reactor fuel development in KAERI

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  20. Research program and uses of the solution fueled reactor SILENE

    International Nuclear Information System (INIS)

    Barbry, F.; Ratel, R.

    1985-09-01

    Designed and operated by the Nuclear Protection and Safety Institute of the CEA, SILENE is an original small sized reactor fueled with an uranyl nitrate solution. The reactor is capable to operate in three modes: ''Pulse'' operation (high power levels up to 1000 Megawatts during several millisecond), ''Free evolution'' operation (simulation of criticality accident excursions), ''Steady state'' operation in a power range of 0.01 W to 1 kW. The core can be surrounded by appropriate shields (lead, polyethylene) to vary the leakage radiations and the gamma to neutron dose ratio. It's possible to insert in the central cavity of the annular core vessel some capsules, devices or samples to be submitted to very high radiations levels. The research activities are mainly devoted towards nuclear safety studies: the criticality accident studies, and the behavior of oxide fuels under transient conditions. Some examples of tests are presented. As to other applications of the SILENE facility, the main studies now in progress deal with: designing and calibration of Health physics intrumentation, neutron and gamma dosimetry, and, radiobiology. Once the characteristics of radiation field are qualified by calculations and experimental techniques, SILENE will be proposed as a reference source [fr

  1. Studies on fuel failure detection in Rikkyo Research Reactor

    International Nuclear Information System (INIS)

    Matsuura, T.; Hayashi, S.H.; Harasawa, S.; Tomura, K.

    1992-01-01

    Studies on fuel failure detection have been made since 1986 in Rikkyo Research Reactor. One of the methods is the monitoring of the trace concentration of fission products appearing in the air on the surface of the water tank of the reactor. The interested radionuclides here are 89 Rb and 138 Cs, which are the daughter nuclides of the FP rare gas nuclides, 89 Kr and 138 Xe, respectively and have the half lives of 15.2 min and 32.2 min respectively. They are detected on a filter paper attached on a conventional dust sampler, by sucking the air of the surface of the water for 15 ∼ 30 min during reactor operation (100 kW). In this presentation are reported the results of an attempt to increase the sensitivity of detecting these nuclides by introducing nitrogen gas bubbles into the water. The bubbling of the gas increased the sensitivity as much as several times compared with the case without bubbling. These measurements are giving us the 'background' concentration, the order of which is almost unchanged for these several years, --in 10 -6 Bq/cm 3 . The origin of these nuclides is considered to be not from the fuel but from the uranium contained as an impurity in the reactor material in the core. (author)

  2. State of the art: Multi-fuel reformers for automotive fuel cell applications. Problem identification and research needs

    Energy Technology Data Exchange (ETDEWEB)

    Westerholm, R. [Stockholm Univ. (Sweden). Dept. of Analytical Chemistry; Pettersson, L.J. [Royal Inst. of Tech., Stockholm (Sweden). Dept. of Chemical Engineering and Technology

    1999-12-01

    On an assignment from the Transport and Communications Research Board (KFB) a literature study and a study trip to the USA and Great Britain have been performed. The literature study and the study trip was made during late spring and autumn 1999.The purpose of the project was to collect available information about the chemical composition of the product gas from a multi-fuel reformer for a fuel cell vehicle. It was furthermore to identify problems and research needs. The report recommends directions for future major research efforts. The results of the literature study and the study trip led to the following general conclusions: With the technology available today it does not seem feasible to develop a highly efficient and reliable multi-fuel reformer for automotive applications, i. e. for applications where all types of fuels ranging from natural gas to heavy diesel fuels can be used. The potential for developing a durable and reliable system is considerably higher if dedicated fuel reformers are used.The authors propose that petroleum-derived fuels should be designed for potential use in mobile fuel cell applications. In the present literature survey and the site visit discussions we found that there are relatively low emissions from fuel cell engines compared to internal combustion engines. However, the major research work on reformers/fuel cells have been performed during steady-state operation. Emissions during start-up, shutdown and transient operation are basically unknown and must be investigated in more detail. The conclusions and findings in this report are based on open/available information, such as discussions at site visits, reports, scientific publications and symposium proceedings.

  3. Overview of current research and development programmes for fuel in Japan

    International Nuclear Information System (INIS)

    Shiozawa, S.

    1991-01-01

    The Research and Development (R and D) programmes for HTGR fuel have been performed since 1969 by Japan Atomic Energy Research Institute (JAERI) as a leading organization in Japan. The R and D covers all fields necessary for the construction of the High Temperature Engineering Test Reactor (HTTR), which is the first HTGR in Japan. This R and D includes fuel fabrication, fuel property data, irradiation performance under normal operating conditions, safety-related research and fuel inspection technology. The R and D for the HTTR has been completed from a licensing point of view. Some R and D including future advanced fuel development continue. 2 figs, 3 tabs

  4. Studies and research concerning BNFP: LWR spent fuel storage

    International Nuclear Information System (INIS)

    Shallo, F.A.

    1978-08-01

    This report describes potential spent fuel storage expansion programs using the Barnwell Nuclear Fuel Plant--Fuel Receiving and Storage Station (BNFP-FRSS) as a model. Three basic storage arrangements are evaluated with cost and schedule estimates being provided for each configuration. A general description of the existing facility is included with emphasis on the technical and equipment requirements which would be necessary to achieve increased spent fuel storage capacity at BNFP-FRSS

  5. International guidelines for fire protection at nuclear installations including nuclear fuel plants, nuclear fuel stores, teaching reactors, research establishments

    International Nuclear Information System (INIS)

    The guidelines are recommended to designers, constructors, operators and insurers of nuclear fuel plants and other facilities using significant quantities of radioactive materials including research and teaching reactor installations where the reactors generally operate at less than approximately 10 MW(th). Recommendations for elementary precautions against fire risk at nuclear installations are followed by appendices on more specific topics. These cover: fire protection management and organization; precautions against loss during construction alterations and maintenance; basic fire protection for nuclear fuel plants; storage and nuclear fuel; and basic fire protection for research and training establishments. There are numerous illustrations of facilities referred to in the text. (U.K.)

  6. Recent emissions research in southwestern shrub and grassland fuels

    Science.gov (United States)

    David R. Weise; Wayne Miller; David R. Cocker; Heejung Jung; Seyedehsan Hosseini; Marko Princevac; Robert J. Yokelson; Ian Burling; Sheryl Akagi; Shawn Urbanski; WeiMin Hao

    2015-01-01

    While it is currently challenging to use prescribed burning in chaparral and other southwestern shrub fuel types due to many constraints, any such activities require smoke management planning. Information on fuels and emissions from chaparral were limited and based on older sampling systems. The DoD SERDP program funded a project to measure fuels and smoke emissions in...

  7. 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.)

  8. Introducing close-range photogrammetry for characterizing forest understory plant diversity and surface fuel structure at fine scales

    Science.gov (United States)

    Benjamin C. Bright; E. Louise Loudermilk; Scott M. Pokswinski; Andrew T. Hudak; Joseph J. O' Brien

    2016-01-01

    Methods characterizing fine-scale fuels and plant diversity can advance understanding of plant-fire interactions across scales and help in efforts to monitor important ecosystems such as longleaf pine (Pinus palustris Mill.) forests of the southeastern United States. Here, we evaluate the utility of close-range photogrammetry for measuring fuels and plant...

  9. Authentication and Interpretation of Weight Data Collected from Accountability Scales at Global Nuclear Fuels

    International Nuclear Information System (INIS)

    Fitzgerald, Peter; Laughter, Mark D.; Martyn, Rose; Richardson, Dave; Rowe, Nathan C.; Pickett, Chris A.; Younkin, James R.; Shephard, Adam M.

    2010-01-01

    Accountability scale data from the Global Nuclear Fuels (GNF) fuel fabrication facility in Wilmington, NC has been collected and analyzed as a part of the Cylinder Accountability and Tracking System (CATS) field trial in 2009. The purpose of the data collection was to demonstrate an authentication method for safeguards applications, and the use of load cell data in cylinder accountability. The scale data was acquired using a commercial off-the-shelf communication server with authentication and encryption capabilities. The authenticated weight data was then analyzed to determine facility operating activities. The data allowed for the determination of the number of full and empty cylinders weighed and the respective weights along with other operational activities. Data authentication concepts, practices and methods, the details of the GNF weight data authentication implementation and scale data interpretation results will be presented.

  10. Migration from Gasoline to Gaseous Fuel for Small-scale Electricity Generation Systems

    Directory of Open Access Journals (Sweden)

    Sukandar Sukandar

    2013-03-01

    Full Text Available This paper describes a study that gives a consideration to change fuel source for electricity generator from gasoline to combustible gas. A gaseous fuel conversion technology is presented and its performance is compared with gasoline. In the experiment, two types of load were tested, resistive and resistive-inductive. By using both fuels mostly the power factor (Cos ? of resistive-inductive load variations were greater than 0.8, and they had slight difference on operational voltage. The drawback of using gaseous fuel is the frequency of the electricity might be up to 10 Hz deviated from the standard frequency (i.e. 50 Hz. In the lab scale experiment, the gasoline consumption increased proportionally with the load increase, while using gaseous fuel the consumption of gas equal for two different load value in the range of 50% maximum load, which is 100 gram per 15 minutes operation. Therefore, the use of gaseous generation system should have average power twice than the required load. The main advantage using gaseous fuel (liquefied petroleum gas or biogas compared to gasoline is a cleaner emitted gas after combustion.

  11. Uncertainties in modeling and scaling in the prediction of fuel stored energy and thermal response

    International Nuclear Information System (INIS)

    Wulff, W.

    1987-01-01

    The steady-state temperature distribution and the stored energy in nuclear fuel elements are computed by analytical methods and used to rank, in the order of importance, the effects on stored energy from statistical uncertainties in modeling parameters, in boundary and in operating conditions. An integral technique is used to calculate the transient fuel temperature and to estimate the uncertainties in predicting the fuel thermal response and the peak clad temperature during a large-break loss of coolant accident. The uncertainty analysis presented here is an important part of evaluating the applicability, the uncertainties and the scaling capabilities of computer codes for nuclear reactor safety analyses. The methods employed in this analysis merit general attention because of their simplicity. It is shown that the blowdown peak is dominated by fuel stored energy alone or, equivalently, by linear heating rate. Gap conductance, peaking factors and fuel thermal conductivity are the three most important fuel modeling parameters affecting peak clad temperature uncertainty. 26 refs., 10 figs., 6 tabs

  12. Design of a full scale model fuel assembly for full power production reactor flow excursion experiments

    International Nuclear Information System (INIS)

    Nash, C.A.; Blake, J.E.; Rush, G.C.

    1990-01-01

    A novel full scale production reactor fuel assembly model was designed and built to study thermal-hydraulic effects of postulated Savannah River Site (SRS) nuclear reactor accidents. The electrically heated model was constructed to simulate the unique annular concentric tube geometry of fuel assemblies in SRS nuclear production reactors. Several major design challenges were overcome in order to produce the prototypic geometry and thermal-hydraulic conditions. The two concentric heater tubes (total power over 6 MW and maximum heat flux of 3.5 MW/m 2 ) (1.1E+6 BTU/(ft 2 hr)) were designed to closely simulate the thermal characteristics of SRS uranium-aluminum nuclear fuel. The paper discusses the design of the model fuel assembly, which met requirements of maintaining prototypic geometric and hydraulic characteristics, and approximate thermal similarity. The model had a cosine axial power profile and the electrical resistance was compatible with the existing power supply. The model fuel assembly was equipped with a set of instruments useful for code analysis, and durable enough to survive a number of LOCA transients. These instruments were sufficiently responsive to record the response of the fuel assembly to the imposed transient

  13. Large scale experiments simulating hydrogen distribution in a spent fuel pool building during a hypothetical fuel uncovery accident scenario

    Energy Technology Data Exchange (ETDEWEB)

    Mignot, Guillaume; Paranjape, Sidharth; Paladino, Domenico; Jaeckel, Bernd; Rydl, Adolf [Paul Scherrer Institute, Villigen (Switzerland)

    2016-08-15

    Following the Fukushima accident and its extended station blackout, attention was brought to the importance of the spent fuel pools' (SFPs) behavior in case of a prolonged loss of the cooling system. Since then, many analytical works have been performed to estimate the timing of hypothetical fuel uncovery for various SFP types. Experimentally, however, little was done to investigate issues related to the formation of a flammable gas mixture, distribution, and stratification in the SFP building itself and to some extent assess the capability for the code to correctly predict it. This paper presents the main outcomes of the Experiments on Spent Fuel Pool (ESFP) project carried out under the auspices of Swissnuclear (Framework 2012–2013) in the PANDA facility at the Paul Scherrer Institut in Switzerland. It consists of an experimental investigation focused on hydrogen concentration build-up into a SFP building during a predefined scaled scenario for different venting positions. Tests follow a two-phase scenario. Initially steam is released to mimic the boiling of the pool followed by a helium/steam mixture release to simulate the deterioration of the oxidizing spent fuel. Results shows that while the SFP building would mainly be inerted by the presence of a high concentration of steam, the volume located below the level of the pool in adjacent rooms would maintain a high air content. The interface of the two-gas mixture presents the highest risk of flammability. Additionally, it was observed that the gas mixture could become stagnant leading locally to high hydrogen concentration while steam condenses. Overall, the experiments provide relevant information for the potentially hazardous gas distribution formed in the SFP building and hints on accident management and on eventual retrofitting measures to be implemented in the SFP building.

  14. Fuel research for subcritical and critical GEN-IV systems cooled by heavy liquid metal

    International Nuclear Information System (INIS)

    Sobolev, V.; Verwerft, M.

    2009-01-01

    The participation of the Belgian Nuclear Research Centre SCK-CEN in the worldwide GEN-IV research can be considered as an opportunity. Today's GEN-IV research at SCK-CEN is mainly driven by the interests of the project MYRRHA (Multipurpose hYbrid Research Reactor for High-tech Applications). The main goal of this project is to build at SCK-CEN in Mol a new generation fast spectrum, subcritical, research and materials testing reactor MYRRHA driven by a high-energy proton accelerator. This GEN-IV MTR is cooled by heavy liquid metal (Pb-Bi) and will be used for the ADS concept demonstration, testing and qualification of new fuels, transmutation targets and innovative materials. On the European scale, MYRRHA is integrated in the Euratom FP6 Integrated Project (IP) EUROTRANS (EUROpean research programme for TRANSmutation of high level nuclear waste in an accelerator driven system), as the small-scale experimental machine for transmutation demonstration called XT-ADS. Last but not least, this experimental facility will also demonstrate the technological feasibility of the LFR (Lead-cooled Fast Reactor) GEN-IV concept; in EU the LFR design studies are performed in the framework of the Euratom FP6 ELSY (European Lead-cooled SYstem) project, where SCK-CEN is a partner. Among the research needed to ensure a safe and reliable operation of the MYRRHA/XT ADS reactor, the development and qualification of fuel and cladding materials have been recognized as one of the main key issues to be addressed

  15. Modeling and simulation of nuclear fuel in scenarios with long time scales

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa, Carlos E.; Bodmann, Bardo E.J., E-mail: eduardo.espinosa@ufrgs.br, E-mail: bardo.bodmann@ufrgs.br [Universidade Federal do Rio Grande do Sul (DENUC/PROMEC/UFRGS), Porto Alegre, RS (Brazil). Departamento de Engenharia Nuclear. Programa de Pos Graduacao em Engenharia Mecanica

    2015-07-01

    Nuclear reactors play a key role in defining the energy matrix. A study by the Fraunhofer Society shows in different time scales for long periods of time the distribution of energy sources. Regardless of scale, the use of nuclear energy is practically constant. In these scenarios, the nuclear fuel behavior over time is of interest. For kinetics of long-term scales, changing the chemical composition of fuel is significant. Thus, it is appropriate to consider fission products called neutron poisons. Such products are of interest in the nuclear reactor, since they become parasitic neutron absorbers and result in long thermal heat sources. The objective of this work is to solve the kinetics system coupled to neutron poison products. To solve this system, we use similar ideas to the method of Adomian decomposition. Initially, one separates the system of equations as the sum of a linear part and a non-linear part in order to solve a recursive system. The nonlinearity is treated as Adomian polynomial. We present numerical results of the effects of changing the power of a reactor, scenarios such as start-up and shut-down. For these results we consider time dependent reactivity, such as linear reactivity, quadratic polynomial and oscillatory. With these results one can simulate the chemical composition of the fuel due to the reuse of the spent fuel in subsequent cycles. (author)

  16. Modeling and simulation of nuclear fuel in scenarios with long time scales

    International Nuclear Information System (INIS)

    Espinosa, Carlos E.; Bodmann, Bardo E.J.

    2015-01-01

    Nuclear reactors play a key role in defining the energy matrix. A study by the Fraunhofer Society shows in different time scales for long periods of time the distribution of energy sources. Regardless of scale, the use of nuclear energy is practically constant. In these scenarios, the nuclear fuel behavior over time is of interest. For kinetics of long-term scales, changing the chemical composition of fuel is significant. Thus, it is appropriate to consider fission products called neutron poisons. Such products are of interest in the nuclear reactor, since they become parasitic neutron absorbers and result in long thermal heat sources. The objective of this work is to solve the kinetics system coupled to neutron poison products. To solve this system, we use similar ideas to the method of Adomian decomposition. Initially, one separates the system of equations as the sum of a linear part and a non-linear part in order to solve a recursive system. The nonlinearity is treated as Adomian polynomial. We present numerical results of the effects of changing the power of a reactor, scenarios such as start-up and shut-down. For these results we consider time dependent reactivity, such as linear reactivity, quadratic polynomial and oscillatory. With these results one can simulate the chemical composition of the fuel due to the reuse of the spent fuel in subsequent cycles. (author)

  17. Roadmap to an Engineering-Scale Nuclear Fuel Performance and Safety Code

    International Nuclear Information System (INIS)

    Turner, John A.; Clarno, Kevin T.; Hansen, Glen A.

    2009-01-01

    Developing new fuels and qualifying them for large-scale deployment in power reactors is a lengthy and expensive process, typically spanning a period of two decades from concept to licensing. Nuclear fuel designers serve an indispensable role in the process, at the initial exploratory phase as well as in analysis of the testing results. In recent years fuel performance capabilities based on first principles have been playing more of a role in what has traditionally been an empirically dominated process. Nonetheless, nuclear fuel behavior is based on the interaction of multiple complex phenomena, and recent evolutionary approaches are being applied more on a phenomenon-by-phenomenon basis, targeting localized problems, as opposed to a systematic approach based on a fundamental understanding of all interacting parameters. Advanced nuclear fuels are generally more complex, and less understood, than the traditional fuels used in existing reactors (ceramic UO 2 with burnable poisons and other minor additives). The added challenges are primarily caused by a less complete empirical database and, in the case of recycled fuel, the inherent variability in fuel compositions. It is clear that using the traditional approach to develop and qualify fuels over the entire range of variables pertinent to the U.S. Department of Energy (DOE) Office of Nuclear Energy on a timely basis with available funds would be very challenging, if not impossible. As a result the DOE Office of Nuclear Energy has launched the Nuclear Energy Advanced Modeling and Simulation (NEAMS) approach to revolutionize fuel development. This new approach is predicated upon transferring the recent advances in computational sciences and computer technologies into the fuel development program. The effort will couple computational science with recent advances in the fundamental understanding of physical phenomena through ab initio modeling and targeted phenomenological testing to leapfrog many fuel

  18. Testing of a Transport Cask for Research Reactor Spent Fuel - 13003

    International Nuclear Information System (INIS)

    Mourao, Rogerio P.; Leite da Silva, Luiz; Miranda, Carlos A.; Mattar Neto, Miguel; Quintana, Jose F.A.; Saliba, Roberto O.; Novara, Oscar E.

    2013-01-01

    Since the beginning of the last decade three Latin American countries that operate research reactors - Argentina, Brazil and Chile - have been joining efforts to improve the regional capability in the management of spent fuel elements from the TRIGA and MTR reactors operated in the region. A main drive in this initiative, sponsored by the International Atomic Energy Agency, is the fact that no definite solution regarding the back end of the research reactor fuel cycle has been taken by any of the participating country. However, any long-term solution - either disposition in a repository or storage away from reactor - will involve at some stage the transportation of the spent fuel through public roads. Therefore, a licensed cask that provides adequate shielding, assurance of subcriticality, and conformance to internationally accepted safety, security and safeguards regimes is considered a strategic part of any future solution to be adopted at a regional level. As a step in this direction, a packaging for the transport of irradiated fuel for MTR and TRIGA research reactors was designed by the tri-national team and a half-scale model equipped with the MTR version of the internal basket was constructed in Argentina and Brazil and tested in Brazil. Three test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions. After failing the tests in the first two test series, the specimen successfully underwent the last test sequence. A second specimen, incorporating the structural improvements in view of the previous tests results, will be tested in the near future. Numerical simulations of the free drop and thermal tests are being carried out in parallel, in order to validate the computational modeling that is going to be used as a support for the package certification. (authors)

  19. Tuneable diode laser gas analyser for methane measurements on a large scale solid oxide fuel cell

    Science.gov (United States)

    Lengden, Michael; Cunningham, Robert; Johnstone, Walter

    2011-10-01

    A new in-line, real time gas analyser is described that uses tuneable diode laser spectroscopy (TDLS) for the measurement of methane in solid oxide fuel cells. The sensor has been tested on an operating solid oxide fuel cell (SOFC) in order to prove the fast response and accuracy of the technology as compared to a gas chromatograph. The advantages of using a TDLS system for process control in a large-scale, distributed power SOFC unit are described. In future work, the addition of new laser sources and wavelength modulation will allow the simultaneous measurement of methane, water vapour, carbon-dioxide and carbon-monoxide concentrations.

  20. ROSA-V large scale test facility (LSTF) system description for the third and fourth simulated fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Mitsuhiro; Nakamura, Hideo; Ohtsu, Iwao [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

    2003-03-01

    The Large Scale Test Facility (LSTF) is a full-height and 1/48 volumetrically scaled test facility of the Japan Atomic Energy Research Institute (JAERI) for system integral experiments simulating the thermal-hydraulic responses at full-pressure conditions of a 1100 MWe-class pressurized water reactor (PWR) during small break loss-of-coolant accidents (SBLOCAs) and other transients. The LSTF can also simulate well a next-generation type PWR such as the AP600 reactor. In the fifth phase of the Rig-of-Safety Assessment (ROSA-V) Program, eighty nine experiments have been conducted at the LSTF with the third simulated fuel assembly until June 2001, and five experiments have been conducted with the newly-installed fourth simulated fuel assembly until December 2002. In the ROSA-V program, various system integral experiments have been conducted to certify effectiveness of both accident management (AM) measures in beyond design basis accidents (BDBAs) and improved safety systems in the next-generation reactors. In addition, various separate-effect tests have been conducted to verify and develop computer codes and analytical models to predict non-homogeneous and multi-dimensional phenomena such as heat transfer across the steam generator U-tubes under the presence of non-condensable gases in both current and next-generation reactors. This report presents detailed information of the LSTF system with the third and fourth simulated fuel assemblies for the aid of experiment planning and analyses of experiment results. (author)

  1. Race for a Better Fuel Begins with NREL Researchers | News | NREL

    Science.gov (United States)

    produced at NREL registers an estimated octane rating of 107. NREL's fuel is better suited for race cars by Dennis Schroeder Look! Up in the Sky! In addition to race cars, the 85% triptane could find a use Race for a Better Fuel Begins with NREL Researchers Race for a Better Fuel Begins with NREL

  2. Development of nuclear fuel materials for research reactor

    International Nuclear Information System (INIS)

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

    1997-09-01

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

  3. Strengthening industry-research linkage for small scale industrial ...

    African Journals Online (AJOL)

    Strengthening industry-research linkage for small scale industrial development in Ghana - the relevance of scientific and technological information. ... Journal of Applied Science and Technology. Journal Home · ABOUT · Advanced Search ...

  4. Economical analysis to utilize MTR fuel elements using silicides in research reactors

    International Nuclear Information System (INIS)

    Bergallo, Juan E.; Novara, Oscar E.; Adelfang, Pablo

    2000-01-01

    According to international programs on reducing enrichment in research reactors and the necessity to maintain their operation, new fuel elements have been developed in order to meet both objectives. Thus, U-Si alloy fuel elements for research reactors are becoming of greater interest for the international markets. It became necessary to make an economic study about the convenience of introducing this type of fuel elements in the RA-3 reactor and to know the potentiality of this fuel. The economical behavior of the reactor operation has been evaluated comparing the actual U 3 O 8 nuclear fuel cycle with U 3 Si 2 nuclear fuels. Results obtained show that the main economical factor to determine the change of fuels is the cost of fabrication, and the change is advisable up to an 80% difference. The other factors related to the cost of nuclear fuel cycle are not relevant or have real minor impacts. (author)

  5. Analyses of the transportation of spent research reactor fuel in the United States

    International Nuclear Information System (INIS)

    Cashwell, J.W.; Neuhauser, K.S.

    1989-01-01

    The Transportation Technology Center at Sandia National Laboratories has analyzed the impacts of transportation of research reactor spent fuel from US and foreign reactors for the US Department of Energy (DOE) Office of Defense Programs. This effort represents the first comprehensive analytical evaluation of the risks of transporting high-, medium-, and low-enriched uranium spent research reactor fuel by both sea and land. Two separate shipment programs have been analyzed: the shipment of research reactor spent fuel from Taiwan to the US (Fuel Movement Program), and the return of research reactor spent fuels of US origin from foreign and domestic reactors (Research Reactor Fuel Return Program). In order to perform these analyses, a comprehensive methodology for analyzing the probabilities and consequences of transportation in coastal waters and port facilities, handling at the port, and shipment by truck to reprocessing facilities was developed. The Taiwanese fuel consists of low-burnup aluminum-clad metallic uranium research reactor spent fuel; the other fuels are primarily aluminum-clad oxide fuels. The Fuel Movement Program is ongoing, while the Fuel Return Program addresses future shipments over a ten-year period. The operational aspects of the Taiwanese shipments have been uniform, but several possible shipping configurations are possible for the Fuel Return Program shipments. Comprehensive assessments, which bound the impacts of spent fuel transport, demonstrate that when shipments are made in compliance with applicable regulations, the risks for all such transport are low. For comparison with previously licensed transport activities and to provide continuity with earlier analyses, the results for shipment of 150-day-old commercial pressurized water reactor (PWR) spent fuel are presented as part of this study

  6. Engineering scale tests of mechanical disassembly and short stroke shearing systems for FBR fuel assembly

    International Nuclear Information System (INIS)

    Higuchi, Hidetoshi; Kitagaki, Toru; Koizumi, Kenji; Hirano, Hiroyasu; Takeuchi, Masayuki; Washiya, Tadahiro; Kawabe, Yukinari; Kobayashi, Tsuguyuki

    2011-01-01

    Japan Atomic Energy Agency (JAEA) and The Japan Atomic Power Company (JAPC) have been developing an advanced head-end process based on mechanical disassembly and short stroke shearing systems as a part of Fast Reactor Cycle Technology Development (FaCT). Fuel pins for a fast reactor are installed within a hexagonal shaped wrapper tube made of stainless steel. In order to reprocess the fast reactor fuel pins, they must be removed from the wrapper tube and transported to the shearing system without failure. In addition, the advanced aqueous reprocessing process, called 'NEXT' (New Extraction System for TRU Recovery) process requires a solution of the spent fuel with relatively high concentration (500g/L). JAEA and JAPC have developed the mechanical disassembly and the short stroke shearing technology which is expected to make fragmented fuel to satisfy these requirements. This paper reports the results of engineering scale tests on the mechanical disassembly and short stroke shearing systems. These tests were carried out with simulated FBR fuel assembly and removed pins. The mechanical cutting method has been developed to avoid fuel pin failure during disassembly operation. The cutting process is divided into two modes, so called 'slit-cut' for cutting the wrapper tube and 'crop-cut' for the end plug region of the fuel pin bundle. In the slit-cut mode, the depth of cutting was automatically controlled based on the calculated wastage of the cutting tool and deformation of the wrapper tube which had been measured before the cutting. This procedure was confirmed to minimize the fuel pin failure which was hard to prevent in the case of laser cutting. The cutting speed was also controlled automatically by the electric current of the cutting motor to lower the load of the cutting tool. The removed fuel pins were transported to the shearing machine, whose fuel shearing magazine width was set to be narrow to realize the suitable configuration for the short stroke shearing

  7. Subcritical Measurements Research Program for Fresh and Spent Materials Test Reactor Fuels

    International Nuclear Information System (INIS)

    Blanchard, A.

    1999-01-01

    'A series of subcritical noise measurements were performed on fresh and spent University of Missouri Research Reactor fuel assemblies. These experimental measurements were performed for the purposes of providing benchmark quality data for validating transport theory computer codes and nuclear cross-section data used to perform criticality safety analyses for highly enriched, uranium-aluminum Material Test Reactor fuel assemblies. A mechanical test rig was designed and built to hold up to four fuel assemblies and neutron detectors in a subcritical array. The rig provided researchers with the ability to evaluate the reactivity effects of variable fuel/detector spacing, fuel rotation, and insertion of metal reflector plates into the lattice.'

  8. Thermal-Hydraulic Research Review and Cooperation Outcome for Light Water Reactor Fuel

    Energy Technology Data Exchange (ETDEWEB)

    In, Wang Kee; Shin, Chang Hwan; Lee, Chan; Chun, Tae Hyun; Oh, Dong Seok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Chi Young [Pukyong Nat’l Univ., Busan (Korea, Republic of)

    2016-12-15

    The fuel assembly for pressurized water reactor (PWR) consists of fuel rod bundle, spacer grid and bottom/top end fittings. The cooling water in high pressure and temperature is introduced in lower plenum of reactor core and directed to upper plenum through the subchannel which is formed between the fuel rods. The main thermalhydraulic performance parameters for the PWR fuel are pressure drop and critical heat flux in normal operating condition, and quenching time in accident condition. The Korea Atomic Energy Research Institute (KAERI) has been developing an advanced PWR fuel, dual-cooled annular fuel and accident tolerant fuel for the enhancement of fuel performance and the localization. For the key thermal-hydraulic technology development of PWR fuel, the KAERI LWR fuel team has conducted the experiments for pressure drop, turbulent flow mixing and heat transfer, critical heat flux(CHF) and quenching. The computational fluid dynamics (CFD) analysis was also performed to predict flow and heat transfer in fuel assembly including the spent fuel assembly in dry cask for interim repository. In addition, the research cooperation with university and nuclear fuel company was also carried out to develop a basic thermalhydraulic technology and the commercialization.

  9. Analyses of the transportation of spent research reactor fuel in the United States

    International Nuclear Information System (INIS)

    Cashwell, J.W.; Neuhauser, K.S.

    1989-01-01

    We analyzed the impacts of transportation of research reactor spent fuel from US and foreign reactors for the US Department of Energy's (DOE) Office of Defense Programs. Two separate shipment programs were analyzed. The shipment of research reactor spent fuel from Taiwan to the US (Fuel Movement Program), and the return of research reactor spent fuels of US origin from foreign and domestic reactors (Research Reactor Fuel Return Program). To perform these analyses, a comprehensive methodology for analyzing the probabilities and consequences of transportation in coastal waters and port facilities, handling at the port, and shipment by truck to reprocessing facilities was developed. The Taiwanese fuel consists of low-burnup aluminum-clad metallic uranium research reactor spent fuel; the other fuels are primarily aluminum-clad oxide fuels. The Fuel Movement Program is ongoing, while the Fuel Return Program addresses future shipments over a ten-year period. The operational aspects of the Taiwanese shipments have been uniform, but several possible shipping configurations are possible for the Fuel Return Program shipments. The risks of transporting spent nuclear fuel and other radioactive materials by all modes have been analyzed extensively. Comprehensive assessments, which bound the impacts of spent fuel transport, demonstrate that when shipments are made in compliance with applicable regulations, the risks for all such transport are low. For comparison with previously licensed transport activities and to provide continuity with earlier analyses, the results for shipment of 150-day-old commercial pressurized water reactor (PWR) spent fuel are presented as part of this study

  10. Fuel Cells: Power System Option for Space Research

    Science.gov (United States)

    Shaneeth, M.; Mohanty, Surajeet

    2012-07-01

    Fuel Cells are direct energy conversion devices and, thereby, they deliver electrical energy at very high efficiency levels. Hydrogen and Oxygen gases are electrochemically processed, producing clean electric power with water as the only by product. A typical, Fuel Cell based power system involve a Electrochemical power converter, gas storage and management systems, thermal management systems and relevant control units. While there exists different types of Fuel cells, Proton Exchange Membrane (PEM) Fuel Cells are considered as the most suitable one for portable applications. Generally, Fuel Cells are considered as the primary power system option in space missions requiring high power ( > 5kW) and long durations and also where water is a consumable, such as manned missions. This is primarily due to the advantage that fuel cell based power systems offer, in terms of specific energy. Fuel cells have the potential to attain specific energy > 500Wh/kg, specific power >500W/kg, energy density > 400Whr/L and also power density > 200 W/L. This apart, a fuel cell system operate totally independent of sun light, whereas as battery based system is fully dependent on the same. This uniqueness provides added flexibility and capabilities to the missions and modularity for power system. High power requiring missions involving reusable launch vehicles, manned missions etc are expected to be richly benefited from this. Another potential application of Fuel Cell would be interplanetary exploration. Unpredictable and dusty atmospheres of heavenly bodies limits sun light significantly and there fuel cells of different types, eg, Bio-Fuel Cells, PEMFC, DMFCs would be able to work effectively. Manned or unmanned lunar out post would require continuous power even during extra long lunar nights and high power levels are expected. Regenerative Fuel Cells, a combination of Fuel Cells and Electrolysers, are identified as strong candidate. While application of Fuel Cells in high power

  11. Experience on wet storage spent fuel sipping at IEA-R1 Brazilian research reactor

    International Nuclear Information System (INIS)

    Perrotta, J.A.; Terremoto, L.A.A.; Zeituni, C.A.

    1998-01-01

    The IEA-R1 research reactor of the Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP) is a pool type reactor of B and W design, that has been operating since 1957 at a power of 2 MW. Irradiated (spent) fuels have been stored at the facility during the various years of operation. At present there are 40 spent fuel assemblies at dry storage, 79 spent fuel assemblies at wet storage and 30 fuel assemblies in the core. The oldest fuels are of United States origin, made with U-Al alloy, both of LEU and HEU MTR fuel type. Many of these fuel assemblies have corrosion pits along their lateral fuel plates. These pits originate by galvanic corrosion between the fuel plate and the stainless steel storage racks. As a consequence of the possibility of sending the irradiated old fuels back the U.S.A., sipping tests were performed with the spent fuel assemblies. The reason for this was to evaluate their 137 Cs leaking rate, if any. This work describes the procedure and methodology used to perform the sipping tests with the fuel assemblies at the storage pool, and presents the results obtained for the 137 Cs sipping water activity for each fuel assembly. A correlation is made between the corrosion pits and the activity values measured. A 137 Cs leaking rate is determined and compared to the criteria established for canning spent fuel assemblies before shipment

  12. Thermo-mechanical behaviour modelling of particle fuels using a multi-scale approach

    International Nuclear Information System (INIS)

    Blanc, V.

    2009-12-01

    Particle fuels are made of a few thousand spheres, one millimeter diameter large, compound of uranium oxide coated by confinement layers which are embedded in a graphite matrix to form the fuel element. The aim of this study is to develop a new simulation tool for thermo-mechanical behaviour of those fuels under radiations which is able to predict finely local loadings on the particles. We choose to use the square finite element method, in which two different discretization scales are used: a macroscopic homogeneous structure whose properties in each integration point are computed on a second heterogeneous microstructure, the Representative Volume Element (RVE). First part of this works is concerned by the definition of this RVE. A morphological indicator based in the minimal distance between spheres centers permit to select random sets of microstructures. The elastic macroscopic response of RVE, computed by finite element has been compared to an analytical model. Thermal and mechanical representativeness indicators of local loadings has been built from the particle failure modes. A statistical study of those criteria on a hundred of RVE showed the significance of choose a representative microstructure. In this perspective, a empirical model binding morphological indicator to mechanical indicator has been developed. Second part of the work deals with the two transition scale method which are based on the periodic homogenization. Considering a linear thermal problem with heat source in permanent condition, one showed that the heterogeneity of the heat source involve to use a second order method to localized finely the thermal field. The mechanical non-linear problem has been treats by using the iterative Cast3M algorithm, substituting to integration of the behavior law a finite element computation on the RVE. This algorithm has been validated, and coupled with thermal resolution in order to compute a radiation loading. A computation on a complete fuel element

  13. Nuclear liability and research reactor fuel. A plant supplier's view

    International Nuclear Information System (INIS)

    Roegler, H.-J.; Hetzmann, A.

    2000-01-01

    Contracts on Research Reactors are normally entered into by the owner and - very often - later user and the supplier of such plants. They are not concluded by the fuel supplier, except fuel supplier and plant supplier are identical. Thus, the fuel supplier mostly has no direct influence into the contract negotiations and the clauses which are eventually agreed upon between the parties. So has any other subcontractor for any other system or component. Any such subsupplier can and will negotiate a subsidiary supply contract (subcontract) with the supplier of the plant. The supplier drafts the related clauses so as to pass on to the subsupplier as many risks out of his contracts as possible. The subsupplier, on the other hand, tries to protect himself, tries to limit the risks he takes over, e.g. to the worth of his subcontract maximum. A critical issue in such negotiations is those concerning the risks the supplier had to accept and the subsupplier, although he may be responsible later for the risk changing to reality; i.e. the occurrence of a loss, cannot be hold liable for in full because the subcontract limits his liability, e.g. to the subcontract value or a certain delay penalty. A typical example for this conflict are delays of the entire project caused by one subsupplier. A very specific case in this context is the so-called nuclear liability. Nuclear liability means being hold responsible for the consequences or damages originating from a nuclear event in the plant. Those consequences or damages may be suffered by third parties, which are neither the owner/operator nor the supplier and result in a liability to such party (third party liability). Several of the aspects below may be related to the nuclear liability issue: The supplier often has its registered office not in the country where the plant is; The supplier may have far bigger assets than the owner/operator. The legal system of the supplier's country may be more favourable for enforcing claims of

  14. Present condition of survey research on actualization strategy of fast breeder reactor (FBR) cycling. Design research on fuel production system

    International Nuclear Information System (INIS)

    Tanaka, Kenya

    2001-01-01

    The fuel production system design investigation was performed for construction of fuel production process concept and plant image searching for the targets such as economics, environmental loading reduction, and so on required for practical use of FBR fuel recycling at a premise of safety security. By expectation of economics as a fuel cycling system, enhancement of nuclear proliferation resistance, and so on, it becomes more important to investigate on a fuel cycling system suitable for raw materials with low decontamination and high radiation intensity. In addition, it is also necessary to carry out investigation on fuel production system concept accompanies with MA recycling system for reduction of environmental loading. Therefore, investigation objects on the system were laid their fundamental processes on denitrification conversion/pelletizing process and gelation/vibration filling process for raw material solution from advancing wet reprocessing and on vibration filling process for oxide granules obtained from dry reprocessing system and casting method for metal fuels. As a result, for the pollution removal fuel production system suitable for either of wet/dry reprocessing, a mass-production scale production plant image was elucidated at a premise of production yield, realizability of remote automation system, and so on. On candidate concepts of every fuel production system, no fatal defect was found on results of outline evaluation on features of system such as production facility scale and so on before present stage. (G.K.)

  15. Water reactor fuel research at the Joint Research Centre of the Commission of the European Union

    Energy Technology Data Exchange (ETDEWEB)

    Markgraf, J [HFR Unit, Inst. for Advanced Materials, Petten (Netherlands)

    1997-12-01

    The JRC programmes 1995-1998 are in progress within 8 JRC Institutes located within different research centres situated in five member states of the European Union. Except for the area of reactor safety and waste management, there are no JRC research programmes related to water reactor fuel directly. However, the JRC is providing support on basis of contracts and cooperations to bilateral and international R and D programmes through access to its large test facilities like the High Flux Reactor at the JRC Petten and the hot cell facilities of the Transuranium Institute at the JRC Karlsruhe. Access is available to customers from all over the world. 3 tabs.

  16. [Research and workshop on alternative fuels for aviation. Final report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-09-01

    The Renewable Aviation Fuels Development Center (RAFDC) at Baylor University was granted U. S. Department of Energy (US DOE) and Federal Aviation Administration (FAA) funds for research and development to improve the efficiency in ethanol powered aircraft, measure performance and compare emissions of ethanol, Ethyl Tertiary Butyl Ether (ETBE) and 100 LL aviation gasoline. The premise of the initial proposal was to use a test stand owned by Engine Components Inc. (ECI) based in San Antonio, Texas. After the grant was awarded, ECI decided to close down its test stand facility. Since there were no other test stands available at that time, RAFDC was forced to find additional support to build its own test stand. Baylor University provided initial funds for the test stand building. Other obstacles had to be overcome in order to initiate the program. The price of the emission testing equipment had increased substantially beyond the initial quote. Rosemount Analytical Inc. gave RAFDC an estimate of $120,000.00 for a basic emission testing package. RAFDC had to find additional funding to purchase this equipment. The electronic ignition unit also presented a series of time consuming problems. Since at that time there were no off-the-shelf units of this type available, one had to be specially ordered and developed. FAA funds were used to purchase a Super Flow dynamometer. Due to the many unforeseen obstacles, much more time and effort than originally anticipated had to be dedicated to the project, with much of the work done on a volunteer basis. Many people contributed their time to the program. One person, mainly responsible for the initial design of the test stand, was a retired engineer from Allison with extensive aircraft engine test stand experience. Also, many Baylor students volunteered to assemble the. test stand and continue to be involved in the current test program. Although the program presented many challenges, which resulted in delays, the RAFDC's test

  17. The Wechsler Memory Scale: A Review of Research.

    Science.gov (United States)

    Ivison, David

    1990-01-01

    Research on the standardization, reliability, validity, factor structure, and subtests of the Wechsler Memory Scale (WMS) (1945) and its revised version (1987) is reviewed. Much research relating to the WMS appears to be relevant to the revised version. Use of the instrument in Australia is discussed. (SLD)

  18. Proceedings of the meeting on large scale computer simulation research

    International Nuclear Information System (INIS)

    2004-04-01

    The meeting to summarize the collaboration activities for FY2003 on the Large Scale Computer Simulation Research was held January 15-16, 2004 at Theory and Computer Simulation Research Center, National Institute for Fusion Science. Recent simulation results, methodologies and other related topics were presented. (author)

  19. New Visions for Large Scale Networks: Research and Applications

    Data.gov (United States)

    Networking and Information Technology Research and Development, Executive Office of the President — This paper documents the findings of the March 12-14, 2001 Workshop on New Visions for Large-Scale Networks: Research and Applications. The workshops objectives were...

  20. Design report small-scale fuel alcohol plant. Volume 2: Detailed construction information

    Science.gov (United States)

    1980-12-01

    The objectives are to provide potential alcohol producers with a reference design and provide a complete, demonstrated design of a small scale fuel alcohol plant. The plant has the capability for feedstock preparation, cooking, saccharification, fermentation, distillation, by-product dewatering, and process steam generation. An interesting feature is an instrumentation and control system designed to allow the plant to run 24 hours per day with only four hours of operator attention.

  1. Large-Scale Production of Fuel and Feed from Marine Microalgae

    Energy Technology Data Exchange (ETDEWEB)

    Huntley, Mark [Cornell Univ., Ithaca, NY (United States)

    2015-09-30

    In summary, this Consortium has demonstrated a fully integrated process for the production of biofuels and high-value nutritional bioproducts at pre-commercial scale. We have achieved unprecedented yields of algal oil, and converted the oil to viable fuels. We have demonstrated the potential value of the residual product as a viable feed ingredient for many important animals in the global food supply.

  2. Study on the fuel cycle cost of gas turbine high temperature reactor (GTHTR300). Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Takei, Masanobu; Katanishi, Shoji; Nakata, Tetsuo; Kunitomi, Kazuhiko [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Oda, Takefumi; Izumiya, Toru [Nuclear Fuel Industries, Ltd., Tokyo (Japan)

    2002-11-01

    In the basic design of gas turbine high temperature reactor (GTHTR300), reduction of the fuel cycle cost has a large benefit of improving overall plant economy. Then, fuel cycle cost was evaluated for GTHTR300. First, of fuel fabrication for high-temperature gas cooled reactor, since there was no actual experience with a commercial scale, a preliminary design for a fuel fabrication plant with annual processing of 7.7 ton-U sufficient four GTHTR300 was performed, and fuel fabrication cost was evaluated. Second, fuel cycle cost was evaluated based on the equilibrium cycle of GTHTR300. The factors which were considered in this cost evaluation include uranium price, conversion, enrichment, fabrication, storage of spent fuel, reprocessing, and waste disposal. The fuel cycle cost of GTHTR300 was estimated at about 1.07 yen/kWh. If the back-end cost of reprocessing and waste disposal is included and assumed to be nearly equivalent to LWR, the fuel cycle cost of GTHTR300 was estimated to be about 1.31 yen/kWh. Furthermore, the effects on fuel fabrication cost by such of fuel specification parameters as enrichment, the number of fuel types, and the layer thickness were considered. Even if the enrichment varies from 10 to 20%, the number of fuel types change from 1 to 4, the 1st layer thickness of fuel changes by 30 {mu}m, or the 2nd layer to the 4th layer thickness of fuel changes by 10 {mu}m, the impact on fuel fabrication cost was evaluated to be negligible. (author)

  3. Detection of fission products release in the research reactor 'RA' spent fuel storage pool

    International Nuclear Information System (INIS)

    Matausek, M.V.; Vukadin, Z.; Pavlovic, S.; Maksin, T.; Idakovic, Z.; Marinkovic, N.

    1997-05-01

    Spent fuel resulting from 25 years of operating the 6.5/10 MW thermal heavy water moderated and cooled research reactor RA at the VINCA Institute is presently all stored in the temporary spent fuel storage pool in the basement of the reactor building. In 1984, the reactor was shut down for refurbishment, which for a number of reasons has not yet been completed. Recent investigations show that independent of the future status of the research reactor, safe disposal of the so far irradiated fuel must be the subject of primary concern. The present status of the research reactor RA spent fuel storage pool at the VINCA Institute presents a serious safety problem. Action is therefore initiated in two directions. First, safety of the existing spent fuel storage should be improved. Second, transferring spent fuel into another, presumably dry storage space should be considered. By storing the previously irradiated fuel of the research reactor RA in a newly built storage space, sufficient free space will be provided in the existing spent fuel storage pool for the newly irradiated fuel when the reactor starts operation again. In the case that it would be decided to decommission the research reactor RA, the newly built storage space would provide safe disposal for the fuel irradiated so far

  4. Re-thinking china's densified biomass fuel policies: Large or small scale?

    International Nuclear Information System (INIS)

    Shan, Ming; Li, Dingkai; Jiang, Yi; Yang, Xudong

    2016-01-01

    Current policies and strategies related to the utilization of densified biomass fuel (DBF) in China are mainly focused on medium- or large-scale manufacturing modes, which cannot provide feasible solutions to solve the household energy problems in China's rural areas. To simplify commercial processes related to the collection of DBF feedstock and the production and utilization of fuel, a novel village-scale DBF approach is proposed. Pilot demonstration projects have shown the feasibility and flexibility of this new approach in realizing sustainable development in rural China. Effective utilization of DBF in rural China will lead to gains for global, regional, and local energy savings, environmental protection, sustainable development, and related social benefits. It could also benefit other developing countries for better utilization of biomass as a viable household energy source. This proposal therefore delivers the possibility of reciprocal gains, and as such deserves the attention of policy makers and various stakeholders. - Highlights: •A field survey of Chinese densified biomass fuel (DBF) development is conducted. •The current situation and problems related to China's DBF industry are analyzed. •A novel and viable village-scale DBF utilization mode is proposed. •Further actions are suggested to boost the utilization of DBF in rural China.

  5. SCALE 5.1 Predictions of PWR Spent Nuclear Fuel Isotopic Compositions

    Energy Technology Data Exchange (ETDEWEB)

    Radulescu, Georgeta [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL

    2010-03-01

    The purpose of this calculation report is to document the comparison to measurement of the isotopic concentrations for pressurized water reactor (PWR) spent nuclear fuel determined with the Standardized Computer Analysis for Licensing Evaluation (SCALE) 5.1 (Ref. ) epletion calculation method. Specifically, the depletion computer code and the cross-section library being evaluated are the twodimensional (2-D) transport and depletion module, TRITON/NEWT,2, 3 and the 44GROUPNDF5 (Ref. 4) cross-section library, respectively, in the SCALE .1 code system.

  6. The neutron beam intensity increase by in-core fuel management enhancement in multipurpose research reactors

    International Nuclear Information System (INIS)

    Martinc, R.; Vukadin, Z.; Konstantinovic, J.

    1986-01-01

    The exploitation characteristics of an existing multipurpose research reactor can be increased not only by great reconstruction, but also, to the considerable extent, by the in-core fuel management sophistication. The optimisation of the in-core fuel management procedure in such reactors is governed (among others) by the identified reactor utilisation goals, i.e. by weighting factors dedicated to different utilisation goals, which are often (regarding the in-core fuel management procedure) highly controversial. In this work the best solution for in-core fuel management is sought, with the highest weighting factor dedicated to the neutron beam usage, rather than sample irradiation in the reactor core. The term in-core fuel management includes: the core configuration, the locations of the fresh fuel inflow zone and spent fuel excite zone, and the fuel transfers between these two zones (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

  8. Assessment of fuel damage of pool type research reactor in the case of fuel plates blockage

    Energy Technology Data Exchange (ETDEWEB)

    Jalil, Jafari; Samad, Khakshournia [AEOI, Karegar Ave. School of R and D of Nuclear Reactors and Accelerators, Teheran (Iran, Islamic Republic of); D' Auria, F. [Pisa Univ., DIMNP (Italy)

    2007-07-01

    Tehran Research Reactor (TRR) is a pool type 5 MW research reactor. It is assumed that external objects or debris that may fall down to reactor core cause obstruction of coolant flow through one of the fuel assemblies. Thermal hydraulic analysis of this event, using the RELAP5 system code has been studied. The reported transient is related to the partial and total obstruction of a single Fuel Element (FE) cooling channel of 27 FE equilibrium core of TRR. Such event constitutes a severe accident for this type of reactor since it may lead to local dryout and eventually to loss of the FE integrity. Two scenarios are analysed to emphasize the severity of the accident. The first one is a partial blockage of an average FE considering four different obstruction levels: 25%, 50%, 75% and 97% of nominal flow area. The second one is an extreme scenario consisting of total blockage of the same FE. This study constitutes the first step of a larger work which consists of performing a 3-dimensional simulation using the Best Estimate coupled code technique. However, as a first approach the instantaneous reactor power is derived through the point kinetic calculation included in the RELAP5 code. Main results obtained from the RELAP5 calculations are as following. First, in the case of flow blockage under 97% of the nominal flow area of an average FE, only an increase of the coolant and clad temperatures is observed without any consequences for the integrity of the FE. The mass flow rate remains sufficient to cool the clad safely. Secondly, in the case of total obstruction of the nominal flow area, it is seen that transient turns out to be a severe accident due to the dryout conditions are reached shortly and melting of the cladding occurs. Thirdly, the use of the point kinetic approach leads to conservative results. A best estimate simulation of such kind of transients requires the use of 3-dimensional kinetic calculations, which could be done using the current Coupled Codes

  9. SCALE Validation Experience Using an Expanded Isotopic Assay Database for Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    Gauld, Ian C.; Radulescu, Georgeta; Ilas, Germina

    2009-01-01

    The availability of measured isotopic assay data to validate computer code predictions of spent fuel compositions applied in burnup-credit criticality calculations is an essential component for bias and uncertainty determination in safety and licensing analyses. In recent years, as many countries move closer to implementing or expanding the use of burnup credit in criticality safety for licensing, there has been growing interest in acquiring additional high-quality assay data. The well-known open sources of assay data are viewed as potentially limiting for validating depletion calculations for burnup credit due to the relatively small number of isotopes measured (primarily actinides with relatively few fission products), sometimes large measurement uncertainties, incomplete documentation, and the limited burnup and enrichment range of the fuel samples. Oak Ridge National Laboratory (ORNL) recently initiated an extensive isotopic validation study that includes most of the public data archived in the Organization for Economic Cooperation and Development/Nuclear Energy Agency (OECD/NEA) electronic database, SFCOMPO, and new datasets obtained through participation in commercial experimental programs. To date, ORNL has analyzed approximately 120 different spent fuel samples from pressurized-water reactors that span a wide enrichment and burnup range and represent a broad class of assembly designs. The validation studies, completed using SCALE 5.1, are being used to support a technical basis for expanded implementation of burnup credit for spent fuel storage facilities, and other spent fuel analyses including radiation source term, dose assessment, decay heat, and waste repository safety analyses. This paper summarizes the isotopic assay data selected for this study, presents validation results obtained with SCALE 5.1, and discusses some of the challenges and experience associated with evaluating the results. Preliminary results obtained using SCALE 6 and ENDF

  10. Proceedings of the international meeting on development, fabrication and application of reduced enrichment fuels for research and test reactors

    International Nuclear Information System (INIS)

    1983-08-01

    Separate abstracts were prepared for each of the papers presented in the following areas: (1) Reduced Enrichment Fuels for Research and Test Reactors (RERTR) Program Status; (2) Fuel Development; (3) Fuel Demonstrations; (4) General Topics; and (5) Specific Reactor Applications

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

    International Nuclear Information System (INIS)

    Hu Xiaodan

    2010-01-01

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

  12. Application of powder metallurgy in production of nuclear fuels for research and power reactors

    International Nuclear Information System (INIS)

    Fukuda, Kosaku

    2000-01-01

    Powder metallurgy has been applied in many of the processes of nuclear fuel fabrication, which has contributed, to a great progress of the nuclear technology to date. Evolution of nuclear fuels still continues to meet various emerging demands in terms of enhanced safety, economical effectiveness, non-proliferation and environmental mitigation. This paper reviews recent progress of nuclear fuels of research and power reactors, in particular, focusing on the powder metallurgy application. First, the review is made on plate type fuels for research reactors, inter alia, silicide fuel which is prevailing worldwide from the viewpoint of non-proliferation. The relation between fabrication and irradiation behavior is also discussed. Next, oxide fuels including MOX are reviewed. Recent interests of UO 2 are directed toward large grain pellets and burnable absorber pellets, both of which arise from requirement of extended burnup. Finally, the MOX fuel for thermal reactors is reviewed. (author)

  13. International collaboration, the route to fuel cycle research and development

    International Nuclear Information System (INIS)

    Tinsley, T.; Mathers, D.; Rayment, F.

    2013-01-01

    In hindsight, involvement with European Framework projects such as GoFastR (Gas-cooled Fast Reactors) and ACSEPT (Actinide Recycling by Separation and Transmutation) was a crucial and, at the time, an innovative step in maintaining the UK skills base during a period of major changes in the UK nuclear industry. It has undoubtedly delivered the objectives intended in terms of maintenance of the key skills, developing and training new staff, regenerating facilities and building strong links with the European nuclear research community. Over the last 2-3 years NNL's participation in European projects has moved forward such that NNL (National Nuclear Laboratory) is an integral partner of several major projects, fully engaged with delivering the core objectives of the projects and intent on forging deep collaborations with key organisations across Europe. With the renewed interest in nuclear energy and future fuel cycle options in the UK, NNL is now well positioned to contribute at an even deeper level in European level programmes

  14. International collaboration, the route to fuel cycle research and development

    Energy Technology Data Exchange (ETDEWEB)

    Tinsley, T.; Mathers, D.; Rayment, F. [National Nuclear Laboratory, Central Laboratory, Sellafield, Seascale, Cumbria, CA20 1PG (United Kingdom)

    2013-07-01

    In hindsight, involvement with European Framework projects such as GoFastR (Gas-cooled Fast Reactors) and ACSEPT (Actinide Recycling by Separation and Transmutation) was a crucial and, at the time, an innovative step in maintaining the UK skills base during a period of major changes in the UK nuclear industry. It has undoubtedly delivered the objectives intended in terms of maintenance of the key skills, developing and training new staff, regenerating facilities and building strong links with the European nuclear research community. Over the last 2-3 years NNL's participation in European projects has moved forward such that NNL (National Nuclear Laboratory) is an integral partner of several major projects, fully engaged with delivering the core objectives of the projects and intent on forging deep collaborations with key organisations across Europe. With the renewed interest in nuclear energy and future fuel cycle options in the UK, NNL is now well positioned to contribute at an even deeper level in European level programmes.

  15. Research of fuel temperature control in fuel pipeline of diesel engine using positive temperature coefficient material

    Directory of Open Access Journals (Sweden)

    Xiaolu Li

    2016-01-01

    Full Text Available As fuel temperature increases, both its viscosity and surface tension decrease, and this is helpful to improve fuel atomization and then better combustion and emission performances of engine. Based on the self-regulated temperature property of positive temperature coefficient material, this article used a positive temperature coefficient material as electric heating element to heat diesel fuel in fuel pipeline of diesel engine. A kind of BaTiO3-based positive temperature coefficient material, with the Curie temperature of 230°C and rated voltage of 24 V, was developed, and its micrograph and element compositions were also analyzed. By the fuel pipeline wrapped in six positive temperature coefficient ceramics, its resistivity–temperature and heating characteristics were tested on a fuel pump bench. The experiments showed that in this installation, the surface temperature of six positive temperature coefficient ceramics rose to the equilibrium temperature only for 100 s at rated voltage. In rated power supply for six positive temperature coefficient ceramics, the temperature of injection fuel improved for 21°C–27°C within 100 s, and then could keep constant. Using positive temperature coefficient material to heat diesel in fuel pipeline of diesel engine, the injection mass per cycle had little change, approximately 0.3%/°C. This study provides a beneficial reference for improving atomization of high-viscosity liquids by employing positive temperature coefficient material without any control methods.

  16. Channel geometric scales effect on performance and optimization for serpentine proton exchange membrane fuel cell (PEMFC)

    Science.gov (United States)

    Youcef, Kerkoub; Ahmed, Benzaoui; Ziari, Yasmina; Fadila, Haddad

    2017-02-01

    A three dimensional computational fluid dynamics model is proposed in this paper to investigate the effect of flow field design and dimensions of bipolar plates on performance of serpentine proton exchange membrane fuel cell (PEMFC). A complete fuel cell of 25 cm2 with 25 channels have been used. The aim of the work is to investigate the effect of flow channels and ribs scales on overall performance of PEM fuel cell. Therefore, geometric aspect ratio parameter defined as (width of flow channel/width of rib) is used. Influences of the ribs and openings current collector scales have been studied and analyzed in order to find the optimum ratio between them to enhance the production of courant density of PEM fuel cell. Six kind of serpentine designs have been used in this paper included different aspect ratio varying from 0.25 to 2.33 while the active surface area and number of channels are keeping constant. Aspect ratio 0.25 corresponding of (0.4 mm channel width/ 1.6mm ribs width), and Aspect ratio2.33 corresponding of (0.6 mm channel width/ 1.4mm ribs width. The results show that the best flow field designs (giving the maximum density of current) are which there dimensions of channels width is minimal and ribs width is maximal (Γ≈0.25). Also decreasing width of channels enhance the pressure drop inside the PEM fuel cell, this causes an increase of gazes velocity and enhance convection process, therefore more power generation.

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  18. Studies and research concerning BNFP: spent fuel dry storage studies at the Barnwell Nuclear Fuel Plant

    International Nuclear Information System (INIS)

    Anderson, K.J.

    1980-09-01

    Conceptual designs are presented utilizing the Barnwell Nuclear Fuel Plant for the dry interim storage of spent light water reactor fuel. Studies were conducted to determine feasible approaches to storing spent fuel by methods other than wet pool storage. Fuel that has had an opportunity to cool for several years, or more, after discharge from a reactor is especially adaptable to dry storage since its thermal load is greatly reduced compared to the thermal load immediately following discharge. A thermal analysis was performed to help in determining the feasibility of various spent fuel dry storage concepts. Methods to reject the heat from dry storage are briefly discussed, which include both active and passive cooling systems. The storage modes reviewed include above and below ground caisson-type storage facilities and numerous variations of vault, or hot cell-type, storage facilities

  19. Nuclear nonproliferation: Concerns with US delays in accepting foregin research reactors' spent fuel

    International Nuclear Information System (INIS)

    1994-01-01

    One key US nonproliferation goal is to discourage use of highly enriched uranium fuel (HEU), which can be used to make nuclear bombs, in civilian nuclear programs worldwide. DOE's Off-Site Fuels Policy for taking back spent HEU from foreign research reactors was allowed to expire due to environmental reasons. This report provides information on the effects of delays in renewing the Off-Site Fuels Policy on US nonproliferation goals and programs (specifically the reduced enrichment program), DOE's efforts to renew the fuels policy, and the price to be charged to the operators of foreign reactors for DOE's activities in taking back spent fuel

  20. Research reactors. Problems of fuel element enrichment reduction. Deliberations and comments

    International Nuclear Information System (INIS)

    1978-10-01

    This paper summarises the main data from the major research reactors in the Federal Republic of Germany utilising highly enriched uranium (HEU) and presently available fuel technology for their fuel elements. The required modification for an adaption of the fabrication to lower enriched fuel are considered as well as the consequences on reactor performance operation and licensing. On the basis of past experience with reactor modifications a rough estimate of 82 months is given for the conversion of a reactor to a modified type of fuel and of 70 months for a fuel test program. The conclusions reflect the own calculations and data from other papers submitted to INFCE-WG 8C

  1. MANAGEMENT OF RESEARCH AND TEST REACTOR ALUMINUM SPENT NUCLEAR FUEL - A TECHNOLOGY ASSESSMENT

    Energy Technology Data Exchange (ETDEWEB)

    Vinson, D.

    2010-07-11

    The Department of Energy's Environmental Management (DOE-EM) Program is responsible for the receipt and storage of aluminum research reactor spent nuclear fuel or used fuel until ultimate disposition. Aluminum research reactor used fuel is currently being stored or is anticipated to be returned to the U.S. and stored at DOE-EM storage facilities at the Savannah River Site and the Idaho Nuclear Technology and Engineering Center. This paper assesses the technologies and the options for safe transportation/receipt and interim storage of aluminum research reactor spent fuel and reviews the comprehensive strategy for its management. The U.S. Department of Energy uses the Appendix A, Spent Nuclear Fuel Acceptance Criteria, to identify the physical, chemical, and isotopic characteristics of spent nuclear fuel to be returned to the United States under the Foreign Research Reactor Spent Nuclear Fuel Acceptance Program. The fuel is further evaluated for acceptance through assessments of the fuel at the foreign sites that include corrosion damage and handleability. Transport involves use of commercial shipping casks with defined leakage rates that can provide containment of the fuel, some of which are breached. Options for safe storage include wet storage and dry storage. Both options must fully address potential degradation of the aluminum during the storage period. This paper focuses on the various options for safe transport and storage with respect to technology maturity and application.

  2. Fuel Cycle Research and Development. Technical Monthly - March 2012

    International Nuclear Information System (INIS)

    Miller, Michael C.

    2012-01-01

    meeting. Kenneth Dayman, the graduate student from University of Texas, completed an initial draft of his master's thesis. His research will contribute to the multivariate classifier currently under development. Sarah Bender, the graduate student from Pennsylvania State University, presented her work on a poster and in a conference paper at the MARC IX meeting. A mass balance flowsheet for the fast reactor fuel was completed and a model simulation is scheduled to begin construction next month. The development of a mass balance flowsheet for light water reactor fuel will predict the behavior of the separation process using mathematical functions. The completed flowsheet will be utilized as the basis for constructing the model simulation for the electrochemical separations. Comments and review of the model from the MPACT Working Group meeting have been used to evaluate updates to the EChem model. A preliminary physical security layout has been developed in ATLAS. Thermal stability tests for high temperature microfluidic interconnections were completed on all compounds tested for bonding strength. An interconnection strategy was determined based on these results that we expect will allow for operation at 400C in the first generation of sampling systems. Design of the sampling system using the chosen interconnections was initiated, with handoff to an external foundry for fabrication based on ANL specified process conditions expected by the middle of the month. Monte Carlo simulations of the sampling system were conducted under conditions of realistic sampling size distributions, electrorefiner inhomogeneity distributions, and detector efficiencies. These simulations were used to establish a baseline limit of detection for system operation, assuming an on-line separation step is conducted before detection. Sensor for measuring density and depth of molten electrolyte - The procurement effort continued. 80% of the components ordered to assemble the double bubbler have arrived at

  3. Calculations of 3D full-scale VVER fuel assembly and core models using MCU and BIPR-7A codes

    Energy Technology Data Exchange (ETDEWEB)

    Aleshin, Sergey S.; Bikeev, Artem S.; Bolshagin, Sergey N.; Kalugin, Mikhail A.; Kosourov, Evgeniy K.; Pavlovichev, Aleksandr M.; Pryanichnikov, Aleksandr V.; Sukhino-Khomenko, Evgenia A.; Shcherenko, Anna I.; Shcherenko, Anastasia I.; Shkarovskiy, Denis A. [Nuclear Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation)

    2015-09-15

    Two types of calculations were made to compare BIPR-7A and MCU results for 3D full-scale models. First EPS (emergency protection system) efficiency and in-core power distributions were analyzed for an equilibrium fuel load of VVER-1000 assuming its operation within an 18-month cycle. Computations were performed without feedbacks and with fuel burnup distributed over the core. After 3D infinite lattices of full-scale VVER-1000 fuel assemblies (A's) with uranium fuel 4.4% enrichment and uranium-erbium fuel 4.4% enrichment and Er{sub 2}O{sub 3} 1 % wt were considered. Computations were performed with feedbacks and fuel burnup at the constant power level. For different time moments effective multiplication factor and power distribution were obtained. EPS efficiency and reactivity effects at chosen time moments were analyzed.

  4. Analysis, scale modeling, and full-scale test of a railcar and spent-nuclear-fuel shipping cask in a high-velocity impact against a rigid barrier

    International Nuclear Information System (INIS)

    Huerta, M.

    1981-06-01

    This report describes the mathematical analysis, the physical scale modeling, and a full-scale crash test of a railcar spent-nuclear-fuel shipping system. The mathematical analysis utilized a lumped-parameter model to predict the structural response of the railcar and the shipping cask. The physical scale modeling analysis consisted of two crash tests that used 1/8-scale models to assess railcar and shipping cask damage. The full-scale crash test, conducted with retired railcar equipment, was carefully monitored with onboard instrumentation and high-speed photography. Results of the mathematical and scale modeling analyses are compared with the full-scale test. 29 figures

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1983-08-01

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

  6. Studies and research concerning BNFP. Nuclear spent fuel transportation studies

    International Nuclear Information System (INIS)

    Anderson, R.T.; Maier, J.B.

    1979-11-01

    Currently, there are a number of institutional problems associated with the shipment of spent fuel assemblies from commercial nuclear power plants: new and conflicting regulations, embargoing of certain routes, imposition of transport safeguards, physical security in-transit, and a lack of definition of when and where the fuel will be moved. This report presents a summary of these types and kinds of problems. It represents the results of evaluations performed relative to fuel receipt at the Barnwell Nuclear Fuel Plant. Case studies were made which address existing reactor sites with near-term spent fuel transportation needs. Shipment by either highway, rail, water, or intermodal water-rail was considered. The report identifies the impact of new regulations and uncertainty caused by indeterminate regulatory policy and lack of action on spent fuel acceptance and storage. This stagnant situation has made it impossible for industry to determine realistic transportation scenarios for business planning and financial risk analysis. A current lack of private investment in nuclear transportation equipment is expected to further prolong the problems associated with nuclear spent fuel and waste disposition. These problems are expected to intensify in the 1980's and in certain cases will make continuing reactor plant operation difficult or impossible

  7. Fundamentals of applied multidimensional scaling for educational and psychological research

    CERN Document Server

    Ding, Cody S

    2018-01-01

    This book explores the fundamentals of multidimensional scaling (MDS) and how this analytic method can be used in applied setting for educational and psychological research. The book tries to make MDS more accessible to a wider audience in terms of the language and examples that are more relevant to educational and psychological research and less technical so that the readers are not overwhelmed by equations. The goal is for readers to learn the methods described in this book and immediately start using MDS via available software programs. The book also examines new applications that have previously not been discussed in MDS literature. It should be an ideal book for graduate students and researchers to better understand MDS. Fundamentals of Applied Multidimensional Scaling for Educational and Psychological Research is divided into three parts. Part I covers the basic and fundamental features of MDS models pertaining to applied research applications. Chapters in this section cover the essential features of da...

  8. Summary of entire research achievements of creative engineering research program on nuclear fuel cycle

    International Nuclear Information System (INIS)

    Takenaka, Shingo; Ikegami, Tetsuo

    2008-03-01

    Creative Engineering Research Program on Nuclear Fuel Cycle (former In-house Innovative Research Encouraging Program) was implemented from FY 2001 to FY 2007 in order to support such in-house researches that create innovative new concepts and aim technical break-through. Totally 37 applications have been received and 14 research themes have been accepted and been performed in this program. As for the research achievements of the 14 research themes, first author papers accepted by scientific journals and by science councils were 47 and 32, respectively, and oral presentations at scientific societies were 99. Furthermore, interpretive articles for scientific journals, requested lectures, patents, and prize winnings were 13, 30, 8, and 3, respectively. Consequently, it can be evaluated that the research achievements resulted from this program are generally in high level and that the expectations, at the starting point of this program, to activate the innovative research activities have been accomplished. In this report, the final reports of the 14 research themes together with the outline of this program are included. (author)

  9. Research program on conditions to failure of high burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    Regarding the power ramp test to verify the out-of-pile test results on hydrogen-induced cladding failure, situation of the shipping port restoration after the earthquake disaster was investigated for the overseas transportation of test fuel rods which had been interrupted. Its reopening schedule was still currently uncertain and the power ramp test plan also remained suspended. The information about the fuel irradiation performance obtained from JNES projects and international projects, etc. is prepared as database, and based on the recent findings, the fuel irradiation performance models and analysis codes are developed and/or improved. (author)

  10. 2011 RENEWABLE ENERGY: SOLAR FUELS GORDON RESEARCH CONFERENCE

    Energy Technology Data Exchange (ETDEWEB)

    Joseph Hupp

    2011-01-21

    The conference will present and discuss current science that underlies solar fuels production, and will focus on direct production pathways for production. Thus, recent advances in design and understanding of molecular systems and materials for light capture and conversion of relevance for solar fuels will be discussed. An important set of topics will be homogeneous, heterogeneous and biological catalysts for the multi-electron processes of water oxidation, hydrogen production and carbon dioxide reduction to useful fuels. Also, progress towards integrated and scalable systems will be presented. Attached is a copy of the formal schedule and speaker program and the poster program.

  11. Numerical Research on Hybrid Fuel Locking Device for Upward Flow Core-Research Reactor

    International Nuclear Information System (INIS)

    Huh, Hyung; Cho, Yeong-Garp; Yoo, Yeon-Sik; Ryu, Jeong-Soo

    2016-01-01

    The assembly must be held firmly against these forces, but cannot be permanently attached to the support stand because periodic refueling of the reactor requires removal or relocation of each assembly. There are so many kinds of fuel locking device, but they are operated manually. As a part of a new project, we have investigated a hybrid fuel locking device (HFLD) for research reactor which is operated automatically. Prior method of holding down the fuel assembly includes a hybrid zero electromagnet consisting of an electromagnet and a permanent magnet. The role of an electromagnet is converged to zero power for overcoming the lifting power of a permanent magnet by controlling the coil current. At this time, a HFLD is an unlocking state. On the contrary, it is locking state that only a permanent magnet works when the power of an electromagnet is off. The results of a FEM in this work lead to the following conclusions: (1) It is possible that an electromagnet is converged to zero power for overcoming the lifting power of a permanent magnet by remote controlling the coil current. (2) At this time, it is able to detect remotely using proximity sensor whether a HFLD is latched or not

  12. A study of UO2 wafer fuel for very high-power research reactors

    International Nuclear Information System (INIS)

    Hsieh, T.C.; Jankus, V.Z.; Rest, J.; Billone, M.C.

    1983-01-01

    The Reduced Enrichment Research and Test Reactor Program is aimed at reducing fuel enrichment to 2 caramel fuel is one of the most promising new types of reduced-enrichment fuel for use in research reactors with very high power density. Parametric studies have been carried out to determine the maximum specific power attainable without significant fission-gas release for UO 2 wafers ranging from 0.75 to 1.50 mm in thickness. The results indicate that (1) all the fuel designs considered in this study are predicted not to fail under full power operation up to a burnup, of 1.9x10 21 fis/cm 3 ; (2) for all fuel designs, failure is predicted at approximately the same fuel centerline temperature for a given burnup; (3) the thinner the wafer, the wider the margin for fuel specific power between normal operation and increased-power operation leading to fuel failure; (4) increasing the coolant pressure in the reactor core could improve fuel performance by maintaining the fuel at a higher power level without failure for a given burnup; and (5) for a given power level, fuel failure will occur earlier at a higher cladding surface temperature and/or under power-cycling conditions. (author)

  13. Full scale simulations of accidents on spent-nuclear-fuel shipping systems

    International Nuclear Information System (INIS)

    Yoshimura, H.R.

    1978-01-01

    In 1977 and 1978, five first-of-a-kind full scale tests of spent-nuclear-fuel shipping systems were conducted at Sandia Laboratories. The objectives of this broad test program were (1) to assess and demonstrate the validity of current analytical and scale modeling techniques for predicting damage in accident conditions by comparing predicted results with actual test results, and (2) to gain quantitative knowledge of extreme accident environments by assessing the response of full scale hardware under actual test conditions. The tests were not intended to validate the present regulatory standards. The spent fuel cask tests fell into the following configurations: crashes of a truck-transport system into a massive concrete barrier (100 and 130 km/h); a grade crossing impact test (130 km/h) involving a locomotive and a stalled tractor-trailer; and a railcar shipping system impact into a massive concrete barrier (130 km/h) followed by fire. In addition to collecting much data on the response of cask transport systems, the program has demonstrated thus far that current analytical and scale modeling techniques are valid approaches for predicting vehicular and cask damage in accident environments. The tests have also shown that the spent casks tested are extremely rugged devices capable of retaining their radioactive contents in very severe accidents

  14. Manufacturing and Construction of Fresh Fuel Storage Rack for a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Jinho; Lee, Sangjin; Lee, Jongmin; Ryu, Jeong-Soo [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The objective of this study is to provide the manufacturing and construction information regarding the Fresh Fuel Storage Rack (FFSR). The main function of a FFSR is to store and protect the 2 core new fuel assemblies for the operation of the research reactor. The fresh fuel assemblies are stored in a rack made of stainless steel and the storage rack is installed in the fresh fuel storage room. The fresh fuel facility provides fresh fuel assemblies with dry storage space. General design requirements of the fresh fuel storage facilities are given in the ANSI-57.3. Design, manufacturing, and construction of the fresh fuel storage rack are introduced. The analysis is performed to confirm the structural intensity of the fresh fuel storage rack under the seismic loads. The fresh fuel storage rack designed for storage of fresh fuel assemblies should be manufactured and installed with consideration of predicted number of fresh fuel assemblies, structural integrity, resistivity to corrosion and radiation, cleaning, and workability.

  15. Manufacturing and Construction of Fresh Fuel Storage Rack for a Research Reactor

    International Nuclear Information System (INIS)

    Oh, Jinho; Lee, Sangjin; Lee, Jongmin; Ryu, Jeong-Soo

    2016-01-01

    The objective of this study is to provide the manufacturing and construction information regarding the Fresh Fuel Storage Rack (FFSR). The main function of a FFSR is to store and protect the 2 core new fuel assemblies for the operation of the research reactor. The fresh fuel assemblies are stored in a rack made of stainless steel and the storage rack is installed in the fresh fuel storage room. The fresh fuel facility provides fresh fuel assemblies with dry storage space. General design requirements of the fresh fuel storage facilities are given in the ANSI-57.3. Design, manufacturing, and construction of the fresh fuel storage rack are introduced. The analysis is performed to confirm the structural intensity of the fresh fuel storage rack under the seismic loads. The fresh fuel storage rack designed for storage of fresh fuel assemblies should be manufactured and installed with consideration of predicted number of fresh fuel assemblies, structural integrity, resistivity to corrosion and radiation, cleaning, and workability

  16. Improved oxidative stability of biodiesel fuels : antioxidant research and development.

    Science.gov (United States)

    2011-01-01

    Biodiesel is a domestic, renewable fuel that is gaining wide acceptance, especially in Europe. : When blended with conventional petroleum diesel, biodiesel reduces hydrocarbon, particulate : and carbon monoxide emissions, while having minimal to no e...

  17. Research on plant of metal fuel fabrication using casting process

    International Nuclear Information System (INIS)

    Senda, Yasuhide; Mori, Yukihide

    2003-12-01

    This document presents the plant concept of metal fuel fabrication system (38tHM/y) using casting process in electrolytic recycle, which based on recent studies of its equipment design and quality control system. And we estimate the cost of its construction and operation, including costs of maintenance, consumed hardware and management of waste. The content of this work is as follows. (1) Designing of fuel fabrication equipment: We make material flow diagrams of the fuel fabrication plant and rough designs of the injection casting furnace, demolder and inspection equipment. (2) Designing of resolution system of liquid waste, which comes from analytical process facility. Increased analytical items, we rearrange analytical process facility, estimate its chemicals and amount of waste. (3) Arrangement of equipments: We made a arrangement diagram of the metal fuel fabrication equipments in cells. (4) Estimation of cost data: We estimated cost to construct the facility and to operate it. (author)

  18. Investigation of research and development subjects for the Very High Burnup Fuel

    International Nuclear Information System (INIS)

    Hayashi, Kimio; Amano, Hidetoshi; Suzuki, Yasufumi; Furuta, Teruo; Nagase, Fumihisa; Suzuki, Masahide

    1993-06-01

    A concept of the Very High Burnup Fuel aiming at a maximum fuel assembly burnup of 100 GWd/t has been proposed in terms of burnup extension, utilization of Pu and transmutation of transuranium elements (TRU: Np, Am and Cm). The authors have investigated research and development (R and D) subjects of the fuel pellet and the cladding material of the Fuel. The present report describes the results on the fuel pellet. First, the chemical state of the Fuel and fission products (FP) was inferred through an FP-inventory and an equilibrium-thermodynamics calculations. Besides, knowledge obtained from post-irradiation examinations was surveyed. Next, an investigation was made on irradiation behavior of U/Pu mixed oxide (MOX) fuel with high enrichment of Pu, as well as on fission-gas release and swelling behavior of high burnup fuels. Reprocessibility of the Fuel, particularly solubility of the spent fuel, was also examined. As for the TRU-added fuel, material property data on TRU oxides were surveyed and summarized as a database. And the subjects on the production and the irradiation behavior were examined on the basis of experiences of MOX fuel production and TRU-added fuel irradiation. As a whole, the present study revealed the necessity of accumulating fundamental data and knowledge required for design and assessment of the fuel pellet, including the information on properties and irradiation performance of the TRU-added fuel. Finally, the R and D subjects were summarized, and a proposal was made on the way of development of the fuel pellet and cladding materials. (author)

  19. Health physics aspects of a research reactor fuel shipment

    International Nuclear Information System (INIS)

    Dodd, B.; Johnson, A.G.; Anderson, T.V.

    1984-01-01

    In June 1982, 92 irradiated fuel elements were shipped from the Oregon State University TRIGA Reactor to Westinghouse Hanford Corporation to be used in the Fuel Materials Examination Facility, This paper describes some of the health physics aspects of the planning, preparation and procedures associated with that shipment. In particular, the lessons learned are described in order that the benefits of the experience gained may be readily available to other small institutions. (author)

  20. Research in Supercritical Fuel Properties and Combustion Modeling

    Science.gov (United States)

    2015-09-18

    identified reactions needing further study and C-2 and C-3 species to add to the mechanism . 15. SUBJECT TERMS Supercritical fluids , Brillouin scattering...kinetics mechanism for combustion of hydrocarbon fuels containing up to 2 carbon atoms, including uncertainties. • We identified key reactions and...safety. The chemical mechanisms for combustion of all of these fuels share the same set of elementary reactions of smaller-fragment hydrocarbons , and

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

    Science.gov (United States)

    2009-04-27

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

  2. Large-scale Labeled Datasets to Fuel Earth Science Deep Learning Applications

    Science.gov (United States)

    Maskey, M.; Ramachandran, R.; Miller, J.

    2017-12-01

    Deep learning has revolutionized computer vision and natural language processing with various algorithms scaled using high-performance computing. However, generic large-scale labeled datasets such as the ImageNet are the fuel that drives the impressive accuracy of deep learning results. Large-scale labeled datasets already exist in domains such as medical science, but creating them in the Earth science domain is a challenge. While there are ways to apply deep learning using limited labeled datasets, there is a need in the Earth sciences for creating large-scale labeled datasets for benchmarking and scaling deep learning applications. At the NASA Marshall Space Flight Center, we are using deep learning for a variety of Earth science applications where we have encountered the need for large-scale labeled datasets. We will discuss our approaches for creating such datasets and why these datasets are just as valuable as deep learning algorithms. We will also describe successful usage of these large-scale labeled datasets with our deep learning based applications.

  3. Status and prospects of safety research of fuel cycle facilities in France

    International Nuclear Information System (INIS)

    Auchere, H.; Mercier, J.P.

    1996-01-01

    The following themes of research are discussed: prolonged loss of cooling in concentrated fission product solution storage tanks; dewatering of a spent fuel storage pond; explosion risks in nuclear fuel cycle laboratories and plants; dissemination of radioactive materials in case of fire in fuel manufacturing plants and in spent fuel analysis laboratories; contamination transfer; phenomenology of liquid uranium hexafluoride vaporization into the atmosphere; ways and means of intervention in the event of liquid ClF 3 leakage; offsite explosion; seismic research. (K.A.)

  4. Bioenergy Research Programme, Yearbook 1995. Production of wood fuels; Bioenergian tutkimusohjelma, vuosikirja 1995. Puupolttoaineen tuotantotekniikka

    Energy Technology Data Exchange (ETDEWEB)

    Alakangas, E. [ed.

    1996-12-31

    Bioenergy Research Programme is one of the energy technology research programmes of the Technology Development Center TEKES. The aim of the Bioenergy Research Programme is to increase, by using technical research and development, the economically profitable and environmentally sound utilisation of bioenergy, to improve the competitiveness of present peat and wood fuels, and to develop new competitive fuels and equipment related to bioenergy. The funding for 1995 was nearly 52 million FIM and the number of projects 66. The main goal of the wood fuels research area is to develop new production methods in order to decrease the production costs to the level of imported fuels. The total potential of the wood fuel use should be at least 1.0 million toe/a (5.5 million m{sup 3}). During the year 1995 There were over 30 projects concerning the production of wood derived fuels going on. Nearly half of them focused on integrated production of pulp wood and wood fuel. About ten projects was carried out to promote the wood fuel production from logging residues. Other topics were firewood production, production logistics and wood fuel resources. For production of fuel chips from logging residues, a new chipper truck, MOHA-SISU, was introduced. The new machine gives a new logistic solution resulting in high productivity and reasonable operating costs. In Mikkeli region three years of active work promoted the usage of wood fuel in a district power plant to the level of over 110 000 m{sup 3} of fuel chips. The production costs tend to be a little high in average, and the production chain still needs to be improved

  5. Bioenergy Research Programme, Yearbook 1995. Production of wood fuels; Bioenergian tutkimusohjelma, vuosikirja 1995. Puupolttoaineen tuotantotekniikka

    Energy Technology Data Exchange (ETDEWEB)

    Alakangas, E [ed.

    1997-12-31

    Bioenergy Research Programme is one of the energy technology research programmes of the Technology Development Center TEKES. The aim of the Bioenergy Research Programme is to increase, by using technical research and development, the economically profitable and environmentally sound utilisation of bioenergy, to improve the competitiveness of present peat and wood fuels, and to develop new competitive fuels and equipment related to bioenergy. The funding for 1995 was nearly 52 million FIM and the number of projects 66. The main goal of the wood fuels research area is to develop new production methods in order to decrease the production costs to the level of imported fuels. The total potential of the wood fuel use should be at least 1.0 million toe/a (5.5 million m{sup 3}). During the year 1995 There were over 30 projects concerning the production of wood derived fuels going on. Nearly half of them focused on integrated production of pulp wood and wood fuel. About ten projects was carried out to promote the wood fuel production from logging residues. Other topics were firewood production, production logistics and wood fuel resources. For production of fuel chips from logging residues, a new chipper truck, MOHA-SISU, was introduced. The new machine gives a new logistic solution resulting in high productivity and reasonable operating costs. In Mikkeli region three years of active work promoted the usage of wood fuel in a district power plant to the level of over 110 000 m{sup 3} of fuel chips. The production costs tend to be a little high in average, and the production chain still needs to be improved

  6. Research and development of spent-fuel shipping casks and the criteria for sea-going vessels carrying them

    International Nuclear Information System (INIS)

    Aoki, S.; Ando, Y.

    1977-01-01

    Since the transport of spent fuel will increase rapidly and extensively in the near future, the Japanese Atomic Energy Committee enacted the Technical Standard for Transportation of Radioactive Materials, based on the IAEA Regulation for the Safe Transport of Radioactive Materials, 1973 Revised Edition. The authorities concerned have begun to review the former ordinances for transporting radioactive materials and to develop a unified system of relevant laws and standards. For ten years the Atomic Energy Bureau has invested in research and development to obtain data for the design and licence of a spent-fuel shipping cask. Different scale models of a prototype weighing 80t were used to clarify the scale effect of drop, puncture and fire tests, which are a feature of Japanese research and development. Also an immersion test in water at pressures up to about 500 bar is now carried out to investigate the integrity of the cask body and sealing structure to prevent leakage of radioactive contents to the surroundings should the cask fall into deep sea. In Japan, depending on the site of nuclear plants, almost all transport of unirradiated and spent fuels is by sea. Therefore, to secure safe transport, the design criteria of ships for the exclusive transport of spent-fuel shipping casks, namely full-load shipping, have been enacted, which aim to make the likelihood of sinking on collision, stranding, and other unforeseen accidents at sea highly improbable and also to keep radiation exposure of the crew as low as possible. (author)

  7. Transportation of failed or damaged foreign research reactor spent nuclear fuel

    International Nuclear Information System (INIS)

    Messick, Charles E.; Mustin, Tracy P.; Massey, Charles D.

    1999-01-01

    Since initiating the Foreign Research Reactor Spent Nuclear Fuel (FRR SNF) Acceptance Program in 1996, the Program has had to deal with difficult issues associated with the transportation of failed or damaged spent fuel. In several instances, problems with failed or damaged fuel have prevented the acceptance of the fuel at considerable cost to both the Department of Energy and research reactor operators. In response to the problems faced by the Acceptance Program, DOE has undertaken significant steps to better define the spent fuel acceptance criteria. DOE has worked closely with the U.S. Nuclear Regulatory Commission to address failed or damaged research reactor spent fuel causing a degradation of the fuel assembly exposing fuel meat and to identify cask certificate issues which must be resolved by cask owners and foreign regulatory authorities. The specific issues and implementation challenges associated with the transport of MTR type FRR SNF will be discussed. The information presented will include U.S. Nuclear Regulatory Commission regulatory issues, cask certificate issues, technical constraints, implementation status, and lessons learned. Specific information will also be provided on the latest efforts to revise DOE's Appendix B, Transport Package (Cask) Acceptance Criteria. The information presented in this paper will be of interest to foreign research reactor operators, shippers, and cask vendors in evaluating the condition of their fuel to ensure it can be transported in accordance with appropriate cask certificate requirements. (author)

  8. Evaluation of spent fuel isotopics, radiation spectra and decay heat using the scale computational system

    International Nuclear Information System (INIS)

    Parks, C.V.; Hermann, O.W.; Ryman, J.C.

    1986-01-01

    In order to be a self-sufficient system for transport/storage cask shielding and heat transfer analysis, the SCALE system developers included modules to evaluate spent fuel radiation spectra and decay heat. The primary module developed for these analyses is ORIGEN-S which is an updated verision of the original ORIGEN code. The COUPLE module was also developed to enable ORIGEN-S to easily utilize multigroup cross sections and neutron flux data during a depletion analysis. Finally, the SAS2 control module was developed for automating the depletion and decay via ORIGEN-S while using burnup-dependent neutronic data based on a user-specified fuel assembly and reactor history. The ORIGEN-S data libraries available for depletion and decay have also been significantly updated from that developed with the original ORIGEN code

  9. Transmutation research and fuel cycle (report on discussion at Research Reactor Institute, Kyoto University)

    International Nuclear Information System (INIS)

    Yamana, Hajimu

    1999-01-01

    A symposium was held on a topic of 'Transmutation Research' on Dec. 21 and 22, 1999 at Research Reactor Institute, Kyoto University. This meeting was held as a joint-meeting of KUR's specialist meeting and Tokyo University's activity supported by the Grant-in-Aid for Scientific Research of Ministry of Education, Sport and Culture of Japan. This paper describes the overview of the discussions of this joint-meeting, and interprets their significance. Major themes discussed are, needed discussions on the transmutation research, policy and concepts of the organizations doing transmutation researches, a view from university side, transmutation researches in the oversea countries, opinions from various standpoints of the nuclear fuel cycle, conclusive discussions. 'the meanings of the transmutation research should be discussed together with the geological disposal and fast reactor system', 'transmutation may be a cooperative option for the disposal, thus, they should not be in a independent relation', and Balance evaluation will be needed' are the examples of the conclusive remarks of this meeting. (author)

  10. Power ramp testing method for PWR fuel rod at research reactor

    International Nuclear Information System (INIS)

    Zhou Yidong; Zhang Peisheng; Zhang Aimin; Gao Yongguang; Wang Huarong

    2003-01-01

    A tentative power ramp test for short PWR fuel rod has been conducted at the Heavy Water Research Reactor (HWRR) in China Institute of Atomic Energy (CIAE). The test fuel rod was cooled by the circulating water in the test loop. The power ramp was realized by moving solid neutron-absorbing screen around the fuel rod. The linear power of the fuel rod increased from 220 W/cm to 340 W/cm with a power ramp rate of 20 W/cm/min. The power of the fuel rod was monitored by both in-core thermal and nuclear measurement sensors in the test rig. This test provides experiences for further developing the power ramp test methods for PWR fuel rods at research reactor. (author)

  11. Foreign research reactor irradiated nuclear fuel inventories containing HEU and LEU of United States origin

    International Nuclear Information System (INIS)

    Matos, J.E.

    1994-12-01

    This report provides estimates of foreign research reactor inventories of aluminum-based and TRIGA irradiated nuclear fuel elements containing highly enriched and low enriched uranium of United States origin that are anticipated in January 1996, January 2001, and January 2006. These fuels from 104 research reactors in 41 countries are the same aluminum-based and TRIGA fuels that were eligible for receipt under the Department of Energy's Offsite Fuels Policy that was in effect in 1988. All fuel inventory and reactor data that were available as of December 1, 1994, have been included in the estimates of approximately 14,300 irradiated fuel elements in January 1996, 18,800 in January 2001, and 22,700 in January 2006

  12. Outline of results of safety research (in nuclear fuel cycle field in fiscal year 1996)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-01

    The safety research in Power Reactor and Nuclear Fuel Development Corporation in fiscal year 1996 has been carried out based on the basic plan of safety research (from fiscal year 1996 to 2000) which was decided in March, 1996. In this report, on nuclear fuel cycle field, namely all the subjects in the fields of nuclear fuel facilities, environmental radioactivity and waste disposal, and the subjects related to nuclear fuel facilities among the fields of aseismatic and probabilistic safety assessments, the results of research in fiscal year 1996, the first year of the 5-year project, are summarized together with the outline of the basic plan of safety research. The basic policy, objective and system for promotion of the safety research are described. The objectives of the safety research are the advancement of safety technology, the safety of facilities, stable operation techniques, the safety design and the evaluation techniques of next generation facilities, and the support of transferring nuclear fuel cycle to private businesses. The objects of the research are uranium enrichment, fuel fabrication and reprocessing, and waste treatment and storage. 52 investigation papers of the results of the safety research in nuclear fuel cycle field in fiscal year 1996 are collected in this report. (K.I.)

  13. Small-scale hot facility for reprocessing and alpha research

    International Nuclear Information System (INIS)

    Abdel-Rassoul, A.A.

    1976-01-01

    The experimental hot facility at Inchas is planned for research activities related to the decontamination of radioactive wastes, analytical chemistry of alpha emitters and chemical treatment of spent UO 2 -Mg fuel samples. The design concept permits safe handling of source materials with radioactivity levels up to 10000Ci. The laboratory includes a reception area, process hall, a number of research laboratories and other facilities for chemical and physical analysis, nuclear measurements and health physics control. The radioactive waste management plant allows for control and decontamination of intermediate- and low-level laboratory effluents. Fixation of radioactive residues will be carried out in the sludge immobilization plant. High-level fission-product waste liquors are subject to preconcentration and transformation to a glassy matrix before ultimate storage. (author)

  14. Research reactors spent fuel management in the Nuclear Research Institute Rez

    International Nuclear Information System (INIS)

    Rychecky, J.

    2001-01-01

    In Czech Republic 3 research and testing nuclear reactors are operated at present time, with the biggest one being the Nuclear Research Institute (NRI) reactor LVR-15, operated with maximum power 10 MW. This reactor serves as a radiation source for material testing, producing of ionizing radiation sources, theoretical studies, and, most recently, for boron neutron capture therapy. Another NRI reactor LR-0 is a reactor of zero power used mainly for the studies of WWER 1000 spent fuel criticality. For training of students the reactor called VRABEC (VR-1), operated also with very low power, serves since 1990 at the Faculty of Nuclear Engineering, of Czech Technical University. The similar testing type reactor (SR-0), already decommissioned, was also used since 1974 to 1989 in Skoda, Nuclear Machinery, Plzen. This contribution summarizes the present state of the spent fuel (SF) management of these nuclear reactors. As the SF management is different for very low or zero power reactors and power reactors, the first type will be only briefly discussed, and then the main attention will be devoted to SF management of the NRI experimental reactor LVR-15

  15. Research reactors spent fuel management in the Nuclear Research Institute Rez

    Energy Technology Data Exchange (ETDEWEB)

    Rychecky, J. [Nuclear Research Institute, 25068 Rez (Czech Republic)

    2001-07-01

    In Czech Republic 3 research and testing nuclear reactors are operated at present time, with the biggest one being the Nuclear Research Institute (NRI) reactor LVR-15, operated with maximum power 10 MW. This reactor serves as a radiation source for material testing, producing of ionizing radiation sources, theoretical studies, and, most recently, for boron neutron capture therapy. Another NRI reactor LR-0 is a reactor of zero power used mainly for the studies of WWER 1000 spent fuel criticality. For training of students the reactor called VRABEC (VR-1), operated also with very low power, serves since 1990 at the Faculty of Nuclear Engineering, of Czech Technical University. The similar testing type reactor (SR-0), already decommissioned, was also used since 1974 to 1989 in Skoda, Nuclear Machinery, Plzen. This contribution summarizes the present state of the spent fuel (SF) management of these nuclear reactors. As the SF management is different for very low or zero power reactors and power reactors, the first type will be only briefly discussed, and then the main attention will be devoted to SF management of the NRI experimental reactor LVR-15.

  16. Criticality safety of storage barrels for enriched uranium fresh fuel at the RB research reactor

    International Nuclear Information System (INIS)

    Pesic, M. P.

    1997-01-01

    Study on criticality safety of fresh low and high enriched uranium (LEU and HEU) fuel elements in the storage/transport barrels at the RB research reactor is carried out by using the well-known MCNP computer code. It is shown that studied arrays of tightly closed fuel barrels, each entirely loaded with 100 fresh (HEU or LEU) fuel slugs, are far away from criticality, even in cases of an unexpected flooding by light water.(author)

  17. Increased power to heat ratio of small scale CHP plants using biomass fuels and natural gas

    International Nuclear Information System (INIS)

    Savola, Tuula; Fogelholm, Carl-Johan

    2006-01-01

    In this paper, we present a systematic study of process changes for increased power production in 1-20 MW e combined heat and power (CHP) plants. The changes are simulated, and their economic feasibility evaluated by using existing small scale CHP case plants. Increasing power production in decentralised CHP plants that operate according to a certain heat demand could reduce the fuel consumption and CO 2 emissions per power unit produced and improve the feasibility of CHP plant investments. The CHP plant process changes were simulated under design and off design conditions and an analysis of power and heat production, investment costs and CO 2 emissions was performed over the whole annual heat demand. The results show that using biomass fuels, there are profitable possibilities to increase the current power to heat ratios, 0.23-0.48, of the small scale CHP plants up to 0.26-0.56, depending on the size of the plant. The profitable changes were a two stage district heat exchanger and the addition of a steam reheater and a feed water preheater. If natural gas is used as an additional fuel, the power to heat ratio may be increased up to 0.35-0.65 by integrating a gas engine into the process. If the CO 2 savings from the changes are also taken into account, the economic feasibility of the changes increases. The results of this work offer useful performance simulation and investment cost knowledge for the development of more efficient and economically feasible small scale CHP processes

  18. UK MOD research on H{sub 2} storage and generation methods for soldier-portable fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Browning, D.J. [Defence Evaluation Research Agency, Hampshire (United Kingdom)

    2000-05-01

    The research which the Defence Evaluation and Research Agency (DERA) has conducted on hydrogen storage and production methods was presented and their applicability to soldier portable applications was discussed. Polymer electrolyte membrane (PEM) fuel cells are considered to be the future power source for solider-portable equipment such as army radios. Advances in fuel cell technology has resulted in improvements in specific power and power density as well as reductions in Pt loading and cost. However, a viable in-service power source is still required. This fuel source is likely to be hydrogen because it has a high energy per unit weight and has a simple reaction with oxygen to produce water as the only by-product. In the future, methanol may be used as the fuel source but currently methanol fuel cells demonstrate poorer performance than those running on hydrogen. A vital part of a fuel cell system, therefore, is hydrogen storage, of which there are two basic types; physical and chemical storage. In addition to weight, other important factors to be considered in a storage system are cost, reliability and safety. This paper described each physical storage method including hydrogen stored as a compressed gas, as a cryogenic liquid, reversibly absorbed in intermetallic alloys and reversibly adsorbed onto carbon materials such as nonofibres. Liquid hydrogen storage is only feasible for larger scale applications such as electric vehicles and is not practical for portable applications. A man-portable fuel cell is likely to be either pressure cylinders or reversible metal hydrides. 6 refs., 1 tab., 10 figs.

  19. Hydrogen Research for Spaceport and Space-Based Applications: Fuel Cell Projects

    Science.gov (United States)

    Anderson, Tim; Balaban, Canan

    2008-01-01

    The activities presented are a broad based approach to advancing key hydrogen related technologies in areas such as fuel cells, hydrogen production, and distributed sensors for hydrogen-leak detection, laser instrumentation for hydrogen-leak detection, and cryogenic transport and storage. Presented are the results from research projects, education and outreach activities, system and trade studies. The work will aid in advancing the state-of-the-art for several critical technologies related to the implementation of a hydrogen infrastructure. Activities conducted are relevant to a number of propulsion and power systems for terrestrial, aeronautics and aerospace applications. Fuel cell research focused on proton exchange membranes (PEM), solid oxide fuel cells (SOFC). Specific technologies included aircraft fuel cell reformers, new and improved electrodes, electrolytes, interconnect, and seals, modeling of fuel cells including CFD coupled with impedance spectroscopy. Research was conducted on new materials and designs for fuel cells, along with using embedded sensors with power management electronics to improve the power density delivered by fuel cells. Fuel cell applications considered were in-space operations, aviation, and ground-based fuel cells such as; powering auxiliary power units (APUs) in aircraft; high power density, long duration power supplies for interplanetary missions (space science probes and planetary rovers); regenerative capabilities for high altitude aircraft; and power supplies for reusable launch vehicles.

  20. Treatment of spent fuels from research reactors and reactor development programs in Germany

    International Nuclear Information System (INIS)

    Closs, K.D.

    1999-01-01

    Quite a great number of different types of spent fuel from research reactors and development programs exists in Germany. The general policy is to send back to the USA as long as possible fuel from MTRs and TRIGAs of USA origin. An option is reprocessing in Great Britain or France. This option is pursued as long as reprocessing and reuse of the recovered material is economically justifiable. For those fuels which cannot be returned to the USA or which will not be reprocessed, a domestic back-up solution of spent fuel management has been developed in Germany, compatible with the management of spent fuel from power reactors. It consists in dry storage in special casks and, later on, direct disposal. Preliminary results from experimental R and D investigations with research reactor fuel and experience from LWR fuel lead to the conclusion that the direct disposal option even for research reactor fuel or exotic fuel does not impose major technical difficulties for the German waste management and disposal concept. (author)

  1. PWR core and spent fuel pool analysis using scale and nestle

    International Nuclear Information System (INIS)

    Murphy, J. E.; Maldonado, G. I.; St Clair, R.; Orr, D.

    2012-01-01

    The SCALE nuclear analysis code system [SCALE, 2011], developed and maintained at Oak Ridge National Laboratory (ORNL) is widely recognized as high quality software for analyzing nuclear systems. The SCALE code system is composed of several validated computer codes and methods with standard control sequences, such as the TRITON/NEWT lattice physics sequence, which supplies dependable and accurate analyses for industry, regulators, and academia. Although TRITON generates energy-collapsed and space-homogenized few group cross sections, SCALE does not include a full-core nodal neutron diffusion simulation module within. However, in the past few years, the open-source NESTLE core simulator [NESTLE, 2003], originally developed at North Carolina State Univ. (NCSU), has been updated and upgraded via collaboration between ORNL and the Univ. of Tennessee (UT), so it now has a growingly seamless coupling to the TRITON/NEWT lattice physics [Galloway, 2010]. This study presents the methodology used to couple lattice physics data between TRITON and NESTLE in order to perform a three-dimensional full-core analysis employing a 'real-life' Duke Energy PWR as the test bed. The focus for this step was to compare the key parameters of core reactivity and radial power distribution versus plant data. Following the core analysis, following a three cycle burn, a spent fuel pool analysis was done using information generated from NESTLE for the discharged bundles and was compared to Duke Energy spent fuel pool models. The KENO control module from SCALE was employed for this latter stage of the project. (authors)

  2. PWR core and spent fuel pool analysis using scale and nestle

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, J. E.; Maldonado, G. I. [Dept. of Nuclear Engineering, Univ. of Tennessee, Knoxville, TN 37996-2300 (United States); St Clair, R.; Orr, D. [Duke Energy, 526 S. Church St, Charlotte, NC 28202 (United States)

    2012-07-01

    The SCALE nuclear analysis code system [SCALE, 2011], developed and maintained at Oak Ridge National Laboratory (ORNL) is widely recognized as high quality software for analyzing nuclear systems. The SCALE code system is composed of several validated computer codes and methods with standard control sequences, such as the TRITON/NEWT lattice physics sequence, which supplies dependable and accurate analyses for industry, regulators, and academia. Although TRITON generates energy-collapsed and space-homogenized few group cross sections, SCALE does not include a full-core nodal neutron diffusion simulation module within. However, in the past few years, the open-source NESTLE core simulator [NESTLE, 2003], originally developed at North Carolina State Univ. (NCSU), has been updated and upgraded via collaboration between ORNL and the Univ. of Tennessee (UT), so it now has a growingly seamless coupling to the TRITON/NEWT lattice physics [Galloway, 2010]. This study presents the methodology used to couple lattice physics data between TRITON and NESTLE in order to perform a three-dimensional full-core analysis employing a 'real-life' Duke Energy PWR as the test bed. The focus for this step was to compare the key parameters of core reactivity and radial power distribution versus plant data. Following the core analysis, following a three cycle burn, a spent fuel pool analysis was done using information generated from NESTLE for the discharged bundles and was compared to Duke Energy spent fuel pool models. The KENO control module from SCALE was employed for this latter stage of the project. (authors)

  3. Research on using depleted uranium as nuclear fuel for HWR

    International Nuclear Information System (INIS)

    Zhang Jiahua; Chen Zhicheng; Bao Borong

    1999-01-01

    The purpose of our work is to find a way for application of depleted uranium in CANDU reactor by using MOX nuclear fuel of depleted U and Pu instead of natural uranium. From preliminary evaluation and calculation, it was shown that MOX nuclear fuel consisting of depleted uranium enrichment tailings (0.25% 235 U) and plutonium (their ratio 99.5%:0.5%) could replace natural uranium in CANDU reactor to sustain chain reaction. The prospects of application of depleted uranium in nuclear energy field are also discussed

  4. Combustion and Heat Transfer Studies Utilizing Advanced Diagnostics: Fuels Research

    Science.gov (United States)

    1992-11-01

    competing reactions (Reactions (6) and (7)) and their respective rate equations (Eqs. 8 and 9). Reaction (6) has the advantage of no activation energy and...cartridges were J&W (diol, cyano and C-18) J.T. Baker (silica gel) and Alltech (IC/Ag). All were conditioned and used according to the vendors...hot fuel through the test section. At the same time the water supply to the fuel cooler is turned on. The main advantage of this procedure is that it

  5. Environmental impact study on a degassing and scaling of cistern trucks for fuel transportation plant.

    OpenAIRE

    Tupia, Elmer

    2014-01-01

    The present Study of Environmental Impact (EIA) it is carried out by the Company Ecoplanet Group from the Peru CORP, to application of Servisya CORP, proprietor of a degassing plant and scaling of trucks cistern for transports of derived liquid fuels of the hydrocarbons, the environment of the location area undestood in the Country of Lima of the district of Villa El Salvador being. El presente estudio de Impacto Ambiental(EIA) es realizado por la empresa Ecoplanet group del Perú S.A. a s...

  6. Perspectives on research and development of microbial fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Ortega-Martinez, A.; Vazquez Larios, A.L.; Solorza-Feria, O.; Poggi Varaldo, H.M. [Centro de Investigacion y de Estudios Avanzados del IPN, Mexico D.F. (Mexico)]. E-mail: hectorpoggi2001@gmail.com; rareli@hotmail.com

    2009-09-15

    Microbial fuel cells (MFC), is an anoxic electrochemical bioreactor where bacteria grow in the absence of oxygen in a chamber containing an anode which it may be covered by a biofilm. Microorganisms anoxically oxidize the organic substrate and electrons generated are released to the anode. Released protons are transferred to the cathode. Natural or forced aeration of the cathode supplies the oxygen for the final reaction 2H{sup +} + 2e{sup -} + (1/2) O{sub 2} = H{sub 2}O. In this work, we present a critical review on MFC focused on subjects that are receiving a growing interest from the research and technological communities: (i) types of MFC, their relative advantages and disadvantages and ranges of application; (ii) development of biocathodes; (iii) enrichment procedures of microbial communities in MFC. Recent research shows that one-chamber fitted with cathode aerated by natural aeration, and other special types of high performance MFC, have displaced the historical two-chamber MFC. Recent studies showed that electrochemically active bacteria (EAB) can be successfully enriched in MFC. The cost and eventual poisoning of the platinum catalyst used at the cathode is a major limitation to MFC application and economic viability. Researchers have started working on the concept of biocathodes that would use bacteria instead of platinum as a biocatalyst. Microbial enrichment of inocula seeded to MFC may provide a way to enrich the consortium with EAB, thus substantially increasing the transfer of electrons to the anode. Bioaugmentation of consortia in MFC with strains EAB, could contribute to the same goal. [Spanish] Las celdas de combustible microbianas (CCM) son un biorrector anoxico donde las bacterias crecen en ausencia de oxigeno en una camara que contiene un anodo que puede cubrirse con una biopelicula. Los microorganismos oxidan onoxicamente el sustrato organico y los electrones generados se liberan al anodo. Los protones liberados se transfieren al catodo. La

  7. Corrosion surveillance programme for Latin American research reactor Al-clad spent fuel in water

    International Nuclear Information System (INIS)

    Ramanathan, L.V.; Haddad, R.; Ritchie, I.

    2002-01-01

    The objectives of the IAEA sponsored Regional Technical Co-operation Project for Latin America (Argentina, Brazil, Chile, Mexico, and Peru) are to provide the basic conditions to define a regional strategy for managing spent fuel and to provide solutions, taking into consideration the economic and technological realities of the countries involved. In particular, to determine the basic conditions for managing research reactor spent fuel during operation and interim storage as well as final disposal, and to establish forms of regional cooperation in the four main areas: spent fuel characterization, safety, regulation and public communication. This paper reports the corrosion surveillance activities of the Regional Project and these are based on the IAEA sponsored co-ordinated research project (CRP) on 'Corrosion of research reactor Al-clad spent fuel in water'. The overall test consists of exposing corrosion coupon racks at different spent fuel basins followed by evaluation. (author)

  8. Application of room temperature ionic liquids in advanced fuel cycles RIAR research concept program users

    International Nuclear Information System (INIS)

    Bychkov, Alexander V.; Kormilitsyn, Michael V.; Savochkin, Yuri P.; Osipenko, Alexander G.; Smolensky, Valeri V.; Shadrin, Alexander Yu.; Babain, Vladimir A.

    2005-01-01

    The paper reviews briefly the research program on application of Room Temperature Ionic Liquids (RTILs) in some processes of the nuclear fuel reprocessing, particularly in the uranium-aluminum fuel reprocessing and separation of TPEs and REEs from the PUREX process liquid waste, and approaches to implementation of this program. (author)

  9. Researches of real observation geometry in monitoring fuel-containing materials' subcriticality

    International Nuclear Information System (INIS)

    Vysotskij, E.D.; Shevchenko, V.G.; Shevchenko, M.V.

    2002-01-01

    The effectiveness of fuel-containing materials monitoring is discussed in the part related to the feasibilities of researches and realization of optimal geometry (detectors - source) of survey of neutron activity dynamics in nuclearly hazardous areas with clusters of fuel-containing materials concentrated in the premises 305/2

  10. Social science to improve fuels management: a synthesis of research on collaboration.

    Science.gov (United States)

    Victoria Sturtevant; Margaret Ann Moote; Pamela Jakes; Anthony S. Cheng

    2005-01-01

    A series of syntheses were commissioned by the USDA Forest Service to aid in fuels mitigation project planning. This synthesis focuses on collaboration research, and offers knowledge and tools to improve collaboration in the planning and implementation of wildland fire and fuels management projects. It covers a variety of topics including benefits of collaboration,...

  11. Conversion and start up of Tehran Research Reactor with LEU fuel

    International Nuclear Information System (INIS)

    Zaker, M.

    2004-01-01

    The MW Tehran Research Reactor, Highly Enriched Uranium (HEU) fuel has been converted to Low Enriched Uranium (LEU) fuel using U 3 0 8 -Al with less than 20% enriched uranium. Measured value of excess reactivity, control rod worth and other parameters indicate good agreement with computational predictions. (author)

  12. Program for fundamental and applied research of fuel cells in VNIIEF

    Energy Technology Data Exchange (ETDEWEB)

    Anisin, A.V.; Borisseonock, V.A.; Novitskii, Y.Z.; Potyomckin, G.A.

    1996-04-01

    According to VNIIEF the integral part of development of fuel cell power plants is fundamental and applied research. This paper describes areas of research on molten carbonate fuel cells. Topics include the development of mathematical models for porous electrodes, thin film electrolytes, the possibility of solid nickel anodes, model of activation polarization of anode, electrolyte with high solubility of oxygen. Other areas include research on a stationary mode of stack operation, anticorrosion coatings, impedance diagnostic methods, ultrasound diagnostics, radiation treatments, an air aluminium cell, and alternative catalysts for low temperature fuel cells.

  13. Nuclear fuel cycle: research and development and push technologies

    International Nuclear Information System (INIS)

    Oliveira, Wagner dos Santos

    2002-01-01

    The scope of this work is to show the importance of 'push technologies in the development of the Nuclear Fuel Cycle more specifically the so called 'Projeto Conversao' PROCON. This R and D activities lead to the design of special equipment, new metallic and polymer materials. (author)

  14. Research and development issues for molten carbonate fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Krumpelt, M.

    1996-04-01

    This paper describes issues pertaining to the development of molten carbonate fuel cells. In particular, the corrosion resistance and service life of nickel oxide cathodes is described. The resistivity of lithium oxide/iron oxides and improvement with doping is addressed.

  15. Alternative Fuel Fleet Vehicle Evaluations | Transportation Research | NREL

    Science.gov (United States)

    delivery, transit, and freight vehicles. Although biodiesel is the most commonly used alternative fuel in Diesel and Biodiesel Renewable diesel is a conventional petroleum diesel substitute produced from alternative to conventional diesel and does not require any vehicle modifications. Biodiesel is an oxygenated

  16. Estimation of PWR spent fuel composition using SCALE and SWAT code systems

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Hee Sung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Kenya, Suyama; Hiroshi, Okuno [Japan Atomic Energy Research Institute, Tokyo (Japan)

    2001-05-01

    The isotopic composition calculations were performed for 26 spent fuel samples from Obrigheim PWR reactor and 55 spent fuel samples from 7 PWR reactors using SCALE4.4 SAS2H with 27, 44 and 238 group cross-section libraries and SWAT with 107 group cross-section library. For convenience, the ratio of the measured to calculated value was used as a parameter. The four kinds of the calculation results were compared with the measured data. For many important nuclides for burnup credit criticality safety evaluation, the four methods applied in this study showed good coincidence with measurements in general. More precise observations showed the following results. Less unity ratios were found for Pu-239 and -241 for selected 16 samples out of the 26 samples from Obrigheim reactor. Larger than unity ratios were found for Am-241 for both the 16 and 55 samples. Larger than unity ratios were found for Sm-149 for the 55 samples. In the case of 26 sample SWAT was generally accompanied by larger ratios than those of SAS2H with some exceptions. Based on the measured-to-calculated ratios for 71 samples of a combined set in which 16 selected samples and 55 samples were included, the correction factors that should be multiplied to the calculated isotopic compositions were generated for a conservative estimate of the neutron multiplication factor of a system containing PWR spent fuel, taking burnup credit into account.

  17. Small Scale SOFC Demonstration Using Bio-Based and Fossil Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Petrik, Michael [Technology Management Inc., Cleveland, OH (United States); Ruhl, Robert [Technology Management Inc., Cleveland, OH (United States)

    2012-05-01

    Technology Management, Inc. (TMI) of Cleveland, Ohio, has completed the project entitled Small Scale SOFC Demonstration using Bio-based and Fossil Fuels. Under this program, two 1-kW systems were engineered as technology demonstrators of an advanced technology that can operate on either traditional hydrocarbon fuels or renewable biofuels. The systems were demonstrated at Patterson's Fruit Farm of Chesterland, OH and were open to the public during the first quarter of 2012. As a result of the demonstration, TMI received quantitative feedback on operation of the systems as well as qualitative assessments from customers. Based on the test results, TMI believes that > 30% net electrical efficiency at 1 kW on both traditional and renewable fuels with a reasonable entry price is obtainable. The demonstration and analysis provide the confidence that a 1 kW entry-level system offers a viable value proposition, but additional modifications are warranted to reduce sound and increase reliability before full commercial acceptance.

  18. Brief report of the JNC cooperative research (C) on the nuclear fuel cycle for 2002

    International Nuclear Information System (INIS)

    2003-08-01

    The Japan Nuclear Cycle Development Institute (JNC) started the JNC Cooperative Research Scheme (C) on the Nuclear Fuel Cycle in 1999 in order to promote research collaboration with universities and other research institutes (referred to as 'universities, etc. 'below'), which means that the fiscal year 2002 was the fourth year of the scheme. This scheme is to invite foresighted and original themes of basic and fundamental research, to be performed using JNC's facilities and equipment, in principle, for the research and development of JNC in the field of the nuclear fuel cycle, from researchers at universities, etc. Those researchers who propose research themes in response to the invitation are expected to lead the research. The purpose of the scheme is to promote the efficiency of basic and fundamental research and development by enhancing the research environment of JNC through collaboration between JNC's own researchers and other researchers, and exchange of information and publication of achievements, etc. Research themes proposed by researchers at universities, etc. are screened by a screening committee consisting mainly of experts independent of JNC. This report includes a summary of the results of the research carried out in fiscal year 2002 on twenty selected themes related to the JNC's collaborative research for fast breeder reactors, the nuclear fuel cycle, radiation safety and geological disposal/geoscience; eight of which ended in 2002. Of these, three were related to fast breeder reactors, two to the nuclear fuel cycle and three to geological disposal/geoscience. (author)

  19. Conversion of research and test reactors to low enriched uranium fuel: technical overview and program status

    International Nuclear Information System (INIS)

    Roglans-Ribas, J.

    2008-01-01

    Many of the nuclear research and test reactors worldwide operate with high enriched uranium fuel. In response to worries over the potential use of HEU from research reactors in nuclear weapons, the U.S Department of Energy (DOE) initiated a program - the Reduced Enrichment for Research and Test Reactors (RERTR) - in 1978 to develop the technology necessary to reduce the use of HEU fuel by converting research reactors to low enriched uranium (LEU) fuel. The Reactor Conversion program is currently under the DOE's National Nuclear Security Administration's Global Threat Reduction Initiative (GTRI). 55 of the 129 reactors included in the scope have been already converted to LEU fuel or have shutdown prior to conversion. The major technical activities of the Conversion Program include: (1) the development of advanced LEU fuels; (2) conversion analysis and conversion support; and (3) technology development for the production of Molybdenum-99 (Mo 99 ) with LEU targets. The paper provides an overview of the status of the program, the technical challenges and accomplishments, and the role of international collaborations in the accomplishment of the Conversion Program objectives. Nuclear research and test reactors worldwide have been in operation for over 60 years. Many of these facilities operate with high enriched uranium fuel. In response to increased worries over the potential use of HEU from research reactors in the manufacturing of nuclear weapons, the U.S Department of Energy (DOE) initiated a program - the Reduced Enrichment for Research and Test Reactors (RERTR) - in 1978 to develop the technology necessary to reduce the use of HEU fuel in research reactors by converting them to low enriched uranium (LEU) fuel. The reactor conversion program was initially focused on U.S.-supplied reactors, but in the early 1990s it expanded and began to collaborate with Russian institutes with the objective of converting Russian supplied reactors to the use of LEU fuel.

  20. DISMANTLING OF THE FUEL CELL LABORATORY AT RESEARCH CENTRE JUELICH

    International Nuclear Information System (INIS)

    Stahn, B.; Matela, K.; Bensch, D.; Ambos, Frank

    2003-01-01

    The fuel cell laboratory was constructed in three phases and taken into operation in the years 1962 to 1966. The last experimental work was carried out in 1996. After all cell internals had been disassembled, the fuel cell laboratory was transferred to shutdown operation in 1997. Three cell complexes, which differed, in particular, by the type of shielding (lead, cast steel, concrete), were available until then for activities at nuclear components. After approval by the regulatory authority, the actual dismantling of the fuel cell laboratory started in March 2000. The BZ I laboratory area consisted of 7 cells with lead shieldings of 100 to 250 mm thickness. This area was dismantled from April to September 2000. Among other things, approx. 30,000 lead bricks with a total weight of approx. 300 Mg were dismantled and disposed of. The BZ III laboratory area essentially consisted of cells with concrete shieldings of 1200 to 1400 mm thickness. The dismantling of this area started in the fir st half of 2001 and was completed in November 2002. Among other things, approx. 900 Mg of concrete was dismantled and disposed of. Since more than 90 % of the dismantled materials was measurable for clearance, various clearance measurement devices were used during dismantling. The BZ II laboratory area essentially consists of cells with cast steel shieldings of 400 to 460 mm thickness. In September 2002 it was decided to continue using this laboratory area for future tasks. The dismantling of the fuel cell laboratory was thus completed. After appropriate refurbishment, the fuel cell laboratory will probably take up operation again in late 2003

  1. Wildland Fire Behaviour Case Studies and Fuel Models for Landscape-Scale Fire Modeling

    Directory of Open Access Journals (Sweden)

    Paul-Antoine Santoni

    2011-01-01

    Full Text Available This work presents the extension of a physical model for the spreading of surface fire at landscape scale. In previous work, the model was validated at laboratory scale for fire spreading across litters. The model was then modified to consider the structure of actual vegetation and was included in the wildland fire calculation system Forefire that allows converting the two-dimensional model of fire spread to three dimensions, taking into account spatial information. Two wildland fire behavior case studies were elaborated and used as a basis to test the simulator. Both fires were reconstructed, paying attention to the vegetation mapping, fire history, and meteorological data. The local calibration of the simulator required the development of appropriate fuel models for shrubland vegetation (maquis for use with the model of fire spread. This study showed the capabilities of the simulator during the typical drought season characterizing the Mediterranean climate when most wildfires occur.

  2. Investigation of Micro- and Macro-Scale Transport Processes for Improved Fuel Cell Performance

    Energy Technology Data Exchange (ETDEWEB)

    Gu, Wenbin [General Motors LLC, Pontiac, MI (United States)

    2014-08-29

    This report documents the work performed by General Motors (GM) under the Cooperative agreement No. DE-EE0000470, “Investigation of Micro- and Macro-Scale Transport Processes for Improved Fuel Cell Performance,” in collaboration with the Penn State University (PSU), University of Tennessee Knoxville (UTK), Rochester Institute of Technology (RIT), and University of Rochester (UR) via subcontracts. The overall objectives of the project are to investigate and synthesize fundamental understanding of transport phenomena at both the macro- and micro-scales for the development of a down-the-channel model that accounts for all transport domains in a broad operating space. GM as a prime contractor focused on cell level experiments and modeling, and the Universities as subcontractors worked toward fundamental understanding of each component and associated interface.

  3. Nuclear fuels technologies fiscal year 1996 research and research development test results

    International Nuclear Information System (INIS)

    Beard, C.A.; Blair, H.T.; Buksa, J.J.; Butt, D.P.

    1996-01-01

    During fiscal year 1996, the Department of Energy's Office of Fissile Materials Disposition (OFMD) funded Los Alamos National Laboratory (LANL) to investigate issues associated with the fabrication of plutonium from dismantled weapons into mixed-oxide (MOX) nuclear fuel for disposition in nuclear power reactors. These issues can be divided into two main categories: issues associated with the fact that the plutonium from dismantled weapons contains gallium, and issues associated with the unique characteristics of the PuO 2 produced by the dry conversion process that OFMD is proposing to convert the weapons material. Initial descriptions of the experimental work performed in fiscal year 1996 to address these issues can be found in Nuclear Fuels Technologies Fiscal Year 1996 Research and Development Test Matrices'. However, in some instances the change in programmatic emphasis towards the Parallex program either altered the manner in which some of these experiments were performed (i.e., the work was done as part of the Parallex fabrication development and not as individual separate-effects tests as originally envisioned) or delayed the experiments into Fiscal Year 1997. This report reviews the experiments that were conducted and presents the results. 7 figs., 14 tabs

  4. Validation concerns for dry storage of foreign research reactor spent nuclear fuel

    International Nuclear Information System (INIS)

    Trumble, E.F.

    1994-01-01

    Recent decisions by the Department of Energy have accelerated the need for storage options to support the return of foreign research reactor (FRR) fuel to the United States. Many of these returns consist of fuel types which contain highly enriched uranium and are aluminum clad. These attributes present many challenges not experienced in the fuel storage designs for commercial nuclear fuels where the fuels have lower enrichment and the cladding is more robust. Historically, returned FRR fuel has been stored for short periods in basins where it is cooled and then sent to be reprocessed. However, a severe lack of basin space and questionable availability of reprocessing facilities necessitates the development of other proposals. One proposed option is to store the FRR fuel in a dry state, thus reducing the corrosion problems associated with aluminum cladding. A drawback to this type of storage, however, is the lack of experimental data for this type of fuel under dry storage conditions. This lack of data has led to recent discussions over the accuracy of some of the current multigroup cross section libraries when applied to dry, fast systems of uranium and aluminum. This concern is evaluated for the specific case of Material Test Reactor (MTR) fuel (MTR is >60% of FRR fuel), a review of applicable experiments is presented and a new experiment is proposed

  5. Swelling Estimation of Multi-wire U-Mo Monolithic Fuel for HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yoon-Sang; Ryu, Ho-Jin; Park, Jong-Man; Oh, Jong-Myeong; Kim, Chang-Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-10-15

    In order to use low-enriched uranium (LEU) instead of highly enriched uranium (HEU) for high performance research reactors, the reduced enrichment for research and test reactors (RERTR) program is developing high uranium density fuel such as U-Mo/Al dispersion fuel. U-Mo alloys have an excellent irradiation performance when compared to other uranium alloys or compounds. But the results from the post-irradiation examination of the U-Mo/Al dispersion fuels indicate that an interaction between the U-Mo alloy fuel and the Al matrix phases occurs readily during an irradiation and it is sensitively dependent on the temperature. In order to lessen these severe interactions, a concept of a multi-wire type fuel was proposed. The fuel configuration is that three to six U-Mo fuel wires (1.5 mm - 2 mm in diameter) are symmetrically arranged at the periphery side in the Al matrix. In this study temperature calculations and a swelling estimation of a multi-wire monolithic fuel were carried out. Also the results of a post irradiation analysis of this fuel will be introduced.

  6. Subcritical Noise Analysis Measurements with Fresh and Spent Research Reactor Fuels Elements

    International Nuclear Information System (INIS)

    Valentine, T.E.; Mihalczo, J.T.; Kryter, R.C.; Miller, V.C.

    1999-01-01

    The verification of the subcriticality is of utmost importance for the safe transportation and storage of nuclear reactor fuels. Transportation containers and storage facilities are designed such that nuclear fuels remain in a subcritical state. Such designs often involve excess conservatism because of the lack of relevant experimental data to verify the accuracy of Monte Carlo codes used in nuclear criticality safety analyses. A joint experimental research program between Oak Ridge National Laboratory, Westinghouse Safety Management Solutions, Inc., and the University of Missouri was initiated to obtain measured quantities that could be directly related to the subcriticality of simple arrays of Missouri University Research Reactor (MURR) fuel elements. A series of measurement were performed to assess the reactivity of materials such as BORAL, stainless steel, aluminum, and lead that are typically used in the construction of shipping casks. These materials were positioned between the fuel elements. In addition, a limited number of measurements were performed with configurations of fresh and spent (irradiated) fuel elements to ascertain the reactivity of the spent fuel elements. In these experiments, fresh fuel elements were replaced by spent fuel elements such that the subcritical reactivity change could be measured. The results of these measurements were used by Westinghouse Safety Management Solutions to determine the subcriticality of MURR fuel elements isolated by absorbing materials. The measurements were interpreted using the MCNP-DSP Monte Carlo code to obtain the subcritical neutron multiplication factor k(sub eff), and the bias in K(sub eff) that are used in criticality safety analyses

  7. Status report on conversion of the Georgia Tech Research Reactor to low enrichment fuel

    International Nuclear Information System (INIS)

    Karam, R.A.; Matos, J.E.; Mo, S.C.; Woodruff, W.L.

    1995-01-01

    The 5 MW Georgia Tech Research Reactor (GTRR) is a heterogeneous, heavy water moderated and cooled reactor, fueled with highly-enriched uranium aluminum alloy fuel plates. The GTRR is required to convert to low enrichment (LEU) fuel in accordance with USNRC policy. The US Department of Energy is funding a program to compare reactor performance with high and low enrichment fuels. The goals of the program are: (1) to amend the SAR and the technical specifications of the GTRR so that LEU U 3 Si 2 -Al dispersion fuel plates can replace the current HEU U-Al alloy fuel, and (2) to optimize the LEU core such that maximum value neutron beams can be extracted for possible neutron capture therapy application. This paper presents a status report on the LEU conversion effort. (author)

  8. Status report on conversion of the Georgia Tech Research Reactor to low enrichment fuel

    International Nuclear Information System (INIS)

    Karam, R.A.; Matos, J.E.; Mo, S.C.; Woodruff, W.L.

    1991-01-01

    The 5 MW Georgia Tech Research Reactor (GTRR) is a heterogeneous, heavy water moderated and cooled reactor, fueled with highly-enriched uranium aluminum alloy fuel plates. The GTRR is required to convert to low enrichment (LEU) fuel in accordance with USNRC policy. The US Department of Energy is funding a program to compare reactor performance with high and low enrichment fuels. The goals of the program are: (1) to amend the SAR and the Technical Specifications of the GTRR so that LEU U 3 Si 2 -Al dispersion fuel plates can replace the current HEU U-Al alloy fuel, and (2) to optimize the LEU core such that maximum value neutron beams can be extracted for possible neutron capture therapy application. This paper presents a status report on the LEU conversion effort

  9. Behaviour of power and research reactor fuel in wet and dry storage

    Energy Technology Data Exchange (ETDEWEB)

    Freire-Canosa, J [Nuclear Waste Management Organization (Canada)

    2012-07-01

    Canada has developed extensive experience in both wet and dry storage of CANDU fuel. Fuel has been stored in water pools at CANDU reactor sites for approximately 45 years, and in dry storage facilities for a large part of the past decade. Currently, Canada has 38 450 t U of spent fuel in storage, of which 8850 t U are in dry storage. In June 2007, the Government of Canada selected the Adaptive Phased Management (APM) approach, recommended by the Nuclear Waste Management Organization (NWMO), for the long-term management of Canada's nuclear-fuel waste. The Canadian utilities and AECL are conducting development work in extended storage systems as well as research on fuel behaviour under storage conditions. Both activities have as ultimate objective to establish a technical basis for assuring the safety of long-term fuel storage.

  10. Validation of PWR core seismic models with shaking table tests on interacting scale 1 fuel assemblies

    International Nuclear Information System (INIS)

    Viallet, E.; Bolsee, G.; Ladouceur, B.; Goubin, T.; Rigaudeau, J.

    2003-01-01

    The fuel assembly mechanical strength must be justified with respect to the lateral loads under accident conditions, in particular seismic loads. This justification is performed by means of time-history analyses with dynamic models of an assembly row in the core, allowing for assembly deformations, impacts at grid locations and reactor coolant effects. Due to necessary simplifications, the models include 'equivalent' parameters adjusted with respect to dynamic characterisation tests of the fuel assemblies. Complementing such tests on isolated assemblies by an overall model validation with shaking table tests on interacting assemblies is obviously desirable. Seismic tests have been performed by French CEA (Commissariat a l'Energie Atomique) on a row of six full scale fuel assemblies, including two types of 17 x 17 12ft design. The row models are built according to the usual procedure, with preliminary characterisation tests performed on a single assembly. The test-calculation comparisons are made for two test configurations : in air and in water. The relatively large number of accelerograms (15, used for each configuration) is also favourable to significant comparisons. The results are presented for the impact forces at row ends, displacements at mid assembly, and also 'statistical' parameters. Despite a non-negligible scattering in the results obtained with different accelerograms, the calculations prove realistic, and the modelling process is validated with a good confidence level. This satisfactory validation allows to evaluate precisely the margins in the seismic design methodology of the fuel assemblies, and thus to confirm the safety of the plants in case of seismic event. (author)

  11. Small-scale reforming of diesel and jet fuels to make hydrogen and syngas for fuel cells: A review

    International Nuclear Information System (INIS)

    Xu, Xinhai; Li, Peiwen; Shen, Yuesong

    2013-01-01

    Highlights: • Issues of reforming of heavy hydrocarbon fuels are reviewed. • The advantages of autothermal reforming over other types of reforming are discussed. • The causes and solutions of the major problems for reforming reactors are studied. • Designs and startup strategies for autothermal reforming reactors are proposed. - Abstract: This paper reviews the technological features and challenges of autothermal reforming (ATR) of heavy hydrocarbon fuels for producing hydrogen and syngas onboard to supply fuels to fuel cells for auxiliary power units. A brief introduction at the beginning enumerates the advantages of using heavy hydrocarbon fuels onboard to provide hydrogen or syngas for fuel cells such as solid oxide fuel cells (SOFCs). A detailed review of the reforming and processing technologies of diesel and jet fuels is then presented. The advantages of ATR over steam reforming (SR) and partial oxidation reforming (POX) are summarized, and the ATR reaction is analyzed from a thermodynamic point of view. The causes and possible solutions to the major problems existing in ATR reactors, including hot spots, formation of coke, and inhomogeneous mixing of fuel, steam, and air, are reviewed and studied. Designs of ATR reactors are discussed, and three different reactors, one with a fixed bed, one with monoliths, and one with microchannels are investigated. Novel ideas for design and startup strategies for ATR reactors are proposed at the end of the review

  12. Overview of lower length scale model development for accident tolerant fuels regarding U3Si2 fuel and FeCrAl cladding

    International Nuclear Information System (INIS)

    Zhang, Yongfeng

    2016-01-01

    U3Si2 and FeCrAl have been proposed as fuel and cladding concepts, respectively, for accident tolerance fuels with higher tolerance to accident scenarios compared to UO2. However, a lot of key physics and material properties regarding their in-pile performance are yet to be explored. To accelerate the understanding and reduce the cost of experimental studies, multiscale modeling and simulation are used to develop physics-based materials models to assist engineering scale fuel performance modeling. In this report, the lower-length-scale efforts in method and material model development supported by the Accident Tolerance Fuel (ATF) high-impact-problem (HIP) under the NEAMS program are summarized. Significant progresses have been made regarding interatomic potential, phase field models for phase decomposition and gas bubble formation, and thermal conductivity for U3Si2 fuel, and precipitation in FeCrAl cladding. The accomplishments are very useful by providing atomistic and mesoscale tools, improving the current understanding, and delivering engineering scale models for these two ATF concepts.

  13. Overview of lower length scale model development for accident tolerant fuels regarding U3Si2 fuel and FeCrAl cladding

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Yongfeng [Idaho National Laboratory

    2016-09-01

    U3Si2 and FeCrAl have been proposed as fuel and cladding concepts, respectively, for accident tolerance fuels with higher tolerance to accident scenarios compared to UO2. However, a lot of key physics and material properties regarding their in-pile performance are yet to be explored. To accelerate the understanding and reduce the cost of experimental studies, multiscale modeling and simulation are used to develop physics-based materials models to assist engineering scale fuel performance modeling. In this report, the lower-length-scale efforts in method and material model development supported by the Accident Tolerance Fuel (ATF) high-impact-problem (HIP) under the NEAMS program are summarized. Significant progresses have been made regarding interatomic potential, phase field models for phase decomposition and gas bubble formation, and thermal conductivity for U3Si2 fuel, and precipitation in FeCrAl cladding. The accomplishments are very useful by providing atomistic and mesoscale tools, improving the current understanding, and delivering engineering scale models for these two ATF concepts.

  14. Enterprise SRS: Leveraging Ongoing Operations To Advance Nuclear Fuel Cycles Research And Development Programs

    Energy Technology Data Exchange (ETDEWEB)

    Murray, Alice M.; Marra, John E.; Wilmarth, William R.; Mcguire, Patrick W.; Wheeler, Vickie B.

    2013-07-03

    The Savannah River Site (SRS) is repurposing its vast array of assets to solve future national issues regarding environmental stewardship, national security, and clean energy. The vehicle for this transformation is Enterprise SRS which presents a new, radical view of SRS as a united endeavor for ''all things nuclear'' as opposed to a group of distinct and separate entities with individual missions and organizations. Key among the Enterprise SRS strategic initiatives is the integration of research into facilities in conjunction with on-going missions to provide researchers from other national laboratories, academic institutions, and commercial entities the opportunity to demonstrate their technologies in a relevant environment and scale prior to deployment. To manage that integration of research demonstrations into site facilities, The Department of Energy, Savannah River Operations Office, Savannah River Nuclear Solutions, the Savannah River National Laboratory (SRNL) have established a center for applied nuclear materials processing and engineering research (hereafter referred to as the Center). The key proposition of this initiative is to bridge the gap between promising transformational nuclear fuel cycle processing discoveries and large commercial-scale-technology deployment by leveraging SRS assets as facilities for those critical engineering-scale demonstrations necessary to assure the successful deployment of new technologies. The Center will coordinate the demonstration of R&D technologies and serve as the interface between the engineering-scale demonstration and the R&D programs, essentially providing cradle-to-grave support to the research team during the demonstration. While the initial focus of the Center will be on the effective use of SRS assets for these demonstrations, the Center also will work with research teams to identify opportunities to perform research demonstrations at other facilities. Unique to this approach is the fact

  15. Enterprise SRS: Leveraging Ongoing Operations To Advance Nuclear Fuel Cycles Research And Development Programs

    International Nuclear Information System (INIS)

    Murray, Alice M.; Marra, John E.; Wilmarth, William R.; Mcguire, Patrick W.; Wheeler, Vickie B.

    2013-01-01

    The Savannah River Site (SRS) is repurposing its vast array of assets to solve future national issues regarding environmental stewardship, national security, and clean energy. The vehicle for this transformation is Enterprise SRS which presents a new, radical view of SRS as a united endeavor for ''all things nuclear'' as opposed to a group of distinct and separate entities with individual missions and organizations. Key among the Enterprise SRS strategic initiatives is the integration of research into facilities in conjunction with on-going missions to provide researchers from other national laboratories, academic institutions, and commercial entities the opportunity to demonstrate their technologies in a relevant environment and scale prior to deployment. To manage that integration of research demonstrations into site facilities, The Department of Energy, Savannah River Operations Office, Savannah River Nuclear Solutions, the Savannah River National Laboratory (SRNL) have established a center for applied nuclear materials processing and engineering research (hereafter referred to as the Center). The key proposition of this initiative is to bridge the gap between promising transformational nuclear fuel cycle processing discoveries and large commercial-scale-technology deployment by leveraging SRS assets as facilities for those critical engineering-scale demonstrations necessary to assure the successful deployment of new technologies. The Center will coordinate the demonstration of R&D technologies and serve as the interface between the engineering-scale demonstration and the R&D programs, essentially providing cradle-to-grave support to the research team during the demonstration. While the initial focus of the Center will be on the effective use of SRS assets for these demonstrations, the Center also will work with research teams to identify opportunities to perform research demonstrations at other facilities. Unique to this approach is the fact that these SRS

  16. Corrosion of research reactor aluminium clad spent fuel in water. Additional information

    International Nuclear Information System (INIS)

    2009-12-01

    A large variety of research reactor spent fuel with different fuel meats, different geometries and different enrichments in 235 U are presently stored underwater in basins located around the world. More than 90% of these fuels are clad in aluminium or aluminium based alloys that are notoriously susceptible to corrosion in water of less than optimum quality. Some fuel is stored in the reactor pools themselves, some in auxiliary pools (or basins) close to the reactor and some stored at away-from-reactor pools. Since the early 1990s, when corrosion induced degradation of the fuel cladding was observed in many of the pools, corrosion of research reactor aluminium clad spent nuclear fuel stored in light water filled basins has become a major concern, and programmes were implemented at the sites to improve fuel storage conditions. The IAEA has since then established a number of programmatic activities to address corrosion of research reactor aluminium clad spent nuclear fuel in water. Of special relevance was the Coordinated Research Project (CRP) on Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water (Phase I) initiated in 1996, whose results were published in IAEA Technical Reports Series No. 418. At the end of this CRP it was considered necessary that a continuation of the CRP should concentrate on fuel storage basins that had demonstrated significant corrosion problems and would therefore provide additional insight into the fundamentals of localized corrosion of aluminium. As a consequence, the IAEA started a new CRP entitled Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water (Phase II), to carry out more comprehensive research in some specific areas of corrosion of aluminium clad spent nuclear fuel in water. In addition to this CRP, one of the activities under IAEA's Technical Cooperation Regional Project for Latin America Management of Spent Fuel from Research Reactors (2001-2006) was corrosion monitoring and surveillance of research

  17. Status of Research on Pebble Bed HTR Fuel Fabrication Technology in Indonesia

    International Nuclear Information System (INIS)

    Rachmawati, M.; Sarjono; Ridwan; Langenati, R.

    2014-01-01

    Research on pebble bed HTR fuel fabrication is conducted in Indonesia. One of the aims is to build a knowledge base on pebble bed HTR fuel element fabrication technology for fuel procurement. The steps of research strategies are firstly to understand the basic design research of TRISO fuel, properties, and requirements, and secondly to understand the TRISO fuel manufacturing technology, which comprises fabrication and quality control, including its facility. Both steps are adopted from research and experiences of the countries with HTR fuel element fabrication technology. From the knowledge gained in the research, an experimental design of the process and a set of prototype process equipment for fabrication are developed, namely kernels production using external gelation process, TRISO coating of the kernel, and pebble compacting. Experiments using the prototypes have been conducted. Characterization of the kernel product, i.e. diameter, sphericity, density and O/U ratio, shows that the kernel product is still not in compliance with the specification requirements. These are deemed to be caused mainly by the selected vibrating system and the viscosity adjustment. Another major cause is the selected NH3 and air feeding method for both NH3 and air layer in the preparation for spherical droplets of liquid. The FB-CVD TRISO coating of the kernel has been experimented but unsuccessful by using an FB-CVD once‐through continuous coating process. For the pebble compacting, the process is still in the early stage of setting-up compaction equipment. This paper summarizes the current status of research on HTR fuel fabrication technology in Indonesia, the proposed process and its equipment setting-up for improvement of the kernel production. The knowledge and lessons learned gained from the research is useful and can be an assistance in planning for fuel development laboratory facilities procurement, formulating User Requirement Document and Bid Invitation Specification for

  18. An improved reactor system for small-scale fuel processing - the Optiformer concept

    Energy Technology Data Exchange (ETDEWEB)

    Silversand, Fredrik; Karlsson, Charlotte (Catator AB, Lund (Sweden))

    2008-08-15

    Catator AB (CAT) has evaluated a revised design of the previously described Ultraformer concept. In comparison to the original Ultraformer, the new reformer concept (Optiformer) shows enhanced performances with respect to thermo-mechanic durability and conversion efficiencies. The unit is also easy to manufacture and ordinary high-temperature steel alloys may be used. The new concept is based on a helix-shaped tubular heat-exchanger reactor designed by ICI Caldaie (ICI) fitted with CATs proprietary wire-mesh catalyst. The concept has currently been evaluated for production rates between 0.5 and 30 nm3/hr of hydrogen and this report describes a detailed study of a unit for small-scale CHP-applications. Such systems will involve a fuel processor together with a suitable fuel cell, e.g. a solid-oxide fuel cell (SOFC) or a high-temperature PEFC (HT-PEFC). The evaluations performed in this study indicate stable operation over a wide window of capacities with negligible emissions of hydrocarbons. Since it is possible to operate the redesigned unit at a higher temperature (>900 deg C) than the original Ultraformer unit (750- 800 deg C), the conversion degree is much higher for thermo-dynamical reasons. The redesigned unit contains all necessary structures for vaporization, recuperation, effect supply and gas purification in a highly integrated structure. Furthermore, the unit is equipped with an internal insulation and a cooling jacked to reduce the skin temperature of the unit. The reactor has undergone about 100 full thermal cycles without any thermo-mechanical issues or catalyst degradation. Natural gas and different kerosene qualities have so far been evaluated with respect to conversion degree and possible slip of hydrocarbons. The conversion degree at rated load (100%) was above 99%, which enable us to reach superior efficiencies. If the unit were to be used together with a SOFC, the WGS-step could be omitted, reducing the size and weight further from about 2.2 l

  19. Small-scale Specimen Testing of Monolithic U-Mo Fuel Foils

    Energy Technology Data Exchange (ETDEWEB)

    Ramprashad Prabhakaran; Douglas E. Burkes; James I. Cole; Indrajit Charit; Daniel M. Wachs

    2008-10-01

    The objective of this investigation is to develop a shear punch testing (SPT) procedure and standardize it to evaluate the mechanical properties of irradiated fuels in a hot-cell so that the tensile behavior can be predicted using small volumes of material and at greatly reduced irradiation costs. This is highly important in the development of low-enriched uranium fuels for nuclear research and test reactors. The load-displacement data obtained using SPT can be interpreted in terms of and correlated with uniaxial mechanical properties. In order to establish a correlation between SPT and tensile data, sub-size tensile and microhardness testing were performed on U-Mo alloys. In addition, efforts are ongoing to understand the effect of test parameters (such as specimen thickness, surface finish, punch-die clearance, crosshead velocity and carbon content) on the measured mechanical properties, in order to rationalize the technique, prior to employing it on a material of unknown strength.

  20. Small-scale Specimen Testing of Monolithic U-Mo Fuel Foils

    International Nuclear Information System (INIS)

    Ramprashad Prabhakaran; Douglas E. Burkes; James I. Cole; Indrajit Charit; Daniel M. Wachs

    2008-01-01

    The objective of this investigation is to develop a shear punch testing (SPT) procedure and standardize it to evaluate the mechanical properties of irradiated fuels in a hot-cell so that the tensile behavior can be predicted using small volumes of material and at greatly reduced irradiation costs. This is highly important in the development of low-enriched uranium fuels for nuclear research and test reactors. The load-displacement data obtained using SPT can be interpreted in terms of and correlated with uniaxial mechanical properties. In order to establish a correlation between SPT and tensile data, sub-size tensile and microhardness testing were performed on U-Mo alloys. In addition, efforts are ongoing to understand the effect of test parameters (such as specimen thickness, surface finish, punch-die clearance, crosshead velocity and carbon content) on the measured mechanical properties, in order to rationalize the technique, prior to employing it on a material of unknown strength

  1. Caramel fuel for research reactors: experience acquired in the fabrication, monitoring and irradiation of Osiris core

    International Nuclear Information System (INIS)

    Contenson, Ghislain de; Foulquier, Henri; Trotabas, Maria; Vignesoult, Nicole; Cerles, J.-M.; Delafosse, Jacques.

    1981-06-01

    A plate type nuclear fuel (Caramel fuel) has been developed in France in the framework of the various activities pursued in the design, fabrication and development of nuclear fuels by the CEA. This fuel can be adapted to various different categories of water cooled reactor (power reactors, marine propulsion reactors, urbain heating reactors, research reactors). The successful work conducted in this field led the realization of a complete core and reloads for the high performance research reactor, Osiris, at Saclay. The existing highly enriched U-Al alloy fuel was replaced by a non-proliferating low enrichment (7%) caramel fuel. This new core has been operating successfully since january 1980. A brief description of Caramel and its main advantages is given. The way in which it is fabricated is described together with the quality controls to which it is subjected. The qualification program and the main results deduced from it are also presented. The program used to monitor its in-pile behavior is described. The essential purpose of this program is to ensure the high performance of the fuel under irradiation. The successful operation of Osiris, which terminated 11 irradiation cycles on the 21st of April 1981 confirmed the correctness of the decisions made and the excellent performance that could be achieved with these fuel elements under the severe conditions encountered in a high performance research reactor [fr

  2. Shipment of Taiwanese research reactor spent nuclear fuel (Phase 2): Environmental assessment

    International Nuclear Information System (INIS)

    1988-06-01

    The proposed action is to transport approximately 1100 spent fuel rods from a foreign research reactor in Taiwan by sea to Hampton Roads, Virginia, and then overland by truck to the receiving basin for offsite fuels at the Savannah River Plant (SRP) for reprocessing to recover uranium and plutonium. The analysis of the impacts of the proposed action have been evaluated and shown to have negligible impact on the local environments. The calculations have been completed using the RADTRAN III code. PWR spent fuel was analyzed as a benchmark to link the calculations in this analysis to those in earlier environmental documentation. Cumulative total, maximum annual, and per shipment risks were calculated. The results indicate that the PWR spent fuel shipment risks are somewhat lower than those previously estimated. The cumulative and maximum annual normal, or incident-free, risks associated with the shipment of Taiwanese research reactor spent fuel is a factor of 10 lower than that for PWR fuel, and the cumulative and maximum annual accident radiological risks are a factor of about 2.2 lower than that for PWR spent fuel. As a result, the port risks are about a factor of 10 larger than the risk of overland transport. All of the risks calculated are small. The PWR risk values are similar to those judged by the NRC to be small enough not to warrant increased stringency in regulations. The Taiwanese research reactor spent fuel shipment risk values are smaller yet. 51 refs., 22 tabs

  3. Replacement of the Pumps for Fuel Channel Cooling Circuit of the Maria Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Krzysztoszek, G.; Mieleszczenko, W.; Moldysz, A. [National Centre for Nuclear Research, Otwock–Świerk (Poland)

    2014-08-15

    The high flux Maria research reactor is operated by the National Centre for Nuclear Research in Świerk. It is a pool type reactor with pressurized fuel channels located in the beryllium matrix. According to the Global Threat Reduction Initiative programme our goal is to convert the Maria reactor from HEU to LEU fuel. Hydraulic losses in the new LEU fuel produced by CERCA are about 30% higher than the existing HEU fuel of type MR-6. For the MR-6 fuel were installed four two speed pumps. These pumps performed the function of the main circulations pumps during reactor operation with residual pumping power provided by emergency pumps. In the new system four main pumps will be used for circulating coolant while the reactor is operation with three auxiliary pumps for decay heat removal after reactor shutdown, meaning that the conversion of Maria research reactor will be possible after increasing flow in the primary cooling circuit of the fuel channels. The technical design of replacement of the pumps in the primary fuel channel cooling circuit was finished in April 2011 and accepted by the Safety Committee. After delivery of the new pumps we are planning to upgrade the primary fuel channel cooling circuit during October–November 2012. (author)

  4. Design, Fabrication, and Testing of a Laboratory-Scale Voloxidation System for Removal of Tritium and Other Volatile Fission Products from Used Nuclear Fuel

    International Nuclear Information System (INIS)

    Spencer, Barry B; DelCul, Guillermo D; Bradley, Eric Craig; Jubin, Robert Thomas; Hylton, Tom D; Collins, Emory D

    2008-01-01

    Advanced nuclear fuel processing methodologies are being demonstrated at the Oak Ridge National Laboratory (ORNL) as part of the Global Nuclear Energy Partnership (GNEP) program. A coupled end-to-end (CETE) research and development (R and D) capability is being installed to provide all primary processing operations, ranging from spent fuel receipt to production of products and waste forms. This R and D capability is designed for small, laboratory-scale throughput and will permit conduct of experiments in the range of 20 kg of spent fuel per year. The head-end processing segment includes single-pin shearing, voloxidation to remove tritium from the fuel before it enters the aqueous based separations systems, cleanup of the cladding hulls for disposition, and transfer of the fuel powder to the dissolution process. This paper describes the voloxidation system design and presents results from the cold checkout of the hardware. Preliminary results of the initial processing campaign with spent fuel is presented as well

  5. Possible research program on a large scale nuclear pressure vessel

    International Nuclear Information System (INIS)

    1983-01-01

    The nuclear pressure vessel structural integrity is actually one of the main items in the nuclear plants safety field. An international study group aimed at investigating the feasibility of a ''possible research program'' on a scale 1:1 LWR pressure vessel. This report presents the study group's work. The different research programs carried out or being carried out in various countries of the European Community are presented (phase I of the study). The main characteristics of the vessel considered for the program and an evaluation of activities required for making them available are listed. Research topic priorities from the different interested countries are summarized in tables (phase 2); a critical review by the study group of the topic is presented. Then, proposals for possible experimental programs and combination of these programs are presented, only as examples of possible useful research activities. The documents pertaining to the results of phase I inquiry performed by the study group are reported in the appendix

  6. Criticality evaluation of long term for spent fuel, using Scale; Evaluacion de criticidad a largo plazo para combustible gastado, utilizando SCALE

    Energy Technology Data Exchange (ETDEWEB)

    Esquivel E, J.; Vargas E, S.; Ramirez S, J. R., E-mail: jaime.esquivel@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    Once carried out the spent fuel discharge, of the reactor core, this continues generating decay heat and diverse fission products, reason why is important to store this fuel inside containers able to dissipate the heat generated by the isotopes decay of the fuel and to maintain the fuels arrangement in subcritical condition. This means that: is necessary to assure the sub-criticality of those fuel assemblies in the time. This work, presents a criticality evaluation of fuel assemblies type PWR in a storage generic container. For this purpose have been used two codes: GeeWiz, to carry out the geometric model of the container with the fuel assemblies, and Keno, with which, the criticality of the full container with fuel is determined until a 10{sup 6} years period. These codes are part of the package Scale. The specifications for each one of the analyzed components are based on a Benchmark document of the Nea/OECD, of where, the results that reports are compared with the obtained results by the realized analysis. (Author)

  7. Fuel cycle safety research. Role and development in future

    International Nuclear Information System (INIS)

    Matsumoto, Shiro

    2005-01-01

    The report stresses important roles played by risk informed methodology in chemical process safety problems as encountered in fuel cycle facility such as a reprocessing plant. Abnormal situation management is a safety issue - The lesson learned from the pesticide manufacturing facility in Bhopal, India. Little attention has been given to understanding the issues regarding performance during normal versus abnormal situation. The first step is in abnormal situation management is to define what really is abnormal. The second step is to ensure that everyone understands the difference between normal and abnormal, and the root causes of abnormal events. The third step is to be aware of current practices that support abnormal situation management, and the procedures, practices, and techniques used to respond to abnormal conditions. Dynamic simulation will become to useful techniques for abnormal situation management as well as probabilistic safety assessment for process facilities including nuclear fuel cycle. (S. Ohno)

  8. Fossil fuel power plant combustion control: Research in Italy

    International Nuclear Information System (INIS)

    Pasini, S.; Trebbi, G.

    1991-01-01

    Electric power demand forecasts for Italy to the year 2000 indicate an increase of about 50% which, due to the current moratorium on nuclear energy, should be met entirely by fossil fuel power plants. Now, there is growing public concern about possible negative health impacts due to the air pollution produced through the combustion of fossil fuels. In response to these concerns, ENEL (Italian National Electricity Board) is investing heavily in air pollution abatement technology R ampersand D. The first phase involves the investigation of pollution mechanisms in order to develop suitable mathematical models and diagnostic techniques. The validity of the models is being tested through through measurements made by sophisticated instrumentation placed directly inside the combustion chambers of steam generator systems. These are allowing engineers to develop improved combustion control methods designed to reduce air pollution at source

  9. Numerical simulations of a full-scale polymer electrolyte fuel cell with analysing systematic performance in an automotive application

    International Nuclear Information System (INIS)

    Park, Heesung

    2015-01-01

    Highlights: • A 3-D full-scale fuel cell performance is numerically simulated. • Generated and consumed power in the system is affected by operating condition. • Systematic analysis predicts the net power of conceptual PEFC stack. - Abstract: In fuel cell powered electric vehicles, the net power efficiency is a critical factor in terms of fuel economy and commercialization. Although the fuel cell stack produces enough power to drive the vehicles, the transferred power to the power train could be significantly reduced due to the power consumption to operate the system components of air blower and cooling module. Thus the systematic analysis on the operating condition of the fuel cell stack is essential to predict the net power generation. In this paper numerical simulation is conducted to characterize the fuel cell performance under various operating conditions. Three dimensional and full-scale fuel cell of the active area of 355 cm 2 is numerically modelled with 47.3 million grids to capture the complexities of the fluid dynamics, heat transfer and electrochemical reactions. The proposed numerical model requires large computational time and cost, however, it can be powerful to reasonably predict the fuel cell system performance at the early stage of conceptual design without requiring prototypes. Based on the model, it has been shown that the net power is reduced down to 90% of the gross power due to the power consumption of air blower and cooling module

  10. Plans and status for accepting spent nuclear fuel from foreign research reactors

    International Nuclear Information System (INIS)

    Zeitoun, Abe; Williams, John; Brown, Keith; Chacey, Kenneth

    1996-01-01

    In May 1996, the Department of Energy, acting with the cooperation of the Department of State, announced adoption of a policy that will have a significant influence on international efforts to prevent the spread of nuclear weapons. This policy is concerned with the management of spent nuclear fuel from foreign research reactors. Spent nuclear fuel, unirradiated fuel, and target material accepted under the policy must contain uranium enriched in the U.S. Although such spent fuel will comprise a relatively small part of the Department of Energy's (the Department's) overall inventory of spent nuclear fuel, the policy invokes actions that provide a cornerstone of U.S. nonproliferation activities. Implementation of this policy is now underway. This paper describes the Department's implementation strategy with the emphasis on those actions that will affect foreign research reactor operators. (author)

  11. Planar Solid-Oxide Fuel Cell Research and Development

    Science.gov (United States)

    2013-03-28

    electrodes and the electrolyte. The effect of the reduction in concentrations can be seen from the well-known Nernst potential equation , given by...reactions is modeled as a jump in the electric potential, which is determined using Nernst potential ( equation (18)) and activation polarization ( equation ...derivatives of structural cost functions. 2. Solution Methodology 2.1 Governing Equations (Fuel Cell) The three-dimensional SOFC model [30,31] utilized in

  12. Research on nuclear energy in the fields of fuel cycle, PWR reactors and LMFBR reactors

    International Nuclear Information System (INIS)

    Barre, B.; Camarcat, N.

    1995-01-01

    In this article we present the CEA research programs to improve the safety of the next generation of reactors, to manage the Plutonium and the wastes of the fuel cycle end and to ameliorate the competitiveness. 6 refs

  13. BNFL's experience in the sea transport of irradiated research reactor fuel to the USA

    International Nuclear Information System (INIS)

    Hudson, I.A.; Porter, I.

    2000-01-01

    BNFL provides worldwide transport for a wide range of nuclear materials. BNFL Transport manages an unique fleet of vessels, designed, built, and operated to the highest safety standards, including the highest rating within the INF Code recommended by the International Maritime Organisation. The company has some 20 years of experience of transporting irradiated research reactor fuel in support of the United States' programme for returning US obligated fuel from around the world. Between 1977 and 1988 BNFL performed 11 shipments of irradiated research reactor fuel from the Japan Atomic Energy Research Institute to the US. Since 1997, a further 3 shipments have been performed as part of an ongoing programme for Japanese research reactor operators. Where possible, shipments of fuel from European countries such as Sweden and Spain have been combined with those from Japan for delivery to the US. (author)

  14. Experimental determination of fuel surface temperature in the Dalat Nuclear Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Khang, Ngo Phu; Huy, Ngo Quang; An, Tran Khac; Lam, Pham Van [Nuclear Research Inst., Da Lat (Viet Nam)

    1994-10-01

    Measured fuel surface temperatures, obtained at various locations of the core of the Dalat Nuclear Research Reactor under normal operating conditions, are presented, and some thermal characteristics of the reactor are discussed. (author). 2 refs., 11 figs., 2 tabs.

  15. Research and development of a phosphoric acid fuel cell/battery power source integrated in a test-bed bus. Final report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-05-30

    This project, the research and development of a phosphoric acid fuel cell/battery power source integrated into test-bed buses, began as a multi-phase U.S. Department of Energy (DOE) project in 1989. Phase I had a goal of developing two competing half-scale (25 kW) brassboard phosphoric acid fuel cell systems. An air-cooled and a liquid-cooled fuel cell system were developed and tested to verify the concept of using a fuel cell and a battery in a hybrid configuration wherein the fuel cell supplies the average power required for operating the vehicle and a battery supplies the `surge` or excess power required for acceleration and hill-climbing. Work done in Phase I determined that the liquid-cooled system offered higher efficiency.

  16. Fuel Testing for Sylvatex: Cooperative Research and Development Final Report, CRADA Number CRD-16-636

    Energy Technology Data Exchange (ETDEWEB)

    Burton, Jonathan L. [National Renewable Energy Lab. (NREL), Golden, CO (United States)

    2018-02-07

    Sylvatex is a green nano-chemistry company that has developed a platform technology utilizing renewable, non-toxic inputs to create a stable nanoparticle that can be used in multiple applications. Their mission is to increase the use of renewables globally, to empower a cleaner and healthier future. The main application is a fuel technology product - MicroX - that utilizes proprietary knowledge to scale low-cost, cleaner-burning renewable diesel fuel and additives by using a co-location commercial model. The aspects of this project will include testing of two Sylvatex MicroX fuels on an engine dynamometer platform. Industry standard ultra-low sulfur diesel (ULSD) B3 fuel and a ULSD B20 will both be used for comparison of the Sylvatex fuels (U.S. standard diesel fuel at the pump contains an average of approximately 3% biodiesel; this is why B3 would be used as a baseline comparison). Sylvatex is currently using a prototype formulation (MicroX 1) that applies a high cost surfactant. An experimental formulation (MicroX 2) that uses lower cost materials is under development. The MicroX 1 will be blended at a 10% level into the B3 ULSD fuel and the MicroX 2 will be blended at a 10% level into both the B3 and the B20 ULSD fuels for study on the engine dynamometer test platform. All fuel blends will be tested over the FTP transient engine test cycle and a steady state ramped modal engine test cycle. Each test cycle will be performed a minimum of 3 times for each fuel. Tailpipe and/or engine out gaseous exhaust emissions (CO2, CO, NOx, THC, O2,), engine out PM emissions, and brake-specific fuel consumption rates will be evaluated for all test cycles.

  17. Reduction of fuel enrichment for research reactors built-up in accordance with Russian (Soviet) projects

    International Nuclear Information System (INIS)

    Aleksandrov, A.B.; Enin, A.A.; Tkachyov, A.A.

    2001-01-01

    In accordance with the Russian program of reduced enrichment for research and test reactors (RERTR) built-up in accordance with Russian (Soviet) projects, AO 'NCCP' performs works on FA fabrication with reduced enrichment fuel. The main trends and results of performed works on research reactors FEs and FAs based on UO 2 and U-9%Mo fuel with U 235 19.7% enrichment are described. (author)

  18. Transportation of failed or damaged foreign research reactor spent nuclear fuel

    International Nuclear Information System (INIS)

    Messick, C.E.; Mustin, T.P.; Massey, C.D.

    1998-01-01

    Since resuming the Foreign Research Reactor Spent Nuclear Fuel (FRR SNF) Acceptance Program in 1996, the Program has had to deal with difficult issues associated with the transportation of failed or damaged spent fuel. In several instances, problems with failed or damaged fuel have prevented the acceptance of the fuel at considerable cost to both the Department of Energy (DOE) and research reactor operators. In response to the problems faced by the Acceptance Program, DOE has undertaken significant steps to better define the spent fuel acceptance criteria. DOE has worked closely with the U.S. Nuclear Regulatory Commission to address failed or damaged research reactor spent fuel and to identify cask certificate issues which must be resolved by cask owners and foreign regulatory authorities. The specific issues associated with the transport of Materials Testing Reactor (MTR)-type FRR SNF will be discussed. The information presented will include U.S. Nuclear Regulatory Commission regulatory issues, cask certificate issues, technical constraints, and lessons learned. Specific information will also be provided on the latest efforts to revise DOE's Appendix B, Transport Package (Cask) Acceptance Criteria. The information presented in this paper will be important to foreign research reactor operators, shippers, and cask vendors, so that appropriate amendments to the Certificate of Compliance for spent fuel casks can be submitted in a timely manner to facilitate the safe and scheduled transport of FRR SNF

  19. Research report on development of spacer grid strap for AFA 3G fuel assembly

    International Nuclear Information System (INIS)

    Ye Yuandong

    2004-11-01

    The current development and tendency for fuel assemblies being of low leakage, high burn-up and long cycle fuel reload in the world are presented, and the necessity and feasibility to develop the spacer grid for high burn-up fuel assembly are elaborated. Considering all the activities in implementing of spacer grid and the technical difficulties in machining of tools, the major technological processes are introduced; the research program and the approaches to develop the spacer grid while research targets and overall schedule are defined and some key technical points and applicable practices are discussed. Finally the requirements and the conditions necessary for developing of spacer grid are proposed. (authors)

  20. Management and storage of spent research reactor fuel within the Federal Republic of Germany

    International Nuclear Information System (INIS)

    Krull, W.

    1996-01-01

    Research reactors in the Federal Republic of Germany and their needs for the interim storage of spent nuclear fuel are described. Existing long-term interim storage facilities are described. Special licensing and legal restrictions imposed by the German Atomic Energy Act are outlined. Possible final solutions for the back end of the nuclear fuel cycle for research reactors, including reprocessing in the United Kingdom or France, return of US-origin fuel and a home-grown German solution are discussed. (author). 2 refs, 5 figs, 4 tabs

  1. Hydrogen as a fuel for today and tomorrow: expectations for advanced hydrogen storage materials/systems research.

    Science.gov (United States)

    Hirose, Katsuhiko

    2011-01-01

    History shows that the evolution of vehicles is promoted by several environmental restraints very similar to the evolution of life. The latest environmental strain is sustainability. Transport vehicles are now facing again the need to advance to use sustainable fuels such as hydrogen. Hydrogen fuel cell vehicles are being prepared for commercialization in 2015. Despite intensive research by the world's scientists and engineers and recent advances in our understanding of hydrogen behavior in materials, the only engineering phase technology which will be available for 2015 is high pressure storage. Thus industry has decided to implement the high pressure tank storage system. However the necessity of smart hydrogen storage is not decreasing but rather increasing because high market penetration of hydrogen fuel cell vehicles is expected from around 2025 onward. In order to bring more vehicles onto the market, cheaper and more compact hydrogen storage is inevitable. The year 2025 seems a long way away but considering the field tests and large scale preparation required, there is little time available for research. Finding smart materials within the next 5 years is very important to the success of fuel cells towards a low carbon sustainable world.

  2. Summary of NRC LWR safety research programs on fuel behavior, metallurgy/materials and operational safety

    International Nuclear Information System (INIS)

    Bennett, G.L.

    1979-09-01

    The NRC light-water reactor safety-research program is part of the NRC regulatory program for ensuring the safety of nuclear power plants. This paper summarizes the results of NRC-sponsored research into fuel behavior, metallurgy and materials, and operational safety. The fuel behavior research program provides a detailed understanding of the response of nuclear fuel assemblies to postulated off-normal or accident conditions. Fuel behavior research includes studies of basic fuel rod properties, in-reactor tests, computer code development, fission product release and fuel meltdown. The metallurgy and materials research program provides independent confirmation of the safe design of reactor vessels and piping. This program includes studies on fracture mechanics, irradiation embrittlement, stress corrosion, crack growth, and nondestructive examination. The operational safety research provides direct assistance to NRC officials concerned with the operational and operational-safety aspects of nuclear power plants. The topics currently being addressed include qualification testing evaluation, fire protection, human factors, and noise diagnostics

  3. Mass and charge transfer on various relevant scales in polymer electrolyte fuel cells[Dissertation 16991

    Energy Technology Data Exchange (ETDEWEB)

    Freunberger, S. A.

    2007-07-01

    This dissertation is concerned with the development, experimental diagnostics and mathematical modelling and simulation of polymer electrolyte fuel cells (PEFC). The central themes throughout this thesis are the closely interlinked phenomena of mass and charge transfer. In the face of developing a PEFC system for vehicle propulsion these phenomena are scrutinized on a broad range of relevant scales. Starting from the material related level of the membrane and the gas diffusion layer (GDL) we turn to length scales, where structural features of the cell additionally come into play. These are the scale of flow channels and ribs, the single cell and the cell stack followed by the cell, stack, and system development for an automotive power train. In Chapter 3 selected fundamental material models and properties, respectively, are explored that are crucial for the mathematical modelling and simulation of PEFC, as needed in some succeeding parts of this work. First, established mathematical models for mass and charge transfer in the membrane are compared within the framework of the membrane electrode assembly (MEA), which represents the electrochemical unit. Second, reliable values for effective diffusivities in the GDLs which are vital for the simulation of gaseous mass transport are measured. Therefore, a method is developed that allows measuring this quantity both as a function of compression and direction as this is a prerequisite of sophisticated more-dimensional numerical PEFC-models. Besides the cross section of the catalyst layer (CL) mass transfer under channels and ribs is considered as a major source of losses in particular under high load operation. As up to now there have been solely non-validated theoretical investigations, in Chapter 4 an experimental method is developed that is for the first time capable of resolving the current density distribution on the this scale. For this, the electron conductors in the cell are considered as 2-dimensional shunt

  4. Nuclear mass inventory, photon dose rate and thermal decay heat of spent research reactor fuel assemblies

    International Nuclear Information System (INIS)

    Pond, R.B.; Matos, J.E.

    1996-05-01

    As part of the Department of Energy's spent nuclear fuel acceptance criteria, the mass of uranium and transuranic elements in spent research reactor fuel must be specified. These data are, however, not always known or readily determined. It is the purpose of this report to provide estimates of these data for some of the more common research reactor fuel assembly types. The specific types considered here are MTR, TRIGA and DIDO fuel assemblies. The degree of physical protection given to spent fuel assemblies is largely dependent upon the photon dose rate of the spent fuel material. These data also, are not always known or readily determined. Because of a self-protecting dose rate level of radiation (dose rate greater than 100 ren-x/h at I m in air), it is important to know the dose rate of spent fuel assemblies at all time. Estimates of the photon dose rate for spent MTR, TRIGA and DIDO-type fuel assemblies are given in this report

  5. Advanced Automotive Fuels Research, Development, and Commercialization Cluster (OH)

    Energy Technology Data Exchange (ETDEWEB)

    Linkous, Clovis; Hripko, Michael; Abraham, Martin; Balendiran, Ganesaratnam; Hunter, Allen; Lovelace-Cameron, Sherri; Mette, Howard; Price, Douglas; Walker, Gary; Wang, Ruigang

    2013-08-31

    Technical aspects of producing alternative fuels that may eventually supplement or replace conventional the petroleum-derived fuels that are presently used in vehicular transportation have been investigated. The work was centered around three projects: 1) deriving butanol as a fuel additive from bacterial action on sugars produced from decomposition of aqueous suspensions of wood cellulose under elevated temperature and pressure; 2) using highly ordered, openly structured molecules known as metal-organic framework (MOF) compounds as adsorbents for gas separations in fuel processing operations; and 3) developing a photocatalytic membrane for solar-driven water decomposition to generate pure hydrogen fuel. Several departments within the STEM College at YSU contributed to the effort: Chemistry, Biology, and Chemical Engineering. In the butanol project, sawdust was blended with water at variable pH and temperature (150 – 250{degrees}C), and heated inside a pressure vessel for specified periods of time. Analysis of the extracts showed a wide variety of compounds, including simple sugars that bacteria are known to thrive upon. Samples of the cellulose hydrolysate were fed to colonies of Clostridium beijerinckii, which are known to convert sugars to a mixture of compounds, principally butanol. While the bacteria were active toward additions of pure sugar solutions, the cellulose extract appeared to inhibit butanol production, and furthermore encouraged the Clostridium to become dormant. Proteomic analysis showed that the bacteria had changed their genetic code to where it was becoming sporulated, i.e., the bacteria were trying to go dormant. This finding may be an opportunity, as it may be possible to genetically engineer bacteria that resist the butanol-driven triggering mechanism to stop further fuel production. Another way of handling the cellulosic hydrolysates was to simply add the enzymes responsible for butanol synthesis to the hydrolytic extract ex-vivo. These

  6. 2010 Thin Film & Small Scale Mechanical Behavior Gordon Research Conference

    Energy Technology Data Exchange (ETDEWEB)

    Dr. Thomas Balk

    2010-07-30

    Over the past decades, it has been well established that the mechanical behavior of materials changes when they are confined geometrically at least in one dimension to small scale. It is the aim of the 2010 Gordon Conference on 'Thin Film and Small Scale Mechanical Behavior' to discuss cutting-edge research on elastic, plastic and time-dependent deformation as well as degradation mechanisms like fracture, fatigue and wear at small scales. As in the past, the conference will benefit from contributions from fundamental studies of physical mechanisms linked to material science and engineering reaching towards application in modern applications ranging from optical and microelectronic devices and nano- or micro-electrical mechanical systems to devices for energy production and storage. The conference will feature entirely new testing methodologies and in situ measurements as well as recent progress in atomistic and micromechanical modeling. Particularly, emerging topics in the area of energy conversion and storage, such as material for batteries will be highlighted. The study of small-scale mechanical phenomena in systems related to energy production, conversion or storage offer an enticing opportunity to materials scientists, who can provide new insight and investigate these phenomena with methods that have not previously been exploited.

  7. Distributed system for large-scale remote research

    International Nuclear Information System (INIS)

    Ueshima, Yutaka

    2002-01-01

    In advanced photon research, large-scale simulations and high-resolution observations are powerfull tools. In numerical and real experiments, the real-time visualization and steering system is considered as a hopeful method of data analysis. This approach is valid in the typical analysis at one time or low cost experiment and simulation. In research of an unknown problem, it is necessary that the output data be analyzed many times because conclusive analysis is difficult at one time. Consequently, output data should be filed to refer and analyze at any time. To support research, we need the automatic functions, transporting data files from data generator to data storage, analyzing data, tracking history of data handling, and so on. The supporting system will be a functionally distributed system. (author)

  8. International topical meeting on research reactor fuel management (RRFM) - United States foreign research reactor (FRR) spent nuclear fuel (SNF) acceptance program: 2010 update

    International Nuclear Information System (INIS)

    Messick, C.E.; Taylor, J.L.; Niehus, M.T.; Landers, C.

    2010-01-01

    The Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel, adopted by the United States Department of Energy (DOE), in consultation with the Department of State (DOS) in May 1996, scheduled to expire May 12, 2016, to return research reactor fuel until May 12, 2019 to the U.S. is in its fourteenth year. This paper provides a brief update on the program, part of the National Nuclear Security Administration (NNSA), and discusses program initiatives and future activities. The goal of the program continues to be recovery of U.S.-origin nuclear materials, which could otherwise be used in weapons, while assisting other countries to enjoy the benefits of nuclear technology. The NNSA is seeking feedback from research reactor operators to help us understand ways to include eligible research reactors who have not yet participated in the program. (author)

  9. Available Reprocessing and Recycling Services for Research Reactor Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    2017-01-01

    The high enriched uranium (HEU) take back programmes will soon have achieved their goals. When there are no longer HEU inventories at research reactors and no commerce in HEU for research reactors, the primary driver for the take back programmes will cease. However, research reactors will continue to operate in order to meet their various mission objectives. As a result, inventories of low enriched uranium spent nuclear fuel will continue to be created during the research reactors' lifetime and, therefore, there is a need to develop national final disposition routes. This publication is designed to address the issues of available reprocessing and recycling services for research reactor spent fuel and discusses the various back end management aspects of the research reactor fuel cycle.

  10. A summary report on recruitment type researches on nuclear fuel cycle in fiscal year of 2001

    International Nuclear Information System (INIS)

    2002-07-01

    The promotion system on recruitment type researches on nuclear fuel cycle begun on fiscal year of 1999, aims to intend to activate researching environment of the Japan Nuclear Cycle Development Institute (JNC) through intercourses, information exchanges, publication of research results, and so on between researchers in other organizations and JNC, as a result, to effectively promote fundamental and basic R and Ds. This report contains summaries of 28 items of research results on the recruitment type researches on nuclear fuel cycle as 9 items relating to fast breeder reactors, 8 items relating to nuclear fuel cycle, 1 item relating to radiation safety, and 10 items relating to geological disposal and science, carried out on fiscal year of 2001. (G.K.)

  11. Alternative Fuels Data Center: Electricity Research and Development

    Science.gov (United States)

    wind, respectively. PV panels convert sunlight directly to electricity using semiconductors. The , small-scale renewables, like solar panels on a rooftop, can both provide clean energy for vehicles and

  12. Status of the DOE's foreign research reactor spent nuclear fuel acceptance program

    International Nuclear Information System (INIS)

    Chacey, K.; Saris, E.C.

    1997-01-01

    In May 1996, the U.S. Department of Energy (DOE), in consultation with the U.S. Department of State (DOS), adopted a policy to accept and manage in the United States ∼20 tonnes of spent nuclear fuel from research reactors in up to 41 countries. This spent fuel is being accepted under the nuclear weapons non-proliferation policy concerning foreign research reactor spent nuclear fuel. Only spent fuel containing uranium enriched in the United States is covered under this policy. Implementing this policy is a top priority of the DOE. The purpose of this paper is to provide the current status of the foreign research reactor acceptance program, including achievements to date and future challenges

  13. Neutronics conceptual design of the innovative research reactor core using uranium molybdenum fuel

    International Nuclear Information System (INIS)

    Tukiran S; Surian Pinem; Tagor MS; Lily S; Jati Susilo

    2012-01-01

    The multipurpose of research reactor utilization make many countries build the new research reactor. Trend of this reactor for this moment is multipurpose reactor type with a compact core to get high neutron flux at the low or medium level of power. The research newest. Reactor in Indonesia right now is already 25 year old. Therefore, it is needed to design a new research reactor, called innovative research reactor (IRR) and then as an alternative to replace the old research reactor. The aim of this research is to get the optimal configuration of equilibrium core with the acceptance criteria are minimum thermal neutron flux is 2.5E14 n/cm 2 s at the power level of 20 MW (minimum), length of cycle of more than 40 days, and the most efficient of using fuel in the core. Neutronics design has been performed for new fuel of U-9Mo-AI with various fuel density and reflector. Design calculation has been performed using WIMSD-5B and BATAN-FUEL computer codes. The calculation result of the conceptual design shows four core configurations namely 5x5, 5x7, 6x5 and 6x6. The optimalization result for equilibrium core of innovative research reactor is the 5x5 configuration with 450 gU fuel loading, berilium reflector, maximum thermal neutron flux at reflector is 3.33E14 n/cm 2 sand length of cycle is 57 days is the most optimal of IRR. (author)

  14. Small watershed-scale research and the challenges ahead

    Science.gov (United States)

    Larsen, M. C.; Glynn, P. D.

    2008-12-01

    For the past century, Federal mission science agencies (eg. USFS, NRCS, ARS, USGS) have had the long- term agency goals, infrastructure, and research staff to conduct research and data collection in small watersheds as well as support these activities for non-Federal partners. The National Science Foundation has been a strong partner with the Federal mission science agencies, through the LTER network, which is dependent on Federally supported research sites, and more recently with the emerging CUAHSI, WATERS, CZEN, and NEON initiatives. Much of the NSF-supported research builds on the foundations provided by their Federally supported partners, who sustain the long-term, extensive monitoring activity and research sites, including making long-term data available to all users via public interfaces. The future of these programs, and their enhancement/expansion to face the intensifying concurrent challenges of population growth, land-use change, and climate change, is dependent on a well-funded national commitment to basic science. Such a commitment will allow the scientific community to advance our understanding of these scientific challenges and to synthesize our understanding among research sites and at the national scale. Small watersheds serve as essential platforms where hypotheses can be tested, as sentinels for climate change, and as a basis for comparing and scaling up local information and syntheses to regional and continental scales. The science guides resource management and mitigation decisions and is fundamental to the development of predictive models. Furthermore, small-watershed research and monitoring programs are generally undervalued because many research questions that can be addressed now or in the future were not anticipated when the sites were initiated. Some examples include: 1) the quantification, characterization, and understanding of how emerging contaminants, personal care products, and endocrine disruptors affect organisms - substances that

  15. Methods of Scientific Research: Teaching Scientific Creativity at Scale

    Science.gov (United States)

    Robbins, Dennis; Ford, K. E. Saavik

    2016-01-01

    We present a scaling-up plan for AstroComNYC's Methods of Scientific Research (MSR), a course designed to improve undergraduate students' understanding of science practices. The course format and goals, notably the open-ended, hands-on, investigative nature of the curriculum are reviewed. We discuss how the course's interactive pedagogical techniques empower students to learn creativity within the context of experimental design and control of variables thinking. To date the course has been offered to a limited numbers of students in specific programs. The goals of broadly implementing MSR is to reach more students and early in their education—with the specific purpose of supporting and improving retention of students pursuing STEM careers. However, we also discuss challenges in preserving the effectiveness of the teaching and learning experience at scale.

  16. Studies on capacity management for factories of nuclear fuel for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Negro, Miguel Luiz Miotto; Durazzo, Michelangelo; Mesquita, Marco Aurélio de; Carvalho, Elita Fontenele Urano de; Andrade, Delvonei Alves de, E-mail: mlnegro@ipen.br, E-mail: mdurazzo@ipen.br, E-mail: elitaucf@ipen.br, E-mail: delvonei@ipen.br, E-mail: mamesqui@usp.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Universidade de São Paulo (USP), SP (Brazil). Escola Politécnica. Departamento de Engenharia de Produção

    2017-11-01

    The use and the power of nuclear reactors for research and materials testing is increasing worldwide. That implies the demand for nuclear fuel for this kind of reactors is rising. Thus, the production facilities of this kind of fuel need reliable guidance on how to augment their production in order to meet the increasing demand efficiently, safely and keeping good quality. Focus is given to factories that produce plate type fuel elements loaded with LEU U{sub 3}Si{sub 2}-Al fuel, which are typically used in nuclear research reactors. Of the various production routes for this kind of fuel, we chose the route which uses hydrolysis of uranium hexafluoride. Raising the capacity of this kind of plants faces several problems, especially regarding safety against nuclear criticality. Some of these problems are briefly addressed. The new issue of the paper is the application of knowledge from the area of production administration to the fabrication of nuclear fuel for research reactors. A specific method for the increase in production capacity is proposed. That method was tested by means of discrete event simulation. The data were collected from the nuclear fuel factory at IPEN. The results indicated the proposed method achieved its goal as well as ways of raising production capacity in up to 50%. (author). (author)

  17. Studies on capacity management for factories of nuclear fuel for research reactors

    International Nuclear Information System (INIS)

    Negro, Miguel Luiz Miotto; Durazzo, Michelangelo; Mesquita, Marco Aurélio de; Carvalho, Elita Fontenele Urano de; Andrade, Delvonei Alves de; Universidade de São Paulo

    2017-01-01

    The use and the power of nuclear reactors for research and materials testing is increasing worldwide. That implies the demand for nuclear fuel for this kind of reactors is rising. Thus, the production facilities of this kind of fuel need reliable guidance on how to augment their production in order to meet the increasing demand efficiently, safely and keeping good quality. Focus is given to factories that produce plate type fuel elements loaded with LEU U_3Si_2-Al fuel, which are typically used in nuclear research reactors. Of the various production routes for this kind of fuel, we chose the route which uses hydrolysis of uranium hexafluoride. Raising the capacity of this kind of plants faces several problems, especially regarding safety against nuclear criticality. Some of these problems are briefly addressed. The new issue of the paper is the application of knowledge from the area of production administration to the fabrication of nuclear fuel for research reactors. A specific method for the increase in production capacity is proposed. That method was tested by means of discrete event simulation. The data were collected from the nuclear fuel factory at IPEN. The results indicated the proposed method achieved its goal as well as ways of raising production capacity in up to 50%. (author). (author)

  18. Brief report of the JNC Cooperative Research (A) on the Nuclear Fuel Cycle for 2004

    International Nuclear Information System (INIS)

    2005-06-01

    The Japan Nuclear Cycle Development Institute (JNC) started the JNC Cooperative Research Scheme (A) on the Nuclear Fuel Cycle in 1995 in order to promote research collaboration with universities and other research institutes (referred to as 'universities, etc.' below), which means that the fiscal year 2004 was the tenth year of the scheme. The purpose of this scheme is to promote the basic and fundamental research that precedes the research and development projects of JNC through collaboration with universities, etc. by using mainly JNC's facilities and equipment. Under the scheme, universities, etc. propose methods and ideas, etc. to lead to the achievement of the goals of research collaboration themes which are set by JNC as research collaboration subjects. Then a screening committee consisting mainly of experts independent of JNC screens the research collaboration subjects. Research collaboration is performed by carrying out cooperative research with universities, etc. or by accepting researchers from universities, etc. as Visiting Research Fellows at JNC. The scheme allows students studying for doctorates at postgraduate schools either to participate in the cooperative research or to be accepted as trainee researchers. This report includes a summary of the results of the research carried out in fiscal year 2004 on 42 research collaboration subjects for preceding basic engineering research related to fast breeder reactors, the nuclear fuel cycle, radiation safety and geological disposal/geoscience; seventeen of which ended in 2004. Of these, eight were related to fast breeder reactors, eight to the nuclear fuel cycle, and one to geological disposal/geoscience. (author)

  19. Brief report of the JNC Cooperative Research (C) on the Nuclear Fuel Cycle for 2004

    International Nuclear Information System (INIS)

    2005-06-01

    The Japan Nuclear Cycle Development Institute (JNC) started the JNC Cooperative Research Scheme (C) on the Nuclear Fuel Cycle in 1999 in order to promote research collaboration with universities and other research institutes (referred to as 'universities, etc.' below), which means that the fiscal year 2004 was the sixth year of the scheme and ended in 2004. This scheme is to invite foresighted and original themes of basic and fundamental research, to be performed using JNC's facilities and equipment, in principle, for the research and development of JNC in the field of the nuclear fuel cycle, from researchers at universities, etc. Those researchers who propose research themes in response to the invitation are expected to lead the research. The purpose of the scheme is to promote the efficiency of basic and fundamental research and development by enhancing the research environment of JNC through collaboration between JNC's own researchers and other researchers, and exchange of information and publication of achievements, etc. Research themes proposed by researchers at universities, etc. are screened by a screening committee consisting mainly of experts independent of JNC. This report includes a summary of the result of the research carried out in fiscal year 2004 on four selected themes related the JNC's collaborative research for fast breeder reactors, the nuclear fuel cycle, radiation safety and geological disposal/geoscience. (author)

  20. Brief report of the JNC cooperative research (C) on the nuclear fuel cycle for 2003

    International Nuclear Information System (INIS)

    2004-06-01

    The Japan Nuclear Cycle Development Institute (JNC) started the JNC Cooperative Research Scheme (C) on the Nuclear Fuel Cycle in 1999 in order to promote research collaboration with universities and other research institutes (referred to as 'universities, etc.' below), which means that the fiscal year 2003 was the fifth year of the scheme. This scheme is to invite foresighted and original themes of basic and fundamental research, to be performed using JNC's facilities and equipment, in principle, for the research and development of JNC in the field of the nuclear fuel cycle, from researchers at universities, etc. Those researchers who propose research themes in response to the invitation are expected to lead the research. The purpose of the scheme is to promote the efficiency of basic and fundamental research and development by enhancing the research environment of JNC through collaboration between JNC's own researchers and other researchers, and exchange of information and publication of achievements, etc. Research themes proposed by researchers at universities, etc. are screened by a screening committee consisting mainly of experts independent of JNC. This report includes a summary of the results of the research carried out in fiscal year 2003 on twelve selected themes related to the JNC's collaborative research for fast breeder reactors, the nuclear fuel cycle, radiation safety and geological disposal/geoscience; eight of which ended in 2003. Of these, four were related to fast breeder reactors, one to radiation safety and three to geological disposal/geoscience. (author)

  1. Research on weathering and biomarkers in heavy fuel oil

    International Nuclear Information System (INIS)

    Ma, Q.; Li, Z.; Yu, Z.

    2008-01-01

    The fate of oil spilled in the ocean depends on several physicochemical and biological factors such as evaporation, dissolution, microbial degradation and photo-oxidation. These weathering processes decrease the low molecules in spilled oils which reduces the harmful effects of spilled oil to the ocean and biota near the spill. In addition to changing the composition of the oil, some weathering processes are key to identifying the spilled oil. As such, the relationship between the weathering processes and the changes in oil composition must be well understood. This paper used gas chromatography and mass spectrometry (GC/MS) to analyze changes of chemical components in heavy fuel oil by weathering in static seawater. The major alkanes of heavy fuel oil include C8 to C33, while the major aromatics include benzene, naphthalene, phenanthrene and dibenzothiophene. After 24 weeks of weathering in seawater, the alkanes from n-C8 to n-C15 evaporated in order of increasing carbon number. The susceptibility of n-alkanes was correlated with carbon numbers. The aromatics evaporated in order of increasing carbon and ring number as weathering time increased. 8 refs., 3 tabs., 5 figs

  2. 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)

  3. 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)

  4. Assessment of the implementation issues for fuel cells in domestic and small scale stationary power generation and CHP applications

    Energy Technology Data Exchange (ETDEWEB)

    Graham, G.; Cruden, A.; Hart, J.

    2002-07-01

    This report discusses implementation issues associated with the use of fuel cells in <10 kW domestic, small-scale power generation and combined heat and power (CHP) operations in the UK. The report examines the key issues (fuel cell system standards and certification, fuel infrastructure, commercial issues and competing CHP technologies), before discussing non-technical issues including finance, ownership, import and export configuration, pricing structure, customer acceptability, installation, operation and training of servicing and commissioning personnel. The report goes on to discuss market and technical drivers, grid connection issues and solutions, operations and maintenance. Recommendations for the future are made.

  5. Bench-scale studies of reactor-based treatment of fuel-contaminated soils

    International Nuclear Information System (INIS)

    Truax, D.D.; Britto, R.; Sherrard, J.H.

    1995-01-01

    Biological treatment of hazardous wastes from accidental spills or underground storage tank leaks has generated interest in bioremediation as a natural, economical mechanism for site decontamination. Because of drawbacks of batch systems, and the successful use of continuous flow treatment of wastewater for several decades, it was felt that continuous treatment of such soils would be a feasible alternative treatment technique. Therefore, bench-scale bioreactor treatability studies were conducted and used contaminated soil made in the laboratory using No. 2 diesel fuel and sand. Contamination levels studied were from 1,335--6,675 mg (TPH) as derived from No. 2 fuel oil per kg sand. Variation in mean cell age was obtained between reactors, with sufficient nutrients and oxygen made available to ensure the fuel oil organics were the only limit to microbial growth. A theoretical biokinetic model was formulated based on Monod's theory of limiting substrate and continuous cultures. Biokinetic constants and removal efficiencies were evaluated. The off-gases, CO 2 , and volatile hydrocarbons were monitored for mass balance analysis of the process. The solids retention times for evaluating final TPH concentration of 100 mg/kg were also calculated. Removal efficiencies of up to 91% were attained at a loading of 1,335 mg TPH/kg wet sand, operated at a biological solid retention time (BSRT) of 60 days. Experiments also showed that TPH desorption and volatilization were not rate-limiting in the overall removal process. Sand-to-moisture ratios in excess of 3:1 were also shown to retard TPH removal rates very little. However, biokinetic constants were found to vary over a range of values. This was particularly true at varying diesel loading levels. Nevertheless, significant removal efficiency (up to 86%) was noted at the highest loading level tested, 6,675 mg TPH/kg wet sand

  6. Corrosion and protection of spent Al-clad research reactor fuel during extended wet storage

    International Nuclear Information System (INIS)

    Ramanathan, Lalgudi V.

    2009-01-01

    A variety of spent research reactor fuel elements with different fuel meats, geometries and 235 U enrichments are presently stored under water in basins throughout the world. More than 90% of these fuels are clad in aluminum (Al) or its alloy and are susceptible to corrosion. This paper presents an overview of the influence of Al alloy composition, galvanic effects (Al alloy/stainless steel), crevice effects, water parameters and synergism between these parameters as well as settled solids on the corrosion of typical Al alloys used as fuel element cladding. Pitting is the main form of corrosion and is affected by water conductivity, chloride ion content, formation of galvanic couples with rack supports and settled solid particles. The extent to which these parameters influence Al corrosion varies. This paper also presents potential conversion coatings to protect the spent fuel cladding. (author)

  7. Plan of studies on fuel failure detection in Rikkyo Research Reactor

    International Nuclear Information System (INIS)

    Matsuura, T.; Nagahara, T.; Hattori, M.; Kawaguchi, K.

    1987-01-01

    Studies on fuel failure detection in Rikkyo Research Reactor have recently been begun in the following four approaches. (1) Accumulation of the data on the concentration of the short-lived radioactivity originating from FP rare gases contained in the air on the water surface of the reactor tank. (2) Accumulation of the data on the concentration of FP (especially 131 I) in the water of the reactor tank. (3) Design and preparation of a ''sniffer'' by which the location of the failed fuel element can be detected, when some anomaly is found in the above two routine measurements. (4) Design and preparation of a vessel containing a fuel element, which can be useful both for ''sipping'' inspection of the fuel element and for storage of the damaged fuel element. In this paper, an outline of the above approaches and the results of some preliminary experiments are reported. (author)

  8. A study for the research trends of membranes for proton exchange membrane fuel cells

    International Nuclear Information System (INIS)

    Sener, T.

    2004-01-01

    'Full text:' A single PEM fuel cell is comprised of a membrane electrode assembly, two bipolar plates and two fields. Membrane electrode assembly is the basic component of PEM fuel cell due to its cost and function, and it consists a membrane sandwiched between two electrocatalyst layers/electrodes and two gas diffusion layers. Increasing the PEM fuel cell operation temperature from 80 o C to 150-200 o C will prevent electrocatalysts CO poisoning and increase the fuel cell performance. Therefore, membranes must have chemical and mechanical resistance and must keep enough water at high temperatures. The aim of membrane studies through fuel cell commercialization is to produce a less expensive thin membrane with high operation temperature, chemical and mechanical resistance and water adsorption capacity. Within this frame, alternative membrane materials, membrane electrode assembly manufacture and evaluation methods are being studied. In this paper, recent studies are reviewed to give a conclusion for research trends. (author)

  9. Pore-scale investigation of mass transport and electrochemistry in a solid oxide fuel cell anode

    Energy Technology Data Exchange (ETDEWEB)

    Grew, Kyle N.; Joshi, Abhijit S.; Peracchio, Aldo A.; Chiu, Wilson K.S. [Department of Mechanical Engineering, University of Connecticut, 191 Auditorium Road, Storrs, CT 06269-3139 (United States)

    2010-04-15

    The development and validation of a model for the study of pore-scale transport phenomena and electrochemistry in a Solid Oxide Fuel Cell (SOFC) anode are presented in this work. This model couples mass transport processes with a detailed reaction mechanism, which is used to model the electrochemical oxidation kinetics. Detailed electrochemical oxidation reaction kinetics, which is known to occur in the vicinity of the three-phase boundary (TPB) interfaces, is discretely considered in this work. The TPB regions connect percolating regions of electronic and ionic conducting phases of the anode, nickel (Ni) and yttria-stabilized zirconia (YSZ), respectively; with porous regions supporting mass transport of the fuel and product. A two-dimensional (2D), multi-species lattice Boltzmann method (LBM) is used to describe the diffusion process in complex pore structures that are representative of the SOFC anode. This diffusion model is discretely coupled to a kinetic electrochemical oxidation mechanism using localized flux boundary conditions. The details of the oxidation kinetics are prescribed as a function of applied activation overpotential and the localized hydrogen and water mole fractions. This development effort is aimed at understanding the effects of the anode microstructure within TPB regions. This work describes the methods used so that future studies can consider the details of SOFC anode microstructure. (author)

  10. SCALE6.1 Hybrid Shielding Methodology For The Spent Fuel Dry Storage

    International Nuclear Information System (INIS)

    Matijevic, M.; Pevec, D.; Trontl, K.

    2015-01-01

    The SCALE6.1/MAVRIC hybrid deterministic-stochastic shielding methodology was used for dose rates calculation of the generic spent fuel dry storage installation. The neutron-gamma dose rates around the cask array were calculated over a large problem domain in order to determine the boundary of the controlled area. The FW-CADIS methodology, based on the deterministic forward and adjoint solution over the phase - space, was used for optimized, global Monte Carlo results over the mesh tally. The cask inventory was modeled as homogenized material corresponding to 20 fuel assemblies from a standard mid - sized PWR reactor. The global simulation model was an array of 32 casks in 2 rows with concrete foundations and external air, which makes a large spatial domain for shielding calculations. The dose rates around the casks were determined using FW-CADIS method with weighted adjoint source and mesh tally covering a portion of spatial domain of interest. The conservatively obtained dose rates give the upper boundary, since the activation reduction of sources was not taken into account when sequential filling of the dry storage will start. The effective area of the dry storage installation can be additionally reduced with lowering concrete foundation under the ground, embankment raising, and with extra concrete walls, that would additionally lower the dominant gamma dose rates. (author).

  11. Micro-electro-mechanical systems (MEMS)-based micro-scale direct methanol fuel cell development

    International Nuclear Information System (INIS)

    Yao, S.-C.; Tang Xudong; Hsieh, C.-C.; Alyousef, Yousef; Vladimer, Michael; Fedder, Gary K.; Amon, Cristina H.

    2006-01-01

    This paper describes a high-power density, silicon-based micro-scale direct methanol fuel cell (DMFC), under development at Carnegie Mellon. Major issues in the DMFC design include the water management and energy-efficient micro fluidic sub-systems. The air flow and the methanol circulation are both at a natural draft, while a passive liquid-gas separator removes CO 2 from the methanol chamber. An effective approach for maximizing the DMFC energy density, pumping the excess water back to the anode, is illustrated. The proposed DMFC contains several unique features: a silicon wafer with arrays of etched holes selectively coated with a non-wetting agent for collecting water at the cathode; a silicon membrane micro pump for pumping the collected water back to the anode; and a passive liquid-gas separator for CO 2 removal. All of these silicon-based components are fabricated using micro-electro-mechanical systems (MEMS)-based processes on the same silicon wafer, so that interconnections are eliminated, and integration efforts as well as post-fabrication costs are both minimized. The resulting fuel cell has an overall size of one cubic inch, produces a net output of 10 mW, and has an energy density three to five times higher than that of current lithium-ion batteries

  12. Fine scale vegetation classification and fuel load mapping for prescribed burning

    Science.gov (United States)

    Andrew D. Bailey; Robert Mickler

    2007-01-01

    Fire managers in the Coastal Plain of the Southeastern United States use prescribed burning as a tool to reduce fuel loads in a variety of vegetation types, many of which have elevated fuel loads due to a history of fire suppression. While standardized fuel models are useful in prescribed burn planning, those models do not quantify site-specific fuel loads that reflect...

  13. Brief of the joint research with universities, etc. for 2002. Except the research by the JNC cooperative research scheme on the nuclear fuel cycle

    International Nuclear Information System (INIS)

    2003-09-01

    The Japan Nuclear Cycle Development Institute (JNC) promote the basic and fundamental research in relation to the research and development projects of JNC through collaboration with universities and other research institutes (referred to as 'universities, etc.' below). This report includes a summary of the results of the research carried out in fiscal year 2002 on 15 joint researches with universities, etc. In addition, this report removes the research by the JNC Cooperative Research Scheme on the Nuclear Fuel Cycle. (author)

  14. Spanish collaboration in the OECD Halden Reactor Project research on Gadolinia Fuel

    International Nuclear Information System (INIS)

    Horvath, M.; Munoz-Reja, C.; Tverberg, T.; Jenssen, H. K.

    2010-01-01

    Safe and reliable operation of nuclear power plants benefit from research and development advances and related technical solutions. One research platform is the OECD Halden Reactor Project (HRP). HRP is a joint undertaking of national organisations in 18 countries sponsoring a jointly financed programme under the auspices of the OECD - Nuclear Energy Agency (NEA). As a member state, Spain is participating HRP research programs with ENUSA as a partner in the fuel research programs. Improving the NPP operations, fuel cycles were designed to increase fuel burnup. Higher fuel burnup reduces the number of spent fuel assemblies and thus the costs of new fuel as well as the costs of back-end management. Higher burnup is reached either by prolonging the reactor cycles or by increasing the number of reactor cycles for the fuel in the core. Both ways entail additional requirements concerning fuel enrichment and burnable absorbers as additives and adjustments on the cladding material properties, such as mechanical treatment and chemical composition of the alloys. For these demands and needs ENUSA promotes the research on high burnup effects, gadolinium doped fuels and cladding material behaviour under irradiation. Various experiments, called IFA, are developed and performed also by providing materials. ENUSA collaborates with HRP on various experiments investigating the fuel densification and swelling, fission gas release, pressure limits on UO 2 and (U,Gd)O 2 fuels (IFA-504, -515, -636, -681); the cladding creep, lift-off, corrosion and hydrides on different tubing materials (IFA-567, -610, -638); instrumentation of the experiments, especially on pre-irradiated materials (IFA-533). These experiments are combined with model calculations to improve predictions for higher burnups and to maintain safety margins (IFA-515, -636, -681). Besides these unique in-pile experiments PIEs are performed as well on fuel and structural materials to complete the scope of these studies (IFA

  15. Separation of Corn Fiber and Conversion to Fuels and Chemicals Phase II: Pilot-scale Operation

    Energy Technology Data Exchange (ETDEWEB)

    Abbas, Charles; Beery, Kyle; Orth, Rick; Zacher, Alan

    2007-09-28

    The purpose of the Department of Energy (DOE)-supported corn fiber conversion project, “Separation of Corn Fiber and Conversion to Fuels and Chemicals Phase II: Pilot-scale Operation” is to develop and demonstrate an integrated, economical process for the separation of corn fiber into its principal components to produce higher value-added fuel (ethanol and biodiesel), nutraceuticals (phytosterols), chemicals (polyols), and animal feed (corn fiber molasses). This project has successfully demonstrated the corn fiber conversion process on the pilot scale, and ensured that the process will integrate well into existing ADM corn wet-mills. This process involves hydrolyzing the corn fiber to solubilize 50% of the corn fiber as oligosaccharides and soluble protein. The solubilized fiber is removed and the remaining fiber residue is solvent extracted to remove the corn fiber oil, which contains valuable phytosterols. The extracted oil is refined to separate the phytosterols and the remaining oil is converted to biodiesel. The de-oiled fiber is enzymatically hydrolyzed and remixed with the soluble oligosaccharides in a fermentation vessel where it is fermented by a recombinant yeast, which is capable of fermenting the glucose and xylose to produce ethanol. The fermentation broth is distilled to remove the ethanol. The stillage is centrifuged to separate the yeast cell mass from the soluble components. The yeast cell mass is sold as a high-protein yeast cream and the remaining sugars in the stillage can be purified to produce a feedstock for catalytic conversion of the sugars to polyols (mainly ethylene glycol and propylene glycol) if desirable. The remaining materials from the purification step and any materials remaining after catalytic conversion are concentrated and sold as a corn fiber molasses. Additional high-value products are being investigated for the use of the corn fiber as a dietary fiber sources.

  16. Thermal stability and filterability of jet fuels containing PDR additives in small-scale tests and realistic rig simulations

    Energy Technology Data Exchange (ETDEWEB)

    Bauldreay, J.M.; Clark, R.H.; Heins, R.J. [Shell Research, Ltd., Chester (United Kingdom)

    1995-05-01

    Specification, small-scale and realistic fuel simulation tests have addressed concerns about the impact of pipeline drag reducer (PDR) flow modifying additives on jet fuel handling and performance. A typical PDR additive tended to block filters which were similar to those used in the specification Jet Fuel Thermal Oxidation Tester (JFTOT) and other thermal stability test apparatus. Blockages reduced flow rates and PDR concentrations downstream of the filters. Consequently two PDR additives (A&B) were tested in JFTOT apparatus without the usual in-line pre-filters as part of a Ministry of Defense (MoD) co-ordinated Round Robin exercise. Some fuel/PDR additive combinations caused decreases in JFTOT breakpoints. Effects were additive- (type, concentration and degree of shear) and fuel-dependent; most failures were caused by filter blockages and not by a failing lacquer rating. In further work at Thornton, the thermal stability characteristics of similar fuel/additive combinations have been examined in non-specification tests. In Flask Oxidation Tests, PDR additives caused no significant increase in the liquid phase oxidation rates of the fuels. Additives were tested in the Single Tube Heat Transfer Rig (STHTR) which duplicates many of the conditions of a heat exchanger element in an engine`s fuel supply system. B produced an average two-fold decrease in thermal stability in a Merox fuel; A had no significant effect. In hydrotreated fuel, B reduced the thermal stability up to five-fold. A had little effect below 205{degrees}C, while at higher temperatures there may have been a marginal improvement in thermal stability. Again, certain jet fuel/PDR combinations were seen to reduce thermal stability.

  17. Standardization of specifications and inspection procedures for LEU plate-type research reactor fuels

    International Nuclear Information System (INIS)

    1988-06-01

    With the transition to high density uranium LEU fuel, fabrication costs of research reactor fuel elements have a tendency to increase because of two reasons. First, the amount of the powder of the uranium compound required increases by more than a factor of five. Second, fabrication requirements are in many cases nearer the fabrication limits. Therefore, it is important that measures be undertaken to eliminate or reduce unnecessary requirements in the specification or inspection procedures of research reactor fuel elements utilizing LEU. An additional stimulus for standardizing specifications and inspection procedures at this time is provided by the fact that most LEU conversions will occur within a short time span, and that nearly all of them will require preparation of new specifications and inspection procedures. In this sense, the LEU conversions offer an opportunity for improving the rationality and efficiency of the fuel fabrication and inspection processes. This report focuses on the standardization of specifications and inspection processes of high uranium density LEU fuels for research reactors. However, in many cases the results can also be extended directly to other research reactor fuels. 15 refs, 1 fig., 3 tabs

  18. Corrosion surveillance for research reactor spent nuclear fuel in wet basin storage

    International Nuclear Information System (INIS)

    Howell, J.P.

    1999-01-01

    Foreign and domestic test and research reactor fuel is currently being shipped from locations over the world for storage in water filled basins at the Savannah River Site (SRS). The fuel was provided to many of the foreign countries as a part of the ''Atoms for Peace'' program in the early 1950's. In support of the wet storage of this fuel at the research reactor sites and at SRS, corrosion surveillance programs have been initiated. The International Atomic Energy Agency (IAEA) established a Coordinated Research Program (CRP) in 1996 on ''Corrosion of Research Reactor Aluminum-Clad Spent Fuel in Water'' and scientists from ten countries worldwide were invited to participate. This paper presents a detailed discussion of the IAEA sponsored CRP and provides the updated results from corrosion surveillance activities at SRS. In May 1998, a number of news articles around the world reported stories that microbiologically influenced corrosion (MIC) was active on the aluminum-clad spent fuel stored in the Receiving Basin for Offsite Fuels (RBOF) at SRS. This assessment was found to be in error with details presented in this paper. A biofilm was found on aluminum coupons, but resulted in no corrosion. Cracks seen on the surface were not caused by corrosion, but by stresses from the volume expansion of the oxide formed during pre-conditioning autoclaving. There has been no pitting caused by MIC or any other corrosion mechanism seen in the RBOF basin since initiation of the SRS Corrosion Surveillance Program in 1993

  19. A Development of Technical Specification of a Research Reactor with Plate Fuels Cooled by Upward Flow

    International Nuclear Information System (INIS)

    Park, Sujin; Kim, Jeongeun; Kim, Hyeonil

    2016-01-01

    The contents of the TS(Technical Specifications) are definitions, safety limits, limiting safety system settings, limiting conditions for operation, surveillance requirements, design features, and administrative controls. TS for Nuclear Power Plants (NPPs) have been developed since many years until now. On the other hands, there are no applicable modernized references of TS for research reactors with many differences from NPPs in purpose and characteristics. Fuel temperature and Departure from Nuclear Boiling Ratio (DNBR) are being used as references from the thermal-hydraulic analysis point of view for determining whether the design of research reactors satisfies acceptance criteria for the nuclear safety or not. Especially for research reactors using plate-type fuels, fuel temperature and critical heat flux, however, are very difficult to measure during the reactor operation. This paper described the outline of main contents of a TS for open-pool research reactor with plate-type fuels using core cooling through passive systems, where acceptance criteria for nuclear safety such as CHF and fuel temperature cannot be directly measured, different from circumstances in NPPs. Thus, three independent variables instead of non-measurable acceptance criteria: fuel temperature and CHF are considered as safety limits, i.e., power, flow, and flow temperature

  20. Irradiation tests on PHWR type fuel elements in TRIGA research reactor of INR Pitesti

    Energy Technology Data Exchange (ETDEWEB)

    Horhoianu, Grigore [Institute for Nuclear Research, Pitesti (Romania). Nuclear Fuel Engineering Lab.; Sorescu, Ion [Institute for Nuclear Research, Pitesti (Romania). TRIGA Reactor Loop Facility; Parvan, Marcel [Institute for Nuclear Research, Pitesti (Romania). Hot Cells Lab.

    2012-12-15

    Nine PHWR type fuel elements with reduced length were irradiated in loop A of the TRIGA Research Reactor of INR Pitesti. The primary objective of the test was to determine the performance of nuclear fuel fabricated at INR Pitesti at high linear powers in pressurized water conditions. Six fuel elements were irradiated with a ramp power history, achieving a maximum power of 45 kW/m during pre-ramp and of 64 kW/m in the ramp. The maximum discharge burnup was of 216 MWh/kgU. Another three fuel elements with reduced length were irradiated with declining power history. At the beginning of irradiation the fuel elements achieved a maximum linear power of 66 kW/m. The maximum fuel power was about 1.3 times the maximum expected in PHWR. The maximum discharge burnup was 205 MWh/kgU. The elements were destructively examined in the hot cells of INR Pitesti. Temperature-sensitive parameters such as UO{sub 2} grain growth, fission-gas release and sheath deformations were examined. The tests proved the feasibility of irradiating PHWR type fuel elements at linear powers up to 66 kW/m under pressurized water conditions and demonstrated the possibility of more flexible operation of this fuel in power reactors. This paper presents the results of the investigation. (orig.)