WorldWideScience

Sample records for fuel mangement program

  1. Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Draft Environmental Impact Statement. Volume 1, Appendix C, Savannah River Site Spent Nuclear Fuel Mangement Program

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

    The US Department of Energy (DOE) is engaged in two related decision making processes concerning: (1) the transportation, receipt, processing, and storage of spent nuclear fuel (SNF) at the DOE Idaho National Engineering Laboratory (INEL) which will focus on the next 10 years; and (2) programmatic decisions on future spent nuclear fuel management which will emphasize the next 40 years. DOE is analyzing the environmental consequences of these spent nuclear fuel management actions in this two-volume Environmental Impact Statement (EIS). Volume 1 supports broad programmatic decisions that will have applicability across the DOE complex and describes in detail the purpose and need for this DOE action. Volume 2 is specific to actions at the INEL. This document, which limits its discussion to the Savannah River Site (SRS) spent nuclear fuel management program, supports Volume 1 of the EIS. Following the introduction, Chapter 2 contains background information related to the SRS and the framework of environmental regulations pertinent to spent nuclear fuel management. Chapter 3 identifies spent nuclear fuel management alternatives that DOE could implement at the SRS, and summarizes their potential environmental consequences. Chapter 4 describes the existing environmental resources of the SRS that spent nuclear fuel activities could affect. Chapter 5 analyzes in detail the environmental consequences of each spent nuclear fuel management alternative and describes cumulative impacts. The chapter also contains information on unavoidable adverse impacts, commitment of resources, short-term use of the environment and mitigation measures.

  2. Radioactive waste mangement in Canada

    International Nuclear Information System (INIS)

    Didyk, J.P.

    1976-01-01

    The objectives of the Canadian radioactive waste management program are to manage the wastes so that the potential hazards of the material are minimized, and to manage the wastes in a manner which places the minimum possible burden on future generations. The Atomic Energy Control Board regulates all activities in the nuclear field in Canada, including radioactive waste management facility licensing. The Atomic Energy Control Act authorizes the Board to make rules for regulating its proceedings and the performance of its functions. The Atomic Energy Control Regulations define basic regulatory requirements for the licensing of facilities, equipment and materials, including requirements for records and inspection, for security and for health and safety

  3. Consolidated fuel reprocessing program

    Science.gov (United States)

    1985-04-01

    A survey of electrochemical methods applications in fuel reprocessing was completed. A dummy fuel assembly shroud was cut using the remotely operated laser disassembly equipment. Operations and engineering efforts have continued to correct equipment operating, software, and procedural problems experienced during the previous uranium compaigns. Fuel cycle options were examined for the liquid metal reactor fuel cycle. In high temperature gas cooled reactor spent fuel studies, preconceptual designs were completed for the concrete storage cask and open field drywell storage concept. These and other tasks operating under the consolidated fuel reprocessing program are examined.

  4. Spent fuels program

    International Nuclear Information System (INIS)

    Shappert, L.B.

    1983-01-01

    The goal of this task is to support the Domestic Spent Fuel Storage Program through studies involving the transport of spent fuel. A catalog was developed to provide authoritative, timely, and accessible transportation information for persons involved in the transport of irradiated reactor fuel. The catalog, drafted and submitted to the Transportation Technology Center, Sandia National Laboratories, for their review and approval, covers such topics as federal, state, and local regulations, spent fuel characteristics, cask characteristics, transportation costs, and emergency response information

  5. Romanian nuclear fuel program

    International Nuclear Information System (INIS)

    Budan, O.

    1999-01-01

    The paper presents and comments the policy adopted in Romania for the production of CANDU-6 nuclear fuel before and after 1990. The CANDU-6 nuclear fuel manufacturing started in Romania in December 1983. Neither AECL nor any Canadian nuclear fuel manufacturer were involved in the Romanian industrial nuclear fuel production before 1990. After January 1990, the new created Romanian Electricity Authority (RENEL) assumed the responsibility for the Romanian Nuclear Power Program. It was RENEL's decision to stop, in June 1990, the nuclear fuel production at the Institute for Nuclear Power Reactors (IRNE) Pitesti. This decision was justified by the Canadian specialists team findings, revealed during a general, but well enough technically founded analysis performed at IRNE in the spring of 1990. All fuel manufactured before June 1990 was quarantined as it was considered of suspect quality. By that time more than 31,000 fuel bundles had already been manufactured. This fuel was stored for subsequent assessment. The paper explains the reasons which provoked this decision. The paper also presents the strategy adopted by RENEL after 1990 regarding the Romanian Nuclear Fuel Program. After a complex program done by Romanian and Canadian partners, in November 1994, AECL issued a temporary certification for the Romanian nuclear fuel plant. During the demonstration manufacturing run, as an essential milestone for the qualification of the Romanian fuel supplier for CANDU-6 reactors, 202 fuel bundles were produced. Of these fuel bundles, 66 were part of the Cernavoda NGS Unit 1 first fuel load (the balance was supplied by Zircatec Precision Industries Inc. ZPI). The industrial nuclear fuel fabrication re-started in Romania in January 1995 under AECL's periodical monitoring. In December 1995, AECL issued a permanent certificate, stating the Romanian nuclear fuel plant as a qualified and authorised CANDU-6 fuel supplier. The re-loading of the Cernavoda NGS Unit 1 started in the middle

  6. Fuel Cell Demonstration Program

    Energy Technology Data Exchange (ETDEWEB)

    Gerald Brun

    2006-09-15

    In an effort to promote clean energy projects and aid in the commercialization of new fuel cell technologies the Long Island Power Authority (LIPA) initiated a Fuel Cell Demonstration Program in 1999 with six month deployments of Proton Exchange Membrane (PEM) non-commercial Beta model systems at partnering sites throughout Long Island. These projects facilitated significant developments in the technology, providing operating experience that allowed the manufacturer to produce fuel cells that were half the size of the Beta units and suitable for outdoor installations. In 2001, LIPA embarked on a large-scale effort to identify and develop measures that could improve the reliability and performance of future fuel cell technologies for electric utility applications and the concept to establish a fuel cell farm (Farm) of 75 units was developed. By the end of October of 2001, 75 Lorax 2.0 fuel cells had been installed at the West Babylon substation on Long Island, making it the first fuel cell demonstration of its kind and size anywhere in the world at the time. Designed to help LIPA study the feasibility of using fuel cells to operate in parallel with LIPA's electric grid system, the Farm operated 120 fuel cells over its lifetime of over 3 years including 3 generations of Plug Power fuel cells (Lorax 2.0, Lorax 3.0, Lorax 4.5). Of these 120 fuel cells, 20 Lorax 3.0 units operated under this Award from June 2002 to September 2004. In parallel with the operation of the Farm, LIPA recruited government and commercial/industrial customers to demonstrate fuel cells as on-site distributed generation. From December 2002 to February 2005, 17 fuel cells were tested and monitored at various customer sites throughout Long Island. The 37 fuel cells operated under this Award produced a total of 712,635 kWh. As fuel cell technology became more mature, performance improvements included a 1% increase in system efficiency. Including equipment, design, fuel, maintenance

  7. Canadian fuel development program

    International Nuclear Information System (INIS)

    Gacesa, M.; Young, E.G.

    1992-11-01

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

  8. Fuel Cell and Hydrogen Technologies Program | Hydrogen and Fuel Cells |

    Science.gov (United States)

    NREL Fuel Cell and Hydrogen Technologies Program Fuel Cell and Hydrogen Technologies Program Through its Fuel Cell and Hydrogen Technologies Program, NREL researches, develops, analyzes, and validates fuel cell and hydrogen production, delivery, and storage technologies for transportation

  9. Consolidated fuel reprocessing program

    International Nuclear Information System (INIS)

    Kuban, D.P.; Noakes, M.W.; Bradley, E.C.

    1987-01-01

    The Advanced Servomanipulator (ASM) System consists of three major components: the ASM slave, the dual arm master controller or master, and the control system. The ASM is a remotely maintainable force-reflecting servomanipulator developed at the Oak Ridge National Laboratory (ORNL) as part of the Consolidated Fuel Reprocessing Program of (CFRP). This new manipulator addresses requirements of advanced nuclear fuel reprocessing with emphasis on force reflection, remote maintainability, and reliability. It uses an all-gear force transmission system. The master arms were designed as a kinematic replica of ASM and use cable force transmission. Special digital control algorithms were developed to improve the system performance. The system is presently operational and undergoing evaluation. Preliminary testing has been completed and is reported. The system is now undergoing commercialization by transferring the technology to the private sector

  10. Counter narratives inplace and city branding and their influence on brand mangement

    DEFF Research Database (Denmark)

    Jørgensen, Ole Have

    2016-01-01

    The study is aimed at describing the types of counter narratives evolving in a city brand process, how they are understood and dealt with and how they may influence brand mangement.......The study is aimed at describing the types of counter narratives evolving in a city brand process, how they are understood and dealt with and how they may influence brand mangement....

  11. Plutonium fuel program

    International Nuclear Information System (INIS)

    1979-09-01

    A review is presented of the development of the (UPu)C sphere-pac fuel project during 1978. In particular, the problems encountered in obtaining good fuel quality in the fabrication process and their solution is discussed. The development of a fabrication pilot plant is considered, and the post-irradiation examination of fuel pins is presented. (Auth.)

  12. Plutonium fuel program

    International Nuclear Information System (INIS)

    1979-01-01

    The work of the Project-Fuel Development reached the apex of its current programme during the course of the year. Notable success was recorded in the area of irradiation testing with the completion of the examination of the MFBS-7 irradiation. The irradiation group also prepared the seventh Filos experiment and this, as well as the DIDO-III test, began irradiation at the end of the year. Consideration was given and plans prepared for a revised pin filling line for bundle tests. Work also began on the conceptual design study for a pilot production line having a nominal capacity of 500 kg fuel per year. (Auth.)

  13. Alcohol fuels program technical review

    Energy Technology Data Exchange (ETDEWEB)

    None

    1981-07-01

    The last issue of the Alcohol Fuels Process R/D Newsletter contained a work breakdown structure (WBS) of the SERI Alcohol Fuels Program that stressed the subcontracted portion of the program and discussed the SERI biotechnology in-house program. This issue shows the WBS for the in-house programs and contains highlights for the remaining in-house tasks, that is, methanol production research, alcohol utilization research, and membrane research. The methanol production research activity consists of two elements: development of a pressurized oxygen gasifier and synthesis of catalytic materials to more efficiently convert synthesis gas to methanol and higher alcohols. A report is included (Finegold et al. 1981) that details the experimental apparatus and recent results obtained from the gasifier. The catalysis research is principally directed toward producing novel organometallic compounds for use as a homogeneous catalyst. The utilization research is directed toward the development of novel engine systems that use pure alcohol for fuel. Reforming methanol and ethanol catalytically to produce H/sub 2/ and CO gas for use as a fuel offers performance and efficiency advantages over burning alcohol directly as fuel in an engine. An application of this approach is also detailed at the end of this section. Another area of utilization is the use of fuel cells in transportation. In-house researchers investigating alternate electrolyte systems are exploring the direct and indirect use of alcohols in fuel cells. A workshop is being organized to explore potential applications of fuel cells in the transportation sector. The membrane research group is equipping to evaluate alcohol/water separation membranes and is also establishing cost estimation and energy utilization figures for use in alcohol plant design.

  14. The DUPIC fuel development program in KAERI

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

  15. 75 FR 37733 - Regulation of Fuels and Fuel Additives: Modifications to Renewable Fuel Standard Program

    Science.gov (United States)

    2010-06-30

    ... Regulation of Fuels and Fuel Additives: Modifications to Renewable Fuel Standard Program AGENCY... direct final rule to amend the Renewable Fuel Standard program requirements on May 10, 2010. Because EPA... Fuel Standard program requirements, published on May 10, 2010. We stated in that direct final rule that...

  16. 1986 fuel cell seminar: Program and abstracts

    Energy Technology Data Exchange (ETDEWEB)

    None

    1986-10-01

    Ninety nine brief papers are arranged under the following session headings: gas industry's 40 kw program, solid oxide fuel cell technology, phosphoric acid fuel cell technology, molten carbonate fuel cell technology, phosphoric acid fuel cell systems, power plants technology, fuel cell power plant designs, unconventional fuels, fuel cell application and economic assessments, and plans for commerical development. The papers are processed separately for the data base. (DLC)

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  18. The Canadian nuclear fuel waste management program

    International Nuclear Information System (INIS)

    Dixon, R.S.

    1984-12-01

    The Canadian Nuclear Fuel Waste Management Program involves research into the storage and transportation of used nuclear fuel, immobilization of fuel waste, and deep geological disposal of the immobilized waste. The program is now in the fourth year of a ten-year generic research and development phase. The objective of this phase of the program is to assess the safety and environmental aspects of the deep underground disposal of immobilized fuel waste in plutonic rock. The objectives of the research for each component of the program and the progress made to the end of 1983 are described in this report

  19. The Canadian fuel waste management program

    International Nuclear Information System (INIS)

    McConnell, D.B.

    1986-04-01

    This report is the sixth in the series of annual reports on the research and development program for the safe management and disposal of Canada's nuclear fuel waste. The report summarizes progress in 1984 for the following activities: storage and transportation of used fuel, immobilization of nuclear fuel waste, geotechnical research, environmental research, and environmental and safety assessment. 186 refs

  20. The Canadian nuclear fuel waste management program

    International Nuclear Information System (INIS)

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

    1984-04-01

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

  1. 76 FR 18066 - Regulation of Fuels and Fuel Additives: Changes to Renewable Fuel Standard Program

    Science.gov (United States)

    2011-04-01

    ... ENVIRONMENTAL PROTECTION AGENCY 40 CFR Part 80 Regulation of Fuels and Fuel Additives: Changes to Renewable Fuel Standard Program CFR Correction In Title 40 of the Code of Federal Regulations, Parts 72 to...-generating foreign producers and importers of renewable fuels for which RINs have been generated by the...

  2. Nuclear-fuel-cycle costs. Consolidated Fuel-Reprocessing Program

    International Nuclear Information System (INIS)

    Burch, W.D.; Haire, M.J.; Rainey, R.H.

    1981-01-01

    The costs for the back-end of the nuclear fuel cycle, which were developed as part of the Nonproliferation Alternative Systems Assessment Program (NASAP), are presented. Total fuel-cycle costs are given for the pressurized-water reactor once-through and fuel-recycle systems, and for the liquid-metal fast-breeder-reactor system. These calculations show that fuel-cycle costs are a small part of the total power costs. For breeder reactors, fuel-cycle costs are about half that of the present once-through system. The total power cost of the breeder-reactor system is greater than that of light-water reactor at today's prices for uranium and enrichment

  3. Ansaldo programs on fuel cell vehicles

    Energy Technology Data Exchange (ETDEWEB)

    Marcenaro, B.G.; Federici, F. [Ansaldo Ricerche Srl, Genova (Italy)

    1996-12-31

    The growth in traffic and the importance of maintaining a stable ecology at the global scale, particularly with regard to atmospheric pollution, raises the necessity to realize a new generation of vehicles which are more efficient, more economical and compatible with the environment. At European level, the Car of Tomorrow task force has identified fuel cells as a promising alternative propulsion system. Ansaldo Ricerche has been involved in the development of fuel cell vehicles since the early nineties. Current ongoing programs relates to: (1) Fuel cell bus demonstrator (EQHEPP BUS) Test in 1996 (2) Fuel cell boat demonstrator (EQHHPP BOAT) Test in 1997 (3) Fuel cell passenger car prototype (FEVER) Test in 1997 (4) 2nd generation Fuel cell bus (FCBUS) 1996-1999 (5) 2nd generation Fuel cell passenger car (HYDRO-GEN) 1996-1999.

  4. Climate Change Fuel Cell Program

    Energy Technology Data Exchange (ETDEWEB)

    Paul Belard

    2006-09-21

    Verizon is presently operating the largest Distributed Generation Fuel Cell project in the USA. Situated in Long Island, NY, the power plant is composed of seven (7) fuel cells operating in parallel with the Utility grid from the Long Island Power Authority (LIPA). Each fuel cell has an output of 200 kW, for a total of 1.4 mW generated from the on-site plant. The remaining power to meet the facility demand is purchased from LIPA. The fuel cell plant is utilized as a co-generation system. A by-product of the fuel cell electric generation process is high temperature water. The heat content of this water is recovered from the fuel cells and used to drive two absorption chillers in the summer and a steam generator in the winter. Cost savings from the operations of the fuel cells are forecasted to be in excess of $250,000 per year. Annual NOx emissions reductions are equivalent to removing 1020 motor vehicles from roadways. Further, approximately 5.45 million metric tons (5 millions tons) of CO2 per year will not be generated as a result of this clean power generation. The project was partially financed with grants from the New York State Energy R&D Authority (NYSERDA) and from Federal Government Departments of Defense and Energy.

  5. The Canadian nuclear fuel waste management program

    International Nuclear Information System (INIS)

    Rummery, T.E.; Rosinger, E.L.J.

    1983-05-01

    The Canadian Nuclear Fuel Waste Management Program is now well established. This report outlines the generic research and technological development underway in this program to assess the concept of immobilization and subsequent disposal of nuclear fuel waste deep in a stable plutonic rock in the Canadian Shield. The program participants, funding, schedule and associated external review processes are briefly outlined. The major scientific and engineering components of the program, namely, immobilization studies, geoscience research and environmental and safety assessment, are described in more detail

  6. The Canadian nuclear fuel waste management program

    International Nuclear Information System (INIS)

    Rummery, T.E.; Rosinger, E.L.J.

    1984-12-01

    The Canadian Nuclear Fuel Waste Management Program is in the fourth year of a ten-year generic research and development phase. The objective of this phase of the program is to assess the basic safety and environmental aspects of the concept of isolating immobilized fuel waste by deep underground disposal in plutonic rock. The major scientific and engineering components of the program, namely immobilization studies, geoscience research, and environmental and safety assessment, are described. Program funding, scheduling and associated external review processes are briefly outlined

  7. French development program on fuel cycle

    International Nuclear Information System (INIS)

    Viala, M.; Bourgeois, M.

    1991-01-01

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

  8. Renewable Fuel Pathways II Final Rule to Identify Additional Fuel Pathways under Renewable Fuel Standard Program

    Science.gov (United States)

    This final rule describes EPA’s evaluation of biofuels derived from biogas fuel pathways under the RFS program and other minor amendments related to survey requirements associated with ULSD program and misfueling mitigation regulations for E15.

  9. The low-enrichment fuel development program

    International Nuclear Information System (INIS)

    Stahl, D.

    1993-01-01

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

  10. DOE Hydrogen and Fuel Cells Program Plan (September 2011)

    Energy Technology Data Exchange (ETDEWEB)

    none,

    2011-09-01

    The Department of Energy Hydrogen and Fuel Cells Program Plan outlines the strategy, activities, and plans of the DOE Hydrogen and Fuel Cells Program, which includes hydrogen and fuel cell activities within the EERE Fuel Cell Technologies Program and the DOE offices of Nuclear Energy, Fossil Energy, and Science.

  11. Program on MOX fuel utilization in light water reactors

    International Nuclear Information System (INIS)

    Kenda, Hirofumi

    2000-01-01

    MOX fuel utilization program by the Japanese electric power companies was released in February, 1997. Principal philosophy for MOX fuel design is that MOX fuel shall be compatible with Uranium fuel and behavior of core loaded with MOX fuel shall be similar to that of conventional core. MOX fuel is designed so that geometry and nuclear capability of MOX fuel are equivalent to Uranium fuel. (author)

  12. The ORNL plasma fueling program

    International Nuclear Information System (INIS)

    Gouge, M.J.; Milora, S.L.; Foster, C.A.

    1989-01-01

    The Oak Ridge National Laboratory (ORNL) has been developing pellet injectors for over ten years. These devices produce frozen hydrogen isotope pellets and then accelerate the projectiles to speeds in the km/s range by either pneumatic or mechanical techniques. A variety of designs have been developed, including single-shot guns, multiple-shot guns, machine guns, and centrifugal accelerators. These injectors have been used to inject hydrogen and deuterium pellets into plasmas on numerous tokamak experiments resulting in improved plasma performance. ORNL has recently provided pellet fueling systems for the Tokamak Fusion Test Reactor (TFTR), the Joint European Torus (JET), and the Tore Supra tokamak. This paper discusses developments on these injector designs. 10 refs

  13. Program summary. Nuclear waste management and fuel cycle programs

    International Nuclear Information System (INIS)

    1982-07-01

    This Program Summary Document describes the US Department of Energy (DOE) Nuclear Waste Management and Fuel Cycle Programs. Particular emphasis is given to near-term, specifically Fiscal Year (FY) 1982, activities. The overall objective of these programs will be achieved by the demonstration of: (1) safe radioactive waste management practices for storage and disposal of high-level waste and (2) advanced technologies necessary to close the nuclear fuel cycle on a schedule which would assure a healthy future for the development of nuclear power in this country

  14. Fuel Flexible Turbine System (FFTS) Program

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2012-12-31

    In this fuel flexible turbine system (FFTS) program, the Parker gasification system was further optimized, fuel composition of biomass gasification process was characterized and the feasibility of running Capstone MicroTurbine(TM) systems with gasification syngas fuels was evaluated. With high hydrogen content, the gaseous fuel from a gasification process of various feed stocks such as switchgrass and corn stover has high reactivity and high flashback propensity when running in the current lean premixed injectors. The research concluded that the existing C65 microturbine combustion system, which is designed for natural gas, is not able to burn the high hydrogen content syngas due to insufficient resistance to flashback (undesired flame propagation to upstream within the fuel injector). A comprehensive literature review was conducted on high-hydrogen fuel combustion and its main issues. For Capstone's lean premixed injector, the main mechanisms of flashback were identified to be boundary layer flashback and bulk flow flashback. Since the existing microturbine combustion system is not able to operate on high-hydrogen syngas fuels, new hardware needed to be developed. The new hardware developed and tested included (1) a series of injectors with a reduced propensity for boundary layer flashback and (2) two new combustion liner designs (Combustion Liner Design A and B) that lead to desired primary zone air flow split to meet the overall bulk velocity requirement to mitigate the risk of core flashback inside the injectors. The new injector designs were evaluated in both test apparatus and C65/C200 engines. While some of the new injector designs did not provide satisfactory performance in burning target syngas fuels, particularly in improving resistance to flashback. The combustion system configuration of FFTS-4 injector and Combustion Liner Design A was found promising to enable the C65 microturbine system to run on high hydrogen biomass syngas. The FFTS-4 injector

  15. Statement on the Consolidated Fuel Reprocessing Program

    International Nuclear Information System (INIS)

    Trauger, D.B.

    1984-01-01

    Oak Ridge National Laboratory has chosen the following objectives for future reprocessing plant design: reduced radiation exposure to workers; minimal environmental impact; improved plant operation and maintenance; improved accountability; no plutonium diversion; and reduced overall capital and operating cost. These objectives lead to a plant with totally remote operation. The Breeder Reactor Engineering Test (BRET) has been designed to perform a key role in demonstrating advanced reprocessing technology. It has been scheduled to be available to reprocess spent fuel from the Fast Flux Test Facility. The principal features of the Consolidated Fuel Reprocessing Program and of the BRET facility are appropriate for all reactor types

  16. Future of alcohol fuels programs in Brasil

    Science.gov (United States)

    Carvalho, A. V., Jr.; Rechtschaffen, E.; Goldstein, L., Jr.

    An updating is given of the Brazilian National Alcohol Program's production and utilization achievements to date in the substitution of ethanol and methanol for imported oil products. A series of Eucalyptus forestry and processing-industry projections are made for fuel output and jobs creation that may be expected by the year 2000. With few exceptions, methanol produced from wood grown on poorer soils than can now be used for sugarcane substitute for oil products and result in jobs creation several orders of magnitude higher than petroleum fuels.

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

    International Nuclear Information System (INIS)

    Barner, J.O.

    1979-04-01

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

  18. GCRA review and appraisal of fuel material development programs

    International Nuclear Information System (INIS)

    1980-09-01

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

  19. Canadian fuel development program in 1997/98

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  20. The American 'severe fuel damage program'

    International Nuclear Information System (INIS)

    Sdouz, G.

    1982-03-01

    The TMI-2 accident has initiated a new phase of safety research. It is necessary to consider severe accidents with degraded or molten core. For NRC there was a need for an improved understanding of this reactor behaviour and the 'Severe Fuel Dage Program' was initiated. Planned are in-pile experiments in PBF, NRU and ESSOR and in addition separate effects tests and results from TMI-2. The analytical component of the program is the development of different versions of the code SCDAP for the detailed analysis during severe accident transients. (Author) [de

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

    International Nuclear Information System (INIS)

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

    1993-09-01

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

  2. Analysis of transient fuel failure mechanisms: selected ANL programs

    International Nuclear Information System (INIS)

    Deitrich, L.W.

    1975-01-01

    Analytical programs at Argonne National Laboratory related to fuel pin failure mechanisms in fast-reactor accident transients are described. The studies include transient fuel pin mechanics, mechanics of unclad fuel, and mechanical effects concerning potential fuel failure propagation. (U.S.).

  3. 14 CFR 121.1113 - Fuel tank system maintenance program.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 3 2010-01-01 2010-01-01 false Fuel tank system maintenance program. 121... Improvements § 121.1113 Fuel tank system maintenance program. (a) Except as provided in paragraph (g) of this... capacity of 7500 pounds or more. (b) For each airplane on which an auxiliary fuel tank is installed under a...

  4. 14 CFR 125.507 - Fuel tank system inspection program.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 3 2010-01-01 2010-01-01 false Fuel tank system inspection program. 125... Airworthiness and Safety Improvements § 125.507 Fuel tank system inspection program. (a) Except as provided in... fuel tank is installed under a field approval, before June 16, 2008, the certificate holder must submit...

  5. 14 CFR 129.113 - Fuel tank system maintenance program.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 3 2010-01-01 2010-01-01 false Fuel tank system maintenance program. 129... Continued Airworthiness and Safety Improvements § 129.113 Fuel tank system maintenance program. (a) Except... on which an auxiliary fuel tank is installed under a field approval, before June 16, 2008, the...

  6. 14 CFR 91.1507 - Fuel tank system inspection program.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 2 2010-01-01 2010-01-01 false Fuel tank system inspection program. 91... Airworthiness and Safety Improvements § 91.1507 Fuel tank system inspection program. (a) Except as provided in... fuel tank is installed under a field approval, before June 16, 2008, the operator must submit to the...

  7. Spent fuel receipt and lag storage facility for the spent fuel handling and packaging program

    International Nuclear Information System (INIS)

    Black, J.E.; King, F.D.

    1979-01-01

    Savannah River Laboratory (SRL) is participating in the Spent Fuel Handling and Packaging Program for retrievable, near-surface storage of spent light water reactor (LWR) fuel. One of SRL's responsibilities is to provide a technical description of the wet fuel receipt and lag storage part of the Spent Fuel Handling and Packaging (SFHP) facility. This document is the required technical description

  8. Alternative transportation fuels in the USA: government hydrogen vehicle programs

    International Nuclear Information System (INIS)

    Cannon, J.S.

    1993-01-01

    The linkage between natural gas-based transportation and hydrogen-based transportation strategies, two clean burning gaseous fuels, provides a strong policy rationale for increased government sponsorship of hydrogen vehicle research and demonstration programs. Existing federal and state government hydrogen vehicle projects are discussed in this paper: research at the NREL, alternate-fueled buses, Renewable Hydrogen for the State of Hawaii program, New York state alternative transportation fuels program, Colorado program. 9 refs

  9. Status and strategy of the U.S. commercial waste management program. Consolidated Fuel Reprocessing Program

    International Nuclear Information System (INIS)

    Croff, A.G.; Jubin, R.T.

    1983-01-01

    Management of airborne waste generally involves the following steps: recovery, treatment, interim storage, transportation, and disposal. The recovery (retention) of airborne radionuclides is generally well developed since the first-generation processes have been used for iodine and particulates for decades by the DOE nuclear materials production plants. Later-generation processes have been carried separately through the cold pilot-plant stage. However, the design and demonstration of a hot, integrated flowsheet for the recovery of all airborne species to the extent necessary to meet applicable regulations are still required. Treatment of the recovered airborne wastes is generally less-well developed. Tentatively preferred processes have been identified: iodine-barium iodate and/or silver zeolites in concrete with additives; krypton-implanted as ions in a metal alloy and encapsulated in concrete; carbon-barium carbonate in concrete with additives; particulates-encapsulation of HEPA filters in concrete; ruthenium-ruthenium traps encapsulated in concrete. The technology for interim storage and transportation appears to be straightforward engineering extensions of existing technology, assuming that the waste forms listed above are to be employed. Waste disposal concepts are the least well-developed aspect of airborne waste mangement technology. It appears that the long-lived materials such as 129 I, 14 C, and particulates will have to be emplaced in a geologic repository and that shorter-lived airborne waste may be acceptable in shallow-land burial grounds. The long-range goal of the program is to determine all of the steps necessary to manage airborne wastes

  10. The spent fuel disposal program in Taiwan

    International Nuclear Information System (INIS)

    Li, K.K.

    1994-01-01

    It is important, especially for countries with plan to develop nuclear power, to recognize that two key factors to the future prosperity of nuclear power are the safety of nuclear power plants and the appropriate management of backend activities. This paper described the financial, managerial, technical, and political status of the spent fuel disposal program in a newly industrialized country. It is concluded that the R ampersand D works and operational practices associated with the backend activities must be carried out in parallel with the development of nuclear power

  11. Tubular solid oxide fuel cell development program

    Energy Technology Data Exchange (ETDEWEB)

    Ray, E.R.; Cracraft, C.

    1995-12-31

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

  12. 75 FR 79964 - Regulation of Fuels and Fuel Additives: Modifications to Renewable Fuel Standard Program

    Science.gov (United States)

    2010-12-21

    ... renewable fuel facility to be considered biogas for purposes of Table 1 to Sec. 80.1426; --Sec. 80.1452(b... explained that many developers of ethanol facilities, including their own, sought to obtain construction permits without going through EPA's New Source Review (NSR) program, and were able to do so by obtaining...

  13. Emergency fuels utilization guidebook. Alternative Fuels Utilization Program

    Energy Technology Data Exchange (ETDEWEB)

    1980-08-01

    The basic concept of an emergency fuel is to safely and effectively use blends of specification fuels and hydrocarbon liquids which are free in the sense that they have been commandeered or volunteered from lower priority uses to provide critical transportation services for short-duration emergencies on the order of weeks, or perhaps months. A wide variety of liquid hydrocarbons not normally used as fuels for internal combustion engines have been categorized generically, including limited information on physical characteristics and chemical composition which might prove useful and instructive to fleet operators. Fuels covered are: gasoline and diesel fuel; alcohols; solvents; jet fuels; kerosene; heating oils; residual fuels; crude oils; vegetable oils; gaseous fuels.

  14. The Canadian nuclear fuel waste management program

    International Nuclear Information System (INIS)

    Dormuth, K.W.; Nuttall, K.

    1987-01-01

    Canada has established an extensive research program to develop and demonstrate the technology for safely disposing of nuclear fuel waste from Canadian nuclear electric generating stations. The program focuses on the concept of disposal deep in plutonic rock, which is abundant in the province of Ontario, Canada's major producer of nuclear electricity. Research is carried out at field research areas in the Canadian Precambrian Shield, and in government and university laboratories. The schedule calls for a document assessing the disposal concept to be submitted to regulatory and environmental agencies in late 1988. This document will form the basis for a review of the concept by these agencies and by the public. No site selection will be carried out before this review is completed. 10 refs.; 2 figs

  15. 76 FR 67287 - Alternative Fuel Transportation Program; Alternative Fueled Vehicle Credit Program (Subpart F...

    Science.gov (United States)

    2011-10-31

    ... additional credits for the use of biodiesel in blends of 20 percent biodiesel or greater and have provided an... discussion in Part II.A), the original program based upon AFV acquisitions and biodiesel use became known as... example, B20 (a 20 percent blend of biodiesel with 80 percent petroleum diesel) is not an alternative fuel...

  16. National fuel cell bus program : proterra fuel cell hybrid bus report, Columbia demonstration.

    Science.gov (United States)

    2011-10-01

    This report summarizes the experience and early results from a fuel cell bus demonstration funded by the Federal Transit Administration (FTA) under the National Fuel Cell Bus Program. A team led by the Center for Transportation and the Environment an...

  17. 75 FR 15893 - Tire Fuel Efficiency Consumer Information Program

    Science.gov (United States)

    2010-03-30

    ... how much the proposed consumer information program would affect consumer tire purchasing behavior and... 575 Tire Fuel Efficiency Consumer Information Program; Final Rule #0;#0;Federal Register / Vol. 75, No... 2127-AK45 Tire Fuel Efficiency Consumer Information Program AGENCY: National Highway Traffic Safety...

  18. Hawaii alternative fuels utilization program. Phase 3, final report

    Energy Technology Data Exchange (ETDEWEB)

    Kinoshita, C.M.; Staackmann, M.

    1996-08-01

    The Hawaii Alternative Fuels Utilization Program originated as a five-year grant awarded by the US Department of Energy (USDOE) to the Hawaii Natural Energy Institute (HNEI) of the University of Hawaii at Manoa. The overall program included research and demonstration efforts aimed at encouraging and sustaining the use of alternative (i.e., substitutes for gasoline and diesel) ground transportation fuels in Hawaii. Originally, research aimed at overcoming technical impediments to the widespread adoption of alternative fuels was an important facet of this program. Demonstration activities centered on the use of methanol-based fuels in alternative fuel vehicles (AFVs). In the present phase, operations were expanded to include flexible fuel vehicles (FFVs) which can operate on M85 or regular unleaded gasoline or any combination of these two fuels. Additional demonstration work was accomplished in attempting to involve other elements of Hawaii in the promotion and use of alcohol fuels for ground transportation in Hawaii.

  19. The Canadian CANDU fuel development program and recent fuel operating experience

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  20. 1990 fuel cell seminar: Program and abstracts

    Energy Technology Data Exchange (ETDEWEB)

    1990-12-31

    This volume contains author prepared short resumes of the presentations at the 1990 Fuel Cell Seminar held November 25-28, 1990 in Phoenix, Arizona. Contained herein are 134 short descriptions organized into topic areas entitled An Environmental Overview, Transportation Applications, Technology Advancements for Molten Carbonate Fuel Cells, Technology Advancements for Solid Fuel Cells, Component Technologies and Systems Analysis, Stationary Power Applications, Marine and Space Applications, Technology Advancements for Acid Type Fuel Cells, and Technology Advancement for Solid Oxide Fuel Cells.

  1. Alcohol Fuels Program technical review, Spring 1984

    Energy Technology Data Exchange (ETDEWEB)

    1984-10-01

    The alcohol fuels program consists of in-house and subcontracted research for the conversion of lignocellulosic biomass into fuel alcohols via thermoconversion and bioconversion technologies. In the thermoconversion area, the SERI gasifier has been operated on a one-ton per day scale and produces a clean, medium-Btu gas that can be used to manufacture methanol with a relatively small gas-water shift reaction requirement. Recent research has produced catalysts that make methanol and a mixture of higher alcohols from the biomass-derived synthetic gas. Three hydrolysis processes have emerged as candidates for more focused research. They are: a high-temperature, dilute-acid, plug-flow approach based on the Dartmouth reactor; steam explosion pretreatment followed by hydrolysis using the RUT-C30 fungal organism; and direct microbial conversion of the cellulose to ethanol using bacteria in a single or mixed culture. Modeling studies, including parametric and sensitivity analyses, have recently been completed. The results of these studies will lead to a better definition of the present state-of-the-art for these processes and provide a framework for establishing the research and process engineering issues that still need resolution. In addition to these modeling studies, economic feasibility studies are being carried out by commercial engineering firms. Their results will supplement and add commercial validity to the program results. The feasibility contractors will provide input at two levels: Technical and economic assessment of the current state-of-the-art in alcohol production from lignocellulosic biomass via thermoconversion to produce methanol and higher alcohol mixtures and bioconversion to produce ethanol; and identification of research areas having the potential to significantly reduce the cost of production of alcohols.

  2. LOFT fuel modules design, characterization, and fabrication program

    International Nuclear Information System (INIS)

    Russell, M.L.

    1977-06-01

    The loss-of-fluid test [LOFT) fuel modules have evolved from a comprehensive five-year design, characterization, and fabrication program which has resulted in the accomplishment of many technical activities of interest in pressurized water reactor fuel design development and safety research. Information is presented concerning: determination of fundamental high-temperature reactor material properties; design invention related to in-core instrumentation attachment; implementation of advanced and/or unique fuel bundle characterization techniques; implementation of improved fuel bundle fabrication techniques; and planning and execution of a multimillion dollar design, characterization, and fabrication program for pressurized water reactor fuel

  3. Overview of the US spent nuclear fuel program

    International Nuclear Information System (INIS)

    Hurt, W.L.

    1999-01-01

    This report, Overview of the United States Spent Nuclear Fuel Program, December, 1997, summarizes the U.S. strategy for interim management and ultimate disposition of spent nuclear fuel from research and test reactors. The key elements of this strategy include consolidation of this spent nuclear fuel at three sites, preparation of the fuel for geologic disposal in road-ready packages, and low-cost dry interim storage until the planned geologic repository is opened. The U.S. has a number of research programs in place that are intended to Provide data and technologies to support both characterization and disposition of the fuel. (author)

  4. Fuel cell energy service Enron`s commerical program

    Energy Technology Data Exchange (ETDEWEB)

    Jacobson, M.W.

    1996-04-01

    Enron, the premier provider of clean fuels worldwide, has launched a unique energy service based on fuel cell technology. The goal of this program is to bring the benefits of fuel cell power to the broad commercial marketplace. Enron`s Energy Service is currently based on a 200 kilowatt phosphoric acid power plant manufactured by ONSI Corporation. This plant is fueled by natural gas or propane, and exhibits superior performance. Enron offers a `no hassle` package that provides customers with immediate benefits with no upfront capital or technical risks. This paper describes Enron`s fuel cell commercial program.

  5. Discussion paper: direction for Canada's alternate fuels program

    Energy Technology Data Exchange (ETDEWEB)

    1982-09-01

    There is a growing need to accelerate the consideration of alternate fuels for use in Canadian vehicle transportation. At the present time various governments and corporations are initiating alternate fuel programs involving ethanol, methanol, CNG, propane, etc. There is a bewildering array of perspectives as to which fuel or fuels will best serve Canada's needs in the future. In response to the 'Discussion Paper on Liquid Fuels Options, 1980', by the Federal Dept. of Energy, Mines and Resources, Ford of Canada has prepared this perspective on each of the alternate fuels from the company's vantage point as a vehicle manufacturer.

  6. 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

  7. Ciclon: A neutronic fuel management program for PWR's consecutive cycles

    International Nuclear Information System (INIS)

    Aragones, J.M.

    1977-01-01

    The program description and user's manual of a new computer code is given. Ciclon performs the neutronic calculation of consecutive reload cycles for PWR's fuel management optimization. Fuel characteristics and burnup data, region or batch sizes, loading schemes and state of previously irradiated fuel are input to the code. Cycle lengths or feed enrichments and burnup sharing for each region or batch are calculate using different core neutronic models and printed or punched in standard fuel management format. (author) [es

  8. RERTR program progress in qualifying reduced-enrichment fuels

    International Nuclear Information System (INIS)

    Snelgrove, James L.

    1983-01-01

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

  9. Fuel Cell Seminar, 1992: Program and abstracts

    Energy Technology Data Exchange (ETDEWEB)

    1992-12-31

    This year`s theme, ``Fuel Cells: Realizing the Potential,`` focuses on progress being made toward commercial manufacture and use of fuel cell products. Fuel cell power plants are competing for market share in some applications and demonstrations of market entry power plants are proceeding for additional applications. Development activity on fuel cells for transportation is also increasing; fuel cell products have potential in energy and transportation industries, with very favorable environmental impacts. This Seminar has the purpose of fostering communication by providing a forum for the international community interested in development, application, and business opportunities related fuel cells. Over 190 technical papers are included, the majority being processed for the data base.

  10. Skidder load capacity and fuel consumption HP-41C program

    Science.gov (United States)

    Ross A. Phillips

    1983-01-01

    This program gives the log weight that the skidder can move and gives fuel consumption either in liters or gallons per turn. Slope of the skid trail, skidder weight, and skid distance must be entered into the program.

  11. PBF severe fuel damage program: results and comparison to analysis

    International Nuclear Information System (INIS)

    McDonald, P.E.; Buescher, B.J.; Gruen, G.E.; Hobbins, R.R.; McCardell, R.K.

    1983-01-01

    The United States Nuclear Regulatory Commission has initiated a severe fuel damage research program in the Power Burst Facility (PBF) to investigate fuel rod and core response, and fission product and hydrogen release and transport under degraded core cooling accident conditions. This paper presents a description of Phase I of the PBF Severe Fuel Damage Program, discusses the results of the first experiment, and compares those results with analysis performed using the TRAC-BD1 computer code

  12. 2015 Annual Progress Report: DOE Hydrogen and Fuel Cells Program

    Energy Technology Data Exchange (ETDEWEB)

    None

    2015-12-23

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

  13. 2016 Annual Progress Report: DOE Hydrogen and Fuel Cells Program

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2017-03-09

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

  14. Spent fuel disassembly and canning programs at the Barnwell Nuclear Fuel Plant (BNFP)

    International Nuclear Information System (INIS)

    Townes, G.A.

    1979-10-01

    Methods of disassembling and canning spent fuel to allow more efficient storage are being investigated at the BNFP. Studies and development programs are aimed at dry disassembly of fuel to allow storage and shipment of fuel pins rather than full fuel assemblies. Results indicate that doubling existing storage capacity or tripling the carrying capacity of existing transportation equipment is achievable. Disassembly could be performed in the BNFP hot cells at rates of about 12 to 15 assemblies per day

  15. Romanian nuclear fuel program: past, present and future

    International Nuclear Information System (INIS)

    Budan, O.; Rotaru, I.; Galeriu, C.A.

    1997-01-01

    The paper presents and comments the policy adopted in Romania for the production of CANDU-6 nuclear fuel before and after 1990. In this paper the word 'past' refers to the period before 1990 and 'present' to the 1990-1997 period. The CANDU-6 nuclear fuel manufacturing started in Romania in December 1983. Neither AECL nor any Canadian nuclear fuel manufacturer were involved in the Romanian industrial nuclear fuel production before 1990. After January 1990, the new created Romanian Electricity Authority (RENEL) assumed the responsibility for the Romanian Nuclear Power Program. It was RENEL's decision to stop, in June 1990, the nuclear fuel production at the Institute for Nuclear Power Reactors (IRNE) Pitesti. This decision was justified by the Canadian specialists team findings, revealed during a general, but well enough technically founded analysis performed at IRNE in the spring of 1990. All fuel manufactured before June 1990 was quarantined as it was considered of suspect quality. By that time more than 31,000 fuel bundles had already been manufactured. This fuel was stored for subsequent assessment. The paper explains the reasons which provoked this decision. The paper also presents the strategy adopted by RENEL after 1990 regarding the Romanian Nuclear Fuel Program. After a complex program done by Romanian and Canadian partners, in November 1994, AECL issued a temporary certification for the Romanian nuclear fuel plant. During the demonstration manufacturing run, as an essential milestone for the qualification of the Romanian fuel supplier for CANDU-6 reactors, 202 fuel bundles were produced. Of these fuel bundles, 66 were part of the Cernavoda NGS Unit 1 first fuel load (the balance was supplied by Zircatec Precision Industries Inc. - ZPI). The industrial nuclear fuel fabrication re-started in Romania in January 1995 under AECL's periodical monitoring. In December 1995, AECL issued a permanent certificate, stating the Romanian nuclear fuel plant as a qualified

  16. Update on the Vancouver Fuel Cell Vehicle Program

    International Nuclear Information System (INIS)

    Rothwell, B.

    2004-01-01

    'Full text:' The Vancouver Fuel Cell Vehicle Program (VFCVP) is a $5.8 million initiative designed to test four Ford Focus Fuel Cell Vehicles for three years in the Lower Mainland of British Columbia. The project is the first of its kind in Canada and is led by Fuel Cells Canada (FCC), the Ford Motor Company (Ford), and the Governments of Canada and British Columbia. This presentation will provide program details and an update on activities leading up to currently planned delivery to Vancouver in November 2004. The VFCVP will test the performance, durability and reliability of the Ford fuel cell vehicle cars in real-world conditions and will examine fuelling issues and solutions, the reduction of greenhouse gas emissions and public acceptance of hydrogen fuel cell vehicles. The program will generate data to help evolve the technology and develop international codes and standards E cents Epnd the implementation and adoption of fuel cell technology. (author)

  17. Methanol supply issues for alternative fuels demonstration programs

    International Nuclear Information System (INIS)

    Teague, J.M.; Koyama, K.K.

    1995-01-01

    This paper surveys issues affecting the supply of fuel-grade methanol for the California Energy Commission's alternative fuels demonstration programs and operations by other public agencies such as transit and school districts. Establishing stable and reasonably priced sources of methanol (in particular) and of alternative fuels generally is essential to their demonstration and commercialization. Development both of vehicle technologies and of fuel supply and distribution are complementary and must proceed in parallel. However, the sequence of scaling up supply and distribution is not necessarily smooth; achievement of volume thresholds in demand and through-put of alternative fuels are marked by different kinds of challenges. Four basic conditions should be met in establishing a fuel supply: (1) it must be price competitive with petroleum-based fuels, at least when accounting for environmental and performance benefits; (2) bulk supply must meet volumes required at each phase; necessitating resilience among suppliers and a means of designating priority for high value users; (3) distribution systems must be reliable, comporting with end users' operational schedules; (4) volatility in prices to the end user for the fuel must be minimal. Current and projected fuel volumes appear to be insufficient to induce necessary economies of scale in production and distribution for fuel use. Despite their benefits, existing programs will suffer absent measures to secure economical fuel supplies. One solution is to develop sources that are dedicated to fuel markets and located within the end-use region

  18. Status of the Canadian Nuclear Fuel Waste Management Program

    International Nuclear Information System (INIS)

    Lyon, R.B.

    1985-10-01

    The Canadian Nuclear Fuel Waste Management Program is in the fifth year of a ten-year generic research and development phase. The major objective of this phase of the program is to assess the basic safety and environmental aspects of the concept of isolating immobilized fuel waste by deep underground disposal in plutonic rock. The major scientific and engineering components of the program, namely immobilization studies, geoscience research, and environmental and safety assessment, are well established

  19. 78 FR 62462 - Regulation of Fuels and Fuel Additives: Modifications to Renewable Fuel Standard Program

    Science.gov (United States)

    2013-10-22

    ... renewable fuel is defined as fuel produced from renewable biomass that is used to replace or reduce the quantity of fossil fuel present in home heating oil or jet fuel.\\3\\ In essence, additional renewable fuel... of ``home heating oil.'' EPA determined that this term was ambiguous, and defined it by incorporating...

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

    International Nuclear Information System (INIS)

    Hakl, A.R.

    1981-10-01

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

  1. Canadian fuel development program and recent operational experience

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  2. Review of the nuclear fuel waste management program

    International Nuclear Information System (INIS)

    Hatcher, S.R.

    1980-06-01

    Progress over the previous year in the nuclear fuel waste management program is reviewed. Universities, industry and consultants have become increasingly involved, and the work is being overseen by a Technical Advisory Committee. The program has also been investigated by Ontario's Porter Commission and Select Committe on Ontario Hydro Affairs. A public information program has been extended to cover most of the Canadian Shield region of Ontario. Ontario Hydro is studying spent fuel storage and transportation, while AECL is covering immobilization of spent fuel or processing wastes, geotechnical and geochemical research in the laboratory and in the field, design of disposal facilities, and environmental and safety assessments. (L.L.)

  3. U-Mo fuel qualification program in HANARO

    International Nuclear Information System (INIS)

    Lee, K.H.; Lee, C.S.; Kim, H.R.; Kuk, I.H.; Kim, C.K.

    2000-01-01

    Atomized U-Mo fuel has shown good performance from the results of previous out-of-pile tests and post-irradiation examinations. A qualification program of rod type U-Mo fuel is in progress and the fuel will be irradiated in HANARO. 6 gU/cm 3 U-7Mo, U-8Mo and U-9Mo are considered in this program. The laboratory test results of porosity, mechanical property, thermal conductivity, and thermal compatibility test are discussed in this paper. In parallel with this qualification program, the feasibility study on the core conversion from the present U 3 Si fuel to U-Mo in HANARO will be initiated to provide technical bases for the policy making. Several options of core conversion for HANARO are proposed and each option will be addressed briefly in terms of the operation policy, fuel management, and licensing of HANARO. (author)

  4. The French development program for a UMo fuel

    International Nuclear Information System (INIS)

    Romano, R.; Nigon, J.L.; Languille, A.; Le Borgne, E.; Freslon, H.

    1999-01-01

    Until now high density U 3 Si 2 fuels were satisfactory for LEU conversion of certain reactors, but their use is limited because their density is physically limited to 5,8 gU/cm3 and they have very poor reprocessing capacities. After the end of the present US return policy in may 2006, the reactor operators will be indeed in a very difficult position with silicides. The international community is thus interested in a very high density fuel with good reprocessing capacities in order to convert most reactors and to find a back end solution. In France, CEA, CERCA, and COGEMA have thus launched an important program in order to sort potential candidates of uranium alloys. UMo is one of the most interesting candidates. After the selection of UMo alloys, France has pooled different skills to start an important program on UMo fuels: CEA has started an important project for a new reactor (Jules Horowitz); CERCA is the main manufacturer for MTR fuel; TECHNICATOME is the design expert for research reactors and associated cores; FRAMATOME is the parent company of CERCA and is interested in the development of new reactors; COGEMA is interested in reprocessing spent fuels. This new fuel has three aims: to allow reactors to benefit from a high performing fuel; to have a reprocessable fuel to limit the fuel storage period and the associate safety problem, and solve the back end issue; to support the international effort for non proliferation involving the end of the use of HEU. This high density fuel will decrease the number of fuel assemblies needed to run the reactors and decrease the global cost of the fuel cycle as the back end management cost is in proportion with the quantity of fuel. Reactor operators will thus derive an advantage from this new fuel, in terms of economy

  5. 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...

  6. 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...

  7. 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...

  8. Computerised programming of the Dragon reactor fuel handling operations

    International Nuclear Information System (INIS)

    Butcher, P.

    1976-11-01

    Two suites of FORTRAN IV computer programs have been written to produce check lists for the operation of the two remote control fuel handling machines of the Dragon Reactor. This document describes the advantages of these programs over the previous manual system of writing check lists, and provides a detailed guide to the programs themselves. (author)

  9. 75 FR 14669 - Regulation of Fuels and Fuel Additives: Changes to Renewable Fuel Standard Program

    Science.gov (United States)

    2010-03-26

    ... RINs from producers of the renewable fuel. The obligated parties do not need lead time for construction... fuels and new limits on renewable biomass feedstocks. This rulemaking marks the first time that... advanced biofuel and multiple cellulosic-based fuels with their 60% threshold. Additional fuel pathways...

  10. Status of Away From Reactor spent fuel storage program

    International Nuclear Information System (INIS)

    King, F.D.

    1979-07-01

    The Away From Reactor (AFR) Spent Fuel Program that the US Department of Energy established in 1977 is intended to preclude the shutting down of commercial nuclear power reactors because of lack of storage space for spent fuel. Legislation now being considered by Congress includes plans to provide storage space for commercial spent fuel beginning in 1983. Utilities are being encouraged to provide as much storage space as possible in their existing storage facilities, but projections indicate that a significant amount of AFR storage will be required. The government is evaluating the use of both existing and new storage facilities to solve this forecasted storage problem for commercial spent fuel

  11. The US Army Foreign Comparative Test fuel cell program

    Science.gov (United States)

    Bostic, Elizabeth; Sifer, Nicholas; Bolton, Christopher; Ritter, Uli; Dubois, Terry

    The US Army RDECOM initiated a Foreign Comparative Test (FCT) Program to acquire lightweight, high-energy dense fuel cell systems from across the globe for evaluation as portable power sources in military applications. Five foreign companies, including NovArs, Smart Fuel Cell, Intelligent Energy, Ballard Power Systems, and Hydrogenics, Inc., were awarded competitive contracts under the RDECOM effort. This paper will report on the status of the program as well as the experimental results obtained from one of the units. The US Army has interests in evaluating and deploying a variety of fuel cell systems, where these systems show added value when compared to current power sources in use. For low-power applications, fuel cells utilizing high-energy dense fuels offer significant weight savings over current battery technologies. This helps reduce the load a solider must carry for longer missions. For high-power applications, the low operating signatures (acoustic and thermal) of fuel cell systems make them ideal power generators in stealth operations. Recent testing has been completed on the Smart Fuel Cell A25 system that was procured through the FCT program. The "A-25" is a direct methanol fuel cell hybrid and was evaluated as a potential candidate for soldier and sensor power applications.

  12. GCFR Fuels and Materials Program at Argonne National Laboratory

    International Nuclear Information System (INIS)

    Neimark, L.A.; Greenberg, S.; Johnson, C.E.; Purohit, A.; Liu, Y.Y.; Rest, J.; Reimann, K.J.; McLennan, G.A.

    1979-01-01

    The F-5 fuel-pin irradiation experiment in EBR-II is a cornerstone of the GCFR program. It is the largest-scale fuel-pin experiment in the present program and will provide data on the performance of pins and a pin-support structure that are prototypic of the GCFR Demonstration Plant. The fuel pins are presently undergoing interim examination after successfully achieving 4.6 at.% burnup. Studies of the thermodynamics and kinetics of the U--Cs--O system, supplemented by analysis of the results of previously irradiated fuel pins, have led to the incorporation of fuel-design modifications in the F-5 experiment to insure adequate performance of the vented fuel. The effect of ribbing, as well as the ribbing process, on the short- and long-term structural performance of fuel-pin cladding is being evaluated via in-reactor and out-of-reactor tests and with the fuel-element modeling code LIFE-GCFR and the finite element program, ADINA

  13. Program on fuels for transmutation: present status and prospects

    Energy Technology Data Exchange (ETDEWEB)

    Rouault, J.; Garnier, J.C.; Chauvin, N.; Pillon, S. [CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. d' Etudes des Combustibles

    2001-07-01

    The performance calculations of appropriate fuel cycle facilities and reactor configurations (scenarios) relying on current reactor technologies (Pressurized Water Reactor and Fast neutrons Reactors) or innovative reactors (Accelerator Driven Systems) have proved the scientific feasibility of some P and T strategies. To insure the technological feasibility, a large program on fuels and materials is underway, including advanced concepts for PWRs and the development of specific targets (dispersed fuels) for transmutation in Fast Reactors. Experiments in different reactors including Phenix are being prepared. The program is presented and recent results are given. (author)

  14. Program on fuels for transmutation: present status and prospects

    International Nuclear Information System (INIS)

    Rouault, J.; Garnier, J.C.; Chauvin, N.; Pillon, S.

    2001-01-01

    The performance calculations of appropriate fuel cycle facilities and reactor configurations (scenarios) relying on current reactor technologies (Pressurized Water Reactor and Fast neutrons Reactors) or innovative reactors (Accelerator Driven Systems) have proved the scientific feasibility of some P and T strategies. To insure the technological feasibility, a large program on fuels and materials is underway, including advanced concepts for PWRs and the development of specific targets (dispersed fuels) for transmutation in Fast Reactors. Experiments in different reactors including Phenix are being prepared. The program is presented and recent results are given. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-09-01

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

  16. On-site fuel cell field test support program

    Science.gov (United States)

    Staniunas, J. W.; Merten, G. P.

    1982-01-01

    In order to assess the impact of grid connection on the potential market for fuel cell service, applications studies were conducted to identify the fuel cell operating modes and corresponding fuel cell sizing criteria which offer the most potential for initial commercial service. The market for grid-connected fuel cell service was quantified using United's market analysis program and computerized building data base. Electric and gas consumption data for 268 buildings was added to our surveyed building data file, bringing the total to 407 buildings. These buildings were analyzed for grid-isolated and grid-connected fuel cell service. The results of the analyses indicated that the nursing home, restaurant and health club building sectors offer significant potential for fuel cell service.

  17. The Canadian long-term experimental used fuel storage program

    International Nuclear Information System (INIS)

    Wasywich, K.M.; Taylor, P.

    1993-01-01

    The Canadian experimental fuel storage program consists of four components: (1) storage of used CANDU (CANadian Deuterium Uranium, registered trademark of AECL) fuel under water, with periodic examination; (2) storage of used CANDU fuel in dry air at seasonally varying temperatures, and in both dry and moisture-saturated air at 150 C, also with periodic examination; (3) underlying research on the oxidation of unused and used UO 2 in dry and moist air at temperatures up to 300 C; and (4) modeling of UO 2 oxidation in dry air. The primary objective of the fuel-storage experiments is to investigate the stability of used CANDU fuel during long-term storage. Burnup of the fuel in these experiments ranges from ∼43 to 582 MW h/kg U, while the outer-element linear power ratings range from 22 to 79 kW/m. The storage behavior of intact and intentionally defected fuel, and fuel that defected in-reactor, is being investigated in the above experiments. Since differences in UO 2 oxidation behavior were observed between dry-air, moisture-saturated air and wet storage of intentionally defected used CANDU fuel, underlying research was initiated on oxidation of unused and used fuel to develop a better understanding of the different mechanisms. Modeling of UO 2 oxidation based on the results of the dry-storage experiments is also under way

  18. Results of 200 KW fuel cell evaluation programs

    Energy Technology Data Exchange (ETDEWEB)

    Torrey, J.M.; Merten, G.P. [SAIC, San Diego, CA (United States); Binder, M.J. [Army Construction Engineering Research Labs., Champaign, IL (United States)] [and others

    1996-12-31

    Science Applications International Corporation (SAIC) has installed six monitoring systems on ONSI Corporation 200 kW phosphoric acid fuel cells. Three of the systems were installed for the U.S. Army Construction Engineering Research Laboratories (USACERL) which is coordinating the Department of Defense (DoD) fuel cell Demonstration Program and three were installed under a contract with the New York State Energy Research and Development Authority (NYSERDA). Monitoring of the three NYSERDA sites has been completed. Monitoring systems for the DoD fuel cells were installed in August, 1996 and thus no operating data was available at the time of this writing, but will be presented at the Fuel Cell Seminar. This paper will present the monitoring configuration and research approach for each program. Additionally, summary performance data is presented for the completed NYSERDA program.

  19. Segmented fuel irradiation program: investigation on advanced materials

    International Nuclear Information System (INIS)

    Uchida, H.; Goto, K.; Sabate, R.; Abeta, S.; Baba, T.; Matias, E. de; Alonso, J.

    1999-01-01

    The Segmented Fuel Irradiation Program, started in 1991, is a collaboration between the Japanese organisations Nuclear Power Engineering Corporation (NUPEC), the Kansai Electric Power Co., Inc. (KEPCO) representing other Japanese utilities, and Mitsubishi Heavy Industries, Ltd. (MHI); and the Spanish Organisations Empresa Nacional de Electricidad, S.A. (ENDESA) representing A.N. Vandellos 2, and Empresa Nacional Uranio, S.A. (ENUSA); with the collaboration of Westinghouse. The objective of the Program is to make substantial contribution to the development of advanced cladding and fuel materials for better performance at high burn-up and under operational power transients. For this Program, segmented fuel rods were selected as the most appropriate vehicle to accomplish the aforementioned objective. Thus, a large number of fuel and cladding combinations are provided while minimising the total amount of new material, at the same time, facilitating an eventual irradiation extension in a test reactor. The Program consists of three major phases: phase I: design, licensing, fabrication and characterisation of the assemblies carrying the segmented rods (1991 - 1994); phase II: base irradiation of the assemblies at Vandellos 2 NPP, and on-site examination at the end of four cycles (1994-1999). Phase III: ramp testing at the Studsvik facilities and hot cell PIE (1996-2001). The main fuel design features whose effects on fuel behaviour are being analysed are: alloy composition (MDA and ZIRLO vs. Zircaloy-4); tubing texture; pellet grain size. The Program is progressing satisfactorily as planned. The base irradiation is completed in the first quarter of 1999, and so far, tests and inspections already carried out are providing useful information on the behaviour of the new materials. Also, the Program is delivering a well characterized fuel material, irradiated in a commercial reactor, which can be further used in other fuel behaviour experiments. The paper presents the main

  20. Hydrogen, Fuel Cells & Infrastructure Technologies Program

    Energy Technology Data Exchange (ETDEWEB)

    2005-03-01

    This plan details the goals, objectives, technical targets, tasks and schedule for EERE's contribution to the DOE Hydrogen Program. Similar detailed plans exist for the other DOE offices that make up the Hydrogen Program.

  1. The KFB Program on Biobased Fuels for Vehicles

    International Nuclear Information System (INIS)

    1996-12-01

    KFB supports research and demonstration projects for bio-based transport fuels, alcohols and biogas. The program started in 1991 and will continue through 1997. The program focuses on heavy vehicles, e.g. buses for public transportation. Projects and intermediate results are described in the brochure. Information is also available at the KFB homepage. //www.kfb.se

  2. 2013 Annual Progress Report: DOE Hydrogen and Fuel Cells Program

    Energy Technology Data Exchange (ETDEWEB)

    none,

    2013-12-01

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

  3. 2014 Annual Progress Report: DOE Hydrogen and Fuel Cells Program

    Energy Technology Data Exchange (ETDEWEB)

    none,

    2014-11-01

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

  4. 2011 Annual Progress Report: DOE Hydrogen and Fuel Cells Program

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-11-01

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

  5. 2016 Annual Progress Report: DOE Hydrogen and Fuel Cells Program

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-02-01

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

  6. 2015 Annual Progress Report: DOE Hydrogen and Fuel Cells Program

    Energy Technology Data Exchange (ETDEWEB)

    Popovich, Neil

    2015-12-01

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

  7. 2012 Annual Progress Report: DOE Hydrogen and Fuel Cells Program

    Energy Technology Data Exchange (ETDEWEB)

    2012-12-01

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

  8. Operations monitoring concept. Consolidated Fuel Reprocessing Program

    International Nuclear Information System (INIS)

    Kerr, H.T.

    1985-01-01

    Operations monitoring is a safeguards concept which could be applied in future fuel cycle facilities to significantly enhance the effectiveness of an integrated safeguards system. In general, a variety of operations monitoring techniques could be developed for both international and domestic safeguards application. The goal of this presentation is to describe specific examples of operations monitoring techniques as may be applied in a fuel reprocessing facility. The operations monitoring concept involves monitoring certain in-plant equipment, personnel, and materials to detect conditions indicative of the diversion of nuclear material. An operations monitoring subsystem should be designed to monitor operations only to the extent necessary to achieve specified safeguards objectives; there is no intent to monitor all operations in the facility. The objectives of the operations monitoring subsystem include: verification of reported data; detection of undeclared uses of equipment; and alerting the inspector to potential diversion activities. 1 fig

  9. 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)

  10. Yugoslav spent nuclear fuel management program and international perspectives

    International Nuclear Information System (INIS)

    Pesic, M.; Subotic, K.; Sotic, O.; Plecas, I.; Ljubenov, V.; Peric, A.; Milosevic, M.

    2002-01-01

    Spent nuclear fuel stored in the Vinca Institute of Nuclear Sciences, Yugoslavia, consists of about 2.5 tons of metal uranium (initial enrichment 2%) and about 20 kg uranium dioxide (dispersed in aluminum matrix, initial fuel uranium enrichment 80%). This spent nuclear fuel is generated in operation of the RA heavy water research reactor during 1959-1984 period. Both types of fuel are of ex-USSR origin, have the same shape and dimensions and approximately the same initial mass of 235 nuclide. They are known as the TVR-S type of fuel elements. The total of 8030 spent fuel elements are stored at the RA research reactor premises, almost all in the spent fuel pool filled by ordinary water. The last used 480 high-enriched uranium spent fuel elements are kept in the drained RA reactor core since 1984. Fuel layer of both enrichments is covered with thin aluminium cladding. Due to non-suitable chemical parameters of water in the spent fuel storage pool, the corrosion processes penetrated aluminium cladding and aluminium walls od storage containers during storage period long from 20 to 40 years. Activity of fission products ( 137 Cs) is detected in water samples during water inspection in 1996 and experts of the lAEA Russia and USA were invited to help. By end of 2001, some remediation of the water transparency of the storage pool and inspections of water samples taken from the storage containers with the spent fuel elements were carried out by the Vinca Institute staff and with the help of experts from the Russia and the IAEA. Following new initiatives on international perspective on spent fuel management, a proposal was set by the IAEA, and was supported by the governments of the USA and the Russian Federation to ship the spent fuel elements of the RA research reactor to Mayak spent fuel processing plant in Russia. This paper describes current status of the reactor RA spent fuel elements, initiative for new Yugoslav spent fuel management program speculates on some of the

  11. Power Burst Facility severe-fuel-damage test program

    International Nuclear Information System (INIS)

    McCardell, R.K.; MacDonald, P.E.

    1982-01-01

    As a result of the Three Mile Island Unit 2 (TMI-2) accident, the United States Nuclear Regulatory Commission (USNRC) has initiated a severe fuel damage research program to investigate fuel rod and core response, and fission product and hydrogen release and transport during degraded core cooling accidents. This paper presents a discussion of the expected benefits of the PBF severe fuel damage tests to the nuclear industry, a description of the first five planned experiments, the results of pretest analysis performed to predict the fuel bundle heatup for the first two experiments, and a discussion of Phase II severe fuel damage experiments. Modifications to the fission product detection system envisioned for the later experiments are also described

  12. Management of radioactive fuel wastes: the Canadian disposal program

    International Nuclear Information System (INIS)

    Boulton, J.

    1978-10-01

    This report describes the research and development program to verify and demonstrate the concepts for the safe, permanent disposal of radioactive fuel wastes from Canadian nuclear reactors. The program is concentrating on deep underground disposal in hard-rock formations. The nature of the radioactive wastes is described, and the options for storing, processing, packaging and disposing of them are outlined. The program to verify the proposed concept, select a suitable site and to build and operate a demonstration facility is described. (author)

  13. The EBR-II spent fuel treatment program

    International Nuclear Information System (INIS)

    Lineberry, M.J.; McFarlane, H.F.

    1995-01-01

    Argonne National Laboratory has refurbished and equipped an existing hot cell facility for demonstrating a high-temperature electrometallurgical process for treating spent nuclear fuel from the Experimental Breeder Reactor-11. Two waste forms will be produced and qualified for geologic disposal of the fission and activation products. Relatively pure uranium will be separated for storage. Following additional development, transuranium elements will be blended into one of the high-level waste streams. The spent fuel treatment program will help assess the viability of electrometallurgical technology as a spent fuel management option

  14. 40 CFR 80.610 - What acts are prohibited under the diesel fuel sulfur program?

    Science.gov (United States)

    2010-07-01

    ... AGENCY (CONTINUED) AIR PROGRAMS (CONTINUED) REGULATION OF FUELS AND FUEL ADDITIVES Motor Vehicle Diesel... supply, store or transport motor vehicle diesel fuel, NRLM diesel fuel, ECA marine fuel or heating oil... transport any diesel fuel for use in motor vehicle or nonroad engines that contains greater than 0.10...

  15. DOE-owned spent nuclear fuel program plan

    International Nuclear Information System (INIS)

    1995-11-01

    The Department of Energy (DOE) has produced spent nuclear fuel (SNF) for many years as part of its various missions and programs. The historical process for managing this SNF was to reprocess it whereby valuable material such as uranium or plutonium was chemically separated from the wastes. These fuels were not intended for long-term storage. As the need for uranium and plutonium decreased, it became necessary to store the SNF for extended lengths of time. This necessity resulted from a 1992 DOE decision to discontinue reprocessing SNF to recover strategic materials (although limited processing of SNF to meet repository acceptance criteria remains under consideration, no plutonium or uranium extraction for other uses is planned). Both the facilities used for storage, and the fuel itself, began experiencing aging from this extended storage. New efforts are now necessary to assure suitable fuel and facility management until long-term decisions for spent fuel disposition are made and implemented. The Program Plan consists of 14 sections as follows: Sections 2--6 describe objectives, management, the work plan, the work breakdown structure, and the responsibility assignment matrix. Sections 7--9 describe the program summary schedules, site logic diagram, SNF Program resource and support requirements. Sections 10--14 present various supplemental management requirements and quality assurance guidelines

  16. DOE-owned spent nuclear fuel program plan

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-11-01

    The Department of Energy (DOE) has produced spent nuclear fuel (SNF) for many years as part of its various missions and programs. The historical process for managing this SNF was to reprocess it whereby valuable material such as uranium or plutonium was chemically separated from the wastes. These fuels were not intended for long-term storage. As the need for uranium and plutonium decreased, it became necessary to store the SNF for extended lengths of time. This necessity resulted from a 1992 DOE decision to discontinue reprocessing SNF to recover strategic materials (although limited processing of SNF to meet repository acceptance criteria remains under consideration, no plutonium or uranium extraction for other uses is planned). Both the facilities used for storage, and the fuel itself, began experiencing aging from this extended storage. New efforts are now necessary to assure suitable fuel and facility management until long-term decisions for spent fuel disposition are made and implemented. The Program Plan consists of 14 sections as follows: Sections 2--6 describe objectives, management, the work plan, the work breakdown structure, and the responsibility assignment matrix. Sections 7--9 describe the program summary schedules, site logic diagram, SNF Program resource and support requirements. Sections 10--14 present various supplemental management requirements and quality assurance guidelines.

  17. Overview of the fast reactors fuels program

    International Nuclear Information System (INIS)

    Evans, E.A.; Cox, C.M.; Hayward, B.R.; Rice, L.H.; Yoshikawa, H.H.

    1980-04-01

    Each nation involved in LMFBR development has its unique energy strategies which consider energy growth projections, uranium resources, capital costs, and plant operational requirements. Common to all of these strategies is a history of fast reactor experience which dates back to the days of the Manhatten Project and includes the CLEMENTINE Reactor, which generated a few watts, LAMPRE, EBR-I, EBR-II, FERMI, SEFOR, FFTF, BR-1, -2, -5, -10, BOR-60, BN-350, BN-600, JOYO, RAPSODIE, Phenix, KNK-II, DFR, and PFR. Fast reactors under design or construction include PEC, CRBR, SuperPhenix, SNR-300, MONJU, and Madras (India). The parallel fuels and materials evolution has fully supported this reactor development. It has involved cermets, molten plutonium alloy, plutonium oxide, uranium metal or alloy, uranium oxide, and mixed uranium-plutonium oxides and carbides

  18. Nuclear criticality safety program at the Fuel Cycle Facility

    International Nuclear Information System (INIS)

    Lell, R.M.; Fujita, E.K.; Tracy, D.B.; Klann, R.T.; Imel, G.R.; Benedict, R.W.; Rigg, R.H.

    1994-01-01

    The Fuel Cycle Facility (FCF) is designed to demonstrate the feasibility of a novel commercial-scale remote pyrometallurgical process for metallic fuels from liquid metal-cooled reactors and to show closure of the Integral Fast Reactor (IFR) fuel cycle. Requirements for nuclear criticality safety impose the most restrictive of the various constraints on the operation of FCF. The upper limits on batch sizes and other important process parameters are determined principally by criticality safety considerations. To maintain an efficient operation within appropriate safety limits, it is necessary to formulate a nuclear criticality safety program that integrates equipment design, process development, process modeling, conduct of operations, a measurement program, adequate material control procedures, and nuclear criticality analysis. The nuclear criticality safety program for FCF reflects this integration, ensuring that the facility can be operated efficiently without compromising safety. The experience gained from the conduct of this program in the Fuel cycle Facility will be used to design and safely operate IFR facilities on a commercial scale. The key features of the nuclear criticality safety program are described. The relationship of these features to normal facility operation is also described

  19. Technical basis for the proposed high efficiency nuclear fuel program

    International Nuclear Information System (INIS)

    MacDonald, P.E.; Herring, J.S.; Crawford, D.C.; Neimark, L.E.

    1999-01-01

    Greenhouse gas emissions from fossil fired electricity generating stations will dramatically increase over the next 20 years. Nuclear energy is the only fully developed technology able to supply large amounts of electricity without generation of greenhouse gases. However, the problem of noncompetitive economics and public concerns about radioactive waste disposal, safety, and nuclear weapons proliferation may prevent the reemergence of nuclear power as a preferred option for new electric energy generation in the U.S. This paper discusses a new research program to help address these issues, by developing fuel designs capable of burnup values in excess of 60 MWD/kgU. The objectives of the program are to: improve the reliability and robustness of light water reactor fuel, thereby improving safety margins; Significantly increase the energy generated by each fuel loading, thereby achieving longer operating cycles, higher capacity factors, and lower cost electric power; Significantly reduce the volume of spent nuclear fuel discharged for disposal by allowing more energy to be extracted from each fuel element prior to discharge; Develop fuel that is much more proliferation resistant. (author)

  20. 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

  1. Irradiation testing of miniature fuel plates for the RERTR program

    Energy Technology Data Exchange (ETDEWEB)

    Senn, R L; Martin, M M [Oak Ridge National Laboratory, Oak Ridge, TN 37830 (United States)

    1983-08-01

    An irradiation test facility, which provides a test bed for irradiating a variety of miniature fuel plates miniplates) for the Reduced Enrichment Research and Test Reactors (RERTR) program, has been placed into operation. The objective of these tests is to screen various candidate fuel materials as to their suitability for replacing the highly enriched uranium fuel materials currently used by the world's test and research reactors with a lower enrichment fuel material, without significantly degrading reactor operating characteristics and power levels. The use of low uranium enrichment of about 20% {sup 235}U in place of highly enriched fuel for these reactors would reduce the potential for {sup 235}U diversion. Fuel materials currently being evaluated in this first phase of these screening tests include aluminum-base dispersion-type fuel plates with fuel cores of 1) high uranium content U{sup 3}){sup 8}-Al being developed by ORNL, 2) high uranium content UAI{sub x}-Al being developed by EG and G Idaho, Inc., and 3) very high uranium content U{sub 3}Si-Al- being developed by ANL. The miniplates are 115-mm long by 50-mm wide with overall plate thicknesses of 1.27 or 1.52 mm. The fuel core dimensions vary according to overall plate thicknesses with a minimal clad thickness requirement of 0.20 mm. Sixty such miniplates (thirty of each thickness) can be irradiated in one test facility. The irradiation test facility, designated as HFED-1 is operating in core position E-7 in the Oak Ridge Research Reactor (ORR), a 30-MW water-moderated reactor. The peak neutron flux measured for this experiment is 1.96 x 10{sup 18} neutrons m{sub -2} s{sub -1}. The various types of miniplates will achieve burnups of up to approximately 2.2x10{sup 27} fissions/m{sup 3} of fuel, which will require approximately eight full power months of irradiation. During reactor shutdown periods, the experiment is removed from the reactor, moved to a special poolside station, disassembled, and inspected

  2. Fuel System: Automotive Mechanics Instructional Program. Block 4.

    Science.gov (United States)

    O'Brien, Ralph D.

    The fourth of six instructional blocks in automotive mechanics, the lessons and supportive information in the document provide a guide for teachers in planning an instructional program in automotive fuel systems at the secondary and post secondary level. The material, as organized, is a suggested sequence of instruction within each block. Each…

  3. Conservation education for Fire, Fuel and Smoke Program

    Science.gov (United States)

    Wayne Cook

    2009-01-01

    The mission of Conservation Education for the Fire, Fuel and Smoke (FFS) Program is to develop and deliver high-quality, science-based education about wildland fire to students, educators, the general public, and agency staff. Goals: 1) Increase awareness of the scope and content of FFS research. 2) Improve understanding of fundamental concepts in wildland fire science...

  4. Role of the consolidated fuel reprocessing program in the United States Breeder Reactor Program

    International Nuclear Information System (INIS)

    Ballard, W.W.; Burch, W.D.

    1980-01-01

    While present US policy precludes the commercial reprocessing of LWR fuels and the recycle of plutonium, the policy does encompass the need to continue a program to develop the technology for reprocessing breeder fuels. Some questions have again risen this year as to the pace of the entire breeder program, including recycle, and the answers are evolving. This paper and the other companion papers which describe several aspects of the Consolidated Fuel Reprocessing Program take a longer-range perspective on the total program. Whether the program is implemented in the general time frame described is dependent on future government actions dedicated to carrying out a systematic program that would permit breeders to be commercialized early in the next century

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

    Energy Technology Data Exchange (ETDEWEB)

    1980-09-01

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

  6. Westinghouse accident tolerant fuel program. Current results and future plans

    Energy Technology Data Exchange (ETDEWEB)

    Ray, Sumit; Xu, Peng; Lahoda, Edward; Hallstadius, Lars; Boylan, Frank [Westinghouse Electric Company LLC, Hopkins, SC (United States)

    2016-07-15

    This paper discusses the current status, results from initial tests, as well as the future direction of the Westinghouse's Accident Tolerant Fuel (ATF) program. The current preliminary testing is addressed that is being performed on these samples at the Massachusetts Institute of Technology (MIT) test reactor, initial results from these tests, as well as the technical learning from these test results. In the Westinghouse ATF approach, higher density pellets play a significant role in the development of an integrated fuel system.

  7. Status of Westinghouse coal-fueled combustion turbine programs

    International Nuclear Information System (INIS)

    Scalzo, A.J.; Amos, D.J.; Bannister, R.L.; Garland, R.V.

    1992-01-01

    Developing clean, efficient, cost effective coal utilization technologies for future power generation is an essential part of our National Energy Strategy. Westinghouse is actively developing power plants utilizing advanced gasification, atmospheric fluidized beds (AFB), pressurized fluidized beds (PFB), and direct firing technology through programs sponsored by the U.S. Dept. of Energy (DOE). The DOE Office of Fossil Energy is sponsoring the Direct Coal-Fired Turbine program. This paper presents the status of current and potential Westinghouse Power Generation Business Unit advanced coal-fueled power generation programs as well as commercial plans

  8. Quality assurance program plan fuel supply shutdown project

    International Nuclear Information System (INIS)

    Metcalf, I.L.

    1998-01-01

    This Quality Assurance Program plan (QAPP) describes how the Fuel Supply Shutdown (FSS) project organization implements the quality assurance requirements of HNF-MP-599, Project Hanford Quality Assurance Program Description (QAPD) and the B and W Hanford Company Quality Assurance Program Plan (QAPP), FSP-MP-004. The QAPP applies to facility structures, systems, and components and to activities (e.g., design, procurement, testing, operations, maintenance, etc.) that could affect structures, systems, and components. This QAPP also provides a roadmap of applicable Project Hanford Policies and Procedures (PHPP) which may be utilized by the FSS project organization to implement the requirements of this QAPP

  9. Program SYVAC, for stochastic assessment of nuclear fuel waste disposal

    International Nuclear Information System (INIS)

    Sherman, G.R.; Hoffman, K.J.; Donahue, D.C.

    1985-01-01

    In this paper, the computer program SYVAC, used to assess concepts for the disposal of nuclear fuel waste, is described with regard to the development approach, the basic program structure, and quality assurance. The interrelationships of these aspects are illustrated by detailed descriptions of two concepts of fundamental importance to the program: the method of selecting parameter values from input probability density functions, and the numerical evaluation of the convolution integral. Quality assurance procedures, including different types of comparisons and peer review, are presented

  10. Status of US program for disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    Smith, R.I.

    1991-04-01

    In this paper, a brief history of the United States' program for the disposal of spent nuclear fuel (SNF) and the legislative acts that have guided the program are discussed. The current plans and schedules for beginning acceptance of SNF from the nuclear utilities for disposal are described, and some of the development activities supporting the program are discussed. And finally, the viability of the SNF disposal fee presently paid into the Nuclear Waste Fund by the owners/generators of commercial SNF and high-level waste (HLW) is examined. 12 refs., 9 figs

  11. Fuel channel in-service inspection programs program design for maximum cost effectiveness

    International Nuclear Information System (INIS)

    Van den Brekel, N.C.

    1995-01-01

    Inspection is an integral part of fuel channel life management strategy. Inspection data is used to assess the state of reactor core integrity and provide the information necessary to optimize long term maintenance programs. This paper will provide an overview of the structured approach to developing fuel channel inspection programs within OHN. The inspection programs are designed to balance the resources utilized (cost, outage time, and dose expenditure) with the benefits provided by the inspection data obtained (improved knowledge of component status, degradation mechanisms and rates, etc..). The CANDU community has yet to have a fuel channel operate for a full 30 year design life. Since research programs can not fully simulate reactor operating conditions, inspections become an essential feature of the life management strategy as the components age. Inspection programs often include activities designed to develop predictive capability for long term fuel channel behaviour and provide early warning of changes in behaviour. It should be noted that although this paper addresses the design of fuel channel inspection programs, the basic principles presented can be applied to the design of inspection programs for any major power plant component or system. (author)

  12. 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)

  13. 75 FR 26049 - Regulation of Fuels and Fuel Additives: Modifications to Renewable Fuel Standard Program

    Science.gov (United States)

    2010-05-10

    ... comment period on this action. Any parties interested in commenting must do so at this time. For further... Technologies for Renewable Fuel Pathways The final RFS2 rule includes two corn ethanol pathways in Table 1 of... construction of the grandfathered facilities commenced would be contained in Sec. 80.1450(b)(vi), since Sec. 80...

  14. Commentary on the Canadian nuclear fuel waste management program

    International Nuclear Information System (INIS)

    Sheng, G.; Shemilt, L.W.

    1981-01-01

    A summary of the first formal review of the Technical Advisory Committee (TAC) to Atomic Energy of Canada Limited on the Nuclear Fuel Waste Management Program is presented. The Program is described briefly and the composition and role of TAC in relation to the Program is outlined. Salient points and major recommendations are presented from the First Annual Report of TAC in which geoscience aspects of the Program were emphasized. It is the view of the Committee that overall, the whole Waste Management Program is well conceived, that there are many impressive accomplishments of high quality, that detailed research objectives are becoming more clearly delineated, that there is growing clarification as to the most critical areas in which research needs to be accomplished and that the increasing participation by university and industry scientists and engineers is reassuring

  15. Copper produced from powder by HIP to encapsulate nuclear fuel elements

    International Nuclear Information System (INIS)

    Ekbom, L.B.; Bogegaard, S.

    1989-02-01

    In the Swedish nuclear waste mangement program, nuclear fuel elements are proposed to be encapsulated in copper canisters. To fill the space between the fuel elements two methods have been proposed. Originally lead was proposed to be cast into the canister. According to a second method the space between the fuel rods is filled with copper powder and hot isostatic pressed (HIP) to seal the canister lid and to densify the powder to homogenous copper. This latter method has the advantage that each fuel rod is individually encapsulated in a very corrosion resistant material. This investigation was performed to find out to what extent pure copper powder can be hot isosatic pressed to full density and to achieve properties comparable to that of the oxygen free high conductivity (OFHC) copper of the canister. OFHC copper was molten under helium gas protection and atomized to a fine spherical powder in a pilot plant. The powder was transfered to a glove box with an argon atmosphere. The powder was filled into a steel container, which was evacuated and sealed. HIP was done at 550 degree C and 200 MPa for one hour. The resulting copper was found to have a good ductility and mechanical properties comparable to that of ordinary copper. The constant strainrate stress corrosion test used to test the canister copper showed that the HIP-ed copper has the same good properties as OFHC copper. (authors)

  16. Quality Assurance Program Plan for SFR Metallic Fuel Data Qualification

    Energy Technology Data Exchange (ETDEWEB)

    Benoit, Timothy [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Hlotke, John Daniel [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Yacout, Abdellatif [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division

    2017-07-05

    This document contains an evaluation of the applicability of the current Quality Assurance Standards from the American Society of Mechanical Engineers Standard NQA-1 (NQA-1) criteria and identifies and describes the quality assurance process(es) by which attributes of historical, analytical, and other data associated with sodium-cooled fast reactor [SFR] metallic fuel and/or related reactor fuel designs and constituency will be evaluated. This process is being instituted to facilitate validation of data to the extent that such data may be used to support future licensing efforts associated with advanced reactor designs. The initial data to be evaluated under this program were generated during the US Integral Fast Reactor program between 1984-1994, where the data includes, but is not limited to, research and development data and associated documents, test plans and associated protocols, operations and test data, technical reports, and information associated with past United States Nuclear Regulatory Commission reviews of SFR designs.

  17. In-core fuel management programs for nuclear power reactors

    International Nuclear Information System (INIS)

    1984-10-01

    In response to the interest shown by Member States, the IAEA organized a co-ordinated research programme to develop and make available in the open domain a set of programs to perform in-core fuel management calculations. This report summarizes the work performed in the context of the CRP. As a result of this programme, complete in-core fuel management packages for three types of reactors, namely PWR's, BWR's and PHWR are now available from the NEA Data Bank. For some reactor types, these program packages are available with three levels of sophistication ranging from simple methods for educational purposes to more comprehensive methods that can be used for reactor design and operation. In addition some operating data have been compiled to allow code validation. (author)

  18. 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

  19. Fuel cycle and waste management demonstration in the IFR Program

    International Nuclear Information System (INIS)

    Lineberry, M.J.; Phipps, R.D.; Benedict, R.W.; Laidler, J.J.; Battles, J.E.; Miller, W.E.

    1992-01-01

    Argonne's National Laboratory's Integral Fast Reactor (IFR) is the main element in the US advanced reactor development program. A unique fuel cycle and waste process technology is being developed for the IFR. Demonstration of this technology at engineering scale will begin within the next year at the EBR-II test facility complex in Idaho. This paper describes the facility being readied for this demonstration, the process to be employed, the equipment being built, and the waste management approach

  20. Current status of Westinghouse tubular solid oxide fuel cell program

    Energy Technology Data Exchange (ETDEWEB)

    Parker, W.G. [Westinghouse Science and Technology Center, Pittsburgh, PA (United States)

    1996-04-01

    In the last ten years the solid oxide fuel cell (SOFC) development program at Westinghouse has evolved from a focus on basic material science to the engineering of fully integrated electric power systems. Our endurance for this cell is 5 to 10 years. To date we have successfully operated at power for over six years. For power plants it is our goal to have operated before the end of this decade a MW class power plant. Progress toward these goals is described.

  1. Nuclear fuel waste management - biosphere program highlights - 1978 to 1996

    Energy Technology Data Exchange (ETDEWEB)

    Zach, R

    1997-07-01

    The biosphere program in support of the development of the disposal concept for Canadian nuclear fuel waste since 1978 is scheduled for close-out. AECL`s Environmental Science Branch (ESB) was mainly responsible for work in this program. In order to preserve as much information as possible, this report highlights many of the key achievements of the program, particularly those related to the development of the BIOTRAC biosphere model and its supporting research. This model was used for the assessment and review of the disposal concept in an environmental impact statement (EIS). The report also treats highlights related to alternative models, external scientific/technical reviews, EIS feedback, and the international BIOMOVS model validation program. Furthermore, it highlights basic aspects of future modelling and research needs in relation to siting a disposal facility. In this, feedback from the various reviews and the EIS is taken into account. Appendices of the report include listings of key ESB staff involved in the program, all the scientific/technical reports and papers produced under the program, contracts let to outside agencies, and issues raised by various participants or intervenors during the EIS review. Although the report is concerned with close-out of the biosphere program, it also provides valuable information for a continuing program concerned with siting a disposal facility. One of the conclusions of the report is that such a program is essential for successfully siting such a facility. (author) Refs.

  2. Nuclear fuel waste management - biosphere program highlights - 1978 to 1996

    International Nuclear Information System (INIS)

    Zach, R.

    1997-07-01

    The biosphere program in support of the development of the disposal concept for Canadian nuclear fuel waste since 1978 is scheduled for close-out. AECL's Environmental Science Branch (ESB) was mainly responsible for work in this program. In order to preserve as much information as possible, this report highlights many of the key achievements of the program, particularly those related to the development of the BIOTRAC biosphere model and its supporting research. This model was used for the assessment and review of the disposal concept in an environmental impact statement (EIS). The report also treats highlights related to alternative models, external scientific/technical reviews, EIS feedback, and the international BIOMOVS model validation program. Furthermore, it highlights basic aspects of future modelling and research needs in relation to siting a disposal facility. In this, feedback from the various reviews and the EIS is taken into account. Appendices of the report include listings of key ESB staff involved in the program, all the scientific/technical reports and papers produced under the program, contracts let to outside agencies, and issues raised by various participants or intervenors during the EIS review. Although the report is concerned with close-out of the biosphere program, it also provides valuable information for a continuing program concerned with siting a disposal facility. One of the conclusions of the report is that such a program is essential for successfully siting such a facility. (author)

  3. Methods and computer programs for PWR's fuel management: Programs Sothis and Ciclon

    International Nuclear Information System (INIS)

    Aragones, J.M.; Corella, M.R.; Martinez-Val, J.M.

    1976-01-01

    Methos and computer programs developed at JEN for fuel management in PWR are discussed, including scope of model, procedures for sistematic selection of alternatives to be evaluated, basis of model for neutronic calculation, methods for fuel costs calculation, procedures for equilibrium and trans[tion cycles calculation with Soth[s and Ciclon codes and validation of methods by comparison of results with others of reference (author) ' [es

  4. The Australian Hydrogen and Fuel Cells Education Program

    International Nuclear Information System (INIS)

    Luigi Bonadio

    2006-01-01

    The next generation of engineers and scientists will face great technical, economic and political challenges to satisfy increasing demands for a secure, reliable and affordable global energy system that maintains and enhances current standards of living. The Australian Hydrogen and Fuel Cells Education Program aims to bolster the quality and relevance of primary and secondary school teaching in emerging areas of science, technology and environmental/sustainability studies using hydrogen, in its capacity as a versatile energy carrier, as the educational basis for teacher and student learning. Critical advances in specific areas of hydrogen production, distribution, storage and end-use technologies arise when students are engaged to develop and apply a broad range of disciplinary and interdisciplinary knowledge and practical skills. A comprehensive hydrogen and fuel cell technology teaching module will be developed to complement existing fuels and energy curricula across Australian schools. The pilot program will be delivered via the collaboration of nine trial schools, a broad range of technical and pedagogy experts and representatives of professional bodies and industry. The program features essential and extensive teacher consultation, a professional learning and development course, industry site visits and a dedicated research and evaluation study. This initiative aims to bolster teacher literacy and student participation in the design, construction and operation of various hydrogen and fuel cell devices and extended activities. Students will reflect on and formally present their learning experiences via several dedicated fora including an awards ceremony where outstanding performance of leading schools, teachers and student groups within the cluster will be acknowledged. (authors)

  5. The Australian Hydrogen and Fuel Cells Education Program

    Energy Technology Data Exchange (ETDEWEB)

    Luigi Bonadio [Senior Consultant Luigi Bonadio and Associates (Australia)

    2006-07-01

    The next generation of engineers and scientists will face great technical, economic and political challenges to satisfy increasing demands for a secure, reliable and affordable global energy system that maintains and enhances current standards of living. The Australian Hydrogen and Fuel Cells Education Program aims to bolster the quality and relevance of primary and secondary school teaching in emerging areas of science, technology and environmental/sustainability studies using hydrogen, in its capacity as a versatile energy carrier, as the educational basis for teacher and student learning. Critical advances in specific areas of hydrogen production, distribution, storage and end-use technologies arise when students are engaged to develop and apply a broad range of disciplinary and interdisciplinary knowledge and practical skills. A comprehensive hydrogen and fuel cell technology teaching module will be developed to complement existing fuels and energy curricula across Australian schools. The pilot program will be delivered via the collaboration of nine trial schools, a broad range of technical and pedagogy experts and representatives of professional bodies and industry. The program features essential and extensive teacher consultation, a professional learning and development course, industry site visits and a dedicated research and evaluation study. This initiative aims to bolster teacher literacy and student participation in the design, construction and operation of various hydrogen and fuel cell devices and extended activities. Students will reflect on and formally present their learning experiences via several dedicated fora including an awards ceremony where outstanding performance of leading schools, teachers and student groups within the cluster will be acknowledged. (authors)

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

    Science.gov (United States)

    Kitzner, E. W.

    1978-01-01

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

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

    International Nuclear Information System (INIS)

    Dixon, R.S.

    1986-08-01

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

  8. Pathways to Commercial Success. Technologies and Products Supported by the Fuel Cell Technologies Program

    Energy Technology Data Exchange (ETDEWEB)

    none,

    2010-08-01

    This report identifies the commercial and near-commercial (emerging) hydrogen and fuel cell technologies and products that resulted from Department of Energy support through the Fuel Cell Technologies Program in the Office of Energy Efficiency and Renewable Energy.

  9. National Jet Fuels Combustion Program – Area #3 : Advanced Combustion Tests

    Science.gov (United States)

    2017-12-31

    The goal of this study is to develop, conduct, and analyze advanced laser and optical measurements in the experimental combustors developed under ASCENT National Fuel Combustion Program to measure sensitivity to fuel properties. We conducted advanced...

  10. Ethanol Research : Alternative Fuels & Life-Cycle Engineering Program : November 29, 2006 to November 28, 2011

    Science.gov (United States)

    2011-12-20

    This report presents the results of the successful ethanol fuel demonstration program conducted from September 2007 to September 2010. This project was a part of the U.S. Department of Transportation (DOT) Alternative Fuels and Life Cycle Engineering...

  11. 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

  12. Post-irradiation examination and R and D programs using irradiated fuels at KAERI

    International Nuclear Information System (INIS)

    Chun, Yong Bum; Min, Duck Kee; Kim, Eun Ka and others

    2000-12-01

    This report describes the Post-Irradiation Examination(PIE) and R and D programs using irradiated fuels at KAERI. The objectives of post-irradiation examination (PIE) for the PWR irradiated fuels, CANDU fuels, HANARO fuels and test fuel materials are to verify the irradiation performance and their integrity as well as to construct a fuel performance data base. The comprehensive utilization program of the KAERI's post-irradiation examination related nuclear facilities such as Post-Irradiation Examination Facility (PIEF), Irradiated Materials Examination Facility (IMEF) and HANARO is described

  13. Post-irradiation examination and R and D programs using irradiated fuels at KAERI

    International Nuclear Information System (INIS)

    Chun, Yong Bum; So, Dong Sup; Lee, Byung Doo; Lee, Song Ho; Min, Duck Kee

    2001-09-01

    This report describes the Post-Irradiation Examination(PIE) and R and D programs using irradiated fuels at KAERI. The objectives of post-irradiation examination (PIE) for the PWR irradiated fuels, CANDU fuels, HANARO fuels and test fuel materials are to verify the irradiation performance and their integrity as well as to construct a fuel performance data base. The comprehensive utilization program of the KAERI's post-irradiation examination related nuclear facilities such as Post-Irradiation Examination Facility (PIEF), Irradiated Materials Examination Facility (IMEF) and HANARO is described

  14. Update on Canada's nuclear fuel waste management program

    International Nuclear Information System (INIS)

    Allan, C.J.

    1992-09-01

    The Canadian Nuclear Fuel Waste Management Program (CNFWMP) was launched in 1978 as a joint initiative by the governments of Canada and Ontario. Under the program, AECL has been developing and assessing a generic concept to dispose of nuclear fuel waste in plutonic rock of the Canadian Shield. The disposal concept has been referred for review under the Environmental Assessment and Review Process. AECL will submit an Environmental Impact Statement (EIS) to an Environmental Assessment Panel, which was appointed in late 1989. Hearings will be held in areas that have a particular interest in the concept and its application. At the end of the review, the Panel will make recommendations as to the acceptability of the concept and the course of future action. The federal government will decide on the next steps to be taken. In the spring of 1990 public open houses were held to tell prospective participants how to enter the process. Sessions designed to assist the Panel in determining the scope of the EIS took place in the autumn of 1990. In June 1991 the Panel issued for comment a set of draft guidelines for the EIS. More than 30 groups and individuals submitted comments. The final guidelines were issued in March 1992, and AECL expects to submit its EIS to the Panel in 1993. If the concept review is completed by 1995 and if the concept is approved, disposal could begin some time after 2025. (L.L.) (12 refs.)

  15. NEPA implementation: The Department of Energy's program to manage spent nuclear fuel

    International Nuclear Information System (INIS)

    Shipler, D.B.

    1994-05-01

    The Department of Energy (DOE) is implementing the National Environmental Protection Act (NEPA) in its management of spent nuclear fuel. The DOE strategy is to address the short-term safety concerns about existing spent nuclear fuel, to study alternatives for interim storage, and to develop a long-range program to manage spent nuclear fuel. This paper discusses the NEPA process, the environmental impact statements for specific sites as well as the overall program, the inventory of DOE spent nuclear fuel, the alternatives for managing the fuel, and the schedule for implementing the program

  16. Determination of the burn-up of TRIGA fuel elements by calculation with new TRIGLAV program

    International Nuclear Information System (INIS)

    Zagar, T.; Ravnik, M.

    1996-01-01

    The results of fuel element burn-up calculations with new TRIGLAV program are presented. TRIGLAV program uses two dimensional model. Results of calculation are compared to results calculated with program, which uses one dimensional model. The results of fuel element burn-up measurements with reactivity method are presented and compared with the calculated results. (author)

  17. Characteristics of used CANDU fuel relevant to the Canadian nuclear fuel waste management program

    Energy Technology Data Exchange (ETDEWEB)

    Wasywich, K M

    1993-05-01

    Literature data on the characteristics of used CANDU power reactor fuel that are relevant to its performance as a waste form have been compiled in a convenient handbook. Information about the quantities of used fuel generated, burnup, radionuclide inventories, fission gas release, void volume and surface area, fuel microstructure, fuel cladding properties, changes in fuel bundle properties due to immobilization processes, radiation fields, decay heat and future trends is presented for various CANDU fuel designs. (author). 199 refs., 39 tabs., 100 figs.

  18. Characteristics of used CANDU fuel relevant to the Canadian nuclear fuel waste management program

    International Nuclear Information System (INIS)

    Wasywich, K.M.

    1993-05-01

    Literature data on the characteristics of used CANDU power reactor fuel that are relevant to its performance as a waste form have been compiled in a convenient handbook. Information about the quantities of used fuel generated, burnup, radionuclide inventories, fission gas release, void volume and surface area, fuel microstructure, fuel cladding properties, changes in fuel bundle properties due to immobilization processes, radiation fields, decay heat and future trends is presented for various CANDU fuel designs. (author). 199 refs., 39 tabs., 100 figs

  19. Extended burnup demonstration: reactor fuel program. Pre-irradiation characterization and summary of pre-program poolside examinations. Big Rock Point extended burnup fuel

    International Nuclear Information System (INIS)

    Exarhos, C.A.; Van Swam, L.F.; Wahlquist, F.P.

    1981-12-01

    This report is a resource document characterizing the 64 fuel rods being irradiated at the Big Rock Point reactor as part of the Extended Burnup Demonstration being sponsored jointly by the US Department of Energy, Consumers Power Company, Exxon Nuclear Company, and General Public Utilities. The program entails extending the exposure of standard BWR fuel to a discharge average of 38,000 MWD/MTU to demonstrate the feasibility of operating fuel of standard design to levels significantly above current limits. The fabrication characteristics of the Big Rock Point EBD fuel are presented along with measurement of rod length, rod diameter, pellet stack height, and fuel rod withdrawal force taken at poolside at burnups up to 26,200 MWD/MTU. A review of the fuel examination data indicates no performance characteristics which might restrict the continued irradiation of the fuel

  20. VANTAGE 5 PWR fuel assembly demonstration program at Virgil C. Summer nuclear station

    International Nuclear Information System (INIS)

    Warner, D.C.; Orr, W.L.

    1985-01-01

    VANTAGE 5 is an improved PWR fuel product designed and manufactured by Westinghouse Electric Corporation. The VANTAGE 5 fuel design features integral fuel burnable absorbers, intermediate flow mixer grids, axial blankets, high burnup capability, and a reconstitutable top nozzle. A demonstration program for this fuel design commenced in late 1984 in cycle 2 of the Virgil C. Summer Nuclear Station. Objectives for VANTAGE 5 fuel are reduced fuel cycle costs, better core operating margins, and increased design and operating flexibility. Inspections of the VANTAGE 5 demonstration assemblies are planned at each refueling outage

  1. The United States foreign research reactor spent nuclear fuel acceptance program: Proposal to modify the program

    International Nuclear Information System (INIS)

    Messick, C.E.

    2005-01-01

    The United States Department of Energy (DOE), in consultation with the Department of State (DOS), adopted the Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel in May 1996. The policy was slated to expire in May 2009. However, in October 2003, a petition requesting a program extension was delivered to the United States Secretary of Energy from a group of research reactor operators from foreign countries. In April 2004, the Secretary directed DOE undertake an analysis, as required by the National Environmental Policy Act (NEPA), to consider potential extension of the Program. On December 1, 2004, a Federal Register Notice was issued approving the program extension. This paper discusses the findings from the NEPA analysis and the potential changes in the program that may result from implementation of the proposed changes. (author)

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

    International Nuclear Information System (INIS)

    Boulton, J.

    1980-12-01

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

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

    International Nuclear Information System (INIS)

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

    1982-12-01

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

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

    International Nuclear Information System (INIS)

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

    1981-12-01

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

  5. Establishing and sustaining a technical program to achieve zero fuel failures

    International Nuclear Information System (INIS)

    Deshon, J.; Whiteside, K.; Burnham, R.

    2015-01-01

    In 2006, Chief Nuclear Officers (CNOs) of electric utilities operating 103 commercial reactors in the United States (U.S.) formally endorsed an initiative to achieve failure-free fuel performance by the end of 2010. This became known as the Zero-by- Ten Fuel Failure Initiative. The endorsement manifested during a meeting at the Institute of Nuclear Power Operations (INPO) while nuclear fuel performance trends were being reviewed. Declining fuel performance generated backing for the Initiative. This paper provides a brief review of some of the drivers that caused those trends, the Initiative elements designed to address them, and a review of the technical program that helped achieve, and thus far sustain, improved fuel performance. (author) Key Words: Nuclear Fuel, Fuel, Fuel Failures, Fuel Failure Initiative, Zero by Ten Initiative

  6. The Canadian program for management of spent fuel and high level wastes

    International Nuclear Information System (INIS)

    Barnes, R.W.; Mayman, S.A.

    A brief history and description of the nuclear power program in Canada is given. Schedules and programs are described for storing spent fuel in station fuel bays, centralized water pool storage facilities, concrete canisters, convection vaults, and rock or salt formations. High-level wastes will be retrievable initially, therefore the focus is on storage in mined cavities. The methods developed for high-level waste storage/disposal will ideally be flexible enough to accommodate spent fuel. (E.C.B.)

  7. PBF Severe Fuel-Damage Program: results and comparison to analysis

    International Nuclear Information System (INIS)

    MacDonald, P.E.; Buescher, B.J.; Hobbins, R.R.; McCardell, R.K.; Gruen, G.E.

    1983-01-01

    The United States Nuclear Regulatory Commission has initiated a severe fuel-damage research program in the Power Burst Facility (PBF) to investigate fuel-rod and core response, and fission-product and hydrogen release and transport under degraded-core-cooling accident conditions. This paper presents a description of Phase I of the PBF Severe Fuel Damage Program, discusses the results of the first experiment, and compares those results with analysis performed using the TRAC-BD1 computer code

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

    Energy Technology Data Exchange (ETDEWEB)

    none,

    2015-10-01

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

  10. 2011 DOE Hydrogen and Fuel Cells Program Annual Merit Review and Peer Evaluation Report

    Energy Technology Data Exchange (ETDEWEB)

    none,

    2011-09-01

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

  11. 2013 DOE Hydrogen and Fuel Cells Program Annual Merit Review and Peer Evaluation Report

    Energy Technology Data Exchange (ETDEWEB)

    none,

    2013-10-01

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

  12. 2014 DOE Hydrogen and Fuel Cells Program Annual Merit Review and Peer Evaluation Report

    Energy Technology Data Exchange (ETDEWEB)

    none,

    2014-10-01

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

  13. 2012 DOE Hydrogen and Fuel Cells Program Annual Merit Review and Peer Evaluation Report

    Energy Technology Data Exchange (ETDEWEB)

    none,

    2012-09-01

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

  14. Spent fuel characterization program in Jose Cabrera nuclear power plant

    International Nuclear Information System (INIS)

    Lloret, M.; Canencia, R.; Blanco, J.; POMAR, C.

    2010-01-01

    Jose Cabrera Nuclear Power Plant (NPP) is a 14x14 PWR reactor built in 1964 in Spain (160 MWe). The commercial operation started in 1969 and finished in 2006. During year 2009, 377 fuel assemblies from cycles 11 to 29 have been stored in 12 containers HI-STORM 100, and positioned in an Interim Spent Fuel Storage Installation built near the NPP. The spent fuel characterization and classification is a critical and complex activity that could impact all the storage process. As every container has a number of positions for damaged fuel, the loading plans and the quantity of containers depends on the total fuels classified as damaged. The classification of the spent fuel in Jose Cabrera has been performed on the basis of the Interim Staff Guidance ISG-1 from USNRC, 'Damaged Fuel'. As the storage system should assure thermal limitations, criticality control, retrievability, confinement and shielding for radioactive protection, the criteria analyzed for every spent fuel have been the existence/non existence of fuel leaks; damage that could affect the criticality analysis (as missing fuel pins) and any situation that could affect the future retrievability, as defects on the top nozzle. The first classification was performed based upon existing core records. If there were no indication of operating leakers during the concerned cycles and the structural integrity was adequate, the fuel was classified as intact or undamaged. When operating records indicated a fuel leaker, an additional inspection by ultrasonic testing of all the fuel in the concerned cycle was performed to determine the fuel leakers. If the examination results indicated that the fuel has cladding cracks, it was classified as damaged fuel without considering if it was a gross breach or a hairline crack. Additionally, it was confirmed that the water chemistry specifications for spent fuel pool has been fulfilled. Finally, a visual inspection before dry cask storage was performed and foreign particles were

  15. Spent fuel characterization program in Jose Cabrera nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Lloret, M.; Canencia, R. [Product Engineering, Enusa Industrias Avanzadas S.A., Santiago Rusinol 12, 28040 Madrid (Spain); Blanco, J.; POMAR, C. [Direction of Nuclear Generation, Gas Natural SDG, Avda. San Luis 77, 28033 Madrid (Spain)

    2010-07-01

    Jose Cabrera Nuclear Power Plant (NPP) is a 14x14 PWR reactor built in 1964 in Spain (160 MWe). The commercial operation started in 1969 and finished in 2006. During year 2009, 377 fuel assemblies from cycles 11 to 29 have been stored in 12 containers HI-STORM 100, and positioned in an Interim Spent Fuel Storage Installation built near the NPP. The spent fuel characterization and classification is a critical and complex activity that could impact all the storage process. As every container has a number of positions for damaged fuel, the loading plans and the quantity of containers depends on the total fuels classified as damaged. The classification of the spent fuel in Jose Cabrera has been performed on the basis of the Interim Staff Guidance ISG-1 from USNRC, 'Damaged Fuel'. As the storage system should assure thermal limitations, criticality control, retrievability, confinement and shielding for radioactive protection, the criteria analyzed for every spent fuel have been the existence/non existence of fuel leaks; damage that could affect the criticality analysis (as missing fuel pins) and any situation that could affect the future retrievability, as defects on the top nozzle. The first classification was performed based upon existing core records. If there were no indication of operating leakers during the concerned cycles and the structural integrity was adequate, the fuel was classified as intact or undamaged. When operating records indicated a fuel leaker, an additional inspection by ultrasonic testing of all the fuel in the concerned cycle was performed to determine the fuel leakers. If the examination results indicated that the fuel has cladding cracks, it was classified as damaged fuel without considering if it was a gross breach or a hairline crack. Additionally, it was confirmed that the water chemistry specifications for spent fuel pool has been fulfilled. Finally, a visual inspection before dry cask storage was performed and foreign particles

  16. Fuel utilization potential in light water reactors with once-through fuel irradiation (AWBA Development Program)

    International Nuclear Information System (INIS)

    Rampolla, D.S.; Conley, G.H.; Candelore, N.R.; Cowell, G.K.; Estes, G.P.; Flanery, B.K.; Duncombe, E.; Dunyak, J.; Satterwhite, D.G.

    1979-07-01

    Current commercial light water reactor cores operate without recylce of fuel, on a once-through fuel cycle. To help conserve the limited nuclear fuel resources, there is interest in increasing the energy yield and, hence, fuel utilization from once-through fuel irradiation. This report evaluates the potential increase in fuel utilization of light water reactor cores operating on a once-through cycle assuming 0.2% enrichment plant tails assay. This evaluation is based on a large number of survey calculations using techniques which were verified by more detailed calculations of several core concepts. It is concluded that the maximum fuel utilization which could be achieved by practical once-through pressurized light water reactor cores with either uranium or thorium is about 17 MWYth/ST U 3 O 8 (Megawatt Years Thermal per Short Ton of U 3 O 8 ). This is about 50% higher than that of current commercial light water reactor cores. Achievement of this increased fuel utilization would require average fuel burnup beyond 50,000 MWD/MT and incorporation of the following design features to reduce parasitic losses of neutrons: reflector blankets to utilize neutrons that would otherwise leak out of the core; fuel management practices in which a smaller fraction of the core is replaced at each refueling; and neutron economic reactivity control, such as movable fuel control rather than soluble boron control. For a hypothetical situation in which all neutron leakage and parasitic losses are eliminated and fuel depletion is not limited by design considerations, a maximum fuel utilization of about 20 MWYth/ST U 3 O 8 is calculated for either uranium or thorium. It is concluded that fuel utilization for comparable reactor designs is better with uranium fuel than with thorium fuel for average fuel depletions of 30,000 to 35,000 MWD/MT which are characteristic of present light water reactor cores

  17. Initiation of depleted uranium oxide and spent fuel testing for the spent fuel sabotage aerosol ratio program

    Energy Technology Data Exchange (ETDEWEB)

    Molecke, M.A.; Gregson, M.W.; Sorenson, K.B. [Sandia National Labs. (United States); Billone, M.C.; Tsai, H. [Argonne National Lab. (United States); Koch, W.; Nolte, O. [Fraunhofer Inst. fuer Toxikologie und Experimentelle Medizin (Germany); Pretzsch, G.; Lange, F. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (Germany); Autrusson, B.; Loiseau, O. [Inst. de Radioprotection et de Surete Nucleaire (France); Thompson, N.S.; Hibbs, R.S. [U.S. Dept. of Energy (United States); Young, F.I.; Mo, T. [U.S. Nuclear Regulatory Commission (United States)

    2004-07-01

    We provide a detailed overview of an ongoing, multinational test program that is developing aerosol data for some spent fuel sabotage scenarios on spent fuel transport and storage casks. Experiments are being performed to quantify the aerosolized materials plus volatilized fission products generated from actual spent fuel and surrogate material test rods, due to impact by a high energy density device, HEDD. The program participants in the U.S. plus Germany, France, and the U.K., part of the international Working Group for Sabotage Concerns of Transport and Storage Casks, WGSTSC have strongly supported and coordinated this research program. Sandia National Laboratories, SNL, has the lead role for conducting this research program; test program support is provided by both the U.S. Department of Energy and Nuclear Regulatory Commission. WGSTSC partners need this research to better understand potential radiological impacts from sabotage of nuclear material shipments and storage casks, and to support subsequent risk assessments, modeling, and preventative measures. We provide a summary of the overall, multi-phase test design and a description of all explosive containment and aerosol collection test components used. We focus on the recently initiated tests on ''surrogate'' spent fuel, unirradiated depleted uranium oxide, and forthcoming actual spent fuel tests. The depleted uranium oxide test rodlets were prepared by the Institut de Radioprotection et de Surete Nucleaire, in France. These surrogate test rodlets closely match the diameter of the test rodlets of actual spent fuel from the H.B. Robinson reactor (high burnup PWR fuel) and the Surry reactor (lower, medium burnup PWR fuel), generated from U.S. reactors. The characterization of the spent fuels and fabrication into short, pressurized rodlets has been performed by Argonne National Laboratory, for testing at SNL. The ratio of the aerosol and respirable particles released from HEDD-impacted spent

  18. Spent fuel characterization for the commercial waste and spent fuel packaging program

    International Nuclear Information System (INIS)

    Fish, R.L.; Davis, R.B.; Pasupathi, V.; Klingensmith, R.W.

    1980-03-01

    This document presents the rationale for spent fuel characterization and provides a detailed description of the characterization examinations. Pretest characterization examinations provide quantitative and qualitative descriptions of spent fuel assemblies and rods in their irradiated conditions prior to disposal testing. This information is essential in evaluating any subsequent changes that occur during disposal demonstration and laboratory tests. Interim examinations and post-test characterization will be used to identify fuel rod degradation mechanisms and quantify degradation kinetics. The nature and behavior of the spent fuel degradation will be defined in terms of mathematical rate equations from these and laboratory tests and incorporated into a spent fuel performance prediction model. Thus, spent fuel characterization is an essential activity in the development of a performance model to be used in evaluating the ability of spent fuel to meet specific waste acceptance criteria and in evaluating incentives for modification of the spent fuel assemblies for long-term disposal purposes

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

    International Nuclear Information System (INIS)

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

    1998-01-01

    Korea is a unique country, having both PWR and CANDU reactors. Korea can therefore exploit the natural synergism between the two reactor types to minimize overall waste production, and maximize energy derived from the fuel, by ultimately burning the spent fuel from its PWR reactors in CANDU reactors. As one of the possible fuel cycles, Recovered Uranium (RU) fuel offers a very attractive alternative to the use of Natural Uranium (NU) and slightly enriched uranium (SEU) in CANDU reactors. Potential benefits can be derived from a number of stages in the fuel cycle: no enrichment required, therefore no enrichment tails, direct conversion to UO 2 , lower sensitivity to 234 U and 236U absorption in the CANDU reactor, and expected lower cost relative to NU and SEU. These benefits all fit well with the PWR-CANDU fuel cycle synergy. RU arising from the conventional reprocessing of European and Japanese oxide spent fuel by 2000 is projected to be approaching 25,000 te. The use of RU fuel in a CANDU 6 reactor should result in no serious radiological difficulties and no requirements for special precautions and should not require any new technologies for the fuel fabrication and handling. The use of the CANDU Flexible Fueling (CANFLEX) bundle as the carrier for RU will be fully compatible with the reactor design, current safety and operational requirements, and there will be improved fuel performance compared with the CANDU 37-element NU fuel bundle. Compared with the 37-element NU bundle, the RU fuel has significantly improved fuel cycle economics derived from increased burnups, a large reduction in both fuel requirements and spent fuel, arisings, and the potential lower cost for RU material. There is the potential for annual fuel cost savings in the range of one-third to two-thirds, with enhanced operating margins using RU in the CANFLEX bundle design. These benefits provide the rationale for justifying R and D efforts on the use of RU fuel for advanced fuel cycles in CANDU

  20. 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)

  1. 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)

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

    International Nuclear Information System (INIS)

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

    1999-01-01

    As one of the possible fuel cycles in Korea, RU (Recycled Uranium) fuel offers a very attractive alternative to the use of NU (Natural Uranium) and SEU in the CANDU reactors, because Korea is a unique country having both PWR and CANDU reactors. Korea can therefore exploit the natural synergism between the two reactor types to minimise overall waste production, and maximise energy derived from the fuel, by burning the spent fuel from its PWR reactors in CANDU reactors. Potential benefits can be derived from a number of stages in the fuel cycle: no enrichment required, no enrichment tails, direct conversion to UO 2 lower sensitivity to 234 U and 236 U absorption in the CANDU reactor, expected lower cost relative to NU and SEU. These benefits all fit well with the PWR-CANDU fuel cycle synergy. RU arising from the reprocessing of European and Japanese oxide spent fuel by 2000 is projected to be approaching 25,000 te. The use of RU fuel in a CANDU-6 reactor should result in no serious radiological difficulties and no requirements for special precautions and should not require any new technologies for the fuel fabrication and handling. A KAERI's feasibility shows that the use of the CANFLEX bundle as the carrier for RU will be compatible with the reactor design, current safety and operational requirements, and there will be no significant fuel performance difference from the CANDU 37-element NU fuel bundle. Compared with the 37-element NU bundle, the RU fuel has significantly improved fuel cycle economics derived from increased burnups, a large reduction in fuel requirements and spent fuel arisings and the potential lower cost for RU material. There is the potential for annual fuel cost savings to be in the range of one-third to two-thirds, with enhanced operating margins using RU in the CANFLEX bundle design. These benefits provide the rationale for justifying R and D effort on the use of RU fuel for advanced fuel cycles in the CANDU reactors of Korea. The RU fuel

  3. Chemical Engineering Division Fuel Cycle Programs: October--December 1976

    International Nuclear Information System (INIS)

    Steindler, M.J.; Ader, M.; Bernstein, G.; Flynn, K.; Gerding, T.; Jardine, L.; Kullen, B.; Mecham, W.; Saunders, B.; Seefeldt, W.; Seitz, M.; Siczek, A.; Trevorrow, L.

    1977-01-01

    Fuel-cycle studies reported for this period include pyrochemical separation of plutonium and americium oxides from contaminated materials of construction such as steel. The actinides are partitioned to a high degree into slags that are contacted by the molten metal. Studies of advanced solvent extraction techniques focussed on the development of centrifugal contactors for use in Purex processes. A miniature contactor is to be used for performance studies applicable to larger units. Review of literature on the process chemistry of zirconium and ruthenium has been carried out to aid in improving the process when fast contactors are used. A review of information on the dispersion of reagents during accidents in reprocessing has been initiated to develop systematic data useful in identifying source terms. A review and evaluation of the encapsulation of high level waste in a metal matrix has been initiated. The data will be used to identify the state of the art and the importance of selected features of this process. Criteria for the handling of hulls are being developed on the basis of past work on the pyrophoricity of zirconium alloys and related criteria from several sources. These suggested criteria will be assembled together with the necessary technical rationalization, into a package for review by interested parties. A brief program to explore the disposal of noble gas fission products by deep-well injection has been started

  4. Status of the Canadian nuclear fuel waste management program

    International Nuclear Information System (INIS)

    Allan, C.J.; Stephens, M.E.

    1992-01-01

    The Canadian Concept for the permanent disposal of nuclear fuel waste has been developed extensively over the past several years, and is now well-advanced. The Concept, which involves the construction of a waste vault 500 to 1000 metres deep in plutonic rock located in the Canadian Precambrian Shield, is supported by an R ampersand D program with the following objectives: (1) to develop and demonstrate technology to site, design, build and operate a disposal facility; (2) to develop and demonstrate a methodology to evaluate the performance of the disposal system; and (3) to demonstrate that sites are likely to exist in the Canadian Precambrian Shield that would meet the regulatory requirements. A combination of engineered and natural barriers will be used to ensure that the vault design will meet rigorous safety standards. Experimental work is being carried out to elucidate all the important phenomena associated with the safety of the vault, including the performance of engineered barriers, natural geological barriers, and the biosphere

  5. Spent Fuel Handling and Packaging Program: a survey of hot cell facilities

    International Nuclear Information System (INIS)

    Menon, M.N.

    1978-07-01

    Hot cell facilities in the United States were surveyed to determine their capabilities for conducting integral fuel assembly and individual fuel rod examinations that are required in support of the Spent Fuel Handling and Packaging Program. The ability to receive, handle, disassemble and reconstitute full-length light water reactor spent fuel assemblies, and the ability to conduct nondestructive and destructive examinations on full-length fuel rods were of particular interest. Three DOE-supported facilities and three commercial facilities were included in the survey. This report provides a summary of the findings

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

    International Nuclear Information System (INIS)

    Crouthamel, C.E.

    1978-10-01

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

  7. HTGR Fuel Technology Program. Semiannual report for the period ending March 31, 1981

    International Nuclear Information System (INIS)

    1981-05-01

    This document reports the technical accomplishments on the HTGR Fuel Technology Program at General Atomic during the first half of FY-81. The activities include the fuel process, fuel materials, fuel cycle, fission product transport, and core component verification testing tasks necessary to support the design and development of a steam cycle/cogeneration (SC/C) version of the HTGR with a follow-on reformer (R) version. An important effort which was initiated during this period was the preparation of input data for a long-range technology program plan

  8. HTGR Fuel-Technology Program. Semiannual report for the period ending September 30, 1982

    International Nuclear Information System (INIS)

    1982-11-01

    This document reports the technical accomplishments on the HTGR Fuel Technology Program at GA Technologies Inc. during the second half of FY-1982. The activities include the fuel process, fuel materials, fuel cycle, fission product transport, and core component verification testing tasks necessary to support the design and development of a steam cycle/cogeneration (SC/C) version of the HTGR with a follow-on reformer (R) version. An important effort which was completed during this period was the preparation of input data for a long-range technology program plan

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

    International Nuclear Information System (INIS)

    Boulton, J.; Gibson, A.R.

    1979-12-01

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

  10. 2011 Annual Progress Report: DOE Hydrogen and Fuel Cells Program (Book)

    Energy Technology Data Exchange (ETDEWEB)

    2011-11-01

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

  11. Certification Report: Army Aviation Alternative Fuels Certification Program

    Science.gov (United States)

    2016-08-01

    Fuel Injector Coking Rig Description • Combustor Section Rig • Full Annular Rig • Sea Level and Simulated Altitude Engine Testing...purpose requirements in ASTM D4054 and MIL-HDBK- 510 for aviation fuel, as modified by the tri-service group to include diesel engine-related properties...atomized and vaporized when passing through the fuel nozzles into the combustor. In the combustor, it is vaporized, ignited and burned to provide the

  12. Site investigations for the disposal of spent fuel - investigation program

    International Nuclear Information System (INIS)

    Aeikaes, Timo

    1985-11-01

    The Industrial Power Company Ltd (TVO) is making preparations for the final disposal of spent nuclear fuel into the Finnish bedrock. The revised site investigation program for the years 1986-2010 is presented in this report. The objectives and activities in the near future are described in more detail. The main objectives and frame programs for the investigations in the more distant future are described. The program planning of these investigations are being developed in the preceding site investigations. The investigations for the site selection are divided into four phases: 1983-1985 selection of the investigation areas, preparations for the field investigations, drilling and investigations in a deep test borehole; 1986-1992 preliminary site investigations in 5-10 investigation areas; 1993-2000 detailed site investigations in 2-3 investigation areas. Site selection in the year 2000; 2001-2010 complementary investigations on the selected site. The first investigation phase will be carried out as planned. In this phase a 1001 m deep test borehole was drilled at Lavia in western Finland. With the investigations in the borehole and related development work, preparations were made for the future field investigations. The equipment and investigation methods are being developed during the site investigations. The equipment for taking groundwater samples and the unit for hydraulic testing have been developed. In the future the emphasis in the work will be in developing equipment for monitoring of the hydraulic head and measuring the volumetric flow. In groundwater sampling the present procedure can be improved by adding the test for the in-situ measurements. The results of the field investigations will be stored and processed in a centralized data base. The data base will transmit the results for the interpretation and then the interpreted results transmitted for model calculations and reporting. The cost estimate for the investigations in 1986-2010 is 110-125 million

  13. Burn-Up Calculation of the Fuel Element in RSG-GAS Reactor using Program Package BATAN-FUEL

    International Nuclear Information System (INIS)

    Mochamad Imron; Ariyawan Sunardi

    2012-01-01

    Calculation of burn lip distribution of 2.96 gr U/cc Silicide fuel element at the 78 th reactor cycle using computer code program of BATAN-FUEL has been done. This calculation uses inputs such as generated power, operation time and a core assumption model of 5/1. Using this calculation model burn up for the entire fuel elements at the reactor core are able to be calculated. From the calculation it is obtained that the minimum burn up of 6.82% is RI-50 at the position of A-9, while the maximum burn up of 57.57% is RI 467 at the position of 8-7. Based on the safety criteria as specified in the Safety Analysis Report (SAR) RSG-GAS reactor, the maximum fuel burn up allowed is 59.59%. It then can be concluded that pattern that elements placement at the reactor core are properly and optimally done. (author)

  14. Voluntary Truck and Bus Fuel-Economy-Program marketing plan. Final technical report, September 29, 1980-January 29, 1982

    Energy Technology Data Exchange (ETDEWEB)

    None

    1982-01-01

    The aim of the program is to improve the utilization of fuel by commercial trucks and buses by updating and implementing specific approaches for educating and monitoring the trucking industry on methods and means of conserving fuels. The following outlines the marketing plan projects: increase use of program logo by voluntary program members and others; solicit trade publication membership and support; brief Congressional delegations on fuel conservation efforts; increase voluntary program presence before trade groups; increase voluntary program presence at truck and trade shows; create a voluntary program display for use at trade shows and in other areas; review voluntary program graphics; increase voluntary program membership; and produce placemats carrying fuel conservation messages; produce a special edition of Fuel Economy News, emphasizing the driver's involvement in fuel conservation; produce posters carrying voluntary program fuel conservation message. Project objectives, activities, and results for each project are summarized.

  15. Correlations between fuel pins irradiated in fast and thermal fluxes using the frump fuel pin modelling program

    International Nuclear Information System (INIS)

    Hayns, M.R.; Adam, J.

    1975-08-01

    There is no experimental facilities in which a fuel pin can be irradiated in a fast environment under well defined conditions of over power or flow run down. Consequently most of the infor mation which is being accumulated on the behaviour of fuel pins under severe conditions is obtained from either capsule or loop rigs in thermal reactors. It is the purpose of this paper to highlight the differences between the behaviour of fuel pins irradiated in a thermal flux and a fast flux. A typical set of conditions is taken from an overpower experiment in a thermal flux and the behaviour of the system is analysed using the fuel modelling program FRUMP. A second numerical experiment is then performed in which the same conditions prevail, except that a fast flux is assumed, the criterion for comparison being that the total power input to the system is the same in both cases. From the many possible correlations which result from such an exercise the fuel tempreature has been selected to highlight various important features of the two irradiations. It is demonstrated that the flux depression can cause differences in the pin behaviour, even to altering the order of events in a transient. For example fuel melting will occur at different times and at different positions in the fuel in the two cases. It is concluded that the techniques of fuel modelling, as typified in the program FRUMP can provide a very useful tool indeed for the analysis of such experiments and for guiding the establishment of the appropriate correlations for the extrapolation to the fast flux case. (author)

  16. Monitoring instrumentation spent fuel management program. Final report

    International Nuclear Information System (INIS)

    1979-01-01

    Preliminary monitoring system methodologies are identified as an input to the risk assessment of spent fuel management. Conceptual approaches to instrumentation for surveillance of canister position and orientation, vault deformation, spent fuel dissolution, temperature, and health physics conditions are presented. In future studies, the resolution, reliability, and uncertainty associated with these monitoring system methodologies will be evaluated

  17. Highlights of the French program on PWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Pages, J P [CEA Centre d` Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Direction des Reacteurs Nucleaires

    1997-12-01

    The presentation reviews the French programme on PWR fuel including the overall results of the year 1996 for nuclear operation; fuel management and economy; French nuclear electricity generation sites; production of nuclear generated electricity; energy availability of the 900 and 1,300 Mw PWR units; average radioactive liquid releases excluding tritium per unit; plutonium recycling experience.

  18. Spent nuclear fuel project quality assurance program plan

    International Nuclear Information System (INIS)

    Lacey, R.E.

    1997-01-01

    This main body of this document describes how the requirements of 10 CFR 830.120 are met by the Spent Nuclear Fuel Project through implementation of WHC-SP-1131. Appendix A describes how the requirements of DOE/RW-0333P are met by the Spent Nuclear Fuel Project through implementation of specific policies, manuals, and procedures

  19. Review of HEDL fuel pin transient analyses analytical programs

    International Nuclear Information System (INIS)

    Scott, J.H.; Baars, R.E.

    1975-05-01

    Methods for analysis of transient fuel pin performance are described, as represented by the steady-state SIEX code and the PECT series of codes used for steady-state and transient mechanical analyses. The empirical fuel failure correlation currently in use for analysis of transient overpower accidents is described. (U.S.)

  20. Remote handling equipment design for the HEDL fuel supply program

    International Nuclear Information System (INIS)

    Metcalf, I.L.

    1984-09-01

    A process line is currently being developed for fabrication of high exposure mixed uranium-plutonium core assemblies. This paper describes the design philosophy, process flow, equipment, and the handling and radiation shielding techniques used for inspection of Fast Flux Test Facility (FFTF) fuel pins and assembly of Driver Fuel Assemblies (DFAs) 6 figures

  1. PND fuel handling decontamination program: specialized techniques and results

    International Nuclear Information System (INIS)

    Pan, R.; Hobbs, K.; Minnis, M.; Graham, K.

    1995-01-01

    The use of various decontamination techniques and equipment has become a critical part of Fuel Handling maintenance work at the Pickering Nuclear Station, an eight unit CANDU station located about 30 km east of Toronto. This paper presents an overview of the set up and techniques used for cleaning in the PND Fuel Handling Maintenance Facility, and the results achieved. (author)

  2. Spent-fuel disassembly and canning programs at the Barnwell Nuclear Fuel Plant (BNFP)

    International Nuclear Information System (INIS)

    Townes, G.A.

    1980-02-01

    Doubling pool storage and tripling truck cask shipping capability appear feasible based on preliminary development work performed at the BNFP. This would be accomplished by disassembly of the fuel assembly and canning of the fuel pins. The dry, remote disassembly and canning in the headend cells offer adequate rates, easily controlled operator exposure, and visual inspection of the fuel pins and operations through shielded windows

  3. Program of monitoring PWR fuel in Spain; Programa de Vigilancia de Combustible pwr en Espana

    Energy Technology Data Exchange (ETDEWEB)

    Martinez Murillo, J. C.; Quecedo, M.; Munoz-Roja, C.

    2015-07-01

    In the year 2000 the PWR utilities: Centrales Nucleares Almaraz-Trillo (CNAT) and Asociacion Nuclear Asco-Vandellos (ANAV), and ENUSA Industrias Avanzadas developed and executed a coordinated strategy named PIC (standing for Coordinated Research Program), for achieving the highest level of fuel reliability. The paper will present the scope and results of this program along the years and will summarize the way the changes are managed to ensure fuel integrity. The excellent performance of the ENUSA manufactured fuel in the PWR Spanish NPPs is the best indicator that the expectations on this program are being met. (Author)

  4. Experimental clean combustor program, alternate fuels addendum, phase 2

    Science.gov (United States)

    Gleason, C. C.; Bahr, D. W.

    1976-01-01

    The characteristics of current and advanced low-emissions combustors when operated with special test fuels simulating broader range combustion properties of petroleum or coal derived fuels were studied. Five fuels were evaluated; conventional JP-5, conventional No. 2 Diesel, two different blends of Jet A and commercial aromatic mixtures - zylene bottoms and haphthalene charge stock, and a fuel derived from shale oil crude which was refined to Jet A specifications. Three CF6-50 engine size combustor types were evaluated; the standard production combustor, a radial/axial staged combustor, and a double annular combustor. Performance and pollutant emissons characteristics at idle and simulated takeoff conditions were evaluated in a full annular combustor rig. Altitude relight characteristics were evaluated in a 60 degree sector combustor rig. Carboning and flashback characteristics at simulated takeoff conditions were evaluated in a 12 degree sector combustor rig. For the five fuels tested, effects were moderate, but well defined.

  5. High-uranium-loaded U3O8--Al fuel element development program

    International Nuclear Information System (INIS)

    Martin, M.M.

    1978-01-01

    The High-Uranium-Loaded U 3 O 8 --Al Fuel Development Program supports Argonne National Laboratory efforts to develop high-uranium-density research and test reactor fuel to accommodate use of low-uranium enrichment. The goal is to fuel most research and test reactors with uranium of less than 20% enrichment for the purpose of lowering the potential for diversion of highly-enriched material for nonpeaceful usages

  6. Summary of non-US national and international fuel cycle and radioactive waste management programs 1982

    International Nuclear Information System (INIS)

    Harmon, K.M.; Kelman, J.A.

    1982-08-01

    Brief program overviews of fuel cycle, spent fuel, and waste management activities in the following countries are provided: Argentina, Australia, Austria, Belgium, Brazil, Canada, China, Denmark, Finland, France, German Federal Republic, India, Italy, Japan, Republic of Korea, Mexico, Netherlands, Pakistan, South Africa, Spain, Sweden, Switzerland, Taiwan, USSR, and the United Kingdom. International nonproliferation activities, multilateral agreements and projects, and the international agencies specifically involved in the nuclear fuel cycle are also described

  7. Summary of non-US national and international fuel cycle and radioactive waste management programs 1982

    Energy Technology Data Exchange (ETDEWEB)

    Harmon, K.M.; Kelman, J.A.

    1982-08-01

    Brief program overviews of fuel cycle, spent fuel, and waste management activities in the following countries are provided: Argentina, Australia, Austria, Belgium, Brazil, Canada, China, Denmark, Finland, France, German Federal Republic, India, Italy, Japan, Republic of Korea, Mexico, Netherlands, Pakistan, South Africa, Spain, Sweden, Switzerland, Taiwan, USSR, and the United Kingdom. International nonproliferation activities, multilateral agreements and projects, and the international agencies specifically involved in the nuclear fuel cycle are also described.

  8. Finite element method programs to analyze irradiation behavior of fuel pellets

    International Nuclear Information System (INIS)

    Yamada, Rayji; Harayama, Yasuo; Ishibashi, Akihiro; Ono, Masao.

    1979-09-01

    For the safety assessment of reactor fuel, it is important to grasp local changes of fuel pins due to irradiation in a reactor. Such changes of fuel result mostly from irradiation of fuel pellets. Elasto-plastic analysis programs based on the finite element method were developed to analyze these local changes. In the programs, emphasis is placed on the analysis of cracks in pellets; the interaction between cracked-pellets and cladding is not taken into consideration. The two programs developed are FEMF3 based on a two-dimensional axially symmetric model (r-z system) and FREB4 on a two-dimensional plane model (r-theta system). It is discussed in this report how the occurrence and distribution of cracks depend on heat rate of the fuel pin. (author)

  9. LEU fuel fabrication program for the RECH-1 reactor. Status report

    International Nuclear Information System (INIS)

    Chavez, J.C.; Barrera, M.; Jimenez, O.; Lisboa, J.; Marin, J.

    2000-01-01

    In 1995 a 50 LEU U 3 Si 2 fuel elements fabrication program for the RECH-1 research reactor was established at the Comision Chilena de Energia Nuclear, CCHEN. After a fabrication process qualification stage, in 1998, four elements were early delivered to the reactor in order to start an irradiation qualification stage. The irradiation has reached an estimated 10% burn-up and no fabrication problems have been detected up to this burn-up level. During 1999 and up to the first quarter of 2000, 19 fuel elements were produced and 7 fuel elements are expected for the end of 2000. This report presents an updated summary of the main results obtained in this fuel fabrication program. A summary of other activities generated by this program, such as in core follow-up of the four leader fuel elements, ISO 9001 implementation for the fabrication process and a fabrication and qualification optimization planning, is also presented here. (author)

  10. Darlington NGD fuel handling head eight acceptance program

    International Nuclear Information System (INIS)

    Skelton, P.H.; Sie, T.

    1996-01-01

    Darlington NGD requires eight fuelling machine heads to fuel the four 932 MW reactors. Six heads are used on the three fuelling machine trolleys for normal fuelling operations. A further two heads are required to allow for maintenance and to provide for such reactor face activities as PIPE and CIGAR. Seven heads were successfully delivered to site from the head supplier. During acceptance testing, stalls on the charge tube screw assembly of the eighth and final head prevented its delivery to site. Replacement of the charge tube screw with a spare screw did not alleviate the problem. An in depth series of tests were undertaken at site, at the supplier and at the screw sub-supplier to determine the root cause of the problem. These tests included taking torque measurements under different operating conditions and using different components to assess the effects of the changes on torque levels. An assessment of the effects of changing chemical conditions (particularly crud levels) was also made. To ensure that the results of the testing were well understood, additional torque testing was also completed on a head and screw assembly at site that was known to work well. Based on all of the above series of tests, a recommendation was made to re-machine the charge tube screw(s). The original charge tube screw from Head eight was subsequently returned to the sub-supplier for re-work. Follow-up torque measurements and acceptance testing showed that the screw rework was effective and that Head eight could be successfully delivered to site. This paper focuses on the results of the head/screw test program. Results of the acceptance testing are also discussed. (author). 2 refs., 4 figs

  11. Darlington NGD fuel handling head eight acceptance program

    Energy Technology Data Exchange (ETDEWEB)

    Skelton, P H; Sie, T [Ontario Hydro, Bowmanville (Canada). Darlington Nuclear Generating Station; Pilgrim, J [Canadian General Electric Co. Ltd., Toronto, ON (Canada)

    1997-12-31

    Darlington NGD requires eight fuelling machine heads to fuel the four 932 MW reactors. Six heads are used on the three fuelling machine trolleys for normal fuelling operations. A further two heads are required to allow for maintenance and to provide for such reactor face activities as PIPE and CIGAR. Seven heads were successfully delivered to site from the head supplier. During acceptance testing, stalls on the charge tube screw assembly of the eighth and final head prevented its delivery to site. Replacement of the charge tube screw with a spare screw did not alleviate the problem. An in depth series of tests were undertaken at site, at the supplier and at the screw sub-supplier to determine the root cause of the problem. These tests included taking torque measurements under different operating conditions and using different components to assess the effects of the changes on torque levels. An assessment of the effects of changing chemical conditions (particularly crud levels) was also made. To ensure that the results of the testing were well understood, additional torque testing was also completed on a head and screw assembly at site that was known to work well. Based on all of the above series of tests, a recommendation was made to re-machine the charge tube screw(s). The original charge tube screw from Head eight was subsequently returned to the sub-supplier for re-work. Follow-up torque measurements and acceptance testing showed that the screw rework was effective and that Head eight could be successfully delivered to site. This paper focuses on the results of the head/screw test program. Results of the acceptance testing are also discussed. (author). 2 refs., 4 figs.

  12. Radioactive waste management: a summary of state laws and administration. National Low-Level Radioactive Waste Mangement Program

    International Nuclear Information System (INIS)

    1983-05-01

    This is the first quarterly update of Radioactive Waste Management: A Summary of State Laws and Administration. Because states have been very active on waste management issues, the whole report is being reissued in this update. It covers the administration, the legislature and the laws in the 50 states related to radioactive waste. The report for each state is divided into four sections: Cover Page; Administrative; Legislative; and Applicable Legislation. The cover page indicates whether or not it is an Agreement State, the low-level waste compacts in which the state is listed as an eligible state, and the high-level waste repository site screening regions in which the state or a portion of it is located. The administrative section provides information on the governor, lead agencies, other involved administrative agencies, relevant commissions, boards and councils and various contacts. The Legislative section provides general information on the legislature and lists legislative leaders, the relevant committees and their chairs and a legislative contact. In the section covering Applicable Legislation, laws related to radiation protection, low-level waste and high-level waste have been summarized. Hazardous waste siting laws are included for states that do not have a siting law covering radioactive waste. The section also contains summaries of relevant bills introduced in 1982 and 1983 legislative sessions and their disposition. In general, the information in this report is accurate as of 15 April 1983

  13. Development of a computer program for the cost analysis of spent fuel management

    International Nuclear Information System (INIS)

    Choi, Heui Joo; Lee, Jong Youl; Choi, Jong Won; Cha, Jeong Hun; Whang, Joo Ho

    2009-01-01

    So far, a substantial amount of spent fuels have been generated from the PWR and CANDU reactors. They are being temporarily stored at the nuclear power plant sites. It is expected that the temporary storage facility will be full of spent fuels by around 2016. The government plans to solve the problem by constructing an interim storage facility soon. The radioactive management act was enacted in 2008 to manage the spent fuels safety in Korea. According to the act, the radioactive waste management fund which will be used for the transportation, interim storage, and the final disposal of spent fuels has been established. The cost for the management of spent fuels is surprisingly high and could include a lot of uncertainty. KAERI and Kyunghee University have developed cost estimation tools to evaluate the cost for a spent fuel management based on an engineering design and calculation. It is not easy to develop a tool for a cost estimation under the situation that the national policy on a spent fuel management has not yet been fixed at all. Thus, the current version of the computer program is based on the current conceptual design of each management system. The main purpose of this paper is to introduce the computer program developed for the cost analysis of a spent fuel management. In order to show the application of the program, a spent fuel management scenario is prepared, and the cost for the scenario is estimated

  14. International safeguards concerns of Spent Fuel Disposal Program

    International Nuclear Information System (INIS)

    Pillay, K.K.S.

    1988-01-01

    The purpose of this paper is to stimulate discussions on the subjects of safeguarding large quantities of plutonium contained in spent fuels to be disposed of in geologic respositories. All the spent fuel disposal scenarios examined here pose a variety of safeguards problems, none of which are adequately addressed by the international safeguards community. The spent fuels from once-through fuel cycles in underground repositories would become an increasingly attractive target for diversion because of their plutonium content and decreasing radioactivity. Current design of the first geologic repository in the US will have the capacity to accommodate wastes equivalent to 70,000 Mt of uranium from commercial and defense fuel cycles. Of this, approximately 62,000 Mt uranium equivalent will be commerical spent fuel, containing over 500 Mt of plutonium. International safeguards commitments may require us to address the safeguards issues of disposing of such large quanities of plutonium in a geologic repository, which has the potential to become a plutonium mine in the future. This paper highlights several issues that should be addressed in the near term by US industries and the DOE before geologic repositories for spent fuels become a reality

  15. EU program fuel cells in 2012 - FCH JU Fuel Cell and Hydrogen Joint Undertaking; EU-program braensleceller 2012 - FCH JU Fuel Cell and Hydrogen Joint Undertaking

    Energy Technology Data Exchange (ETDEWEB)

    Ridell, Bengt

    2013-03-15

    An EU activity in fuel cell and hydrogen field are gathered since 2008 in a so called JU, Joint Undertaking, or as it is also referred to as JTI Joint Technology Initiative. The program will run 2008 - 2013 and covers in total 940 MEUR of which the EU Commission is funding 470 MEUR. The activities of the FCH JU are governed by a Governing Board which has 12 members, five from the Commission, one of the research group and 5 from the Industrial Group. The current agreement for the FCH JU / JTI is coming to an end, and the sixth and final call was released in January 2013 with the deadline of 22 May 2013. Funding from the Commission is made through the Seventh Framework Programme FP7, which ends in 2013. Next the Eighth Framework Programme called Horizon 2020 shall continue for the years 2014 - 2020. Five of the six calls are completed. From the four first calls there are 61 projects started which 6 have been completed. From the fifth announcement is further 27 projects selected for negotiation with the Commission and they will start soon. It is now working intensively to plan Horizon 2020. There are plans to continue the new FCH JU but nothing is decided either for this or for the budget for Horizon 2020. If the FCH Joint Undertaking shall continue in its present form as a Joint Undertaking it will require clear long-term commitments from the private sector and also from the Member States. Another issue is that the long-term research should also get space it has not been the case in the present FCH JU. There are several Swedish participants in the projects and in the working groups of the program. There are Swedish participants in 11 of the 68 projects launched so far. It is in the areas of Stationary systems, Transportation and Early Markets. Project manager for the project FCGEN is Volvo Technology AB. FCH JU has its own website, www.fch-ju.eu, which opened in 2010 when the organization of the program was taken over from the Commission to permanent organisation

  16. PLUS 7TM advanced fuel assembly development program for KSNPs and APR1400

    International Nuclear Information System (INIS)

    Kim, Kyutae; Stucker, David L.

    2002-01-01

    KNFC and Westinghouse have recently completed the development of the PLUS 7 TM advanced 16 X 16 fuel assembly for the Korean Standard Nuclear Plants (KSNPs) and the Advanced Power Reactor 1400 (APR 1400). This fuel design utilized the proven advanced design features including mixing vane spacer grids to increase critical heat flux performance, ZIRLO TM advanced materials to enable high-duty, high burnup fuel management and an optimized fuel rod diameter which improves fuel cycle cost while resulting in significant standardization of Korean fuel manufacture. PLUS 7 TM , also includes a patented spacer grid design with conformal fuel rod support designed to provide superior fuel rod wear/fretting resistance while minimizing pressure drop. This paper will present an overview of the PLUS 7 TM fuel assembly development process including a summary of the three-year design and testing program from a mechanical, neutronic, and thermal/hydraulic perspective. The PLUS 7 TM fuel for the KSNPs and the APR1400 reactors results in multi-million dollar per cycle savings in imported enriched uranium product for the Korean nuclear power program with technology specifically developed for Korea by experienced Korean engineers

  17. Locomotive fuel tank structural safety testing program : passenger locomotive fuel tank jackknife derailment load test.

    Science.gov (United States)

    2010-08-01

    This report presents the results of a passenger locomotive fuel tank load test simulating jackknife derailment (JD) load. The test is based on FRA requirements for locomotive fuel tanks in the Title 49, Code of Federal Regulations (CFR), Part 238, Ap...

  18. Alternative bio-based fuels for aviation: the clean airports program

    International Nuclear Information System (INIS)

    Shauck, M.E.; Zanin, M.G.

    1997-01-01

    The Renewable Aviation Fuels Development Center at Baylor University in Waco, Texas, has been designated as the national coordinator of the Clean Airports Program. The U.S. Dept. of Energy (US DOE) conferred this designation in March 1996. This program, a spin-off of the Clean Cities Program, was initiated to increase the use of alternative fuels in aviation. The two major fuels used in aviation are the current piston engine aviation gasoline and the current turbine engine fuel. The environmental impact of each of these fuels is significant. Aviation gasoline (100LL), currently used in the general aviation piston engine fleet, contributes 100% of the emissions containing lead in the U.S. today. Turbine engine fuel (jet fuel) produces two major environmental impacts: a local one, in the vicinity of the airports, and a global impact on climate change. The Clean Airports Program was established to achieve and maintain clean air at and in the vicinity of airports, through the use of alternative fuel-powered air and ground transportation vehicles. (author)

  19. Optimization of axial enrichment and gadolinia distributions for BWR fuel under control rod programming, (2)

    International Nuclear Information System (INIS)

    Hida, Kazuki; Yoshioka, Ritsuo

    1992-01-01

    A method has been developed for optimizing the axial enrichment and gadolinia distributions for the reload BWR fuel under control rod programming. The problem was to minimize the enrichment requirement subject to the criticality and axial power peaking constraints. The optimization technique was based on the successive linear programming method, each linear programming problem being solved by a goal programming algorithm. A rapid and practically accurate core neutronics model, named the modified one-dimensional core model, was developed to describe the batch-averaged burnup behavior of the reload fuel. A core burnup simulation algorithm, employing a burnup-power-void iteration, was also developed to calculate the rigorous equilibrium cycle performance. This method was applied to the optimization of axial two- and 24-region fuels for demonstrative purposes. The optimal solutions for both fuels have proved the optimality of what is called burnup shape optimization spectral shift. For the two-region fuel with a practical power peaking of 1.4, the enrichment distribution was nearly uniform, because a bottom-peaked burnup shape flattens the axial power shape. Optimization of the 24-region fuel has shown a potential improvement in BWR fuel cycle economics, which will guide future advancement in BWR fuel designs. (author)

  20. Development of fusion fuel cycles: Large deviations from US defense program systems

    Energy Technology Data Exchange (ETDEWEB)

    Klein, James Edward, E-mail: james.klein@srnl.doe.gov; Poore, Anita Sue; Babineau, David W.

    2015-10-15

    Highlights: • All tritium fuel cycles start with a “Tritium Process.” All have similar tritium processing steps. • Fusion tritium fuel cycles minimize process tritium inventories for various reasons. • US defense program facility designs did not minimize in-process inventories. • Reduced inventory tritium facilities will lower public risk. - Abstract: Fusion energy research is dominated by plasma physics and materials technology development needs with smaller levels of effort and funding dedicated to tritium fuel cycle development. The fuel cycle is necessary to supply and recycle tritium at the required throughput rate; additionally, tritium confinement throughout the facility is needed to meet regulatory and environmental release limits. Small fuel cycle development efforts are sometimes rationalized by stating that tritium processing technology has already been developed by nuclear weapons programs and these existing processes only need rescaling or engineering design to meet the needs of fusion fuel cycles. This paper compares and contrasts features of tritium fusion fuel cycles to United States Cold War era defense program tritium systems. It is concluded that further tritium fuel cycle development activities are needed to provide technology development beneficial to both fusion and defense programs tritium systems.

  1. A status report on the integral fast reactor fuels and safety program

    International Nuclear Information System (INIS)

    Pedersen, D.R.; Seidel, B.R.

    1990-01-01

    The integral fast reactor (IFR) is an advanced liquid-metal-cooled reactor (ALMR) concept being developed at Argonne National Laboratory. The IFR program is specifically responsible for the irradiation performance, advanced core design, safety analysis, and development of the fuel cycle for the US Department of Energy's ALMR program. The basic elements of the IFR concept are (a) metallic fuel, (b) liquid-sodium cooling, (c) modular, pool-type reactor configuration, (d) an integral fuel cycle based upon pyrometallurgical processing. The most significant safety aspects of the IFR program result from its unique fuel design, a ternary alloy of uranium, plutonium, and zirconium. This fuel is based on experience gained through > 25 yr operation of the Experimental Breeder Reactor II (EBR-II) with a uranium alloy metallic fuel. The ultimate criteria for fuel pin design is the overall integrity at the target burnup. The probability of core meltdown is remote; however, a theoretical possibility of core meltdown remains. The next major step in the IFR development program will be a full-scale pyroprocessing demonstration to be carried out in conjunction with EBR-II. The IFR fuel cycle closure based on pyroprocessing will also have a dramatic impact on waste management options and on actinide recycling

  2. 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)

  3. TRIGA high wt -% LEU fuel development program. Final report

    International Nuclear Information System (INIS)

    West, G.B.

    1980-07-01

    The principal purpose of this work was to investigate the characteristics of TRIGA fuel where the contained U-235 was in a relatively high weight percent (wt %) of LEU (low enriched uranium - enrichment of less than 20%) rather than a relatively low weight percent of HEU (high enriched uranium). Fuel with up to 45 wt % U was fabricated and found to be acceptable after metallurgical examinations, fission product retention tests and physical property examinations. Design and safety analysis studies also indicated acceptable prompt negative temperature coefficient and core lifetime characteristics for these fuels

  4. Guide to the Canadian nuclear fuel waste management program. 2.ed

    International Nuclear Information System (INIS)

    Rosinger, E.L.J.; Lyon, R.B.; Gillespie, P.; Tamm, J.

    1983-02-01

    This document describes the administrative structure and major research and development components of the Canadian Nuclear Fuel Waste Management Program. It outlines the participating organizations, summarizes the program statistics, and describes the international cooperation and external review aspects of the program

  5. Developments in the Canadian program for geological disposal of nuclear fuel waste

    International Nuclear Information System (INIS)

    Allan, C.J.; Nuttall, K.

    1996-01-01

    The Canadian Nuclear Fuel Waste Management Program is at the end of disposal concept and technology development and is now undergoing a comprehensive environmental review. This paper will review: the history of the Canadian program; the disposal concept and the associated technologies; the program achievements and the lessons learned; and the status of the environmental review. (author)

  6. A cost analysis of Colorado's 1991-92 oxygenated fuels program

    International Nuclear Information System (INIS)

    Manderino, L.A.; Bowles, S.L.

    1993-01-01

    This paper discusses the methodology used to conduct a cost analysis of Colorado's 1991-92 Oxygenated Fuels Program. This program requires the use of oxygenated fuels during the winter season in Denver and surrounding areas. The cost analysis was conducted as part of an overall cost-effectiveness study of the 1991-92 program conducted by PRC Environmental Management, Inc. (PRC). The paper, however, focuses on cost analysis and does not consider potential benefits of the program. The study analyzed costs incurred by different segments of society, including government, industry, and consumers. Because the analysis focused on a specific program year, neither past nor future costs were studied. The discussion of government costs includes the agencies interviewed and the types of costs associated with government administration and enforcement of the program. The methodology used to calculate costs to private industry is also present. The study examined the costs to fuel refineries, pipelines, and blenders, as well as fuel retailers and automobile fleet operators. Finally, the paper discusses the potential costs incurred by the consumer purchasing oxygenated fuels. Costs associated with issues such as vehicle driveability, automobile parts durability and performance, and fuel economy are also examined. A summary of all costs by category is presented along with an analysis of the major cost components. These include costs which are sensitive to specific circumstances and which may vary among programs

  7. 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.

  8. Quality assurance program for surveillance of fast reactor mixed oxide fuel analytical chemistry

    International Nuclear Information System (INIS)

    Rein, J.E.; Zeigler, R.K.; Waterbury, G.R.; McClung, W.E.; Praetorius, P.R.; Delvin, W.L.

    1976-01-01

    An effective quality assurance program for the chemical analysis of nuclear fuel is essential to assure that the fuel will meet the strict chemical specifications required for optimum reactor performance. Such a program has been in operation since 1972 for the fuels manufactured for the Fast Flux Test Facility. This program, through the use of common quality control and calibration standards, has consistently provided high levels of agreement among laboratories in all areas of analysis. The paper presented gives a summary of the chemical specifications for the fuel and source material, an outline of the requirements for laboratory qualifications and the preparation of calibration and quality control materials, general administration details of the plan, and examples where the program has been useful in solving laboratory problems

  9. The hydrogen and the fuel cells in the world. Programs and evolutions

    International Nuclear Information System (INIS)

    Lucchese, P.

    2008-01-01

    HyPac is a french platform on the hydrogen and fuel cells, created in 2008. The author presents the opportunity of such a platform facing the world research programs and other existing platforms. (A.L.B.)

  10. TEMP-M program for thermal-hydraulic calculation of fast reactor fuel assemblies

    International Nuclear Information System (INIS)

    Bogoslovskaya, C.P.; Sorokin, A.P.; Tikhomirov, B.B.; Titov, P.A.; Ushakov, P.A.

    1983-01-01

    TEMP-M program (Fortran, BESM-6 computer) for thermal-hydraulic calculation of fast reactor fuel assemblies is described. Results of calculation of temperature field in a 127 fuel element assembly of BN-600, reactor accomplished according to TEMP-N program are considered as an example. Algorithm, realized in the program, enables to calculate the distributions of coolant heating, fuel element temperature (over perimeter and length) and assembly shell temperature. The distribution of coolant heating in assembly channels is determined from a solution of the balance equation system which accounts for interchannel exchange, nonadiabatic conditions on the assembly shell. The TEMP-M program gives necessary information for calculation of strength, seviceability of fast reactor core elements, serves an effective instrument for calculations when projecting reactor cores and analyzing thermal-hydraulic characteristics of operating reactor fuel assemblies

  11. Optimization programs for reactor core fuel loading exhibiting reduced neutron leakage

    International Nuclear Information System (INIS)

    Darilek, P.

    1991-01-01

    The program MAXIM was developed for the optimization of the fuel loading of WWER-440 reactors. It enables the reactor core reactivity to be maximized by modifying the arrangement of the fuel assemblies. The procedure is divided into three steps. The first step includes the passage from the three-dimensional model of the reactor core to the two-dimensional model. In the second step, the solution to the problem is sought assuming that the multiplying properties, or the reactivity in the zones of the core, vary continuously. In the third step, parameters of actual fuel assemblies are inserted in the ''continuous'' solution obtained. Combined with the program PROPAL for a detailed refinement of the loading, the program MAXIM forms a basis for the development of programs for the optimization of fuel loading with burnable poisons. (Z.M.). 16 refs

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

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  13. Spent LWR fuel leach tests: Waste Isolation Safety Assessment program

    International Nuclear Information System (INIS)

    Katayama, Y.B.

    1979-04-01

    Spent light-water-reactor (LWR) fuels with burnups of 54.5, 28 and 9 MWd/kgU were leach-tested in deionized water at 25 0 C. Fuel burnup has no apparent effect on the calculated leach rates based upon the behavior of 137 Cs and 239+240 Pu. A leach test of 54.5 MWd/kgU spent fuel in synthetic sea brine showed that the cesium-based leach rate is lower in sea brine than in deionized water. A rise in the leach rate was observed after approximately 600 d of cumulative leaching. During the rise, the leach rate for all the measured radionuclides become nearly equal. Evidence suggests that exposure of new surfaces to the leachant may cause the increase. As a result, experimental work to study leaching mechanisms of spent fuel has been initiated. 22 figures

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  15. Fuel Gas Demonstration Plant Program. Volume I. Demonstration plant

    Energy Technology Data Exchange (ETDEWEB)

    1979-01-01

    The objective of this project is for Babcock Contractors Inc. (BCI) to provide process designs, and gasifier retort design for a fuel gas demonstration plant for Erie Mining Company at Hoyt Lake, Minnesota. The fuel gas produced will be used to supplement natural gas and fuel oil for iron ore pellet induration. The fuel gas demonstration plant will consist of five stirred, two-stage fixed-bed gasifier retorts capable of handling caking and non-caking coals, and provisions for the installation of a sixth retort. The process and unit design has been based on operation with caking coals; however, the retorts have been designed for easy conversion to handle non-caking coals. The demonstration unit has been designed to provide for expansion to a commercial plant (described in Commercial Plant Package) in an economical manner.

  16. 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...

  17. National Fuel Cell Bus Program : Accelerated Testing Report, AC Transit

    Science.gov (United States)

    2009-01-01

    This is an evaluation of hydrogen fuel cell transit buses operating at AC Transit in revenue service since March 20, 2006 compared to similar diesel buses operating from the same depot. This evaluation report includes results from November 2007 throu...

  18. Vehicle Technologies and Fuel Cell Technologies Program: Prospective Benefits Assessment Report for Fiscal Year 2016

    Energy Technology Data Exchange (ETDEWEB)

    Stephens, T. S. [Argonne National Lab. (ANL), Argonne, IL (United States); Taylor, C. H. [TA Engineering, Inc., Catonsville, MD (United States); Moore, J. S. [TA Engineering, Inc., Catonsville, MD (United States); Ward, J. [United States Department of Energy, Washington, DC (United States). Office of Energy Efficiency and Renewable Energy

    2016-02-23

    Under a diverse set of programs, the Vehicle Technologies and Fuel Cell Technologies offices of DOE’s Office of Energy Efficiency and Renewable Energy invest in research, development, demonstration, and deployment of advanced vehicle, hydrogen production, delivery and storage, and fuel cell technologies. This report estimates the benefits of successfully developing and deploying these technologies (a “Program Success” case) relative to a base case (the “No Program” case). The Program Success case represents the future with completely successful deployment of Vehicle Technologies Office (VTO) and Fuel Cell Technologies Office (FCTO) technologies. The No Program case represents a future in which there is no contribution after FY 2016 by the VTO or FCTO to these technologies. The benefits of advanced vehicle, hydrogen production, delivery and storage, and fuel cell technologies were estimated on the basis of differences in fuel use, primary energy use, and greenhouse gas (GHG) emissions from light-, medium- and heavy-duty vehicles, including energy and emissions from fuel production, between the base case and the Program Success case. Improvements in fuel economy of various vehicle types, growth in the stock of fuel cell vehicles and other advanced technology vehicles, and decreased GHG intensity of hydrogen production and delivery in the Program Success case over the No Program case were projected to result in savings in petroleum use and GHG emissions. Benefits were disaggregated by individual program technology areas, which included the FCTO program and the VTO subprograms of batteries and electric drives; advanced combustion engines; fuels and lubricants; materials (for reduction in vehicle mass, or “lightweighting”); and, for medium- and heavy-duty vehicles, reduction in rolling and aerodynamic resistance. Projections for the Program Success case indicate that by 2035, the average fuel economy of on-road, light-duty vehicle stock could be 47% to 76

  19. A computer program for structural analysis of fuel elements

    International Nuclear Information System (INIS)

    Hayashi, I.M.V.; Perrotta, J.A.

    1988-01-01

    It's presented the code ELCOM for the matrix analysis of tubular structures coupled by rigid spacers, typical of PWR's fuel elements. The code ELCOM makes a static structural analysis, where the displacements and internal forces are obtained for each structure at the joints with the spacers, and also, the natural frequencies and vibrational modes of an equivalent integrated structure are obtained. The ELCOM result is compared to a PWR fuel element structural analysis obtained in published paper. (author) [pt

  20. Exploration for fossil and nuclear fuels from orbital altitudes. [results of ERTS program for oil exploration

    Science.gov (United States)

    Short, N. M.

    1974-01-01

    Results from the ERTS program pertinent to exploration for oil, gas, and uranium are discussed. A review of achievements in relevant geological studies from ERTS, and a survey of accomplishments oriented towards exploration for energy sources are presented along with an evaluation of the prospects and limitations of the space platform approach to fuel exploration, and an examination of continuing programs designed to prove out the use of ERTS and other space system in exploring for fuel resources.

  1. Spent fuel sabotage test program, characterization of aerosol dispersal : interim final report

    International Nuclear Information System (INIS)

    Gregson, Michael Warren; Brockmann, John E.; Loiseau, Olivier; Klennert, Lindsay A.; Nolte, Oliver; Molecke, Martin Alan; Autrusson, Bruno A.; Koch, Wolfgang; Pretzsch, Gunter Guido; Brucher, Wenzel; Steyskal, Michele D.

    2008-01-01

    This multinational, multi-phase spent fuel sabotage test program is quantifying the aerosol particles produced when the products of a high energy density device (HEDD) interact with and explosively particulate test rodlets that contain pellets of either surrogate materials or actual spent fuel. This program provides source-term data that are relevant to plausible sabotage scenarios in relation to spent fuel transport and storage casks and associated risk assessments. We present details and significant results obtained from this program from 2001 through 2007. Measured aerosol results include: respirable fractions produced; amounts, nuclide content, and produced particle size distributions and morphology; measurements of volatile fission product species enhanced sorption--enrichment factors onto respirable particles; and, status on determination of the spent fuel ratio, SFR, needed for scaling studies. Emphasis is provided on recent Phase 3 tests using depleted uranium oxide pellets plus non-radioactive fission product dopants in surrogate spent fuel test rodlets, plus the latest surrogate cerium oxide results and aerosol laboratory supporting calibration work. The DUO 2 , CeO 2 , plus fission product dopant aerosol particle results are compared with available historical data. We also provide a status review on continuing preparations for the final Phase 4 in this program, tests using individual short rodlets containing actual spent fuel from U.S. PWR reactors, with both high- and lower-burnup fuel. The source-term data, aerosol results, and program design have been tailored to support and guide follow-on computer modeling of aerosol dispersal hazards and radiological consequence assessments. This spent fuel sabotage, aerosol test program was performed primarily at Sandia National Laboratories, with support provided by both the U.S. Department of Energy and the Nuclear Regulatory Commission. This program has significant input from, and is cooperatively supported and

  2. Learning in Ressource Mangement

    DEFF Research Database (Denmark)

    Hoffmann, Birgitte; Agger, Annika

    2005-01-01

    This paper explores the roles of NGOs as intermediaries in societal learning processes. The key question is how NGOs facilitate learning in the water sector, and the paper explores the notion of critical friends to grasp their dual position as both partner in the development and critical...... strategies and finally in relation to changes of the water management discourse. Developing the notion ‘critical friend’ the paper discusses how the NGO facilitates learning processes. Traditional approaches to learning are often based on the ‘petrol station pedagogy’ in which knowledge figuratively ‘are...

  3. Bachelor of Science-Engineering Technology Program and Fuel Cell Education Program Concentration

    Energy Technology Data Exchange (ETDEWEB)

    Block, David L. [Florida Solar Energy Center, Cocoa, FL (United States); Sleiti, Ahmad [Univ. of North Carolina, Charlotte, NC (United States)

    2011-09-19

    The Hydrogen and Fuel Cell Technology education project has addressed DOE goals by supplying readily available, objective, technical, and accurate information that is available to students, industry and the public. In addition, the program has supplied educated trainers and training opportunities for the next generation workforce needed for research, development, and demonstration activities in government, industry, and academia. The project has successfully developed courses and associated laboratories, taught the new courses and labs and integrated the HFCT option into the accredited engineering technology and mechanical engineering programs at the University of North Carolina at Charlotte (UNCC). The project has also established ongoing collaborations with the UNCC energy related centers of the Energy Production & Infrastructure Center (EPIC), the NC Motorsports and Automotive Research Center (NCMARC) and the Infrastructure, Design, Environment and Sustainability Center (IDEAS). The results of the project activities are presented as two major areas – (1) course and laboratory development, offerings and delivery, and (2) program recruitment, promotions and collaborations. Over the project period, the primary activity has been the development and offering of 11 HFCT courses and accompanying laboratories. This process has taken three years with the courses first being developed and then offered each year over the timeframe.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-31

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

  5. Alternative fuels for vehicles fleet demonstration program. Final report, volume 2: Appendices

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-06-01

    The Alternative Fuels for Vehicles Fleet Demonstration Program (AFV-FDP) was a multiyear effort to collect technical data for use in determining the costs and benefits of alternative-fuel vehicles (AFVs) in typical applications in New York State. This report, Volume 2, includes 13 appendices to Volume 1 that expand upon issues raised therein. Volume 1 provides: (1) Information about the purpose and scope of the AFV-FDP; (2) A summary of AFV-FDP findings organized on the basis of vehicle type and fuel type; (3) A short review of the status of AFV technology development, including examples of companies in the State that are active in developing AFVs and AFV components; and (4) A brief overview of the status of AFV deployment in the State. Volume 3 provides expanded reporting of AFV-FDP technical details, including the complete texts of the brochure Garage Guidelines for Alternative Fuels and the technical report Fleet Experience Survey Report, plus an extensive glossary of AFV terminology. The appendices cover a wide range of issues including: emissions regulations in New York State; production and health effects of ozone; vehicle emissions and control systems; emissions from heavy-duty engines; reformulated gasoline; greenhouse gases; production and characteristics of alternative fuels; the Energy Policy Act of 1992; the Clean Fuel Fleet Program; garage design guidelines for alternative fuels; surveys of fleet managers using alternative fuels; taxes on conventional and alternative fuels; and zero-emission vehicle technology.

  6. Alternative-fueled truck demonstration natural gas program: Caterpillar G3406LE development and demonstration

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-06-01

    In 1990, the California Energy Commission, the South Coast Air Quality Management District, and the Southern California Gas Company joined together to sponsor the development and demonstration of compressed natural gas engines for Class 8 heavy-duty line-haul trucking applications. This program became part of an overall Alternative-Fueled Truck Demonstration Program, with the goal of advancing the technological development of alternative-fueled engines. The demonstration showed natural gas to be a technically viable fuel for Class 8 truck engines.

  7. Whole-core damage analysis of EBR-II driver fuel elements following SHRT program

    International Nuclear Information System (INIS)

    Chang, L.K.; Koenig, J.F.; Porter, D.L.

    1987-01-01

    In the Shutdown Heat Removal Testing (SHRT) program in EBR-II, fuel element cladding temperatures of some driver subassemblies were predicted to exceed temperatures at which cladding breach may occur. A whole-core thermal analysis of driver subassemblies was performed to determine the cladding temperatures of fuel elemnts, and these temperatures were used for fuel element damage calculation. The accumulated cladding damage of fuel element was found to be very small and fuel element failure resulting from SHRT transients is unlikely. No element breach was noted during the SHRT transients. The reactor was immediately restarted after the most severe SHRT transient had been completed and no driver fuel breach has been noted to date. (orig.)

  8. Optimization of fuel management and control poison of a nuclear power reactor by dynamic programming

    International Nuclear Information System (INIS)

    Lima, C.A.R. de.

    1977-01-01

    The distribution of fuel and control poison in a nuclear reactor was optimized by the method of Dynamic Programming. A 620 M We Pressurized Water Reactor similar to Angra-1 was studied. The reactor operation was simulated in a IBM-1130 computer. Two fuel shuffling schemes and three poison management schemes were simultaneously employed in the reactor divided into three regions of equal volume and two consecutive stages were studied in order to determine the influence of poison management on the optimum fuel management policy. When uniform poisoning on all the three regions was permitted the traditional out-in fuel management policy proved to be more economic. On introducing simultaneous poison management, the optimum fuel management sequence was found to be different. The results obtained indicate a stronger interaction between the fuel management and the poison management than anticipated in previous works. (author)

  9. HEXBU-3D, a three-dimensional PWR-simulator program for hexagonal fuel assemblies

    International Nuclear Information System (INIS)

    Karvinen, E.

    1981-06-01

    HEXBU-3D is a three-dimensional nodal simulator program for PWR reactors. It is designed for a reactor core that consists of hexagonal fuel assemblies and of big follower-type control assemblies. The program solves two-group diffusion equations in homogenized fuel assembly geometry by a sophisticated nodal method. The treatment of feedback effects from xenon-poisoning, fuel temperature, moderator temperature and density and soluble boron concentration are included in the program. The nodal equations are solved by a fast two-level iteration technique and the eigenvalue can be either the effective multiplication factor or the boron concentration of the moderator. Burnup calculations are performed by tabulated sets of burnup-dependent cross sections evaluated by a cell burnup program. HEXBY-3D has been originally programmed in FORTRAN V for the UNIVAC 1108 computer, but there is also another version which is operable on the CDC CYBER 170 computer. (author)

  10. RERTR program activities related to the development and application of new LEU fuels

    International Nuclear Information System (INIS)

    Travelli, A.

    1983-01-01

    The statue of the U.S. Reduced Enrichment Research and Test Reactor (RERTR) Program is reviewed. After a brief outline of RERTR Program objectives and goals, program accomplishments are discussed with emphasis on the development, demonstration and application of new LEU fuels. Most program activities have proceeded as planned, and a combination of two silicide fuels (U 3 Si 2 -Al and U 3 Si-Al) holds excellent promise for achieving the long-term program goals. Current plans and schedules project the uranium density of qualified RERTR fuels for plate-type reactors to grow by approximately 1 g U/cm 3 each year, from the current 1.7 g U/cm 3 to the 7.0 g U/cm 3 which will be reached in late 1988. The technical needs of research and test reactors for HEU exports are also forecasted to undergo a gradual but dramatic decline in the coming years

  11. Progress of the United States foreign research reactor spent nuclear fuel acceptance program

    International Nuclear Information System (INIS)

    Huizenga, D.G.; Clapper, M.; Thrower, A.W.

    2002-01-01

    The United States Department of Energy (DOE), in consultation with the Department of State (DOS), adopted the Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel in May 1996. To date, the Foreign Research Reactor (FRR) Spent Nuclear Fuel (SNF) Acceptance Program has completed 23 shipments. Almost 5000 spent fuel assemblies from eligible research reactors throughout the world have been accepted into the United States under this program. Over the past year, another cross-country shipment of fuel was accomplished, as well as two additional shipments in the fourth quarter of calendar year 2001. These shipments attracted considerable safeguards oversight since they occurred post September 11. Recent guidance from the Nuclear Regulatory Commission (NRC) pertaining to security and safeguards issues deals directly with the transport of nuclear material. Since the Acceptance Program has consistently applied above regulatory safety enhancements in transport of spent nuclear fuel, this guidance did not adversely effect the Program. As the Program draws closer to its termination date, an increased number of requests for program extension are received. Currently, there are no plans to extend the policy beyond its current expiration date; therefore, eligible reactor operators interested in participating in this program are strongly encouraged to evaluate their inventory and plan for future shipments as soon as possible. (author)

  12. Application of linear programming and perturbation theory in optimization of fuel utilization in a nuclear reactor

    International Nuclear Information System (INIS)

    Zavaljevski, N.

    1985-01-01

    Proposed optimization procedure is fast due to application of linear programming. Non-linear constraints which demand iterative application of linear programming are slowing down the calculation. Linearization can be done by different procedures starting from simple empirical rules for fuel in-core management to complicated general perturbation theory with higher order of corrections. A mathematical model was formulated for optimization of improved fuel cycle. A detailed algorithm for determining minimum of fresh fuel at the beginning of each fuel cycle is shown and the problem is linearized by first order perturbation theory and it is optimized by linear programming. Numerical illustration of the proposed method was done for the experimental reactor mostly for saving computer time

  13. 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

  14. 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.

  15. K Basin Fuel Characterization Program Technical Baseline Summary

    International Nuclear Information System (INIS)

    SUYAMA, R.M.

    1999-01-01

    This document provides a summary of the systematic process used by the SNF Project to characterize K-Basin spent fuel, and to develop and apply the appropriate conservative safety margins to the resulting parameters for technical designs and safety analyses

  16. Broad Specification Fuels Combustion Technology Program. Phase 2

    Science.gov (United States)

    1990-10-01

    4 4C Where: M is the molecular weight of th hxth specie Nt is the mole fraction of the x specie a is the hydrogen to carbon ratio of the fuel...RATIO F’gure 7-15 Idle Emisions Characteristics of Variable Geometry Cornbusuom geometry combustor configurations as well. The remaining performance

  17. Alternative Fuels Data Center: City of Chicago Program Encourages Petroleum

    Science.gov (United States)

    lacked the funds to support such a project, but officials worked with the Chicago Area Clean Cities (CACC networking opportunities. This photo shows a parked compressed natural gas van the City of Chicago's fleet Infrastructure: 11 private E85 fueling stations Related Links Ethanol Natural Gas Chicago Area Clean Cities

  18. FREC-4A: a computer program to predict fuel rod performance under normal reactor operation

    International Nuclear Information System (INIS)

    Harayama, Yasuo; Izumi, Fumio

    1981-10-01

    The program FREC-4A (Fuel Reliability Evaluation Code-version 4A) is used for predicting fuel rod performance in normal reactor operation. The performance is calculated in accordance with the irradiation history of fuel rods. Emphasis is placed on the prediction of the axial elongation of claddings induced by pellet-cladding mechanical interaction, including the influence of initially preloaded springs inserted in fuel rod lower plenums. In the FREC-4A, an fuel rod is divided into axial segments. In each segment, it is assumed that the temperature, stress and strain are axi-symmetrical, and the axial strain in constant in fuel pellets and in a cladding, though the values in the pellets and in the cladding are different. The calculation of the contact load and the clearance along the length of a fuel rod and the stress and strain in each segment is explained. The method adopted in the FREC-4A is simple, and suitable to predict the deformation of fuel rods over their full length. This report is described on the outline of the program, the method of solving the stiffness equations, the calculation models, the input data such as irradiation history, output distribution, material properties and pores, the printing-out of input data and calculated results. (Kako, I.)

  19. Considerations for a national program on spent fuel management

    International Nuclear Information System (INIS)

    Lopez-Perez, B.; Melches-Serrano, C.

    1980-01-01

    The spent fuel discharged from the two LWR's that are in operation (Zorita, 160 MW PWR, and Santa Maria de Garona, 460 MW BWR) is being reprocessed under contracts with BNFL; these contracts will expire in the next few years. The fuel discharged from Vandelos (50 MW GCR) is being reprocessed by Cogema under a long-term contract. No new reprocessing contracts for LWR's in operation, under construction, or planned have been signed or are being considered for the near future. The plutonium and the residual uranium contained in LWR spent fuel are considered important potential energy resources. They are especially valuable for countries such as Spain, which is short of energy resources, and they might be used in the future in fast breeder or thermal reactors. This is the reason that, until reprocessing is justified and appropriate solutions to make reprocessing available are developed, Spain has decided to build the appropriate capacity for the temporary storage of spent fuel. The capacity is being achieved, on short term, by the extension of AR storage capacity. It is being achieved, at medium or longer term, by the construction of centralized AFR facilities to serve all Spanish nuclear power plants. Spanish utilities are undertaking the expansion of reactor storage capacities, using densified racks, to increment capacity to at least 8 to 10 reloads, in addition to full core discharge capacity. Spain has the time and the financial and technical resources to implement a national solution for spent fuel storage. Financial strategy, technology choice, and licensing considerations are under examination in order to make a decision for medium- and long-term storage alternatives

  20. A review of low carbon fuel policies: Principles, program status and future directions

    International Nuclear Information System (INIS)

    Yeh, Sonia; Witcover, Julie; Lade, Gabriel E.; Sperling, Daniel

    2016-01-01

    A low carbon fuel standard (LCFS) is a market-based policy that specifies declining standards for the average lifecycle fuel carbon intensity (AFCI) of transportation fuels sold in a region. This paper: (i) compares transportation fuel carbon policies in terms of their economic efficiency, fuel price impacts, greenhouse gas emission reductions, and incentives for innovation; (ii) discusses key regulatory design features of LCFS policies; and (iii) provides an update on the implementation status of LCFS policies in California, the European Union, British Columbia, and Oregon. The economics literature finds that an intensity standard implicitly taxes emissions and subsidizes output. The output subsidy results in an intensity standard being inferior to a carbon tax in a first-best world, although the inefficiency can be corrected with a properly designed consumption tax (or mitigated by a properly designed carbon tax or cap-and-trade program). In California, from 2011 to 2015 the share of alternative fuels in the regulated transportation fuels pool increased by 30%, and the reported AFCI of all alternative fuels declined 21%. LCFS credit prices have varied considerably, rising to above $100/credit in the first half of 2016. LCFS programs in other jurisdictions share many features with California's, but have distinct provisions as well. - Highlights: • LCFS is a market-based policy that sets standards for carbon intensity of fuels. • We compare efficiency, price impacts, GHG emissions, and innovation of C policies. • In California, reported carbon intensity of alternative fuels declined 21% 2011–2015. • LCFS credit prices have varied considerably, rising to above $100/credit in the first half of 2016. • Other LCFS programs share many features with CA's and have distinct provisions.

  1. Fuel poverty, affordability, and energy justice in England: Policy insights from the Warm Front Program

    International Nuclear Information System (INIS)

    Sovacool, Benjamin K.

    2015-01-01

    Millions of homes around the world suffer from “fuel poverty,” commonly defined as the necessity to spend more than 10 percent of their income paying energy bills. This article first discusses how home energy efficiency schemes, such as those that pay to weatherize doors and windows, install insulation, and give free energy audits, can significantly reduce the prevalence of fuel poverty. It then examines the “Warm Front” program in England, which over the course of 2000–2013 saw 2.3 million “fuel poor” British homes receive energy efficiency upgrades to save them money and improve their overall health. Warm Front not only lessened the prevalence of fuel poverty; it cut greenhouse gas emissions, produced an average extra annual income of £1894.79 per participating household, and reported exceptional customer satisfaction with more than 90 percent of its customers praising the scheme. This study details the history, benefits, and challenges of the program, and it teases out six noteworthy lessons for energy analysts, planners, and policymakers. - Highlights: • Millions of homes around the world suffer from “fuel poverty”. • The “Warm Front” program in England saw 2.3 million “fuel poor” British homes receive energy efficiency upgrades. • Warm Front ran from over the course of 2000–2013. • Warm Front lessened the prevalence of fuel poverty, cut greenhouse gas emissions, and saved households money. • Warm Front offers important lessons for energy analysts, planners, and policymakers.

  2. Studsvik's implementation of the DOE spent fuel return program: The US perspective

    International Nuclear Information System (INIS)

    Grover, Stephen; Browser, Rita C.

    1996-01-01

    The government of Sweden has long been a proponent of nuclear nonproliferation policies. This includes laws governing the conduct of the Studsvik R2 Reactor in Nykoeping, Sweden. Studsvik became a participant in the RERTR program early on and was one of the first research reactors to convert from Highly Enriched Uranium (HEU) to Low Enriched Uranium (LEU) fuel. Since Studsvik operates the reactor on a commercial basis, it is important to fully comply with all Swedish laws and regulations for the operation of a nuclear reactor. This includes development of a proper long term program for the final disposition for the irradiated fuel. For this reason it was very important to Studsvik to be involved in the process for development of the DOE Spent Fuel Return Program as early as possible. This process included input into the Environmental Impact Statement process, early contractual negotiations after the issuance of the Record of Decision, participation in litigation involving the Spent Fuel Return Program, and plans for return of HEU fuel on the first European shipment. Not only was program involvement important, but actual program implementation was necessary for Studsvik to maintain its reactor license. Studsvik needed to have fuel included in the first shipment in the return program not only for licensing considerations, but also to reduce the proliferation risk associated with the storage of non-self protecting fuel. To ensure participation in the first shipment, Studsvik's activities included the early development of Appendix A's, coordinating the shipment, obtaining approvals and authorizations, and other technical aspects. (author)

  3. Spent fuel sabotage aerosol ratio program : FY 2004 test and data summary

    International Nuclear Information System (INIS)

    Brucher, Wenzel; Koch, Wolfgang; Pretzsch, Gunter Guido; Loiseau, Olivier; Mo, Tin; Billone, Michael C.; Autrusson, Bruno A.; Young, F. I.; Coats, Richard Lee; Burtseva, Tatiana; Luna, Robert Earl; Dickey, Roy R.; Sorenson, Ken Bryce; Nolte, Oliver; Thompson, Nancy Slater; Hibbs, Russell S.; Gregson, Michael Warren; Lange, Florentin; Molecke, Martin Alan; Tsai, Han-Chung

    2005-01-01

    This multinational, multi-phase spent fuel sabotage test program is quantifying the aerosol particles produced when the products of a high energy density device (HEDD) interact with and explosively particulate test rodlets that contain pellets of either surrogate materials or actual spent fuel. This program has been underway for several years. This program provides data that are relevant to some sabotage scenarios in relation to spent fuel transport and storage casks, and associated risk assessments. The program also provides significant technical and political benefits in international cooperation. We are quantifying the Spent Fuel Ratio (SFR), the ratio of the aerosol particles released from HEDD-impacted actual spent fuel to the aerosol particles produced from surrogate materials, measured under closely matched test conditions, in a contained test chamber. In addition, we are measuring the amounts, nuclide content, size distribution of the released aerosol materials, and enhanced sorption of volatile fission product nuclides onto specific aerosol particle size fractions. These data are the input for follow-on modeling studies to quantify respirable hazards, associated radiological risk assessments, vulnerability assessments, and potential cask physical protection design modifications. This document includes an updated description of the test program and test components for all work and plans made, or revised, during FY 2004. It also serves as a program status report as of the end of FY 2004. All available test results, observations, and aerosol analyses plus interpretations--primarily for surrogate material Phase 2 tests, series 2/5A through 2/9B, using cerium oxide sintered ceramic pellets are included. Advanced plans and progress are described for upcoming tests with unirradiated, depleted uranium oxide and actual spent fuel test rodlets. This spent fuel sabotage--aerosol test program is coordinated with the international Working Group for Sabotage Concerns of

  4. Spent fuel sabotage aerosol ratio program : FY 2004 test and data summary.

    Energy Technology Data Exchange (ETDEWEB)

    Brucher, Wenzel (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Koch, Wolfgang (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Pretzsch, Gunter Guido (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Loiseau, Olivier (Institut de Radioprotection et de Surete Nucleaire, France); Mo, Tin (U.S. Nuclear Regulatory Commission, Washington, DC); Billone, Michael C. (Argonne National Laboratory, Argonne, IL); Autrusson, Bruno A. (Institut de Radioprotection et de Surete Nucleaire, France); Young, F. I. (U.S. Nuclear Regulatory Commission, Washington, DC); Coats, Richard Lee; Burtseva, Tatiana (Argonne National Laboratory, Argonne, IL); Luna, Robert Earl; Dickey, Roy R.; Sorenson, Ken Bryce; Nolte, Oliver (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Thompson, Nancy Slater (U.S. Department of Energy, Washington, DC); Hibbs, Russell S. (U.S. Department of Energy, Washington, DC); Gregson, Michael Warren; Lange, Florentin (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Molecke, Martin Alan; Tsai, Han-Chung (Argonne National Laboratory, Argonne, IL)

    2005-07-01

    This multinational, multi-phase spent fuel sabotage test program is quantifying the aerosol particles produced when the products of a high energy density device (HEDD) interact with and explosively particulate test rodlets that contain pellets of either surrogate materials or actual spent fuel. This program has been underway for several years. This program provides data that are relevant to some sabotage scenarios in relation to spent fuel transport and storage casks, and associated risk assessments. The program also provides significant technical and political benefits in international cooperation. We are quantifying the Spent Fuel Ratio (SFR), the ratio of the aerosol particles released from HEDD-impacted actual spent fuel to the aerosol particles produced from surrogate materials, measured under closely matched test conditions, in a contained test chamber. In addition, we are measuring the amounts, nuclide content, size distribution of the released aerosol materials, and enhanced sorption of volatile fission product nuclides onto specific aerosol particle size fractions. These data are the input for follow-on modeling studies to quantify respirable hazards, associated radiological risk assessments, vulnerability assessments, and potential cask physical protection design modifications. This document includes an updated description of the test program and test components for all work and plans made, or revised, during FY 2004. It also serves as a program status report as of the end of FY 2004. All available test results, observations, and aerosol analyses plus interpretations--primarily for surrogate material Phase 2 tests, series 2/5A through 2/9B, using cerium oxide sintered ceramic pellets are included. Advanced plans and progress are described for upcoming tests with unirradiated, depleted uranium oxide and actual spent fuel test rodlets. This spent fuel sabotage--aerosol test program is coordinated with the international Working Group for Sabotage Concerns of

  5. Pathways to Commercial Success. Technologies and Products Supported by the Fuel Cell Technologies Program

    Energy Technology Data Exchange (ETDEWEB)

    none,

    2011-09-01

    This FY 2011 report updates the results of an effort to identify and characterize commercial and near-commercial (emerging) technologies and products that benefited from the support of the Fuel Cell Technologies Program and its predecessor programs within DOE's Office of Energy Efficiency and Renewable Energy.

  6. Pathways to Commercial Success. Technologies and Products Supported by the Fuel Cell Technologies Program - 2012

    Energy Technology Data Exchange (ETDEWEB)

    none,

    2012-09-01

    This FY 2012 report updates the results of an effort to identify and characterize commercial and near-commercial (emerging) technologies and products that benefited from the support of the Fuel Cell Technologies Program and its predecessor programs within DOE's Office of Energy Efficiency and Renewable Energy.

  7. A fast-running fuel management program for a CANDU reactor

    International Nuclear Information System (INIS)

    Choi, Hangbok

    2000-01-01

    A fast-running fuel management program for a CANDU reactor has been developed. The basic principle of this program is to select refueling channels such that the reference reactor conditions are maintained by applying several constraints and criteria when selecting refueling channels. The constraints used in this program are the channel and bundle power and the fuel burnup. The final selection of the refueling channel is determined based on the priority of candidate channels, which enhances the reactor power distribution close to the time-average model. The refueling simulation was performed for a natural uranium CANDU reactor and the results were satisfactory

  8. The post-irradiation examination of fuel in support of Bruce A Nuclear Division fueling with flow program

    International Nuclear Information System (INIS)

    Montin, J.; Sagat, S.

    1995-10-01

    Bruce A Nuclear Division (BAND) units are operating at ∼ 75% of full power, because of the potential of a power pulse in the event of an inlet header break. As a result, BAND is converting to fueling with flow, to eliminate the potential of a power pulse and to allow for full-power operation. Concerns regarding the integrity of the end-of-life (EOL) bundles interacting with the latch at the downstream end of the fuel channel were raised. BAND carried out a test program in which EOL bundles in the upstream position of 13 of Unit 2 were cascaded into the downstream latch position 1 of another channel. Six of twelve cascaded bundles and two typical EOL position 13 (benchmark) bundles were selected for post-irradiation examination (PIE). Incipient cracks were found in the benchmark bundles. Metallographic and fractographic examination, along with crack dating, and hydrogen and deuterium analyses, indicated that the incipient cracks were the result of delayed-hydride assisted cracking at the EOL. Consequently, Ontario Hydro changed the design of the outlet shield plug to support all three rings of the fuel bundle, to minimize stress and prevent end plate cracking. Also, an ultrasonic end plate inspection tool (UT) was developed and located in the fuel bay, to inspect fuel-bundle end plates for cracks. A second test was done involving a series of four bundle cascades in BAND Unit 4 channels that had new outlet shield plugs. The latch bundles were discharged after a hot shutdown. The cascaded Unite 2 and Unit 4 latch bundles were checked for cracks using the UT. The PIE found incipient cracks or less-than-ideal welds in the assembly welds of fuel elements from Unit 2 (latch-supported fuel bundles) that had been identified by the UT as having incipient cracks. No incipient cracks were found in the assemble welds of fuel elements from Unit 4 (new outlet shield-supported fuel bundles) confirming the UT results. (author). 5 refs., 8 figs

  9. 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)

  10. Status of the inert matrix fuel program at PSI

    International Nuclear Information System (INIS)

    Ledergerber, G.; Degueldre, C.; Kasemeyer, U.; Stanculescu, A.; Paratte, J.M.; Chawla, R.

    1997-01-01

    Incineration of plutonium by a once-through cycle in LWRs utilising an inert matrix based fuel may prove to be an attractive way of making use of the energy of fissile plutonium and reducing both the hazard potential and the volumes of the waste. Yttria stabilised zirconia forms a solid solution with oxides of rare earth elements (e.g. erbium, cerium) and some actinides. The small absorption cross section, the excellent stability under irradiation, and the insolubility in acids and water recommends this material as an inert matrix. Neutronics calculations with erbium as burnable poison show that these compositions would be optimal from the reactivity point of view. A fuel element with an improved reactivity behaviour over its life cycle has been designed for possible introduction into a heterogeneous LWR core. (author). 16 refs., 1 tab., 10 figs

  11. Evaluation of bioassay program at uranium fuel fabrication plants

    International Nuclear Information System (INIS)

    Biggs, D.

    1981-03-01

    Results of a comprehensive study of urinalysis, lung burden and personal air sample measurements for workers at a uranium fuel fabrication plant are presented. Correlations between measurements were found and regression models used to explain the relationship between lung burden, daily intakes and urinary excretions of uranium. Assuming the ICRP lung model, the lung burden histories of ten workers were used to estimate the amounts in each of the long-term compartments of the lung. Estimates of the half lives of each compartment and of the maximum relative contributions to the urine from each compartment are given. These values were then used to predict urinary excretions from the long-term compartments for workers at another fuel fabrication plant. The standard error of estimate compared well with the daily variation in urinary excretion. (author)

  12. Consolidated fuel reprocessing program. Developments for the future in reprocessing

    International Nuclear Information System (INIS)

    Burch, W.D.

    1982-01-01

    The future reprocessing developments focus on three major areas: (1) the retention of gaseous fission products to reduce off-site doses to very low values; (2) the initial steps of breakdown, shearing, and dissolution of breeder fuels; and (3) advanced facility and equipment concepts, which are expected to lead to a reliable, cost-effective, totally remotely operated and maintained plant. Work in the first area - removal of fission gases (the most important of which is 85 Kr) - is largely completed through tracer and bench-scale engineering equipment. Efforts are now mainly devoted to breeder fuels and advanced remote concepts. A facility, the Integrated Equipment Test Facility, which will be used to carry out much of this work, is nearing completion in Oak Ridge. In it a large, simulated, remote reprocessing cell will house a disassembly-shear machine for either breeder or LWR fuels, a rotary continuous dissolver, a solvent extraction cycle utilizing a new generation of centrifugal contactors, and related equipment

  13. Identification and Analysis of Critical Gaps in Nuclear Fuel Cycle Codes Required by the SINEMA Program

    International Nuclear Information System (INIS)

    Miron, Adrian; Valentine, Joshua; Christenson, John; Hawwari, Majd; Bhatt, Santosh; Dunzik-Gougar, Mary Lou; Lineberry, Michael

    2009-01-01

    The current state of the art in nuclear fuel cycle (NFC) modeling is an eclectic mixture of codes with various levels of applicability, flexibility, and availability. In support of the advanced fuel cycle systems analyses, especially those by the Advanced Fuel Cycle Initiative (AFCI), University of Cincinnati in collaboration with Idaho State University carried out a detailed review of the existing codes describing various aspects of the nuclear fuel cycle and identified the research and development needs required for a comprehensive model of the global nuclear energy infrastructure and the associated nuclear fuel cycles. Relevant information obtained on the NFC codes was compiled into a relational database that allows easy access to various codes' properties. Additionally, the research analyzed the gaps in the NFC computer codes with respect to their potential integration into programs that perform comprehensive NFC analysis.

  14. Identification and Analysis of Critical Gaps in Nuclear Fuel Cycle Codes Required by the SINEMA Program

    Energy Technology Data Exchange (ETDEWEB)

    Adrian Miron; Joshua Valentine; John Christenson; Majd Hawwari; Santosh Bhatt; Mary Lou Dunzik-Gougar: Michael Lineberry

    2009-10-01

    The current state of the art in nuclear fuel cycle (NFC) modeling is an eclectic mixture of codes with various levels of applicability, flexibility, and availability. In support of the advanced fuel cycle systems analyses, especially those by the Advanced Fuel Cycle Initiative (AFCI), Unviery of Cincinnati in collaboration with Idaho State University carried out a detailed review of the existing codes describing various aspects of the nuclear fuel cycle and identified the research and development needs required for a comprehensive model of the global nuclear energy infrastructure and the associated nuclear fuel cycles. Relevant information obtained on the NFC codes was compiled into a relational database that allows easy access to various codes' properties. Additionally, the research analyzed the gaps in the NFC computer codes with respect to their potential integration into programs that perform comprehensive NFC analysis.

  15. Development of a Computer Program for the Analysis Logistics of PWR Spent Fuels

    International Nuclear Information System (INIS)

    Choi, Heui Joo; Choi, Jong Won; Cha, Jeong Hun

    2008-01-01

    It is expected that the temporary storage facilities at the nuclear power plants will be full of the spent fuels within 10 years. Provided that a centralized interim storage facility is constructed along the coast of the Korean peninsula to solve this problem, a substantial amount of spent fuels should be transported by sea or by land every year. In this paper we developed a computer program for the analysis of transportation logistics of the spent fuels from 4 different nuclear power plant sites to the hypothetical centralized interim storage facility and the final repository. Mass balance equations were used to analyze the logistics between the nuclear power plants and the interim storage facility. To this end a computer program, CASK, was developed by using the VISUAL BASIC language. The annual transportation rates of spent fuels from the four nuclear power plant sites were determined by using the CASK program. The parameter study with the program illustrated the easiness of logistics analysis. The program could be used for the cost analysis of the spent fuel transportation as well.

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

    International Nuclear Information System (INIS)

    Dixon, Cathy

    2012-01-01

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

  17. Center for Alternative Fuels Research Program | College of Engineering &

    Science.gov (United States)

    Electrical Engineering Instructional Laboratories Student Resources Industrial & Manufacturing Engineering Industrial & Manufacturing Engineering Academic Programs Industrial & Manufacturing Engineering Major Industrial & Manufacturing Engineering Minor Industrial & Manufacturing Engineering

  18. United States Department of Energy commercial reactor spent fuel programs being conducted at the Idaho National Engineering Laboratory

    International Nuclear Information System (INIS)

    Piscitella, R.R.; Rasmussen, T.L.; Uhl, D.L.

    1987-01-01

    The Idaho National Engineering Laboratory participation in OCRWM programs includes the Spent Fuel Storage Cask Testing Program, Dry Rod Consolidation Technology Program, Prototypical Consolidation Demonstration Program, the Nuclear Fuel Services Project, and the Cask Systems Acquisition Program. The DOE has entered into a cooperative agreement with Virginia Power and the Electric Power Research Institute to demonstrate storage of commercial spent fuel in steel storage casks. The Program conducted heat transfer and shielding tests with three storage casks with intact spent fuel assemblies and two casks with consolidated spent fuel rods, one of which was previously tested with intact fuel, and provides test information in support of Virginia Power's at-reactor dry storage licensing effort. 3 figs., 1 tab

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

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2017-10-16

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

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

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2016-11-01

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

  1. DANCOFF-MC: A program to calculate Dancoff factors in CANDU type fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Feher, S; Valko, J

    1992-12-01

    The objective of DANCOFF-MC is the evaluation of Dancoff factors for cylindrical fuel rods arranged parallel in various and complicated bundle geometries. No interaction with fuel rods in any of the other bundles are considered due to the large distance, in mean free paths, between the buldes. Using a common basic algorithm three versions of the program have been written so far: The DANCOFF-MC-2, the DANCOFF-MC-19 and the DANCOFF-MC-27. (orig./HP).

  2. Summary report on fuel development and miniplate fabrication for the RERTR Program, 1978 to 1990

    International Nuclear Information System (INIS)

    Wiencek, T.C.

    1995-08-01

    This report summarizes the efforts of the Fabrication Technology Section at Argonne National Laboratory in the program of Reduced Enrichment Research and Test Reactors (RERTR). The main objective of this program was to reduce the amount of high enriched (∼93% 235 U) uranium (HEU) used in nonpower reactors. Conversion from low-density (0.8--1.6 g U/cm 3 ) HEU fuel elements to highly loaded (up to 7 g U/cm 3 ) low-enrichment ( 235 U) uranium (LEU) fuel elements allows the same reactor power levels, core designs and sizes to be retained while greatly reducing the possibility of illicit diversion of HEU nuclear fuel. This document is intended as an overview of the period 1978--1990, during which the Section supported this project by fabricating mainly powder metallurgy uranium-silicide dispersion fuel plates. Most of the subjects covered in detail are fabrication-related studies of uranium silicide fuels and fuel plate properties. Some data are included for out-of-pile experiments such as corrosion and compatibility tests. Also briefly covered are most other aspects of the RERTR program such as irradiation tests, full-core demonstrations, and technology transfer. References included are for further information on most aspects of the entire program. A significant portion of the report is devoted to data that were never published in their entirety. The appendices contain a list of previous RERTR reports, ANL fabrication procedures, calculations for phases present in two-phase fuels, chemical analysis of fuels, miniplate characteristics, and a summary of bonding runs made by hot isostatic pressing

  3. Summary report on fuel development and miniplate fabrication for the RERTR Program, 1978 to 1990

    Energy Technology Data Exchange (ETDEWEB)

    Wiencek, T.C. [Argonne National Lab., IL (United States). Energy Technology Div.

    1995-08-01

    This report summarizes the efforts of the Fabrication Technology Section at Argonne National Laboratory in the program of Reduced Enrichment Research and Test Reactors (RERTR). The main objective of this program was to reduce the amount of high enriched ({approx}93% {sup 235}U) uranium (HEU) used in nonpower reactors. Conversion from low-density (0.8--1.6 g U/cm{sup 3}) HEU fuel elements to highly loaded (up to 7 g U/cm{sup 3}) low-enrichment (<20% {sup 235}U) uranium (LEU) fuel elements allows the same reactor power levels, core designs and sizes to be retained while greatly reducing the possibility of illicit diversion of HEU nuclear fuel. This document is intended as an overview of the period 1978--1990, during which the Section supported this project by fabricating mainly powder metallurgy uranium-silicide dispersion fuel plates. Most of the subjects covered in detail are fabrication-related studies of uranium silicide fuels and fuel plate properties. Some data are included for out-of-pile experiments such as corrosion and compatibility tests. Also briefly covered are most other aspects of the RERTR program such as irradiation tests, full-core demonstrations, and technology transfer. References included are for further information on most aspects of the entire program. A significant portion of the report is devoted to data that were never published in their entirety. The appendices contain a list of previous RERTR reports, ANL fabrication procedures, calculations for phases present in two-phase fuels, chemical analysis of fuels, miniplate characteristics, and a summary of bonding runs made by hot isostatic pressing.

  4. The post irradiation examination of fuel in support of Bruce A nuclear division fueling with flow program

    International Nuclear Information System (INIS)

    Montin, J.; Sagat, S.; Day, R.; Novak, J.; Bromfield, H.

    1995-01-01

    Bruce A Nuclear Division (BAND) units are operating at ∼ 75% of full power, because of the potential of a power pulse in the event of an inlet header break. As a result, BAND is converting to fueling with flow, to eliminate the potential of a power pulse and to allow for full-power operation. Concerns regarding the integrity of the end-of-life (EOL) bundles interacting with the latch at the downstream end of the fuel channel were raised. BAND carried out a test program in which EOL bundles in the upstream position 13 of Unit 2 were cascaded into the downstream latch position 1 of another channel. Six of twelve cascaded bundles and two typical EOL position 13 (benchmark) bundles were selected for post-irradiation examination (PIE). Incipient cracks were found in the assembly welds (endplateto-endcap welds) of all six cascaded bundles. No incipient cracks were found in the benchmark bundles. Metallographic and fractographic examination, along with crack dating, and hydrogen and deuterium analyses, indicated that the incipient cracks were the result of delayed-hydride assisted cracking at the EOL. Consequently, Ontario Hydro changed the design of the outlet shield plug to support all three rings of the fuel bundle, to minimize stress and prevent endplate cracking. Also, an ultrasonic endplate inspection tool (UT) was developed and located in the fuel bay. to inspect fuelbundle endplates for cracks. A second test was done involving a series of four bundle cascades in BAND Unit 4 channels that had new outlet shield plugs. The latch bundles were discharged after a hot shutdown. The cascaded Unit 2 and Unit 4 latch bundles were checked for cracks using the UT. The PIE found incipient cracks or less-than-ideal welds in the assembly welds of fuel elements from Unit 2 (latch-supported fuel bundles) that had been identified by the UT as having incipient cracks. No incipient cracks were found in the assembly welds of fuel elements from Unit 4 (new outlet shield

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

    International Nuclear Information System (INIS)

    Felt, R.E.

    1980-05-01

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

  6. Spent fuel sabotage aerosol test program :FY 2005-06 testing and aerosol data summary

    International Nuclear Information System (INIS)

    Gregson, Michael Warren; Brockmann, John E.; Nolte, O.; Loiseau, O.; Koch, W.; Molecke, Martin Alan; Autrusson, Bruno; Pretzsch, Gunter Guido; Billone, M. C.; Lucero, Daniel A.; Burtseva, T.; Brucher, W; Steyskal, Michele D.

    2006-01-01

    This multinational, multi-phase spent fuel sabotage test program is quantifying the aerosol particles produced when the products of a high energy density device (HEDD) interact with and explosively particulate test rodlets that contain pellets of either surrogate materials or actual spent fuel. This program has been underway for several years. This program provides source-term data that are relevant to some sabotage scenarios in relation to spent fuel transport and storage casks, and associated risk assessments. This document focuses on an updated description of the test program and test components for all work and plans made, or revised, primarily during FY 2005 and about the first two-thirds of FY 2006. It also serves as a program status report as of the end of May 2006. We provide details on the significant findings on aerosol results and observations from the recently completed Phase 2 surrogate material tests using cerium oxide ceramic pellets in test rodlets plus non-radioactive fission product dopants. Results include: respirable fractions produced; amounts, nuclide content, and produced particle size distributions and morphology; status on determination of the spent fuel ratio, SFR (the ratio of respirable particles from real spent fuel/respirables from surrogate spent fuel, measured under closely matched test conditions, in a contained test chamber); and, measurements of enhanced volatile fission product species sorption onto respirable particles. We discuss progress and results for the first three, recently performed Phase 3 tests using depleted uranium oxide, DUO 2 , test rodlets. We will also review the status of preparations and the final Phase 4 tests in this program, using short rodlets containing actual spent fuel from U.S. PWR reactors, with both high- and lower-burnup fuel. These data plus testing results and design are tailored to support and guide, follow-on computer modeling of aerosol dispersal hazards and radiological consequence assessments

  7. Industry perspectives on the U.S. used nuclear fuel management program

    International Nuclear Information System (INIS)

    Matzie, R.

    2001-01-01

    The U.S. used nuclear fuel management program is at a critical juncture. It has been 17 years since the U.S. Congress authorized geologic disposal of used nuclear fuel and 12 years since Congress elected to study Yucca Mountain, Nevada, as the nation's lone long-term disposal option. During this period, Yucca Mountain's ability to isolate long-lived radionuclides was the subject of considerable scientific study. Concurrently, specialists expended significant design effort on the engineered barriers that work in conjunction with the proposed repository's natural features to ensure that health protection standards are met. Confidence as to the soundness of the geologic disposal option is high. While technical progress on geologic disposal is substantial, policy level decision-making needed to move the program forward remains lacking. A number of important policy questions concerning used fuel management remain unanswered as this fuel accumulates at 103 operating U.S. commercial nuclear power reactors at 65 sites. Wet fuel storage space at those locations is becoming increasingly scarce. This paper will review both the scientific successes and policy challenges on the road to geologic disposal. It will outline the nation's used nuclear fuel storage needs, highlight what the nation's scientists have learned about geologic disposal, discuss what is being done to meet near-term interim storage requirements, describe the emerging regulatory framework associated with the program and summarize the ongoing policy debate. (author)

  8. High-Uranium-Loaded U3O8-Al fuel element development program. Part 1

    International Nuclear Information System (INIS)

    Martin, M.M.

    1993-01-01

    The High-Uranium-Loaded U 3 O 8 -Al Fuel Element Development Program supports Argonne National Laboratory efforts to develop high-uranium-density research and test reactor fuel to accommodate use of low-uranium enrichment. The goal is to fuel most research and test reactors with uranium of less than 20% enrichment for the purpose of lowering the potential for diversion of highly-enriched material for nonpeaceful usages. The specific objective of the program is to develop the technological and engineering data base for U 3 O 8 -Al plate-type fuel elements of maximal uranium content to the point of vendor qualification for full scale fabrication on a production basis. A program and management plan that details the organization, supporting objectives, schedule, and budget is in place and preparation for fuel and irradiation studies is under way. The current programming envisions a program of about four years duration for an estimated cost of about two million dollars. During the decades of the fifties and sixties, developments at Oak Ridge National Laboratory led to the use of U 3 O 8 -Al plate-type fuel elements in the High Flux Isotope Reactor, Oak Ridge Research Reactor, Puerto Rico Nuclear Center Reactor, and the High Flux Beam Reactor. Most of the developmental information however applies only up to a uranium concentration of about 55 wt % (about 35 vol % U 3 O 8 ). The technical issues that must be addressed to further increase the uranium loading beyond 55 wt % U involve plate fabrication phenomena of voids and dogboning, fuel behavior under long irradiation, and potential for the thermite reaction between U 3 O 8 and aluminum

  9. Results from the DOE Advanced Gas Reactor Fuel Development and Qualification Program

    Energy Technology Data Exchange (ETDEWEB)

    David Petti

    2014-06-01

    Modular HTGR designs were developed to provide natural safety, which prevents core damage under all design basis accidents and presently envisioned severe accidents. The principle that guides their design concepts is to passively maintain core temperatures below fission product release thresholds under all accident scenarios. This level of fuel performance and fission product retention reduces the radioactive source term by many orders of magnitude and allows potential elimination of the need for evacuation and sheltering beyond a small exclusion area. This level, however, is predicated on exceptionally high fuel fabrication quality and performance under normal operation and accident conditions. Germany produced and demonstrated high quality fuel for their pebble bed HTGRs in the 1980s, but no U.S. manufactured fuel had exhibited equivalent performance prior to the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. The design goal of the modular HTGRs is to allow elimination of an exclusion zone and an emergency planning zone outside the plant boundary fence, typically interpreted as being about 400 meters from the reactor. To achieve this, the reactor design concepts require a level of fuel integrity that is better than that claimed for all prior US manufactured TRISO fuel, by a few orders of magnitude. The improved performance level is about a factor of three better than qualified for German TRISO fuel in the 1980’s. At the start of the AGR program, without a reactor design concept selected, the AGR fuel program selected to qualify fuel to an operating envelope that would bound both pebble bed and prismatic options. This resulted in needing a fuel form that could survive at peak fuel temperatures of 1250°C on a time-averaged basis and high burnups in the range of 150 to 200 GWd/MTHM (metric tons of heavy metal) or 16.4 to 21.8% fissions per initial metal atom (FIMA). Although Germany has demonstrated excellent performance of TRISO-coated UO

  10. In-pile irradiation test program and safety analysis report of the KAERI fuel for HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Wan; Ryu, Woo Suck; Byun, Taek Sang; Park, Jong Man; Lee, Byung Chul; Kim, Hack No; Park, Hee Tae; Kim, Chang Kyu [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-05-01

    Localization of HANARO fuel has been carried out successfully, and design and fabrication technologies of the fuel are recently arrived the final stage of development. The performance of the fuel which has been fabricated in KAERI is confirmed through out-of-pile characterization, and the quality assurance procedure and assessment criteria are described. In order to verify the KAERI fuel, thus, in-pile irradiation test program of the KAERI fuel is scheduled in HANARO. This report summarizes the in-pile testing schedule, design documents of test rods and assemblies, fabrication history and out-of-pile characteristics of test rods, irradiation test condition and power history, post-irradiation examination scheme, linear power generation distribution, and safety analysis results. The design code for HANARO fuel is used to analyze the centerline temperature and swelling of the KAERI fuels. The results show that at 120 kW/m of linear power the maximum centerline temperature is 267 deg C which is much lower than the limitation temperature of 350 deg C, and that the swelling is 9.3 % at 95 at% lower than criterion of 20 %. Therefore, the KAERI fuels of this in-pile irradiation test is assessed to show good performance of integrity and safety in HANARO. 10 tabs., 7 figs., 3 refs. (Author).

  11. Extending the foreign spent fuel acceptance program: Policy and implementation issues

    International Nuclear Information System (INIS)

    Lyman, Edwin S.

    2005-01-01

    The May 2006 expiration date of the Foreign Research Reactor Spent Nuclear Fuel (FRR SNF) Acceptance Program is fast approaching. In April 2004, Energy Secretary Spencer Abraham instructed the Energy Department to 'initiate actions necessary to extend .... the fuel acceptance deadline'. However, extending the deadline may not be a simple task. The limits on the original program resulted from a delicate negotiation among many stakeholders. Any proposal to increase the duration and scope of the program will have to be considered in the context of DOE's failure since 1996 to develop viable treatment, packaging and long-term disposal options for FRR SNF. It is also unclear whether accepting additional low-enriched uranium FRR SNF can be justified on security grounds. This paper will propose criteria for acceptance of spent fuel under an extension that are intended to minimize controversy and ensure consistency with a threat-based prioritization of homeland security expenditures. (author)

  12. Transient behaviour study program of research reactors fuel elements at the Hydra Pulse Reactor

    International Nuclear Information System (INIS)

    Khvostionov, V.E.; Egorenkov, P.M.; Malankin, P.V.

    2004-01-01

    Program on behavior study of research reactor Fuel Elements (FE) under transient regimes initiated by excessive reactivity insertion is being presented. Program would be realized at HYDRA pulse reactor at Russian Research Center 'Kurchatov Institute' (RRC 'K1'). HYDRA uses aqueous solution of uranyl sulfate (UO 2 SO 4 ) as a fuel. Up to 30 MJ of energy can be released inside the core during the single pulse, effective power pulse width varying from 2 to 10 ms. Reactor facility allows to investigate behaviour of FE consisting of different types of fuel composition, being developed according to Russian RERTR. First part of program is aimed at transient behaviour studying of FE MR, IRT-3M, WWR-M5 types containing meats based on dioxide uranium in aluminum matrix. Mentioned FEs use 90% and 36% enriched uranium. (author)

  13. Technical Basis - Spent Nuclear Fuels (SNF) Project Radiation and Contamination Trending Program

    International Nuclear Information System (INIS)

    ELGIN, J.C.

    2000-01-01

    This report documents the technical basis for the Spent Nuclear Fuel (SNF) Program radiation and contamination trending program. The program consists of standardized radiation and contamination surveys of the KE Basin, radiation surveys of the KW basin, radiation surveys of the Cold Vacuum Drying Facility (CVD), and radiation surveys of the Canister Storage Building (CSB) with the associated tracking. This report also discusses the remainder of radiological areas within the SNFP that do not have standardized trending programs and the basis for not having this program in those areas

  14. 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

  15. Impacts of the Weatherization Assistance Program in fuel-oil heated houses

    Energy Technology Data Exchange (ETDEWEB)

    Levins, W.P.; Ternes, M.P.

    1994-10-01

    In 1990, the US Department of Energy (DOE) initiated a national evaluation of its lowincome Weatherization Assistance Program. This report, which is one of five parts of that evaluation, evaluates the energy savings and cost-effectiveness of the Program as it had been applied to single-family houses heated primarily by fuel-oil. The study was based upon a representative sample (41 local weatherization agencies, 222 weatherized and 115 control houses) from the nine northeastern states during 1991 and 1992 program years. Dwelling-specific and agency-level data on measures installed, costs, and service delivery procedures were collected from the sampled agencies. Space-heating fuel-oil consumption, indoor temperature, and outdoor temperature were monitored at each house. Dwelling characteristics, air-leakage measurements, space-heating system steady-state efficiency measurements, safety inspections, and occupant questionnaires were also collected or performed at each monitored house. We estimate that the Program weatherized a total of 23,400 single-family fuel-oil heated houses in the nine northeastern states during program years 1991 and 1992. Annual fuel-oil savings were calculated using regression techniques to normalize the savings to standard weather conditions. For the northeast region, annual net fuel-oil savings averaged 160 gallons per house, or 17.7% of pre-weatherization consumption. Although indoor temperatures changed in individual houses following weatherization, there was no average change and no significant difference as compared to the control houses; thus, there was no overall indoor temperature takeback effect influencing fuel-oil savings. The weatherization work was performed cost effectively in these houses from the Program perspective, which included both installation costs and overhead and management costs but did not include non-energy benefits (such as employment and environmental).

  16. The Alternate Technology Program for Aluminum Research Reactor Spent Fuel

    International Nuclear Information System (INIS)

    Barlow, M.W.

    1998-01-01

    This paper describes the program for disposition of aluminum-based RRSNF, including the requirements for road-ready dry storage and repository disposal and the criteria to be considered in selecting among the alternative technologies

  17. National fuel cell seminar. Program and abstracts. [Abstracts of 40 papers

    Energy Technology Data Exchange (ETDEWEB)

    None

    1977-01-01

    Abstracts of 40 papers are presented. Topics include fuel cell systems, phosphoric acid fuel cells, molten carbonate fuel cells, solid fuel and solid electrolyte fuel cells, low temperature fuel cells, and fuel utilization. (WHK)

  18. Selected concrete spent fuel storage cask concepts and the DOE/PSN Cooperative Cask Testing Program

    International Nuclear Information System (INIS)

    Creer, J.M.; McKinnon, M.A.; Collantes, C.E.

    1990-01-01

    To date, water pools, metal casks, horizontal concrete modules, and modular vaults have been used to store the major quantity of commercial light water reactor spent nuclear fuel. Recently, vertical concrete dry storage casks have received consideration for storage of spent nuclear fuel. This paper reviews the evolution of the development of selected vertical concrete dry storage casks and outlines a cooperative cask testing (heat transfer and shielding) program involving the US Department of Energy and Pacific Sierra Nuclear Associates. Others participating in the cooperative program are Pacific Northwest Laboratory; EG ampersand G Idaho, Inc.; Wisconsin Electric Power Company; and the Electric Power Research Institute. 28 refs., 14 figs

  19. Spent fuel sabotage test program, characterization of aerosol dispersal : interim final report.

    Energy Technology Data Exchange (ETDEWEB)

    Gregson, Michael Warren; Brockmann, John E.; Loiseau, Olivier (Institut de Radioprotection et de Surete Nucleaire, France); Klennert, Lindsay A.; Nolte, Oliver (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Molecke, Martin Alan; Autrusson, Bruno A. (Institut de Radioprotection et de Surete Nucleaire, France); Koch, Wolfgang (Fraunhofer Institut fur Toxikologie und Experimentelle Medizin, Germany); Pretzsch, Gunter Guido (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Brucher, Wenzel (Gesellschaft fur Anlagen- und Reaktorsicherheit, Germany); Steyskal, Michele D.

    2008-03-01

    This multinational, multi-phase spent fuel sabotage test program is quantifying the aerosol particles produced when the products of a high energy density device (HEDD) interact with and explosively particulate test rodlets that contain pellets of either surrogate materials or actual spent fuel. This program provides source-term data that are relevant to plausible sabotage scenarios in relation to spent fuel transport and storage casks and associated risk assessments. We present details and significant results obtained from this program from 2001 through 2007. Measured aerosol results include: respirable fractions produced; amounts, nuclide content, and produced particle size distributions and morphology; measurements of volatile fission product species enhanced sorption--enrichment factors onto respirable particles; and, status on determination of the spent fuel ratio, SFR, needed for scaling studies. Emphasis is provided on recent Phase 3 tests using depleted uranium oxide pellets plus non-radioactive fission product dopants in surrogate spent fuel test rodlets, plus the latest surrogate cerium oxide results and aerosol laboratory supporting calibration work. The DUO{sub 2}, CeO{sub 2}, plus fission product dopant aerosol particle results are compared with available historical data. We also provide a status review on continuing preparations for the final Phase 4 in this program, tests using individual short rodlets containing actual spent fuel from U.S. PWR reactors, with both high- and lower-burnup fuel. The source-term data, aerosol results, and program design have been tailored to support and guide follow-on computer modeling of aerosol dispersal hazards and radiological consequence assessments. This spent fuel sabotage, aerosol test program was performed primarily at Sandia National Laboratories, with support provided by both the U.S. Department of Energy and the Nuclear Regulatory Commission. This program has significant input from, and is cooperatively

  20. Chemical Reactivity Testing for the National Spent Nuclear Fuel Program. Quality Assurance Project Plan

    International Nuclear Information System (INIS)

    Newsom, H.C.

    1999-01-01

    This quality assurance project plan (QAPjP) summarizes requirements used by Lockheed Martin Energy Systems, Incorporated (LMES) Development Division at Y-12 for conducting chemical reactivity testing of Department of Energy (DOE) owned spent nuclear fuel, sponsored by the National Spent Nuclear Fuel Program (NSNFP). The requirements are based on the NSNFP Statement of Work PRO-007 (Statement of Work for Laboratory Determination of Uranium Hydride Oxidation Reaction Kinetics.) This QAPjP will utilize the quality assurance program at Y-12, QA-101PD, revision 1, and existing implementing procedures for the most part in meeting the NSNFP Statement of Work PRO-007 requirements, exceptions will be noted

  1. Implementation of the utilization program for the fuel elements of the Atucha I nuclear power plant

    International Nuclear Information System (INIS)

    Martin, H.R.; Serra, O.H.; Parker, Alejandro

    1981-01-01

    The programming operation for the use of the fuel elements in the Atucha-1 nuclear power plant was initially under the responsibility of the KWU Company, as part of the services rendered due for the manufacturing of said elements. This job was done with the help of the TRISIC program, developed in the early seventies by CNEA and SIEMENS staff. From april 21, 1979 on, CNEA took over the responsibility and strategy of the interchange of fuel elements. The several stages carried out for the implementation of this service are detailed. (M.E.L.) [es

  2. Overview of the fast reactors fuels program. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Evans, E.A.; Cox, C.M.; Hayward, B.R.; Rice, L.H.; Yoshikawa, H.H.

    1980-04-01

    Each nation involved in LMFBR development has its unique energy strategies which consider energy growth projections, uranium resources, capital costs, and plant operational requirements. Common to all of these strategies is a history of fast reactor experience which dates back to the days of the Manhatten Project and includes the CLEMENTINE Reactor, which generated a few watts, LAMPRE, EBR-I, EBR-II, FERMI, SEFOR, FFTF, BR-1, -2, -5, -10, BOR-60, BN-350, BN-600, JOYO, RAPSODIE, Phenix, KNK-II, DFR, and PFR. Fast reactors under design or construction include PEC, CRBR, SuperPhenix, SNR-300, MONJU, and Madras (India). The parallel fuels and materials evolution has fully supported this reactor development. It has involved cermets, molten plutonium alloy, plutonium oxide, uranium metal or alloy, uranium oxide, and mixed uranium-plutonium oxides and carbides.

  3. Advanced teleoperation in nuclear applications: consolidated fuel reprocessing program

    International Nuclear Information System (INIS)

    Hamel, W.R.; Feldman, M.J.; Martin, H.L.

    1984-01-01

    A new generation of integrated remote maintenance systems is being developed to meet the needs of future nuclear fuel reprocessing at the Oak Ridge National Laboratory. Development activities cover all aspects of an advanced teleoperated maintenance system with particular emphasis on a new force-reflecting servomanipulator concept. The new manipulator, called the advanced servomanipulator, is microprocessor controlled and is designed to achieve force-reflection performance near that of mechanical master/slave manipulators. The advanced servomanipulator uses a gear-drive transmission which permits modularization for remote maintainability (by other advanced servomanipulators) and increases reliability. Human factors analysis has been used to develop an improved man/machine interface concept based upon colographic displays and menu-driven touch screens. Initial test and evaluation of two advanced servomanipulator slave arms and several other development components have begun. 9 references, 5 figures

  4. CANDU advanced fuel R and D programs for 1997 - 2006 in Korea

    International Nuclear Information System (INIS)

    Suk, H.C.; Yang, M.S.; Sim, K-S.; Yoo, K.J.

    1997-01-01

    KAERI has a comprehensive product development program of CANFLEX and DUPIC fuels to introduce them into CANDU reactors in Korea and a clear vision of how the product will evolve over the next 10 years. CANDU reactors are not the majority of nuclear power plants in Korea, but they produce significant electricity to contribute Korea's economic growth as well as to satisfy the need for energy. The key targets of the development program are safety enhancement, reduction of spent fuel volume, and economic improvements, using the inherent characteristics and advantages of CANDU technology The CANFLEX and DUPIC R and D programs are conducted currently under the second stage of Korea's Nuclear Energy R and D Project as a national mid- and long-term program over the next 10 years from 1997 to 2006. The specific activities of the programs have taken account of the domestic and international environment concerning on non-proliferation in the Peninsula of Korea. As the first of the development products in the short-term, the CANFLEX-NU fuel will be completely developed jointly by KAERI/AECL and will be useful for the older CANDU-6 Wolsong unit 1. As the second product, the CANFLEX-0.9 % equivalent SEU fuel is expected to be completely developed within the next decade. It will be used in CANDU-6 reactors in Korea immediately after the development, if the existing RU in the world is price competitive with natural uranium. The DUPIC R and D program, as a long term program, is expected to demonstrate the possibility of use of used PWR fuel in CANDU reactors in Korea during the next 10 years. The pilot scale fabrication facility would be completed around 2010. (author)

  5. CANDU advanced fuel R and D programs for 1997 - 2006 in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Suk, H.C.; Yang, M.S.; Sim, K-S.; Yoo, K.J. [Korea Atomic Energy Research Inst., Yusong, Taejon (Korea, Republic of)

    1997-07-01

    KAERI has a comprehensive product development program of CANFLEX and DUPIC fuels to introduce them into CANDU reactors in Korea and a clear vision of how the product will evolve over the next 10 years. CANDU reactors are not the majority of nuclear power plants in Korea, but they produce significant electricity to contribute Korea's economic growth as well as to satisfy the need for energy. The key targets of the development program are safety enhancement, reduction of spent fuel volume, and economic improvements, using the inherent characteristics and advantages of CANDU technology The CANFLEX and DUPIC R and D programs are conducted currently under the second stage of Korea's Nuclear Energy R and D Project as a national mid- and long-term program over the next 10 years from 1997 to 2006. The specific activities of the programs have taken account of the domestic and international environment concerning on non-proliferation in the Peninsula of Korea. As the first of the development products in the short-term, the CANFLEX-NU fuel will be completely developed jointly by KAERI/AECL and will be useful for the older CANDU-6 Wolsong unit 1. As the second product, the CANFLEX-0.9 % equivalent SEU fuel is expected to be completely developed within the next decade. It will be used in CANDU-6 reactors in Korea immediately after the development, if the existing RU in the world is price competitive with natural uranium. The DUPIC R and D program, as a long term program, is expected to demonstrate the possibility of use of used PWR fuel in CANDU reactors in Korea during the next 10 years. The pilot scale fabrication facility would be completed around 2010. (author)

  6. SYNBURN: fast-reactor fuel-cycle program

    International Nuclear Information System (INIS)

    Pizzica, P.A.; Meneley, D.A.

    1976-01-01

    The SYNBURN computer program for fast reactors will calculate all the neutronics necessary to completely characterize the equilibrium cycle as well as the startup to equilibrium cycles. The program's run time is very short and this makes the program suitable for survey of parametric studies. It can search on the cycle time for a specified burnup, for the shim control necessary for criticality as well as feed enrichments and the enrichment ratio among core zones. SYNBURN synthesizes in a very simple fashion the one-dimensional fluxes in radial and axial geometry to achieve an approximate two-dimensional solution which agrees very well with the exact two-dimensional solution when measuring regional integrated quantities

  7. Fuel Cell Development for NASA's Human Exploration Program: Benchmarking with "The Hydrogen Economy"

    Science.gov (United States)

    Scott, John H.

    2007-01-01

    The theoretically high efficiency and low temperature operation of hydrogen-oxygen fuel cells has motivated them to be the subject of much study since their invention in the 19th Century, but their relatively high life cycle costs kept them as a "solution in search of a problem" for many years. The first problem for which fuel cells presented a truly cost effective solution was that of providing a power source for NASA's human spaceflight vehicles in the 1960 s. NASA thus invested, and continues to invest, in the development of fuel cell power plants for this application. This development program continues to place its highest priorities on requirements for minimum system mass and maximum durability and reliability. These priorities drive fuel cell power plant design decisions at all levels, even that of catalyst support. However, since the mid-1990's, prospective environmental regulations have driven increased governmental and industrial interest in "green power" and the "Hydrogen Economy." This has in turn stimulated greatly increased investment in fuel cell development for a variety of commercial applications. This investment is bringing about notable advances in fuel cell technology, but, as these development efforts place their highest priority on requirements for minimum life cycle cost and field safety, these advances are yielding design solutions quite different at almost every level from those needed for spacecraft applications. This environment thus presents both opportunities and challenges for NASA's Human Exploration Program

  8. Atomics International fuel fabrication facility and low enrichment program [contributed by T.A. Moss, AI

    International Nuclear Information System (INIS)

    Moss, T.A.

    1993-01-01

    The AI facility is approximately 30,000 square feet in area and consists of four general areas. One area is devoted to the production of UAl x powder. It consists of a series of arc melting furnaces, crushing lines, glove boxes, and compacting presses. The second area is used for the rolling of fuel plates. The third area is used for the machining of the plates to final size and also the machining of the fuel elements. In the fourth area the fuel plates are swaged into assemblies, and all welding and inspection operations are performed. As part of the lower enrichment program we are scheduled to put a second UAl x powder line into operation and we have had to expand some of our storage area. Under the low enrichment program the AI fuel facility will be modified to accommodate a separate low enrichment Al x production line and compacting line. This facility modification should be done by the end of the fiscal year. We anticipate producing fuel with an enrichment slightly less than 20% We anticipate powder being available for plate production shortly after the facility is completed. Atomics International is scheduled to conduct plate LEU verification work using fully enriched material in the June-July time period, at which time we will investigate what level of uranium loadings we can go to using the current process. It is anticipated that 55 volume percent uranium compound in our fuel form can be achieved

  9. The status of uranium-silicon alloy fuel development for the RERTR program

    International Nuclear Information System (INIS)

    Domagala, R.F.; Wiencek, T.C.; Thresh, H.R.; Stahl, D.

    1983-01-01

    As part of the national Reduced Enrichment Research and Test Reactor (RERTR) Program, Argonne National Laboratory (ANL) is engaged in a fuel-alloy development project. The fuel alloys are dispersed in an aluminum matrix and metallurgically roll-bonded within 6061 Al alloy. To date, 'miniplates' with up to 40 vol. fuel alloy have been successfully fabricated. Thirty-one of these plates have been or are being irradiated in the Oak Ridge Reactor (ORR). Three different fuels have been used in the ANL miniplates: U 3 Si (U + 4 wt.% Si), U 3 Si 2 (U + 7.4 wt.% Si), or ''U 3 SiAl'' (U + 3.5 wt.% Si + 1.5 wt.% Al). All three are candidates for permitting higher fuel loadings and thus lower enrichments of 235 U than would be possible with either UAl x or U 3 O 8 , the current fuels for plate-type elements. The enrichment level employed at ANL is ∼19.8%. Continuing effort involves the production of miniplates with up to ∼60 vol. % fuel, the development of a technology for full-size plate fabrication, and post-irradiation examination of miniplates already removed from the ORR. (author)

  10. Corrosion surveillance program of aluminum spent fuel elements in wet storage sites

    International Nuclear Information System (INIS)

    Linardi, E; Haddad, R

    2012-01-01

    Due to different degradation issues observed in aluminum-clad spent fuel during long term storage in water, the IAEA implemented in 1996 a Coordinated Research Project (CRP) and a Regional Project for Latin America, on Corrosion of Research Reactor Aluminum Clad Spent Fuel in Water. Argentine has been among the participant countries of these projects, carrying out spent fuel corrosion surveillance activities in its storage facilities. As a result of the research a large database on corrosion of aluminum-clad fuel has been generated. It was determined that the main types of corrosion affecting the spent fuel are pitting and galvanic corrosion due to contact with stainless steel. It was concluded that the quality of the water is the critical factor to control in a spent fuel storage facility. Another phase of the program is being conducted currently, which began in 2011 with the immersion of test racks in the RA1 reactor pool, and in the Research Reactor Spent Fuel Storage Facility (FACIRI), located in Ezeiza Atomic Center. This paper presents the results of the chemical analysis of the water performed so far, and its relationship with the examination of the coupons extracted from the sites (author)

  11. Fuel cell programs in the United States for stationary power applications

    Energy Technology Data Exchange (ETDEWEB)

    Singer, M.

    1996-04-01

    The Department of Energy (DOE), Office of Fossil Energy, is participating with the private sector in sponsoring the development of molten carbonate fuel cell (MCFC) and solid oxide fuel cell (SOFC) technologies for application in the utility, commercial and industrial sectors. Phosphoric acid fuel cell (PAFC) development was sponsored by the Office of Fossil Energy in previous years and is now being commercialized by the private sector. Private sector participants with the Department of Energy include the Electric Power Research Institute (EPRI), the Gas Research institute (GRI), electric and gas utilities, universities, manufacturing companies and their suppliers. through continued government and private sector support, fuel cell systems are emerging power generation technologies which are expected to have significant worldwide impacts. An industry with annual sales of over a billion dollars is envisioned early in the 21st century. PAFC power plants have begun to enter the marketplace and MCFC and SOFC power plants are expected to be ready to enter the marketplace in the late 1990s. In support of the efficient and effective use of our natural resources, the fuel cell program seeks to increase energy efficiency and economic effectiveness of power generation. This is to be accomplished through effectiveness of power generation. This is accomplished through the development and commercialization of cost-effective, efficient and environmentally desirable fuel cell systems which will operate on fossil fuels in multiple and end use sectors.

  12. A computer program for calculation of the fuel cycle in pressurized water reactors

    International Nuclear Information System (INIS)

    Solanilla, R.

    1976-01-01

    The purpose of the FUCEFURE program is two-fold: first, it is designed to solve the problem of nuclear fuel cycle cost in one pressurized light water reactor calculation. The code was developed primarily for comparative and sensitivity studies. The program contains simple correlations between exposure and available depletion data used to predict the uranium and plutonium content of the fuel as a function of the fuel initial enrichment. Second, it has been devised to evaluate the nuclear fuel demand associated with an expanding nuclear power system. Evaluation can be carried out at any time and stage in the fuel cycle. The program can calculate the natural uranium and separate work requirements of any final and tails enrichment. It also can determine the nuclear power share of each reactor in the system when a decision has been made about the long-term nuclear power installations to be used and the types of PWR and fast breeder reactor characteristics to be involved in them. (author)

  13. Development of a program for evaluating the temperature of SMART-P fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Cheon, Jin Sik; Lee, Byung Ho; Koo, Yang Hyun; Oh, Je Yong; Yim, Jeong Sik; Sohn, Dong Seong

    2003-11-01

    A code for evaluating the temperature of SMART-P fuel rod has been developed. Finite Element (FE) method is adopted for the developed code sharing the user subroutines which has been prepared for the ABAQUS commercial FE code. The developed program for SMART-P fuel rod corresponds to a nonlinear transient heat transfer problem, and uses a sparse matrix solver for FE equations during iterations at every time step. The verifications of the developed program were conducted using the ABAQUS code. Steady state and transient problems were analyzed for 1/8 rod model due to the symmetry of the fuel rod and full model. From the evaluation of temperature for the 1/8 rod model at steady state, maximal error of 0.18 % was present relative to the ABAQUS result. Analysis for the transient problem using the fuel rod model resulted in the same as the variation of centerline temperature from the ABAQUS code during a hypothetical power transient. Also, given a power depression in fuel meat as a function of burnup, its effect on the centerline temperature was more precisely evaluated by the developed program compared to the ABAQUS code. The distribution of heat flux for the entire cross section and surface was almost identical for the two codes.

  14. Critical experiment needs and plans of the consolidated fuel reprocessing program

    International Nuclear Information System (INIS)

    Primm, R.T.

    1984-01-01

    An integral part of the United States Department of Energy (DOE) plan for the development of breeder reactors is the development of the capability for fuel reprocessing. The Consolidated Fuel Reprocessing Program (CFRP) was established by the DOE to identify and conduct research and development activities in this area. The DOE is currently proposing that a capability to reprocess fast reactor fuel be established in the Fuels and Materials Examination Facility at the Hanford Engineering Development Laboratory. This capability would include conversion of plutonium nitrate to plutonium oxide. The reprocessing line is designated the Breeder Reprocessing Engineering Test (BRET). Criticality safety remains an important critetion in the design of the BRET. The different steps in the reprocessing are reviewed and areas where additional critical experiments are needed have been indentified as also areas where revision or clarification of existing criticality safety standards are desirable

  15. Model for incorporating fuel swelling and clad shrinkage effects in diffusion theory calculations (LWBR Development Program)

    International Nuclear Information System (INIS)

    Schick, W.C. Jr.; Milani, S.; Duncombe, E.

    1980-03-01

    A model has been devised for incorporating into the thermal feedback procedure of the PDQ few-group diffusion theory computer program the explicit calculation of depletion and temperature dependent fuel-rod shrinkage and swelling at each mesh point. The model determines the effect on reactivity of the change in hydrogen concentration caused by the variation in coolant channel area as the rods contract and expand. The calculation of fuel temperature, and hence of Doppler-broadened cross sections, is improved by correcting the heat transfer coefficient of the fuel-clad gap for the effects of clad creep, fuel densification and swelling, and release of fission-product gases into the gap. An approximate calculation of clad stress is also included in the model

  16. 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

  17. Development program for fuel elements with low enriched uranium for high temperature reactors

    International Nuclear Information System (INIS)

    1987-12-01

    The results of HTR fuel development taking place at the THTR's can be summarized as follows for the main points of core manufacture coating matrix and fuel emenent manufacture: 1. The well known gel precipitation process was modified for the manufacture of UO 2 cores. 2. The TRISO coating (additional SiC layer between two very dense PyC layers) can be applied with the required quality on an economical 10 kg scale. 3. The particle fracture in the complete fuel element due to manufacture was lowered during the course of the project to below the target values of -6 U/U total. For testing fuel elements, the required irradiation samples were designed in agreement with the reactor constructors, were prepared and the first phase of the irradiation program was successfully completed in the context of the HBK project. (orig./HP) [de

  18. SOLID STATE ENERGY CONVERSION ALLIANCE (SECA) SOLID OXIDE FUEL CELL PROGRAM

    Energy Technology Data Exchange (ETDEWEB)

    Nguyen Minh; Jim Powers

    2003-10-01

    This report summarizes the work performed for April 2003--September 2003 reporting period under Cooperative Agreement DE-FC26-01NT41245 for the U.S. Department of Energy, National Energy Technology Laboratory (DOE/NETL) entitled ''Solid State Energy Conversion Alliance (SECA) Solid oxide Fuel Cell Program''. During this reporting period, the conceptual system design activity was completed. The system design, including strategies for startup, normal operation and shutdown, was defined. Sealant and stack materials for the solid oxide fuel cell (SOFC) stack were identified which are capable of meeting the thermal cycling and degradation requirements. A cell module was tested which achieved a stable performance of 0.238 W/cm{sup 2} at 95% fuel utilization. The external fuel processor design was completed and fabrication begun. Several other advances were made on various aspects of the SOFC system, which are detailed in this report.

  19. Quality Assurance Plans under the Renewable Fuel Standard Program

    Science.gov (United States)

    Quality Assurance Plan or (QAP) is a voluntary program where independent third-parties may audit and verify that RINs have been properly generated and are valid for compliance purposes. RINs verified under a QAP may be purchased by regulated parties.

  20. Computer programs for the in-core fuel management of power reactors

    International Nuclear Information System (INIS)

    1981-08-01

    This document gives a survey of the presently tested and used computer programs applicable to the in-core fuel management of light and heavy water moderated nuclear power reactors. Each computer program is described (provided that enough information was supplied) such that the nature of the physical problem solved and the basic mathematical or calculational approach are evident. In addition, further information regarding computer requirements, up-to-date applications and experiences and specific details concerning implementation, staff requirements, etc., are provided. Program procurement conditions, possible program implementation assistance and commercial conditions (where applicable) are given. (author)

  1. Fuel cell program - Overview reports 2007; Programm Brennstoffzellen inkl. Wasserstoff - Ueberblicksberichte der BFE-Programmleiter 2007

    Energy Technology Data Exchange (ETDEWEB)

    Luzzi, A.; Spirig, M.

    2008-07-01

    This report for the Swiss Federal Office of Energy (SFOE) presents the overview reports made by SFOE Heads of Program on work done in 2007. Projects reported on in the natural gas-fired fuel cell area include the EU-project REAL-SFOC, the long-term testing of anode-supported SOFC stacks, intermediate-temperature fuel cells based on proton conducting electrolytes, the interdisciplinary ONEBAT project and lifetime-enhancement of SOFC stacks for CHP applications. In the polymer-electrolyte fuel cell (PEFC) area, projects concerning proton-conducting polymer membranes, factors limiting the lifetime of fuel cell membranes, a new highly active oxygen reduction electrode for PEM fuel cell and zinc/air battery applications, the enhancement of PEFC durability and reliability, model-based investigation of PEFC performance, and local gas analysis of PE fuel cells are briefly reported on. Long-term research activities in the hydrogen technology area reported on include those concerning the photo-chemical conversion and storage of solar energy and the storage of hydrogen in metallic and complex hydrides. Further projects reported on include those concerning the physical aspects of hydrides for system integration and safety and new, complex metal hydrides. Swiss national and international co-ordination is reviewed in the areas of fuel cell technology and hydrogen technology. Work done in several projects run within the framework of the IEA's Advanced Fuel Cells Program is reviewed. Several pilot and demonstration (P and D) projects are also reported on in the natural-gas SOFC and PEFC areas. Comments on the 2007 results and a review of work to be done in 2008, along with a list of R, D, P and D projects, complete the report.

  2. Engineering program in order to increase the irradiated fuel storage capacity in pool facilities of Juragua

    International Nuclear Information System (INIS)

    Rodriguez R, J.

    1996-01-01

    In 1993, a technical program in the spent fuel storage area of Nuclear Plant Juragua was launched. Such a program tries to carry out an engineering assessment of the possibility of increasing the spent fuel storage capacity in pool storage facilities by using high density racks (re-racking) instead of the original (non-compact) ones. The purpose of the above-mentioned program is to evaluate possible solutions that can be applied to the construction works prior to plant operation. The first stage of the program for the 1994-95 period is an ongoing Engineering-Economic Feasibility Study (EEFS), which endeavors to examine the capabilities of the reloading pool in Unit-1 Reactor building and long-term storage pool in auxiliary building in high density storage conditions. Technical details of the EEFS and reached results and difficulties are described. (author). 5 refs., 2 figs

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

    Energy Technology Data Exchange (ETDEWEB)

    Bryant, J.W.

    1994-01-01

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

  4. New high density MTR fuel. The CEA-CERCA-COGEMA development program

    International Nuclear Information System (INIS)

    Languille, A.; Durand, J.P.; Gay, A.

    1999-01-01

    The development of a new generation of LEU, high in density and with reprocessing capacities MTR fuel, is a key issue to provide reactor operators with a smooth operation which is necessary for a long term development of Nuclear Energy. In the RRFM'98 meeting, a joint contribution of CEA, CERCA and COGEMA presented a technical classification of the potential candidates uranium alloys. In this paper this MTR working group presents the development program of a new high density fuel. This program is composed of three main steps: Basic Data analysis and collection, Plate Tests (Irradiation and Post Irradiation Examinations) and Lead Test Assemblies (Irradiation and Post Irradiation Examinations). The goal to be reached is to make this new fuel available before the end of the present US return policy. (author)

  5. Industrial Fuel Gas Demonstration Plant Program: environmental permit compliance plan

    Energy Technology Data Exchange (ETDEWEB)

    Bodamer, Jr., James W.; Bocchino, Robert M.

    1979-11-01

    This Environmental Permit Compliance Plan is intended to assist the Memphis Light, Gas and Water Division in acquiring the necessary environmental permits for their proposed Industrial Fuel Gas Demonstration Plant in a time frame consistent with the construction schedule. Permits included are those required for installation and/or operation of gaseous, liquid and solid waste sources and disposal areas. Only those permits presently established by final regulations are described. The compliance plan describes procedures for obtaining each permit from identified federal, state and local agencies. The information needed for the permit application is presented, and the stepwise procedure to follow when filing the permit application is described. Information given in this plan was obtained by reviewing applicable laws and regulations and from telephone conversations with agency personnel on the federal, state and local levels. This Plan also presents a recommended schedule for beginning the work necessary to obtain the required environmental permits in order to begin dredging operations in October, 1980 and construction of the plant in September, 1981. Activity for several key permits should begin as soon as possible.

  6. Measuring Success in Your Fuels Program: From the Report Card to Valuable Learning

    Science.gov (United States)

    Paula Nasiatka; David Christenson

    2006-01-01

    How can a unit learn in everyday fuels programs and from program reviews? How can a unit move from living in the “report card” culture to discovering more effective ways to improve what it knows and how it learns? Six specific tasks are critical to organizational learning according to David A. Garvin of Harvard Business School. By engaging in these tasks a unit can...

  7. Development program of two-arm bilateral servomanipulator system for nuclear fuel cycle facilities in PNC

    International Nuclear Information System (INIS)

    Kashihara, H.; Igarashi, M.; Maeda, M.; Nomizu, T.; Uematsu, K.

    1984-01-01

    PNC started the development program of a two-arm bilateral servomanipulator to improve remote maintenance in the planning of pilot plants for HLLW vitrification and FBR fuel reprocessing. In this paper, the development program of two-arm bilateral servomanipulators, signal and power transmissions, high performance viewing system, man-machine interface, and transporter system for manipulators will be presented. Also radiation tests for all components used in a hot cell are being performed to aid remote system design

  8. The Krakow clean fossil fuels and energy efficiency program

    Energy Technology Data Exchange (ETDEWEB)

    Feibus, H.

    1995-12-31

    The joint effort by Polish and American organizations in Krakow has accomplished a great deal in just a few years. In particular, the low emission sources program has had major successes. Poland and America have a lot to learn from each other in the clean and economical use of coal. Both our countries are major producers and users of coal. Both have had to deal with the emissions of particulate and organics from coal combustion. We were fortunate, since our free market economy and democratic government helped us deal with a lot of these problems in the 1950s. In Poland, the freedom to solve these problems has evolved only in the last few years. In the first phase of the program, Polish and American engineers ran combustion tests on boilers and stoves in Krakow. They also performed analyses on the cost and feasibility of various equipment changes. The results of the first phase were used in refining the spreadsheet model to give better estimates of costs emissions. The first phase also included analyses of incentives for proceeding with needed changes. These analyses identified actions needed to create a market for the goods and services which control pollution. Such actions could include privatization, regulation, or financial incentives. The second phase of the program consisted of public meetings in Chicago, Washington, and Krakow. The purpose of the meetings was to inform U.S. and Polish firms about the results of phase 1 and to encourage them to compete to take part in phase 3. The third phase currently underway consists of the commercial ventures that were competitively selected. These ventures were consistent with recommendations unanimously made by the BSC. The three phases of the Polish-American program are discussed.

  9. Building on success. The foreign research reactor spent nuclear fuel acceptance program

    International Nuclear Information System (INIS)

    Huizenga, David G.; Mustin, Tracy P.; Saris, Elizabeth C.; Massey, Charles D.

    1998-01-01

    The second year of implementation of the research reactor spent nuclear fuel acceptance program was marked by significant challenges and achievements. In July 1998, the Department of Energy completed by significant challenges and achievements. In July 1998, the Department of Energy completed its first shipment of spent fuel from Asia via the Concord Naval Weapons Station in California to the Idaho National Engineering and Environmental (INEEL). This shipment, which consisted of three casks of spent nuclear fuel from two research reactors in the Republic of Korea, presented significant technical, legal, and political challenges in the United States and abroad. Lessons learned will be used in the planning and execution of our next significant milestone, a shipment of TRIGA spent fuel from research reactors in Europe to INEEL, scheduled for the summer of 1999. This shipment will include transit across the United States for over 2,000 miles. Other challenges and advances include: clarification of the fee policy to address changes in the economic status of countries during the life of the program; resolution of issues associated with cask certification and the specific types and conditions of spent fuel proposed for transport; revisions to standard contract language in order to more clearly address unique shipping situations; and priorization and scheduling of shipments to most effectively implement the program. As of this meeting, eight shipments, consisting of nearly 2,000 spent fuel assemblies from fifteen countries, have been successfully completed. With the continued cooperation of the international research reactor community, we are committed to building on this success in the remaining years of the program. (author)

  10. Quality Assurance Program Plan for Project W-379: Spent Nuclear Fuels Canister Storage Building Projec

    International Nuclear Information System (INIS)

    Duncan, D.W.

    1995-01-01

    This document describes the Quality Assurance Program Plan (QAPP) for the Spent Nuclear Fuels (SNF) Canister Storage Building (CSB) Project. The purpose of this QAPP is to control project activities ensuring achievement of the project mission in a safe, consistent and reliable manner

  11. Pathways to Commercial Success: Technologies and Products Supported by the Fuel Cell Technologies Program

    Energy Technology Data Exchange (ETDEWEB)

    Weakley, Steven A.; Brown, Scott A.

    2011-09-29

    The purpose of the project described in this report is to identify and document the commercial and emerging (projected to be commercialized within the next 3 years) hydrogen and fuel cell technologies and products that resulted from Department of Energy support through the Fuel Cell Technologies (FCT) Program in the Office of Energy Efficiency and Renewable Energy (EERE). To do this, Pacific Northwest National Laboratory (PNNL) undertook two efforts simultaneously to accomplish this project. The first effort was a patent search and analysis to identify hydrogen- and fuel-cell-related patents that are associated with FCT-funded projects (or projects conducted by DOE-EERE predecessor programs) and to ascertain the patents current status, as well as any commercial products that may have used the technology documented in the patent. The second effort was a series of interviews with current and past FCT personnel, a review of relevant program annual reports, and an examination of hydrogen- and fuel-cell-related grants made under the Small Business Innovation Research and Small Business Technology Transfer Programs, and within the FCT portfolio.

  12. 49 CFR 575.106 - Tire fuel efficiency consumer information program.

    Science.gov (United States)

    2010-10-01

    ... to provide information indicating the relative performance of replacement passenger car tires in the... achieve the level of performance represented by each rating. (A) Ratings. Each tire shall be rated with... 49 Transportation 7 2010-10-01 2010-10-01 false Tire fuel efficiency consumer information program...

  13. Characterization program, management and isotopic inventory calculation, radiological and fuel thermal irradiated in nuclear power Cofrentes

    International Nuclear Information System (INIS)

    Albendea, M.; Diego, J. L. de; Urrea, M.

    2012-01-01

    Characterization is a very detailed and user-friendly program takes into account the history of irradiation individualized and real all the fuel, even taking into account the interim periods are periods of discharge and recharge cycles and which have not been used.

  14. Pathways to Commercial Success: Technologies and Products Supported by the Fuel Cell Technologies Program

    Energy Technology Data Exchange (ETDEWEB)

    Weakley, Steven A.

    2012-09-28

    The purpose of the project described in this report is to identify and document the commercial and emerging (projected to be commercialized within the next 3 years) hydrogen and fuel cell technologies and products that resulted from Department of Energy support through the Fuel Cell Technologies (FCT) Program in the Office of Energy Efficiency and Renewable Energy (EERE). Pacific Northwest National Laboratory (PNNL) undertook two efforts simultaneously to accomplish this project. The first effort was a patent search and analysis to identify patents related to hydrogen and fuel cells that are associated with FCT-funded projects (or projects conducted by DOE-EERE predecessor programs) and to ascertain the patents’ current status, as well as any commercial products that may have used the technology documented in the patent. The second effort was a series of interviews with current and past FCT personnel, a review of relevant program annual reports, and an examination of grants made under the Small Business Innovation Research and Small Business Technology Transfer Programs that are related to hydrogen and fuel cells.

  15. CORCON: a computer program for modelling molten fuel/concrete interactions

    International Nuclear Information System (INIS)

    Muir, J.F.

    1980-01-01

    A computer program modelling the interaction between molten core materials and structural concrete is being developed to provide a capability for making quantitative estimates of reactor fuel-melt accidents. The principal phenomenological models, inter-component heat transfer, concrete erosion, and melt/gas chemical reactions, are described. A code test comparison calculation is discussed

  16. Market impacts of hypothetical fuel treatment thinning programs on federal lands in the western United States

    Science.gov (United States)

    Peter J. Ince; Henry Spelter; Kenneth Skog; Andrew Kramp; Dennis P. Dykstra

    2000-01-01

    This paper addresses the economics of forest fuel thinning programs on federal lands in the U.S. West, and presents a model of regional timber and product market impacts. The issue of economics is vital to the debate about fire management, and this paper presents market implications of two alternative silvicultural strategies, even-aged and uneven-aged...

  17. Characterization program management plan for Hanford K basin spent nuclear fuel

    International Nuclear Information System (INIS)

    TRIMBLE, D.J.

    1999-01-01

    The program management plan for characterization of the K Basin spent nuclear fuel was revised to incorporate corrective actions in response to SNF Project QA surveillance 1K-FY-99-060. This revision of the SNF Characterization PMP replaces Duke Eng

  18. Operational experience using the OSTR flip fuel self-protection program

    International Nuclear Information System (INIS)

    Dodd, B.; Ringle, J.C.; Anderson, T.V.; Johnson, A.G.

    1982-01-01

    Recent changes in NRC Physical Security regulations make it highly desirable for a small number of TRIGA research reactor establishments to maintain each of the fuel elements in their reactor core above the self-protection dose rate criterion. OSTR operations personnel have written a computer program (SPOOF) which calculates the exposure rate (in Rhr -1 ) from an irradiated fuel element at 3 feet in air using the actual operating history of the reactor. The purpose of this current paper is to describe the operational experience gained over the last year and a half while using the SPOOF computer program, and while performing the quarterly dose rate measurements needed to confirm the continuing accuracy of the program, and, most importantly, the self-protection status of the OSTR fuel. The computer program in association with the quarterly dose rate measurements have been accepted by the NRC, and allow the OSTR to take credit for self-protecting FLIP fuel under the current physical security regulations

  19. OVERVIEW OF ADVANCED PETROLEUM-BASED FUELS-DIESEL EMISSIONS CONTROL PROGRAM (APBF-DEC)

    Energy Technology Data Exchange (ETDEWEB)

    Sverdrup, George M.

    2000-08-20

    The Advanced Petroleum-Based Fuels-Diesel Emissions Control Program (APBF-DEC) began in February 2000 and is supported by government agencies and industry. The purpose of the APBF-DEC program is to identify and evaluate the optimal combinations of fuels, lubricants, diesel engines, and emission control systems to meet the projected emission standards for the 2000 to 2010 time period. APBF-DEC is an outgrowth of the earlier Diesel Emission Control-Sulfur Effects Program (DECSE), whose objective is to determine the impact of the sulfur levels in fuel on emission control systems that could lower the emissions of NOx and particulate matter (PM) from diesel powered vehicles in the 2002 to 2004 period. Results from the DECSE studies of two emission control technologies-diesel particle filter (DPF) and NOx adsorber-will be used in the APBF-DEC program. These data are expected to provide initial information on emission control technology options and the effects of fuel properties (including additives) on the performance of emission control systems.

  20. History and update of the U.S. program for the management of used nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Isaacs, T. [US Blue Ribbon Commission, Washington, DC (United States)

    2015-07-01

    This paper discusses the history and gives an update of the US program for the management of used nuclear fuel. The key issue in the deliberation was the option of long-term monitored storage versus disposal capability. The decision was to give future generations the option of disposal.

  1. A proposal for an international program to develop dry recycle of spent nuclear fuel

    International Nuclear Information System (INIS)

    Feinroth, H.

    1999-01-01

    The dry oxidation-reduction process (called OREOX for Oxidation Reduction of Oxide Fuel) being developed by Korea and Canada, in cooperation with IAEA and the US State Department, is limited to recycle of spent LWR fuel into CANDU reactors (DUPIC). When first conceived and demonstrated via irradiation of test elements by Atomics International in 1965, (the process was called AIROX at that time) a wider range of applications was intended, including recycle of spent LWR fuel into LWRs. Studies sponsored by DOE's Idaho Office in 1992 confirmed the applicability of this technology to regions containing LWR's only, and described the potential advantages of such recycle from an environmental, waste management and economic point of view, as compared to the direct disposal option. Recent analyses conducted by the author indicates that such dry recycle may be one of the few acceptable paths remaining for resolution of the US spent fuel storage dilemma that remains consistent with US non-proliferation policy. It is proposed that a new US program be established to develop AIROX dry recycle for use in the US, and this become part of an international cooperative program, including the current Canadian - Korean program, and possibly including participation of other countries wishing to pursue alternatives to the once through cycle, and wet reprocessing. With shared funding of major project elements, such international cooperation would accelerate the demonstration and commercial deployment of dry recycle technology, as compared to separate and independent programs in each country. (author)

  2. Energy poverty reduction by fuel switching. Impact evaluation of the LPG conversion program in Indonesia

    International Nuclear Information System (INIS)

    Andadari, Roos Kities; Mulder, Peter; Rietveld, Piet

    2014-01-01

    In low- and middle-income countries, Liquefied Petroleum Gas (LPG) can be an attractive alternative to the widespread use of traditional kerosene. Not only is LPG a relatively clean, safe and cost-effective fuel for households, its large-scale adoption also reduces the heavy burden of kerosene consumption subsidies on government budgets. Against this background, we evaluate the impact of a large government program to substitute LPG for kerosene in Indonesia. Using a household survey across urban, suburban and rural regions we find that this program was very effective in causing a large scale shift from kerosene to LPG. This shift was positively influenced by level of education, household size and household income. Contradicting the energy-ladder model, the LPG program, reinforced by an increase in the price of kerosene, led to increased stacking of fuels, including increasing consumption of both electricity and traditional biomass. In addition, our analysis shows that the LPG program failed to substantially reduce the overall number of energy-poor people, but it has been effective in alleviating extreme energy-poverty. Finally, we find that medium and higher income households in suburban areas benefitted most from the LPG program. - Highlights: • Impact evaluation of a government program to substitute LPG for kerosene in Indonesia. • The program caused a large scale shift from kerosene to LPG. • Contradicting the energy-ladder model, the program led to increased stacking of fuels. • The program did not substantially reduce overall energy poverty, but alleviated extreme energy-poverty. • Medium and higher income households in suburban areas benefitted most from the LPG program

  3. Program plan for research and development in support of LWR fuel recycle

    International Nuclear Information System (INIS)

    1975-01-01

    The ERDA program that is being planned to assist industry in the commercialization of the LWR fuel cycle will involve a range of activities, including joint programs with industry, R and D to provide technology, conceptual design of fuel recycle facilities, and environmental and economic assessments. A two-part program to begin in 1976 that is a portion of the overall ERDA plan is described. Responsibility for coordination and management of the tasks described in this document has been assigned to Du Pont as prime contractor to the ERDA Savannah River Operations Office. The first part of the program consists of the conceptual design of complete recycle facilities. The second part of the program, which will proceed concurrently, consists of supporting R and D activities, economic and environmental studies, and other studies to assist in the regulatory process. The R and D program will include both near-term activities in support of the conceptual design effort, and other activities aimed at general improvements in fuel cycle technology. The conceptual design will be used to develop current cost information for a complete reprocessing complex. The design will be based initially on current technology with provision for improvements as confirmatory information and advanced technology become available from the R and D program. The conceptual design and cost estimate will be developed by the Du Pont Atomic Energy Division. The R and D program and supporting studies will be directed at uncertainties in current technology as well as toward development of improved technology. It will include such R and D as might be appropriate for ERDA to undertake in support of joint programs with industry. The Savannah River Laboratory will have responsibility for coordinating the program

  4. Enhancing instruction in Fuels and Combustion Laboratory via a developed computer-assisted program for establishing efficient coal-diesel oil mixture (CDOM) fuel proportions

    Energy Technology Data Exchange (ETDEWEB)

    Maglaya, A.B. [La Salle University, Manila (Philippines). Dept. of Mechanical Engineering

    2004-07-01

    This paper discusses the relevance of digital computation in Fuels and Combustion Laboratory experiments used by the senior students of the Department of Mechanical Engineering, De La Salle University-Manila, Philippines. One of the students' experiments involved the determination of the most efficient CDOM fuel proportion as alternative fuel to diesel oil for steam generators and other industrial applications. Theoretical calculations show that it requires tedious and repetitive computations. A computer-assisted program was developed to lessen the time-consuming activities. The formulation of algorithms were based on the system of equations of the heat interaction between the CDOM fuel, combustion air and products of combustion and by applying the principles of mass and energy equations (or the First Law of Thermodynamics) for reacting systems were utilized. The developed computer-assisted program output verified alternative fuel selected through actual experimentation.

  5. Hydropyrolysis of biomass to produce liquid hydrocarbon fuels. Final report. Biomass Alternative-Fuels Program

    Energy Technology Data Exchange (ETDEWEB)

    Fujita, R K; Bodle, W W; Yuen, P C

    1982-10-01

    The ojective of the study is to provide a process design and cost estimates for a biomass hydropyrolysis plant and to establish its economic viability for commercial applications. A plant site, size, product slate, and the most probable feedstock or combination of feedstocks were determined. A base case design was made by adapting IGT's HYFLEX process to Hawaiian biomass feedstocks. The HYFLEX process was developed by IGT to produce liquid and/or gaseous fuels from carbonaceous materials. The essence of the process is the simultaneous extraction of valuable oil and gaseous products from cellulosic biomass feedstocks without forming a heavy hard-to-handle tar. By controlling rection time and temperature, the product slate can be varied according to feedstock and market demand. An optimum design and a final assessment of the applicability of the HYFLEX process to the conversion of Hawaiian biomass was made. In order to determine what feedstocks could be available in Hawaii to meet the demands of the proposed hydropyrolysis plant, various biomass sources were studied. These included sugarcane and pineapple wastes, indigenous and cultivated trees and indigenous and cultivated shrubs and grasses.

  6. Characterization program management plan for Hanford K Basin Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    Lawrence, L.A.

    1995-01-01

    A management plan was developed for Westinghouse Hanford Company (WHC) and Pacific Northwest Laboratories (PNL) to work together on a program to provide characterization data to support removal, conditioning and subsequent dry storage of the spent nuclear fuels stored at the Hanford K Basins. The Program initially supports gathering data to establish the current state of the fuel in the two basins. Data Collected during this initial effort will apply to all SNF Project objectives. N Reactor fuel has been degrading with extended storage resulting in release of material to the basin water in K East and to the closed conisters in K West. Characterization of the condition of these materials and their responses to various conditioning processes and dry storage environments are necessary to support disposition decisions. Characterization will utilize the expertise and capabilities of WHC and PNL organizations to support the Spent Nuclear Fuels Project goals and objectives. This Management Plan defines the structure and establishes the roles for the participants providing the framework for WHC and PNL to support the Spent Nuclear Fuels Project at Hanford

  7. Information on the evolution of severe LWR fuel element damage obtained in the CORA program

    International Nuclear Information System (INIS)

    Schanz, G.; Hagen, S.; Hofmann, P.; Sepold, L.; Schumacher, G.

    1992-01-01

    In the CORA program a series of out-of-pile experiments on LWR severe accidental situations is being performed, in which test bundles of LWR typical components and arrangements (PWR, BWR) are exposed to temperature transients up to about 2400deg C under flowing steam. The individual features of the facility, the test conduct, and the evaluation will be presented. In the frame of the international cooperation in severe fuel damage (SFD) programs the CORA tests are contributing confirmatory and complementary informations to the results from the limited number of in-pile tests. The identification of basic phenomena of the fuel element destruction, observed as a function of temperature, is supported by separate-effects test results. Most important mechanisms are the steam oxidation of the Zircaloy cladding, which determines the temperature escalation, the chemical interaction between UO 2 fuel and cladding, which dominates fuel liquefaction, relocation and resulting blockage formation, as well as chemical interactions with Inconel spacer grids and absorber units ((Ag, In, Cd) alloy or B 4 C), which are leading to extensive low-temperature melt formation around 1200deg C. Interrelations between those basic phenomena, resulting for example in cladding deformation ('flowering') and the dramatic hydrogen formation in response to the fast cooling of a hot bundle by cold water ('quenching') are determining the evolution paths of fuel element destruction, which are to be identified. (orig.)

  8. A continuing success - The United States Foreign Research Reactor Spent Nuclear Fuel Acceptance Program

    International Nuclear Information System (INIS)

    Mustin, Tracy P.; Clapper, Maureen; Reilly, Jill E.

    2000-01-01

    The United States Department of Energy, in consultation with the Department of State, adopted the Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel in May 1996. To date, the Foreign Research Reactor (FRR) Spent Nuclear Fuel (SNF) Acceptance Program, established under this policy, has completed 16 spent fuel shipments. 2,651 material test reactor (MTR) assemblies, one Slowpoke core containing less than 1 kilogram of U.S.-origin enriched uranium, 824 Training, Research, Isotope, General Atomic (TRIGA) rods, and 267 TRIGA pins from research reactors around the world have been shipped to the United States so far under this program. As the FRR SNF Acceptance Program progresses into the fifth year of implementation, a second U.S. cross country shipment has been completed, as well as a second overland truck shipment from Canada. Both the cross country shipment and the Canadian shipment were safely and successfully completed, increasing our knowledge and experience in these types of shipments. In addition, two other shipments were completed since last year's RERTR meeting. Other program activities since the last meeting included: taking pre-emptive steps to avoid license amendment pitfalls/showstoppers for spent fuel casks, publication of a revision to the Record of Decision allowing up to 16 casks per ocean going vessel, and the issuance of a cable to 16 of the 41 eligible countries reminding their governments and the reactor operators that the U.S.-origin uranium in their research reactors may be eligible for return to the United States under the Acceptance Program and urging them to begin discussions on shipping schedules. The FRR SNF program has also supported the Department's implementation of the competitive pricing policy for uranium and resumption of shipments of fresh uranium for fabrication into assemblies for research reactors. The United States Foreign Research Reactor Spent Nuclear Fuel Acceptance Program continues

  9. An economic analysis of a major bio-fuel program undertaken by OECD countries

    International Nuclear Information System (INIS)

    2002-01-01

    Biofuels such as ethanol and bio-diesel are creating a new demand for agricultural output and for agriculture land in Canada. However, the participation of other large countries with a large demand potential is necessary for bio-fuels to have a significant impact on the price of grains and oilseeds. This paper quantified the potential impact that a major bio-fuel program initiated by OECD countries has on grain and oilseed prices. The program was initiated for the period 1999 to 2006. There is considerable interest by Canadian producers to stimulate grain and oilseed prices by increasing demand of biofuels. This renewable energy source produces fewer greenhouse gas emissions than petroleum products. The analysis presented in this paper only considered ethanol from corn or wheat and bio-diesel from vegetable oils. It also focused only on the use of bio-fuels in the OECD transportation sector. The analysis was undertaken with AGLINK, a multi-commodity multi-country policy-specific dynamic model of the international agricultural markets built by the OECD with member countries. It was shown that the increase in world and domestic prices for grains and vegetable oils will remain strong, particularly toward 2006. It was also shown that a major bio-fuel program for all OECD countries would be beneficial to Canadian agriculture. It was concluded that ultimately, an increase in OECD bio-fuels usage has a direct impact on the demand for grains and oilseeds which are important feed-stocks in biofuel production. The analysis presumes an increase in renewable fuel use, but does not consider factors such as financial incentives and regulatory requirements that could bring about this increase. 7 refs., 6 tabs., 4 figs

  10. Coated Particle Fuel and Deep Burn Program Monthly Highlights May 2011

    International Nuclear Information System (INIS)

    Snead, Lance Lewis; Bell, Gary L.; Besmann, Theodore M.

    2011-01-01

    During FY 2011 the CP and DB Program will report Highlights on a monthly basis, but will no longer produce Quarterly Progress Reports. Technical details that were previously included in the quarterly reports will be included in the appropriate Milestone Reports that are submitted to FCRD Program Management. These reports will also be uploaded to the Deep Burn website. The Monthly Highlights report for April 2011, ORNL/TM-2011/125, was distributed to program participants on May 10, 2011. As reported previously, the final Quarterly for FY 2010, Deep Burn Program Quarterly Report for July - September 2010, ORNL/TM-2010/301, was announced to program participants and posted to the website on December 28, 2010. This report discusses the following: (1) Fuel Performance Modeling - Fuel Performance Analysis; (2) Thermochemical Data and Model Development - (a) Thermochemical Modeling, (b) Thermomechanical Modeling, (c) Actinide and Fission Product Transport; (3) TRU (transuranic elements) TRISO (tri-structural isotropic) Development - (a) TRU Kernel Development, (b) Coating Development; and (4) LWR Fully Ceramic Fuel - (a) FCM Fabrication Development, (b) FCM Irradiation Testing.

  11. Coated Particle Fuel and Deep Burn Program Monthly Highlights June 2011

    International Nuclear Information System (INIS)

    Snead, Lance Lewis; Bell, Gary L.; Besmann, Theodore M.

    2011-01-01

    During FY 2011 the CP and DB Program will report Highlights on a monthly basis, but will no longer produce Quarterly Progress Reports. Technical details that were previously included in the quarterly reports will be included in the appropriate Milestone Reports that are submitted to FCRD Program Management. These reports will also be uploaded to the Deep Burn website. The Monthly Highlights report for May 2011, ORNL/TM-2011/126, was distributed to program participants on June 9, 2011. As reported previously, the final Quarterly for FY 2010, Deep Burn Program Quarterly Report for July - September 2010, ORNL/TM-2010/301, was announced to program participants and posted to the website on December 28, 2010. This report discusses the following: (1) Fuel Performance Modeling - Fuel Performance Analysis; (2) Thermochemical Data and Model Development - (a) Thermochemical Behavior, (b) Thermomechanical Modeling, (c) Actinide and Fission Product Transport; (3) TRU (transuranic elements) TRISO (tri-structural isotropic) Development - (a) TRU Kernel Development, (b) Coating Development; and (4) LWR Fully Ceramic Fuel - (a) FCM Fabrication Development, (b) FCM Irradiation Testing.

  12. New version of NPP Krsko Decommissioning program and LILW and spent fuel management

    International Nuclear Information System (INIS)

    Zeleznik, N.; Mele, I.; Jenko, T.; Lokner, V.; Levanat, I.; Rapic, A.

    2004-01-01

    According to the requirements of the bilateral agreement between Republic of Slovenia and Republic of Croatia on the legal and other obligations for Nuclear power plant (NPP) Krsko the Decommissioning program was prepared. The main purpose of the program was to estimate the overall expenses of the future decommissioning, radioactive waste and spent fuel management of the NPP Krsko in order to establish separate fund in Croatia and to correct the rate per kWh collected in the existing decommissioning fund in Slovenia. The program looked at all possible scenarios of dismantling, radioactive waste and spent fuel management and proposed the most plausible two scenarios which are technically possible and financially feasible. (author)

  13. SOLID STATE ENERGY CONVERSION ALLIANCE (SECA) SOLID OXIDE FUEL CELL PROGRAM

    Energy Technology Data Exchange (ETDEWEB)

    Unknown

    2003-06-01

    This report summarizes the progress made during the September 2001-March 2002 reporting period under Cooperative Agreement DE-FC26-01NT41245 for the U. S. Department of Energy, National Energy Technology Laboratory (DOE/NETL) entitled ''Solid State Energy Conversion Alliance (SECA) Solid Oxide Fuel Cell Program''. The program focuses on the development of a low-cost, high-performance 3-to-10-kW solid oxide fuel cell (SOFC) system suitable for a broad spectrum of power-generation applications. The overall objective of the program is to demonstrate a modular SOFC system that can be configured to create highly efficient, cost-competitive, and environmentally benign power plants tailored to specific markets. When fully developed, the system will meet the efficiency, performance, life, and cost goals for future commercial power plants.

  14. Irradiation program of slightly enriched fuel elements at the Atucha I nuclear power plant

    International Nuclear Information System (INIS)

    Casario, J.A.; Cesario, R.H.; Perez, R.A.; Sidelnik, J.I.

    1987-01-01

    An irradiation program of fuel elements with slightly enriched uranium is implemented, tending to the homogenization of core at Atucha I nuclear power plant. The main benefits of the enrichment program are: a) to extend the average discharge burnup of fuel elements, reducing the number of elements used to generate the same amount of energy. This implies a smaller annual consumption of elements and consequently the reduction of transport and replacement operations and of the storage pool systems as well as that of radioactive wastes; b) the saving of uranium and structural materials (Zircaloy and others). In the initial stage of program an homogeneous core enrichment of 0.85% by weight of U-235 is anticipated. The average discharge burnup of fuel elements, as estimated by previous studies, is approximately 11.6 MW d/kg U. The annual consumption of fuel elements is reduced from 396 of natural uranium to 205, with a load factor of 0.85. It is intended to reach the next equilibrium steps with an enrichment of 1.00 and 1.20% in U-235. (Author)

  15. Consolidated Fuel Reprocessing Program. Operating experience with pulsed-column holdup estimators

    International Nuclear Information System (INIS)

    Ehinger, M.H.

    1986-01-01

    Methods for estimating pulsed-column holdup are being investigated as part of the Safeguards Assessment task of the Consolidated Fuel Reprocessing Program (CFRP) at the Oak Ridge National Laboratory. The CFRP was a major sponsor of test runs at the Barnwell Nuclear Fuel plant (BNFP) in 1980 and 1981. During these tests, considerable measurement data were collected for pulsed columns in the plutonium purification portion of the plant. These data have been used to evaluate and compare three available methods of holdup estimation

  16. Nerva Fuel Element Development Program Summary Report - July 1966 through June 1972 Extrusion Studies

    Energy Technology Data Exchange (ETDEWEB)

    Napier, J. M.

    1973-09-21

    This part of the completion report pertaining to the NERVA graphite fuel element program covers data collected during the extrusion studies. The physical properties of the fuel element reached the following values: coefficient of thermal expansion (CTE) - 7.0 x 10-6/o C (25 - l,OOOo C); modulus of elasticity - 1.5 x lo6 psi; flexural strength - - 8,000 psi; ultimate strain to failure - 5,500 pidin; good thermal stress resistance. Matrices were produced which could be vapor coated with crack-free films of zirconium carbide. The CTE of the matrix was almost equal to the CTE of the zirconium carbide coating.

  17. U. S. Fuel Cycle Technologies R and D Program for Next Generation Nuclear Materials Management

    International Nuclear Information System (INIS)

    Miller, M. C.; Vega, D. A.

    2013-01-01

    The U. S. Department of Energy's Fuel Cycle Technologies R and D program under the Office of Nuclear Energy is working to advance technologies to enhance both the existing and future fuel cycles. One thrust area is in developing enabling technologies for next generation nuclear materials management under the Materials Protection, Accounting and Control Technologies (MPACT) Campaign where advanced instrumentation, analysis and assessment methods, and security approaches are being developed under a framework of Safeguards and Security by Design. An overview of the MPACT campaign's activities and recent accomplishments is presented along with future plans

  18. Technology development program for safe shipment of spent fuel from liquid metal fast breeder reactor

    International Nuclear Information System (INIS)

    Freedman, J.M.; Humphreys, J.R.

    1975-10-01

    A comprehensive plan to develop shipping cask technology is described. Technical programs in the disciplines of heat transfer, structures and containment, spent fuel characterization, hot laboratory verification, shielding, and hazards analysis are discussed. Both short- and long-term goals in each discipline are delineated and how the disciplines interrelate is shown. The technologies developed will be used in the design, fabrication, and testing of truck-mounted and rail-car casks. These casks will be used for safely transporting short-cooled, high-burnup Liquid Metal Fast Breeder Reactor (LMFBR) spent fuel from reactors to reprocessing plants

  19. The use of burnup credit in criticality control for the Korean spent fuel management program

    International Nuclear Information System (INIS)

    Koh, Duck Joon; Chon, Je Keun; Park, Chung Ryul; Ji, Pyung Kuk; Kim, Byung Tae; Jo, Chang Keun; Cho, Nam Zin

    1997-01-01

    More than 25% k-eff saving effect is observed in this burnup credit analysis. This mainly comes from the adoption of actinide nuclides and fission products in the criticality analysis. By taking burnup credit, the high capacity of the storage and transportation can be more fully utilized, reducing the space of storage and the number of shipments. Larger storage and fewer shipments for a given inventory of spent fuel result should in remarkable cost savings and more importantly reduce the risks to the public and occupational workers for the Korean Spent Fuel Management Program

  20. U.S. FUEL CYCLE TECHNOLOGIES R&D PROGRAM FOR NEXT GENERATION NUCLEAR MATERIALS MANAGEMENT

    Directory of Open Access Journals (Sweden)

    M.C. MILLER

    2013-11-01

    Full Text Available The U.S. Department of Energy's Fuel Cycle Technologies R&D program under the Office of Nuclear Energy is working to advance technologies to enhance both the existing and future fuel cycles. One thrust area is in developing enabling technologies for next generation nuclear materials management under the Materials Protection, Accounting and Control Technologies (MPACT Campaign where advanced instrumentation, analysis and assessment methods, and security approaches are being developed under a framework of Safeguards and Security by Design. An overview of the MPACT campaign's activities and recent accomplishments is presented along with future plans.

  1. Program of quality management when fabricating fast reactor vibropack oxide fuel pins

    International Nuclear Information System (INIS)

    Mayorshin, A.A.; Kisly, V.A.; Sudakov, L.V.

    2000-01-01

    There are presented main principles of creation and operation of Quality Management Program in fabricating vibropack oxide fuel pins for BOR-60 and BN-600 being in force in SSC RF RIAR. There is given structure of documentation for QS principal elements. Under Quality System there are defined all the procedures, assuring that fuel pin meets the normative requirements. The system model is complied with the standard model IS 9001. There are shown technologic flowchart and check operation, statistic results of pin critical parameter check as well as main results of in-pile tests. (author)

  2. Summary of national and international fuel cycle and radioactive waste management programs, 1984

    Energy Technology Data Exchange (ETDEWEB)

    Harmon, K.M.; Lakey, L.T.; Leigh, I.W.

    1984-07-01

    Worldwide activities related to nuclear fuel cycle and radioactive waste management programs are summarized. Several trends have developed in waste management strategy: All countries having to dispose of reprocessing wastes plan on conversion of the high-level waste (HLW) stream to a borosilicate glass and eventual emplacement of the glass logs, suitably packaged, in a deep geologic repository. Countries that must deal with plutonium-contaminated waste emphasize pluonium recovery, volume reduction and fixation in cement or bitumen in their treatment plans and expect to use deep geologic repositories for final disposal. Commercially available, classical engineering processing are being used worldwide to treat and immobilize low- and intermediate-level wastes (LLW, ILW); disposal to surface structures, shallow-land burial and deep-underground repositories, such as played-out mines, is being done widely with no obvious technical problems. Many countries have established extensive programs to prepare for construction and operation of geologic repositories. Geologic media being studied fall into three main classes: argillites (clay or shale); crystalline rock (granite, basalt, gneiss or gabbro); and evaporates (salt formations). Most nations plan to allow 30 years or longer between discharge of fuel from the reactor and emplacement of HLW or spent fuel is a repository to permit thermal and radioactive decay. Most repository designs are based on the mined-gallery concept, placing waste or spent fuel packages into shallow holes in the floor of the gallery. Many countries have established extensive and costly programs of site evaluation, repository development and safety assessment. Two other waste management problems are the subject of major R and D programs in several countries: stabilization of uranium mill tailing piles; and immobilization or disposal of contaminated nuclear facilities, namely reactors, fuel cycle plants and R and D laboratories.

  3. Summary of national and international fuel cycle and radioactive waste management programs, 1984

    International Nuclear Information System (INIS)

    Harmon, K.M.; Lakey, L.T.; Leigh, I.W.

    1984-07-01

    Worldwide activities related to nuclear fuel cycle and radioactive waste management programs are summarized. Several trends have developed in waste management strategy: All countries having to dispose of reprocessing wastes plan on conversion of the high-level waste (HLW) stream to a borosilicate glass and eventual emplacement of the glass logs, suitably packaged, in a deep geologic repository. Countries that must deal with plutonium-contaminated waste emphasize pluonium recovery, volume reduction and fixation in cement or bitumen in their treatment plans and expect to use deep geologic repositories for final disposal. Commercially available, classical engineering processing are being used worldwide to treat and immobilize low- and intermediate-level wastes (LLW, ILW); disposal to surface structures, shallow-land burial and deep-underground repositories, such as played-out mines, is being done widely with no obvious technical problems. Many countries have established extensive programs to prepare for construction and operation of geologic repositories. Geologic media being studied fall into three main classes: argillites (clay or shale); crystalline rock (granite, basalt, gneiss or gabbro); and evaporates (salt formations). Most nations plan to allow 30 years or longer between discharge of fuel from the reactor and emplacement of HLW or spent fuel is a repository to permit thermal and radioactive decay. Most repository designs are based on the mined-gallery concept, placing waste or spent fuel packages into shallow holes in the floor of the gallery. Many countries have established extensive and costly programs of site evaluation, repository development and safety assessment. Two other waste management problems are the subject of major R and D programs in several countries: stabilization of uranium mill tailing piles; and immobilization or disposal of contaminated nuclear facilities, namely reactors, fuel cycle plants and R and D laboratories

  4. COXPRO-II: a computer program for calculating radiation and conduction heat transfer in irradiated fuel assemblies

    International Nuclear Information System (INIS)

    Rhodes, C.A.

    1984-12-01

    This report describes the computer program COXPRO-II, which was written for performing thermal analyses of irradiated fuel assemblies in a gaseous environment with no forced cooling. The heat transfer modes within the fuel pin bundle are radiation exchange among fuel pin surfaces and conduction by the stagnant gas. The array of parallel cylindrical fuel pins may be enclosed by a metal wrapper or shroud. Heat is dissipated from the outer surface of the fuel pin assembly by radiation and convection. Both equilateral triangle and square fuel pin arrays can be analyzed. Steady-state and unsteady-state conditions are included. Temperatures predicted by the COXPRO-II code have been validated by comparing them with experimental measurements. Temperature predictions compare favorably to temperature measurements in pressurized water reactor (PWR) and liquid-metal fast breeder reactor (LMFBR) simulated, electrically heated fuel assemblies. Also, temperature comparisons are made on an actual irradiated Fast-Flux Test Facility (FFTF) LMFBR fuel assembly

  5. Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs draft environmental impact statement. Volume 1, Appendix B: Idaho National Engineering Laboratory Spent Nuclear Fuel Management Program

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

    The US Department of Energy (DOE) has prepared this report to assist its management in making two decisions. The first decision, which is programmatic, is to determine the management program for DOE spent nuclear fuel. The second decision is on the future direction of environmental restoration, waste management, and spent nuclear fuel management activities at the Idaho National Engineering Laboratory. Volume 1 of the EIS, which supports the programmatic decision, considers the effects of spent nuclear fuel management on the quality of the human and natural environment for planning years 1995 through 2035. DOE has derived the information and analysis results in Volume 1 from several site-specific appendixes. Volume 2 of the EIS, which supports the INEL-specific decision, describes environmental impacts for various environmental restoration, waste management, and spent nuclear fuel management alternatives for planning years 1995 through 2005. This Appendix B to Volume 1 considers the impacts on the INEL environment of the implementation of various DOE-wide spent nuclear fuel management alternatives. The Naval Nuclear Propulsion Program, which is a joint Navy/DOE program, is responsible for spent naval nuclear fuel examination at the INEL. For this appendix, naval fuel that has been examined at the Naval Reactors Facility and turned over to DOE for storage is termed naval-type fuel. This appendix evaluates the management of DOE spent nuclear fuel including naval-type fuel.

  6. Thermal Hydraulic Fortran Program for Steady State Calculations of Plate Type Fuel Research Reactors

    International Nuclear Information System (INIS)

    Khedr, H.

    2008-01-01

    The safety assessment of Research and Power Reactors is a continuous process over their life and that requires verified and validated codes. Power Reactor codes all over the world are well established and qualified against a real measuring data and qualified experimental facilities. These codes are usually sophisticated, require special skills and consume much more running time. On the other hand, most of the Research Reactor codes still requiring more data for validation and qualification. Therefore it is benefit for a regulatory body and the companies working in the area of Research Reactor assessment and design to have their own program that give them a quick judgment. The present paper introduces a simple one dimensional Fortran program called THDSN for steady state best estimate Thermal Hydraulic (TH) calculations of plate type fuel RRs. Beside calculating the fuel and coolant temperature distribution and pressure gradient in an average and hot channel the program calculates the safety limits and margins against the critical phenomena encountered in RR such as the burnout heat flux and the onset of flow instability. Well known TH correlations for calculating the safety parameters are used. THDSN program is verified by comparing its results for 2 and 10 MW benchmark reactors with that published in IAEA publications and good agreement is found. Also the program results are compared with those published for other programs such as PARET and TERMIC. An extension for this program is underway to cover the transient TH calculations

  7. Planning of fuel coal imports using a mixed integer programming method

    International Nuclear Information System (INIS)

    Shih, L.H.

    1997-01-01

    In the public utility and commercial fuel industries, commodities from multiple supply sources are sometimes blended before use to reduce costs and assure quality. A typical example of these commodities is the fuel coal used in coal fired power plants. The diversity of the supply sources for these plants makes the planning and scheduling of fuel coal logistics difficult, especially for a power company that has more than one power plant. This study proposes a mixed integer programming model that provides planning and scheduling of coal imports from multiple suppliers for the Taiwan Power Company. The objective is to minimize total inventory cost by minimizing procurement cost, transportation cost and holding cost. Constraints on the system include company procurement policy, power plant demand, harbor unloading capacity, inventory balance equations, blending requirements, and safety stock. An example problem is presented using the central coal logistics system of the Taiwan Power Company to demonstrate the validity of the proposed model

  8. Planning of fuel coal imports using a mixed integer programming method

    Energy Technology Data Exchange (ETDEWEB)

    Shih, L.H. [National Cheng Kung University, Tainan (Taiwan). Dept. of Mineral and Petroleum Engineering

    1997-12-31

    In the public utility and commercial fuel industries, commodities from multiple supply sources are sometimes blended before use to reduce costs and assure quality. A typical example of these commodities is the fuel coal used in coal fired power plants. The diversity of the supply sources for these plants makes the planning and scheduling of fuel coal logistics difficult, especially for a power company that has more than one power plant. This study proposes a mixed integer programming model that provides planning and scheduling of coal imports from multiple suppliers for the Taiwan Power Company. The objective is to minimize total inventory cost by minimizing procurement cost, transportation cost and holding cost. Constraints on the system include company procurement policy, power plant demand, harbor unloading capacity, inventory balance equations, blending requirements, and safety stock. An example problem is presented using the central coal logistics system of the Taiwan Power Company to demonstrate the validity of the proposed model.

  9. Solid State Energy Conversion Alliance (SECA) Solid Oxide Fuel Cell Program

    Energy Technology Data Exchange (ETDEWEB)

    Nguyen Minh

    2006-07-31

    This report summarizes the work performed for Phase I (October 2001 - August 2006) under Cooperative Agreement DE-FC26-01NT41245 for the U. S. Department of Energy, National Energy Technology Laboratory (DOE/NETL) entitled 'Solid State Energy Conversion Alliance (SECA) Solid Oxide Fuel Cell Program'. The program focuses on the development of a low-cost, high-performance 3-to-10-kW solid oxide fuel cell (SOFC) system suitable for a broad spectrum of power-generation applications. During Phase I of the program significant progress has been made in the area of SOFC technology. A high-efficiency low-cost system was designed and supporting technology developed such as fuel processing, controls, thermal management, and power electronics. Phase I culminated in the successful demonstration of a prototype system that achieved a peak efficiency of 41%, a high-volume cost of $724/kW, a peak power of 5.4 kW, and a degradation rate of 1.8% per 500 hours. . An improved prototype system was designed, assembled, and delivered to DOE/NETL at the end of the program. This prototype achieved an extraordinary peak efficiency of 49.6%.

  10. Fuels planning: science synthesis and integration; social issues fact sheet 01: Developing personal responsibility for fuels reduction: Building a successful program to engage property owners

    Science.gov (United States)

    Rocky Mountain Research Station USDA Forest Service

    2004-01-01

    In the course of work as a land manager, you will no doubt be involved in developing programs to achieve various objectives, including the improvement of fuels management on private lands. This fact sheet describes six steps that will help you plan and conduct a successful program.

  11. Romanian-Canadian joint program for qualification of FCN as a CANDU fuel supplier

    International Nuclear Information System (INIS)

    Galeriu, C.A.; Andrei, G.; Bailescu, A.

    1995-01-01

    RENEL (Romania Power Authority), the co-ordinator of Romanian Nuclear Program, have decided to improve, starting 1990 the existing capability to produce CANDU nuclear fuel at FCN Pitesti. The objective of the program was defined with AAC (AECL - ANSALDO Consortium) for the qualification of FCN fuel plant according to Canadian Z299.2 standard. The Qualification Program was performed under AAC Work Order C-003. The co-ordination was assumed by AECL, as overall Design Authority. ZPI (Zircatec Precision Industries Inc., Canada), were designated to supply technical assistance, equipments and know how where necessary. After a preliminary verification of the FCN fuel plant, including the processes and system investigation, performed under AECL and ZPI assistance, the Qualification Program was defined in all details. The upgrading of documentation on all aspects required by Z299.2 was performed. Few processes needed to be reconsidered and equipment was delivered by ZPI or other suppliers. This includes mainly welding equipments and special inspection equipments. Health Physics was practically fully reconsidered. New equipment and practice were adapted to provide adequate control on health conditions. Every manufacturing and inspection process was checked to determine their performance during a Qualification Run based on acceptance criteria which have been established in the Qualification Plan. Manufacturing Demonstration Run was an important step to prove that all plant functions have been accomplished during the fabrication of 200 fuel bundles. These bundles have been fully accepted and 66 of them have been loaded in the first charge of Unit 1 Cemavoda NPS. The surveillance and audit actions made by AECL and ZPI during this period confirmed the FCN capability to operate an adequate system meeting the to required quality assurance standard. The very open attitude of AECL, Zircatec and FCN staff have stimulated the progress of the project and a successful achievement of the

  12. Program for responsible and safe disposal of spent fuel elements and radioactive wastes (National disposal program)

    International Nuclear Information System (INIS)

    2015-01-01

    The contribution covers the following topics: fundamentals of the disposal policy; amount of radioactive wastes and prognosis; disposal of radioactive wastes - spent fuel elements and wastes from waste processing, radioactive wastes with low heat production; legal framework of the nuclear waste disposal in Germany; public participation, cost and financing.

  13. Synthesis and Characterization of Oxide Feedstock Powders for the Fuel Cycle R and D Program

    International Nuclear Information System (INIS)

    Voit, Stewart L.; Vedder, Raymond James; Johnson, Jared A.

    2010-01-01

    Nuclear fuel feedstock properties, such as physical, chemical, and isotopic characteristics, have a significant impact on the fuel fabrication process and, by extension, the in-reactor fuel performance. This has been demonstrated through studies with UO 2 spanning greater than 50 years. The Fuel Cycle R and D Program with The Department of Energy Office of Nuclear Energy has initiated an effort to develop a better understanding of the relationships between oxide feedstock, fresh fuel properties, and in-reactor fuel performance for advanced mixed oxide compositions. Powder conditioning studies to enable the use of less than ideal powders for ceramic fuel pellet processing are ongoing at Los Alamos National Laboratory (LANL) and an understanding of methods to increase the green density and homogeneity of pressed pellets has been gained for certain powders. Furthermore, Oak Ridge National Laboratory (ORNL) is developing methods for the co-conversion of mixed oxides along with techniques to analyze the degree of mixing. Experience with the fabrication of fuel pellets using co-synthesized multi-constituent materials is limited. In instances where atomically mixed solid solutions of two or more species are needed, traditional ceramic processing methods have been employed. Solution-based processes may be considered viable synthesis options, including co-precipitation (AUPuC), direct precipitation, direct-conversion (Modified Direct Denitration or MDD) and internal/external gelation (sol-gel). Each of these techniques has various advantages and disadvantages. The Fiscal Year 2010 feedstock development work at ORNL focused on the synthesis and characterization of one batch of UO x and one batch of U 80 Ce 20 O x . Oxide material synthesized at ORNL is being shipped to LANL for fuel fabrication process development studies. The feedstock preparation was performed using the MDD process which utilizes a rotary kiln to continuously thermally denitrate double salts of ammonium

  14. Characterization program management plan for Hanford K Basin spent nuclear fuel

    International Nuclear Information System (INIS)

    Lawrence, L.A.

    1998-01-01

    The management plan developed to characterize the K Basin Spent Nuclear Fuel was revised to incorporate actions necessary to comply with the Office of Civilian Radioactive Waste Management Quality Assurance Requirements Document 0333P. This plan was originally developed for Westinghouse Hanford Company and Pacific Northwest National Laboratory to work together on a program to provide characterization data to support removal, conditioning, and subsequent dry storage of the spent nuclear fuels stored at the Hanford K Basins. This revision to the Program Management Plan replaces Westinghouse Hanford Company with Duke Engineering and Services Hanford, Inc., updates the various activities where necessary, and expands the Quality Assurance requirements to meet the applicable requirements document. Characterization will continue to utilize the expertise and capabilities of both organizations to support the Spent Nuclear Fuels Project goals and objectives. This Management Plan defines the structure and establishes the roles for the participants providing the framework for Duke Engineering and Services Hanford, Inc. and Pacific Northwest National Laboratory to support the Spent Nuclear Fuels Project at Hanford

  15. ROBOT3: a computer program to calculate the in-pile three-dimensional bowing of cylindrical fuel rods (AWBA Development Program)

    International Nuclear Information System (INIS)

    Kovscek, S.E.; Martin, S.E.

    1982-10-01

    ROBOT3 is a FORTRAN computer program which is used in conjunction with the CYGRO5 computer program to calculate the time-dependent inelastic bowing of a fuel rod using an incremental finite element method. The fuel rod is modeled as a viscoelastic beam whose material properties are derived as perturbations of the CYGRO5 axisymmetric model. Fuel rod supports are modeled as displacement, force, or spring-type nodal boundary conditions. The program input is described and a sample problem is given

  16. Moving into the 21st century - The United States' Research Reactor Spent Nuclear Fuel Acceptance Program

    International Nuclear Information System (INIS)

    Huizenga, David G.; Mustin, Tracy P.; Saris, Elizabeth C.; Reilly, Jill E.

    1999-01-01

    Since 1996, when the United States Department of Energy and the Department of State jointly adopted the Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel, twelve shipments totaling 2,985 MTR and TRIGA spent nuclear fuel assemblies from research reactors around the world have been accepted into the United States. These shipments have contained approximately 1.7 metric tons of HEU and 0.6 metric tons of LEU. Foreign research reactor operators played a significant role in this success. A new milestone in the acceptance program occurred during the summer of 1999 with the arrival of TRIGA spent nuclear fuel from Europe through the Charleston Naval Weapons Station via the Savannah River Site to the Idaho National Engineering and Environmental Laboratory. This shipment consisted of five casks of TRIGA spent nuclear fuel from research reactors in Germany, Italy, Slovenia, and Romania. These casks were transported by truck approximately 2,400 miles across the United States (one cask packaged in an ISO container per truck). Drawing upon lessons learned in previous shipments, significant technical, legal, and political challenges were addressed to complete this cross-country shipment. Other program activities since the last RERTR meeting have included: formulation of a methodology to determine the quantity of spent nuclear fuel in a damaged condition that may be transported in a particular cask (containment analysis for transportation casks); publication of clarification of the fee policy; and continued planning for the outyears of the acceptance policy including review of reactors and eligible material quantities. The United States Foreign Research Reactor Spent Nuclear Fuel Acceptance Program continues to demonstrate success due to the continuing commitment between the United States and the research reactor community to make this program work. We strongly encourage all eligible research reactors to decide as soon as possible to

  17. Impacts of the Weatherization Assistance Program in Fuel-Oil Heated Houses

    Energy Technology Data Exchange (ETDEWEB)

    Levins, W.P.

    1994-01-01

    In 1990, the U.S. Department of Energy (DOE) initiated a national evaluation of its low-income Weatherization Assistance Program. This report, which is one of five parts of that evaluation, evaluates the energy savings and cost-effectiveness of the Program as it had been applied to single-family houses heated primarily by fuel-oil. The study was based upon a representative sample (41 local weatherization agencies, 222 weatherized and 115 control houses) from the nine northeastern states during 1991 and 1992 program years. Dwelling-specific and agency-level data on measures installed, costs, and service delivery procedures were collected from the sampled agencies. Space-heating fuel-oil consumption, indoor temperature, and outdoor temperature were monitored at each house. Dwelling characteristics, air-leakage measurements, space-heating system steady-state efficiency measurements, safety inspections, and occupant questionnaires were also collected or performed at each monitored house. We estimate that the Program weatherized a total of 23,400 single-family fuel-oil heated houses in the nine northeastern states during program years 1991 and 1992. Annual fuel-oil savings were calculated using regression techniques to normalize the savings to standard weather conditions. For the northeast region, annual net fuel-oil savings averaged 160 gallons per house, or 17.7% of pre-weatherization consumption. Although indoor temperatures changed in individual houses following weatherization, there was no average change and no significant difference as compared to the control houses; thus, there was no overall indoor temperature takeback effect influencing fuel-oil savings. The weatherization work was performed cost effectively in these houses from the Program perspective, which included both installation costs and overhead and management costs but did not include non-energy benefits (such as employment and environmental). Total average costs were $1819 per house ($1192 for

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

    International Nuclear Information System (INIS)

    Maraman, W.J.

    1979-12-01

    This formal monthly report covers the studies related to the use of 238 PuO 2 in radioisotopic power systems carried out for the Advanced Nuclear Systems and Projects Division of the Los Alamos Scientific Laboratory. The two programs involved are general-purpose heat source development and space nuclear safety and fuels. Most of the studies discussed hear are of a continuing nature. Results and conclusions described may change as the work continues

  19. Prototype vibration measurement program for reactor internals (177-fuel assembly plant). Supplement 1

    International Nuclear Information System (INIS)

    Simonis, J.C.; Post, R.C.; Thoren, D.E.

    1976-08-01

    The surveillance specimen holder tubes installed in the Babcock and Wilcox 177-fuel assembly plants have been redesigned. The structural adequacy of this design has been verified through extensive analysis. The design adequacy will be further confirmed by measuring the vibrational response of the surveillance specimen holder tube during normal and transient flow operation. This report describes the vibration measurement program that will be conducted at Toledo Edison's Davis Besse 1 site

  20. PHEBUS program: first results on PWR fuel behaviour in LOCA conditions

    International Nuclear Information System (INIS)

    Del Negro, R.; Reocreux, M.; Pelce, J.; Legrand, B.; Berna, P.

    1982-09-01

    In the first PHEBUS test with pressurized rods some rods burst and clad temperature reached 1100 0 C in the 25 rods bundle. There is now a lot of valuable experimental results and their analysis is in progress. The phase II on fuel behaviour in case of a large LOCA will start at the beginning of 83. The onset of the SFD program is foreseen to take place on the first months of 85

  1. Development of a Computer Program (CASK) for the Analysis of Logistics and Transportation Cost of the Spent Fuels

    International Nuclear Information System (INIS)

    Cha, Jeong-Hun; Choi, Heui-Joo; Cho, Dong-Keun; Kim, Seong-Ki; Lee, Jong-Youl; Choi, Jong-Won

    2008-07-01

    The cost for the spent fuel management includes the costs for the interim storage, the transportation, and the permanent disposal of the spent fuels. The CASK(Cost and logistics Analysis program for Spent fuel transportation in Korea) program is developed to analyze logistics and transportation cost of the spent fuels. And the total amount of PWR spent fuels stored in four nuclear plant sites, a centralized interim storage facility near coast and a permanent disposal facility near the interim storage facility are considered in this program. The CASK program is developed by using Visual Basic language and coupled with an excel sheet. The excel sheet shows a change of logistics and transportation cost. Also transportation unit cost is easily changed in the excel sheet. The scopes of the report are explanation of parameters in the CASK program and a preliminary calculation. We have developed the CASK version 1.0 so far, and will update its parameters in transportation cost and transportation scenario. Also, we will incorporate it into the program which is used for the projection of spent fuels from the nuclear power plants. Finally, it is expected that the CASK program could be a part of the cost estimation tools which are under development at KAERI. And this program will be a very useful tool for the establishment of transportation scenario and transportation cost in Korean situations

  2. Criticality safety assessment on the RSG-GAS spent fuel storage for anticipating the next core conversion program

    International Nuclear Information System (INIS)

    Sembiring, Tagor Malem; Kuntoro, Iman; Zuhair; Liem, Peng Hong

    2003-01-01

    Criticality assessment on the spent fuel storage racks of the RSG-GAS multipurpose reactor has been conducted to support the undergoing core conversion program, in which higher uranium fuel densities of silicide (up to 4.8 gU.cm -3 ) and molybdenum (up to 8.3 gU.cm -3 ) fuel elements are adopted to enhance the reactor performance, core cycle length and reactor utilization. In the assessment, the k eff of the rack as a function of fuel density is calculated for fresh fuel elements which is a very conservative approach recommended by IAEA. Besides fuel densities, effects of water densities due to pool water temperature variation, and the fuel elements' orientation on the k eff are analyzed as well. The criticality calculations are all carried out by using MNCP4B2 Monte Carlo code with ENDF/B-VI library. For the library sensitivity, JENDL-3.3 library is also used and compared. The calculation results show the most reactive condition is for the case when the spent fuel racks are filled with fresh U-6Mo fuel element with meat density of 8.30 gU.cm -3 . For all fuel types, density and operating condition, the calculated k eff with 3 times standard deviations are confirmed less than the allowable value of 0.95. It can be concluded that the existing spent fuel storage racks can be safely used for storing the planned high density uranium fuels. (author)

  3. FUEL Your Life: A Translation of the Diabetes Prevention Program to Worksites.

    Science.gov (United States)

    Wilson, Mark G; DeJoy, David M; Vandenberg, Robert; Padilla, Heather; Davis, Marsha

    2016-01-01

    To evaluate the effectiveness of FUEL Your Life, a translation of the Diabetes Prevention Program for worksites. A randomized control group design was conducted in five worksites of a large transportation company. Measures were collected pretest, posttest (6 months), and follow-up (12 months). Railroad maintenance facilities of Union Pacific Railroad. Participants consisted of 362 workers (227 treatment, 135 control). FUEL Your Life was translated from the Diabetes Prevention Program to better fit within the context of the worksite. The primary difference was the use of peer health coaches to provide social support and reinforcement and an occupational nurse to provide lesson content (six sessions of 10 minutes) to participants instead of the lifestyle coaches employed by the Diabetes Prevention Program, resulting in a less structured meeting schedule. The primary outcomes were weight and body mass index (BMI), with secondary outcomes including eating behaviors, physical activity, and social support. Latent growth modeling was used to measure changes in the outcomes over time. Participants in the intervention group maintained weight/BMI (-.1 pounds/-.1 BMI), whereas the control participants gained weight/BMI (+2.6 pounds/+.3 BMI), resulting in a statistically significant difference between groups. Fifty-five percent of intervention participants lost some weight, whereas only 35% of the control group lost weight. FUEL Your Life, a low intensity intervention, was not effective for promoting weight loss, but was effective for helping workers maintain weight over a 12-month period.

  4. Fission gas release from ThO2 and ThO2--UO2 fuels (LWBR development program)

    International Nuclear Information System (INIS)

    Goldberg, I.; Spahr, G.L.; White, L.S.; Waldman, L.A.; Giovengo, J.F.; Pfennigwerth, P.L.; Sherman, J.

    1978-08-01

    Fission gas release data are presented from 51 fuel rods irradiated as part of the LWBR irradiations test program. The fuel rods were Zircaloy-4 clad and contained ThO 2 or ThO 2 -UO 2 fuel pellets, with UO 2 compositions ranging from 2.0 to 24.7 weight percent and fuel densities ranging from 77.8 to 98.7 percent of theoretical. Rod diameters ranged from 0.25 to 0.71 inches and fuel active lengths ranged from 3 to 84 inches. Peak linear power outputs ranged from 2 to 22 kw/ft for peak fuel burnups up to 56,000 MWD/MTM. Measured fission gas release was quite low, ranging from 0.1 to 5.2 percent. Fission gas release was higher at higher temperature and burnup and was lower at higher initial fuel density. No sensitivity to UO 2 composition was evidenced

  5. Microbial mediation of radionuclide transport -significance for the nuclear fuel waste management program

    International Nuclear Information System (INIS)

    Champ, D.R.

    1984-08-01

    The role that microbially catalyzed processes may play in determining, or altering, radionuclide migration is an unresolved question in the Nuclear Fuel Waste Management Program. This report documents the results of a review of the available information on the existence and biochemical capabilities of micro-organisms and the potential for their involvement in processes affecting the migration of radionuclides of interest. The potential was judged sufficient to warrant conducting experiments to assess their role. The outline of an experimental program to address the role of micro-organisms is presented

  6. Advanced containment research for the Canadian Nuclear Fuel Waste Management Program

    International Nuclear Information System (INIS)

    Onofrei, M.; Mathew, P.M.; McKay, P.; Hosaluk, L.J.; Oscarson, D.W.

    1986-09-01

    This document outlines the program on the development of advanced containment systems for the disposal of used fuel in a vault deep in plutonic rock. Possible advanced containment concepts, the strategy adopted in selecting potential container materials, and experimental programs currently underway or planned are presented. Most effort is currently directed toward developing long-term containment systems based on non-metallic materials and massive metal containers. The use of additional independent barriers to extend the lifetime of simple containment systems is also being evaluated. 58 refs

  7. Development programs on decommissioning technology for reactors and fuel cycle facilities in Japan

    International Nuclear Information System (INIS)

    Fujiki, K.

    1992-01-01

    The Science and Technology Agency (STA) of Japan is promoting technology development for decommissioning of nuclear facilities by entrusting various research programs to concerned research organisations: JAERI, PNC and RANDEC, including first full scale reactor decommissioning of JPDR. According to the results of these programs, significant improvement on dismantling techniques, decontamination, measurement etc. has been achieved. Further development of advanced decommissioning technology has been started in order to achieve reduction of duration of decommissioning work and occupational exposures in consideration of the decommissioning of reactors and fuel cycle facilities. (author) 5 refs.; 7 figs.; 1 tab

  8. Program for in-pile qualification of high density silicide dispersion fuel at IPEN/CNEN-SP

    International Nuclear Information System (INIS)

    Silva, Jose E.R. da; Silva, Antonio T. e; Terremoto, Luis A.A.; Durazzo, Michelangelo

    2009-01-01

    The development of high density nuclear fuel (U 3 Si 2 -Al) with 4,8 gU/cm 3 is on going at IPEN, at this time. This fuel has been considered to be utilized at the new Brazilian Multipurpose Reactor (RMB), planned to be constructed up to 2014. As Brazil does not have hot-cell facilities available for post-irradiation analysis, an alternative qualifying program for this fuel is proposed based on the same procedures used at IPEN since 1988 for qualifying its own U 3 O 8 -Al (1,9 and 2,3 gU/cm 3 ) and U 3 Si 2 -Al (3,0 gU/cm 3 ) dispersion fuels. The fuel miniplates and full-size fuel elements irradiations should be tested at IEA-R1 core. The fuel characterization along the irradiation time should be made by means of non-destructive methods, including periodical visual inspections with an underwater video camera system, sipping tests for fuel elements suspected of leakage, and underwater dimensional measurements for swelling evaluation, performed inside the reactor pool. This work presents the program description for the qualification of the high density nuclear fuel (U 3 Si 2 -Al) with 4,8 gU/cm 3 , and describes the IPEN fuel fabrication infrastructure and some basic features of the available systems for non-destructive tests at IEA-R1 research reactor. (author)

  9. 40 CFR 80.500 - What are the implementation dates for the motor vehicle diesel fuel sulfur control program?

    Science.gov (United States)

    2010-07-01

    ... Information § 80.500 What are the implementation dates for the motor vehicle diesel fuel sulfur control... 40 Protection of Environment 16 2010-07-01 2010-07-01 false What are the implementation dates for the motor vehicle diesel fuel sulfur control program? 80.500 Section 80.500 Protection of Environment...

  10. Foreign programs for the storage of spent nuclear power plant fuels, high-level waste canisters and transuranic wastes

    International Nuclear Information System (INIS)

    Harmon, K.M.; Johnson, A.B. Jr.

    1984-04-01

    The various national programs for developing and applying technology for the interim storage of spent fuel, high-level radioactive waste, and TRU wastes are summarized. Primary emphasis of the report is on dry storage techniques for uranium dioxide fuels, but data are also provided concerning pool storage

  11. Status of the US foreign research reactor spent nuclear fuel program

    International Nuclear Information System (INIS)

    Chacey, K.A.; Zeitoun, A.; Saris, E.C.

    1997-01-01

    A significant step was made in 1996 with the establishment of a new nuclear weapons nonproliferation policy concerning foreign research reactor spent nuclear fuel. Specifically the United States will accept over a 13-year period up to 20 tonnes of spent nuclear fuel from 41 countries. Only spent fuel containing uranium enriched in the United States is covered under this policy. Since the acceptance policy took effect on 13 May 1996, the Department of Energy has undertaken a number of steps to effectively implement the policy. An implementation strategy plan, mitigation action plan, and detailed transportation plans have been developed. Other activities include foreign research reactor assessments, and the determination of shipment priorities and schedules. The first shipment under the acceptance policy was received into the United States in September 1996. A second shipment was received from Canada in December 1996. The next shipment of foreign research reactor spent nuclear fuel is expected from Europe in early March 1997. The primary challenge for DOE is to continue to transport this material in a consistent, cost-effective manner over the 13-year duration of the program. This article covers the following topics: background; acceptance policy; implementation of the acceptance policy; next steps/closing. 6 figs

  12. Alternative Renewable Biomass Tracking Program Document under the Renewable Fuel Standard Program

    Science.gov (United States)

    The approval letters that EPA approves a company's biomass tracking program meeting all the requirements outlined in 40 CFR part 80.1454, including elements determined necessary to achieve the level of quality assurance required under the regulation list.

  13. Development of program for evaluating the temperature of Zr-U metallic fuel rod

    International Nuclear Information System (INIS)

    Chun, J. S.; Lee, B. H.; Ku, Y. H.; Oh, J. Y.; Im, J. S.; Sohn, D. S.

    2003-01-01

    A code for evaluating the temperature of Zr-U metallic rod has been developed. Finite element (FE) method is adopted for the developed code sharing the user subroutines which has been prepared for the ABAQUS commercial FE code. The developed program for the Zr-U metallic fuel rod corresponds to a nonlinear transient heat transfer problem, and uses a sparse matrix solver for FE equations during iterations at every time step. The verifications of the developed program were conducted using the ABAQUS code. Steady state and transient problems were analyzed for 1/8 rod model due to the symmetry of the fuel rod and full model. From the evaluation of temperature for the 1/8 rod model at steady state, maximal error of 0.18 % was present relative to the ABAQUS result. Analysis for the transient problem using the fuel rod model resulted in the same as the variation of centerline temperature from the ABAQUS code during a hypothetical power transient. The distribution of heat flux for the entire cross section and surface was almost identical for the two codes

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

    International Nuclear Information System (INIS)

    Maraman, W.J.

    1980-02-01

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

  15. KUEBEL. A Fortran program for computation of cooling-agent-distribution within reactor fuel-elements

    International Nuclear Information System (INIS)

    Inhoven, H.

    1984-12-01

    KUEBEL is a Fortran-program for computation of cooling-agent-distribution within reactor fuel-elements or -zones of theirs. They may be assembled of max. 40 cooling-channels with laminar up to turbulent type of flow (respecting Reynolds' coefficients up to 2.0E+06) at equal pressure loss. Flow-velocity, dynamic flow-, contraction- and friction-losses will be calculated for each channel and for the total zone. Other computations will present mean heat-up of cooling-agent, mean outlet-temperature of the core, boiling-temperature and absolute pressure at flow-outlet. All characteristic coolant-values, including the factor of safety for flow-instability of the most-loaded cooling gap are computed by 'KUEBEL' too. Absolute pressure at flow-outlet or is-factor may be defined as dependent or independent variables of the program alternatively. In latter case 3 variations of solution will be available: Adapted flow of cooling-agent, inlet-temperature of the core and thermal power. All calculations can be done alternatively with variation of parameters: flow of cooling-agent, inlet-temperature of the core and thermal power, which are managed by the program itself. 'KUEBEL' is able to distinguish light- and heavy-water coolant, flow-direction of coolant and fuel elements with parallel, rectangular, respectively concentric, cylindrical shape of their gaps. Required material specifics are generated by the program. Segments of fuel elements or constructively unconnected gaps can also be computed by means of interposition of S.C. 'phantom channels'. (orig.) [de

  16. Liquefied gaseous fuels safety and environmental control assessment program: third status report

    Energy Technology Data Exchange (ETDEWEB)

    1982-03-01

    This Status Report contains contributions from all contractors currently participating in the DOE Liquefied Gaseous Fuels (LG) Safety and Environmental Control Assessment Program and is presented in two principal sections. Section I is an Executive Summary of work done by all program participants. Section II is a presentation of fourteen individual reports (A through N) on specific LGF Program activities. The emphasis of Section II is on research conducted by Lawrence Livermore National Laboratory (Reports A through M). Report N, an annotated bibliography of literature related to LNG safety and environmental control, was prepared by Pacific Northwest Laboratory (PNL) as part of its LGF Safety Studies Project. Other organizations who contributed to this Status Report are Aerojet Energy Conversion Company; Applied Technology Corporation; Arthur D. Little, Incorporated; C/sub v/ International, Incorporated; Institute of Gas Technology; and Massachusetts Institute of Technology. Separate abstracts have been prepared for Reports A through N for inclusion in the Energy Data Base.

  17. 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)

  18. 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)

  19. Development of Methodology for Spent Fuel Pool Severe Accident Analysis Using MELCOR Program

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Won-Tae; Shin, Jae-Uk [RETech. Co. LTD., Yongin (Korea, Republic of); Ahn, Kwang-Il [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    The general reason why SFP severe accident analysis has to be considered is that there is a potential great risk due to the huge number of fuel assemblies and no containment in a SFP building. In most cases, the SFP building is vulnerable to external damage or attack. In contrary, low decay heat of fuel assemblies may make the accident processes slow compared to the accident in reactor core because of a great deal of water. In short, its severity of consequence cannot exclude the consideration of SFP risk management. The U.S. Nuclear Regulatory Commission has performed the consequence studies of postulated spent fuel pool accident. The Fukushima-Daiichi accident has accelerated the needs for the consequence studies of postulated spent fuel pool accidents, causing the nuclear industry and regulatory bodies to reexamine several assumptions concerning beyond-design basis events such as a station blackout. The tsunami brought about the loss of coolant accident, leading to the explosion of hydrogen in the SFP building. Analyses of SFP accident processes in the case of a loss of coolant with no heat removal have studied. Few studies however have focused on a long term process of SFP severe accident under no mitigation action such as a water makeup to SFP. USNRC and OECD have co-worked to examine the behavior of PWR fuel assemblies under severe accident conditions in a spent fuel rack. In support of the investigation, several new features of MELCOR model have been added to simulate both BWR fuel assembly and PWR 17 x 17 assembly in a spent fuel pool rack undergoing severe accident conditions. The purpose of the study in this paper is to develop a methodology of the long-term analysis for the plant level SFP severe accident by using the new-featured MELCOR program in the OPR-1000 Nuclear Power Plant. The study is to investigate the ability of MELCOR in predicting an entire process of SFP severe accident phenomena including the molten corium and concrete reaction. The

  20. Idaho Chemical Processing Plant Spent Fuel and Waste Management Technology Development Program Plan

    International Nuclear Information System (INIS)

    1993-09-01

    The Department of Energy (DOE) has received spent nuclear fuel (SNF) at the Idaho Chemical Processing Plant (ICPP) for interim storage and reprocessing since 1953. Reprocessing of SNF has resulted in an existing inventory of 1.5 million gallons of radioactive sodium-bearing liquid waste and 3800 cubic meters (m 3 ) of calcine, in addition to the 768 metric tons (MT) of SNF and various other fuel materials in inventory. To date, the major activity of the ICPP has been the reprocessing of SNF to recover fissile uranium; however, recent changes in world events have diminished the demand to recover and recycle this material. As a result, DOE has discontinued reprocessing SNF for uranium recovery, making the need to properly manage and dispose of these and future materials a high priority. In accordance with the Nuclear Waste Policy Act (NWPA) of 1982, as amended, disposal of SNF and high-level waste (HLW) is planned for a geological repository. Preparation of SNF, HLW, and other radioactive wastes for disposal may include mechanical, physical, and/or chemical processes. This plan outlines the program strategy of the ICPP Spent Fuel and Waste Management Technology Development Program (SF ampersand WMTDP) to develop and demonstrate the technology required to ensure that SNF and radioactive waste will properly stored and prepared for final disposal. Program elements in support of acceptable interim storage and waste minimization include: developing and implementing improved radioactive waste treatment technologies; identifying and implementing enhanced decontamination and decommissioning techniques; developing radioactive scrap metal (RSM) recycle capabilities; and developing and implementing improved technologies for the interim storage of SNF

  1. Development of a Computer Program for the Integrated Control of the Fuel Homogeneity Measurement System

    International Nuclear Information System (INIS)

    Shin, H. S.; Jang, J. W.; Lee, Y. H.; Oh, S. J.; Park, H. D.; Kim, C. K.

    2005-11-01

    The computer program is developed based on Visual C++, which is equipped with a user-friendly interface of the input/output(I/O) and a display function for the measuring conditions. This program consists of three parts which are the port communication, PLC(Programmable Logic Controller) and the MCA(Multi Channel Analyzer) control parts. The communication type between the CPU of the PLC module box and the computer is selected as be the Rs-232 asynchronous type and the thread method is adapted in the development of the first part of the program. The PLC-related program has been developed so that the data communication between the PLC CPU and the computer could be harmonized with the unique commands which have already been defined in the PLC. The measuring space and time intervals, the start and end ROI(region of interest) values, and the allowable error limitation are input at each measurement in this program. Finally the controlling MCA program has been developed by using Canberra's programming library which contains several files including the head files in which the variable and the function of C++ are declared according to the MCA function. The performance test has been carried out through an application of the developed computer program to the homogeneity measurement system. The gamma counts at 28 measuring points of a fuel rod of 700 mm in length are measured for 50 sec at each point. It was revealed that the measurement results are better than the previous ones in respects of the measurement accuracy and a measurement time saving could be achieved. It was concluded that the gamma measurement system can be improved through equipping it with the developed control program

  2. Development of very-high-density fuels by the RERTR program

    International Nuclear Information System (INIS)

    Snelgrove, J.L.; Hofman, G.L.; Trybus, C.L.; Wiencek, T.C.

    1996-01-01

    The RERTR program has recently begun an aggressive effort to develop dispersion fuels for research and test reactors with uranium densities of 8 to 9 g U/cm 3 , based on the use of γ-stabilized uranium alloys. Fabrication development teams and facilities are being put into place, and preparations for the first irradiation test are in progress. The first screening irradiations are expected to begin in late April 1997 and the first results should be available by the end of 1997. Discussions with potential international partners in fabrication development and irradiation testing have begun

  3. AUTOSORO: A fuel management study program for Ontario Hydro CANDU reactors

    International Nuclear Information System (INIS)

    Wilk, L.

    1988-01-01

    A computer program, AUTOSORO, has been developed to automatically simulate an Ontario Hydro CANDU reactor core for any time duration according to user-defined on-power refuelling criteria. It is a three-dimensional two-group diffusion code coupled to refuelling decision logic at three screening levels: burnup, coupled neighbor, full-core. A central feature is a projected local-iteration scheme for predicting fuelling-induced local neutron flux changes. Comparisons of AUTOSORO results with actual histories demonstrate that it will be an excellent productivity tool for future in-core fuel management studies, reducing several man-months of effort to several man-hours

  4. DOE program for the management of radioactive waste and spent reactor fuel

    International Nuclear Information System (INIS)

    Cooley, C.R.

    1978-01-01

    The development of nuclear energy is seen by the Administration and the Department of Energy (DOE) as one of the important sources of energy for the country. Nuclear energy now provides a major fraction of the electrical power generation in some parts of the United States. In the northwest, with a wealth of hydroelectric power, nuclear power is expected to provide an increasing share of the total electrical energy. However, a great deal of public concern is being expressed about waste management associated with nuclear power. On current Research and Development programs, this paper considers several of the key activities which include work on disposal of spent fuel. 2 refs

  5. Twelfth symposium on biotechnology for fuels and chemicals: Program and abstracts

    International Nuclear Information System (INIS)

    Scheitlin, F.M.

    1990-01-01

    This report is the program and abstracts of the twelfth symposium on biotechnology for fuels and chemicals, held on May 7--11, 1990, at Gatlinburg, Tennessee. The symposium, sponsored by the Department of Energy, Oak Ridge National Laboratory, Solar Energy Research Institute, Badger Engineers, Inc., Gas Research Institute, and American Chemical Society, consists of five sessions: Session 1, thermal, chemical, and biological processing; Session 2 and 3, applied biological research; Session 4, bioengineering research; and Session 5, biotechnology, bioengineering, and the solution of environmental problems. It also consists of a poster session of the same five subject categories

  6. Twelfth symposium on biotechnology for fuels and chemicals: Program and abstracts

    Energy Technology Data Exchange (ETDEWEB)

    Scheitlin, F.M. (ed.)

    1990-01-01

    This report is the program and abstracts of the twelfth symposium on biotechnology for fuels and chemicals, held on May 7--11, 1990, at Gatlinburg, Tennessee. The symposium, sponsored by the Department of Energy, Oak Ridge National Laboratory, Solar Energy Research Institute, Badger Engineers, Inc., Gas Research Institute, and American Chemical Society, consists of five sessions: Session 1, thermal, chemical, and biological processing; Session 2 and 3, applied biological research; Session 4, bioengineering research; and Session 5, biotechnology, bioengineering, and the solution of environmental problems. It also consists of a poster session of the same five subject categories.

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

    Energy Technology Data Exchange (ETDEWEB)

    1993-05-15

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

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

    International Nuclear Information System (INIS)

    1990-10-01

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

  9. General-purpose heat source project and space nuclear safety fuels program. Progress report, February 1980

    International Nuclear Information System (INIS)

    Maraman, W.J.

    1980-05-01

    This formal monthly report covers the studies related to the use of 238 PuO 2 in radioisotopic power systems carried out for the Advanced Nuclear Systems and Projects Division of the Los Alamos Scientific Laboratory. The two programs involved are: General-Purpose Heat Source Development and Space Nuclear Safety and Fuels. Most of the studies discussed here are of a continuing nature. Results and conclusions described may change as the work continues. Published reference to the results cited in this report should not be made without the explicit permission of the person in charge of the work

  10. 77 FR 13009 - Regulation of Fuels and Fuel Additives: Identification of Additional Qualifying Renewable Fuel...

    Science.gov (United States)

    2012-03-05

    ... Regulation of Fuels and Fuel Additives: Identification of Additional Qualifying Renewable Fuel Pathways Under the Renewable Fuel Standard Program AGENCY: Environmental Protection Agency (EPA). ACTION: Withdrawal... Renewable Fuel Standard program regulations. Because EPA received adverse comment, we are withdrawing the...

  11. New approach to the optimization of nuclear fuel cycle - application of the goal programming and the AHP

    International Nuclear Information System (INIS)

    Kim, Poong Oh

    1998-02-01

    The front-end fuel cycle from mining to enrichment is in the maturity. Unlike the front-end fuel cycle, there are several pathways in the back-end fuel cycle. in this study five fuel cycle scenarios derived from a unique position in Korea of having a two-reactor programme of PWR and PHWR are proposed. In a selection of an optimal fuel cycle in a country, a number of attributes and factors that interact each other should be taken into account. Those factors to be considered in the study are categorized into two groups, one is tangible factor and the other is intangible factor. The major factors consist of minimizing fuel cycle cost, maximizing resource utilization, minimizing environmental impact and satisfying domestic and international politics. The long-term consequences of any decision for the back-end fuel cycle requires some sophisticated decision making tools. In this paper the Goal Programming method in combination with the Analytic Hierarchy Process (AHP) is applied in the decision making process. The Goal Programming is a very useful decision making tool to solve complex and multi-objective problems. The AHP, a method of solving complex decision problems with multiple attributes or objectives shows the strength in measuring the preferences of the attributes. In the study, the AHP is used for quantification of the intangible factors of which the evaluation is done by a team of nuclear experts. A model for fuel cycle selection is established in accordance with the logic of the Goal Programming. Also an interactive computer program is developed to obtain a solution for the most optimal fuel cycle in Korea

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

    Energy Technology Data Exchange (ETDEWEB)

    Satypal, S.

    2011-09-01

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

  13. 77 FR 22061 - FTA Section 5307 Urbanized Area Formula Program: Allocation of Funding Caps for Treating Fuel and...

    Science.gov (United States)

    2012-04-12

    ... Formula Program: Allocation of Funding Caps for Treating Fuel and Electric Utility Costs for Vehicle... closed on February 29. Based on the $100,000,000 cap on use of this provision, FTA has allocated funding caps to program recipients that responded to this announcement based on their relative share of the FY...

  14. Pathways to Commercial Success: Technologies and Products Supported by the Hydrogen, Fuel Cells and Infrastructure Technologies Program

    Energy Technology Data Exchange (ETDEWEB)

    none,

    2009-08-01

    This report documents the results of an effort to identify and characterize commercial and near-commercial (emerging) technologies and products that benefited from the support of the Hydrogen, Fuel Cells and Infrastructure Technologies Program and its predecessor programs within DOE's Office of Energy Efficiency and Renewable Energy.

  15. Characterization of Zircaloy-4 tubing procured for fuel cladding research programs

    International Nuclear Information System (INIS)

    Chapman, R.H.

    1976-01-01

    A quantity of Zircaloy-4 tubing [10.92 mm outside diameter by 0.635 mm wall thickness] was purchased specifically for use in a number of related fuel cladding research programs sponsored by the Division of Reactor Safety Research, Nuclear Regulatory Commission (NRC/RSR). Identical tubing (produced simultaneously and from the same ingot) was purchased concurrently by the Electric Power Research Institute (EPRI) for use in similar research programs sponsored by that organization. In this way, source variability and prior fabrication history were eliminated as parameters, thus permitting direct comparison (as far as as-received material properties are concerned) of experimental results from the different programs. The tubing is representative of current reactor technology. Consecutive serial numbers assigned to each tube identify the sequence of the individual tubes through the final tube wall reduction operation. The report presented documents the procurement activities, provides a convenient reference source of manufacturer's data and tubing distribution to the various users, and presents some preliminary characterization data. The latter have been obtained routinely in various research programs and are not complete. Although the number of analyses, tests, and/or examinations performed to date are insufficient to draw statistically valid conclusions with regard to material characterization, the data are expected to be representative of the as-received tubing. It is anticipated that additional characterizations will be performed and reported routinely by the various research programs that use the tubing

  16. International topical meeting on research reactor fuel management (RRFM) - United States Foreign Research Reactor (FRR) Spent Nuclear Fuel (SNF) acceptance program: 2007 update

    International Nuclear Information System (INIS)

    Messick, C.E.; Taylor, J.L.

    2007-01-01

    The Nuclear Weapons Non-proliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel, adopted by The United States Department of Energy (DOE), in consultation with the Department of State in May 1996, has been extended to expire May 12, 2016, providing an additional 10 years to return fuel to the U.S. This paper provides a brief update on the program, now transferred to the National Nuclear Security Administration (NNSA), and discusses program initiatives and future activities. The goal of the program continues to be recovery of nuclear materials (27 countries have participated so far, returning a total of 7620 spent nuclear fuel elements), which could otherwise be used in weapons, while assisting other countries to enjoy the benefits of nuclear technology. More than ever before, DOE and reactor operators need to work together to schedule shipments as soon as possible, to optimize shipment efficiency over the remaining years of the program. The NNSA is seeking feedback from research reactor operators to help us understand ways to include eligible reactor who have not yet participated in the program

  17. Program description for the program Fuel program sustainability July 1, 2011 through June 30, 2015; Programbeskrivning foer programmet Braensleprogrammet haallbarhet 1 juli 2011 till och med 30 juni 2015

    Energy Technology Data Exchange (ETDEWEB)

    2011-04-15

    The Fuel program sustainability is included as one of three programs in a cohesive commitment to increased, sustainable and efficient production and use of indigenous and renewable fuels that are implemented by the Swedish Energy Agency from July 1, 2011 to June 30, 2015. The program focuses on issues of environment and ecological sustainability of the production of biofuels, and systems and resource issues concerning the use. It does not include purely technical issues related to biofuels production or processes in which fuel is converted to heat, electricity or fuel. Questions about fuel supply and fuel processing / scale combustion are treated in the parallel running programs; the Fuel program supply and the Fuel program conversion. The four-year program will commence on July 1, 2011 and will run until June 30, 2015 and has an annual budget of total SEK 16 millions. Ambiguities in the environmental area may through various regulations mean various forms of barriers on the biofuel market. The Fuel program sustainability aims to sort out such ambiguities and, if possible, eliminate such obstacles, identify solutions and develop opportunities. The availability of biofuels and croplands is limited relative to needs. Thus the program also aims to describe the resource efficiency and climate benefits of current biofuel chains, and the possibility of using instruments to stimulate good practice. The program consists of sub-areas that partly overlap. - The sub-area Environment and sustainability focuses on how biofuel production will be designed to meet national and international environmental objectives and sustainability criteria. - The sub-area Biofuels and greenhouse gases describes different climate aspects related to production and exploitation of biofuels. - The sub-area System and market focuses on resource- and climate-efficient solutions in a system perspective, and how the bio-energy system can be affected by policy instruments

  18. Fuel and target programs for the transmutation at Phenix and other reactors

    International Nuclear Information System (INIS)

    Gaillard-Groleas, G.

    2002-01-01

    The fuels and targets program for transmutation, performed in the framework of the axis 1 of the December 1991 law about the researches on the management of long-lived radioactive wastes, is in perfect consistency with the transmutation scenario studies carried out in the same framework. These studies put forward the advantage of fast breeder reactors (FBR) in the incineration of minor actinides and long-lived fission products. The program includes exploratory and technological demonstration studies covering the different design options. It aims at enhancing our knowledge of the behaviour of materials under irradiation and at ensuring the mastery of processes. The goals of the different experiments foreseen at Phenix reactor are presented. The main goal is to supply a set of results allowing to precise the conditions of the technical feasibility of minor actinides and long-lived fission products incineration in FBRs. (J.S.)

  19. Phebus program main results and status for severe fuel damage studies

    International Nuclear Information System (INIS)

    Duco, J.; Reocreux, M.; Tattegrain, A.

    1986-06-01

    A large experimental in-pile program has been set up at the PHEBUS facility to investigate the actual behavior of .8 m active height, 25-rod PWR-type pressurized fresh fuel bundles under typical accident conditions. The program consists of four stages. Stage 1 was devoted to the adjustment of the operational procedure for stage 2. Stage 2 refers to the simulation of conservatively calculated L.B. LOCA 2 - peak transients. Stages 3/4 refer to four PWR severe accident scenarios retained for in-pile simulation at PHEBUS: a) a large break LOCA with injection failure; b) a small break LOCA associated with an injection failure; c) a prolonged total loss of the steam generator feedwater; and, d) a prolonged core uncovery a few days after reactor shutdown. The main PHEBUS stage 2 results are presented and finally interpreted

  20. United States Program on Spent Nuclear Fuel and High-Level Radioactive Waste Management

    International Nuclear Information System (INIS)

    Stewart, L.

    2004-01-01

    The President signed the Congressional Joint Resolution on July 23, 2002, that designated the Yucca Mountain site for a proposed geologic repository to dispose of the nation's spent nuclear fuel (SNF) and high-level radioactive waste (HLW). The United States (U.S.) Department of Energy's (DOE) Office of Civilian Radioactive Waste Management (OCRWM) is currently focusing its efforts on submitting a license application to the U.S. Nuclear Regulatory Commission (NRC) in December 2004 for construction of the proposed repository. The legislative framework underpinning the U.S. repository program is the basis for its continuity and success. The repository development program has significantly benefited from international collaborations with other nations in the Americas

  1. Status of the joint United States/Japanese criticality data development program for the Fast Breeder fuel cycle

    International Nuclear Information System (INIS)

    Takeda, H.; Kishimoto, Y.; Matsumoto, T.; Haire, M.J.

    1987-01-01

    An experimental program is described to accumulate the basic criticality data on uranium-plutonium aqueous solution systems, including pin and water systems. This experimental program, managed by the Consolidated Fuel Reprocessing Program, utilized existing critical experiment facilities. However, Japan at least for the near future, does not have a facility for conducting critical experiments where fissible solution can be treated and processed. Therefore, Japan has remoted the existing US experimental program as well as enlarged the program so that a close cooperative relationship may exist in this area. This paper is an update of a description of the program presented in September 1985. 9 refs., 5 figs., 1 tab

  2. Fuel rod D07/B15 from Ringhals 2 PWR: Source material for corrosion/leach tests in groundwater. Fuel rod/pellet characterization program. Pt. 1

    International Nuclear Information System (INIS)

    Forsyth, R.

    1987-03-01

    A joint SKB/STUDSVIK experimental program to determine the corrosion rates and to establish the corrosion mechanisms of spent UO 2 fuel in groundwater under both oxidizing and reducing conditions is in progress in the Hot Cell Laboratory of Studsvik Energiteknik AB. High burnup fuel of both BWR and PWR type are studied. Characterization of the spent fuel at both rod and pellet level is an important part of the experimental program. Experiments on PWR fuel have been concentrated so far on specimens from one rod, manufacturer's number 03688, which had occupied position B15 in assembly D07. This assembly had been irradiated for 5 cycles in the Ringhals 2 reactor between 1977 and 1983. The calculated assembly burnup was 41.3 MWd/kg U. The present report is a collection of separate reports describing those items in the characterization program which have been performed so far. No overall summary of the experimental results is given here, and the report should be viewed as a collection of reference data. (orig.)

  3. Analysis of Core Physics Experiments on Irradiated BWR MOX Fuel in REBUS Program

    International Nuclear Information System (INIS)

    Yamamoto, Toru; Ando, Yoshihira; Hayashi, Yamato

    2008-01-01

    As part of analyses of experimental data of a critical core containing a irradiated BWR MOX test bundle in the REBUS program, depletion calculations was performed for the BWR MOX fuel assemblies from that the MOX test rods were selected by using a general purpose neutronics code system SRAC. The core analyses were carried out using SRAC and a continuous energy Monte Carlo code MVP. The calculated k eff s were compared with those of the core containing a fresh MOX fuel bundle in the program. The SRAC-diffusion calculation underestimates k eff s of the both cores by 1.0 to 1.3 %dk and the k eff s of MVP are 1.001. The difference in k eff between the irradiated BWR MOX test bundle core and the fresh MOX one is 0.4 %dk in the SRAC-diffusion calculation and 0.0 %dk in the MVP calculation. The calculated fission rate distributions are in good agreement with the measurement in the SRAC-diffusion and MVP calculations. The calculated neutron flux distributions are also in good agreement with the measurement. The calculated burnup reactivity in the both calculations well reproduce the measurements. (authors)

  4. Fuzzy linear programming based optimal fuel scheduling incorporating blending/transloading facilities

    Energy Technology Data Exchange (ETDEWEB)

    Djukanovic, M.; Babic, B.; Milosevic, B. [Electrical Engineering Inst. Nikola Tesla, Belgrade (Yugoslavia); Sobajic, D.J. [EPRI, Palo Alto, CA (United States). Power System Control; Pao, Y.H. [Case Western Reserve Univ., Cleveland, OH (United States)]|[AI WARE, Inc., Cleveland, OH (United States)

    1996-05-01

    In this paper the blending/transloading facilities are modeled using an interactive fuzzy linear programming (FLP), in order to allow the decision-maker to solve the problem of uncertainty of input information within the fuel scheduling optimization. An interactive decision-making process is formulated in which decision-maker can learn to recognize good solutions by considering all possibilities of fuzziness. The application of the fuzzy formulation is accompanied by a careful examination of the definition of fuzziness, appropriateness of the membership function and interpretation of results. The proposed concept provides a decision support system with integration-oriented features, whereby the decision-maker can learn to recognize the relative importance of factors in the specific domain of optimal fuel scheduling (OFS) problem. The formulation of a fuzzy linear programming problem to obtain a reasonable nonfuzzy solution under consideration of the ambiguity of parameters, represented by fuzzy numbers, is introduced. An additional advantage of the FLP formulation is its ability to deal with multi-objective problems.

  5. Liquefied Gaseous Fuels Safety and Environmental Control Assessment Program: second status report

    Energy Technology Data Exchange (ETDEWEB)

    1980-10-01

    The Assistant Secretary for Environment has responsibility for identifying, characterizing, and ameliorating the environmental, health, and safety issues and public concerns associated with commercial operation of specific energy systems. The need for developing a safety and environmental control assessment for liquefied gaseous fuels was identified by the Environmental and Safety Engineering Division as a result of discussions with various governmental, industry, and academic persons having expertise with respect to the particular materials involved: liquefied natural gas, liquefied petroleum gas, hydrogen, and anhydrous ammonia. This document is arranged in three volumes and reports on progress in the Liquefied Gaseous Fuels (LGF) Safety and Environmental Control Assessment Program made in Fiscal Year (FY)-1979 and early FY-1980. Volume 1 (Executive Summary) describes the background, purpose and organization of the LGF Program and contains summaries of the 25 reports presented in Volumes 2 and 3. Annotated bibliographies on Liquefied Natural Gas (LNG) Safety and Environmental Control Research and on Fire Safety and Hazards of Liquefied Petroleum Gas (LPG) are included in Volume 1.

  6. The Advanced Petroleum-Based Fuels Program DECSE and APBF Overview

    International Nuclear Information System (INIS)

    None

    2000-01-01

    The following topics are summarized: Role of fuel blends in controlling engine-out emissions; Effect of fuels and lubricants on emission control devices; and Effect of fuels and lubricants on vehicle emissions and operations

  7. The Advanced Petroleum-Based Fuels Program DECSE and APBF Overview

    Energy Technology Data Exchange (ETDEWEB)

    None

    2000-04-11

    The following topics are summarized: Role of fuel blends in controlling engine-out emissions; Effect of fuels and lubricants on emission control devices; and Effect of fuels and lubricants on vehicle emissions and operations.

  8. Unreviewed safety question evaluation of 100K East and 100K West in-basin fuel characterization program activities

    International Nuclear Information System (INIS)

    Alwardt, L.D.

    1995-01-01

    The purpose of this report is to provide the basis for answers to an Unreviewed Safety Question (USQ) safety evaluation of the 105K East (KE) and 105K West (KW) in-basin activities associated with the fuel characterization program as described in the characterization shipping plan. The significant activities that are common to both 105 KE and 105 KW basins are the movement of canisters from their main basin storage locations (or potentially from the 105 KE Tech View Pit if a dump table is available) to the south loadout pit transfer channel, hydrogen generation testing in the single element fuel container, loading the single element fuel container into the shipping cask, loading of the shipping cask onto a flat-bed trailer, return of the test fuel elements or element pieces from the 327 facility, placement of the fuel elements back into Mark 2 canisters, and placement of the canisters in the main storage basin. Decapping of canisters in the south loadout pit transfer channel and re-encapsulation of canisters are activities specific to the 105 KW basin. The scope of this safety evaluation includes only those characterization fuel shipment activities performed in the 105 KE and 105 KW fuel storage basin structures up to installation of the overpack. The packaging safety evaluation report governs the shipment of the fuel elements. The K Basins Plant Review Committee has determined that the in-basin activities associated with the fuel characterization program fuel shipments are bounded by the current safety envelop and do not constitute an unreviewed safety question. This determination is documented on Attachment 1

  9. Uranium from Seawater Program Review; Fuel Resources Uranium from Seawater Program DOE Office of Nuclear Energy

    Energy Technology Data Exchange (ETDEWEB)

    None

    2013-07-01

    For nuclear energy to remain sustainable in the United States, economically viable sources of uranium beyond terrestrial ores must be developed. The goal of this program is to develop advanced adsorbents that can extract uranium from seawater at twice the capacity of the best adsorbent developed by researchers at the Japan Atomic Energy Agency (JAEA), 1.5 mg U/g adsorbent. A multidisciplinary team from Oak Ridge National Laboratory, Lawrence Berkeley National Laboratory, Pacific Northwest National Laboratory, and the University of Texas at Austin was assembled to address this challenging problem. Polymeric adsorbents, based on the radiation grafting of acrylonitrile and methacrylic acid onto high surface-area polyethylene fibers followed by conversion of the nitriles to amidoximes, have been developed. These poly(acrylamidoxime-co-methacrylic acid) fibers showed uranium adsorption capacities for the extraction of uranium from seawater that exceed 3 mg U/g adsorbent in testing at the Pacific Northwest National Laboratory Marine Sciences Laboratory. The essence of this novel technology lies in the unique high surface-area trunk material that considerably increases the grafting yield of functional groups without compromising its mechanical properties. This technology received an R&D100 Award in 2012. In addition, high surface area nanomaterial adsorbents are under development with the goal of increasing uranium adsorption capacity by taking advantage of the high surface areas and tunable porosity of carbon-based nanomaterials. Simultaneously, de novo structure-based computational design methods are being used to design more selective and stable ligands and the most promising candidates are being synthesized, tested and evaluated for incorporation onto a support matrix. Fundamental thermodynamic and kinetic studies are being carried out to improve the adsorption efficiency, the selectivity of uranium over other metals, and the stability of the adsorbents. Understanding

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-10-18

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

  11. User's guide to REVERT. A CDC 7600 program for converting Spent Fuel Test - Climax data to engineering units, with corrections

    International Nuclear Information System (INIS)

    Hage, G.

    1984-10-01

    A CDC 7600 computer program, REVERT, can revise Spent Fuel Test - Climax data files using one of several algorithms, depending on the type of data. The algorithms use coefficients from a separate file organized by data type identifiers. REVERT can also make that file of coefficients, using data from tapes made by Hewlett-Packard equipment employed for data acquisition on the spent Fuel Test - Climax at NTS. 12 references

  12. Investigation on structural analysis computer program of spent nuclear fuel shipping cask

    International Nuclear Information System (INIS)

    Yagawa, Ganki; Ikushima, Takeshi.

    1987-10-01

    This report describes the results done by the Sub-Committee of Research Cooperation Committee (RC-62) of the Japan Society of Mechanical Engineers under the trust of the Japan Atomic Energy Research Institute. The principal fulfilments and accomplishments are summarized as follows: (1) Regarding the survey of structural analysis methods of spent fuel shipping cask, several documents, which explain the features and applications of the exclusive computer programs for impact analysis on the basis of 2 or 3 dimensional finite element or difference methods such as HONDO, STEALTH and DYNA-3D, were reviewed. (2) In comparative evaluation of the existing computer programs, the common benchmark test problems for 9 m vertical drop impact of the axisymmetric lead cylinder with and without stainless steel clads were adopted where the calculational evaluations for taking into account the strain rate effect were carried out. (3) Evaluation of impact analysis algorithm of computer programs were conducted and the requirements for computer programs to be developed in future and an index for further studies have been clarified. (author)

  13. Investigation on structural analysis computer program of spent nuclear fuel shipping cask, (2)

    International Nuclear Information System (INIS)

    Yagawa, Ganki; Ikushima, Takeshi.

    1987-10-01

    This report describes the results (II) done by the Sub-Committee of Research Cooperation Committee (RC-62) of the Japan Society of Mechanical Engineers under the trust of the Japan Atomic Energy Research Institute. The principal fulfilments and accomplishments are summarized as follows: (1) Regarding the survey of structural analysis methods of spent fuel shipping cask, several documents, which explain the features and applications of the exclusive computer programs for impact analysis on the basis of 2 or 3 dimensional finite element or difference methods, were reviewed. (2) In comparative evaluation of the existing computer programs, the common benchmark test problems for drop impact of the axisymmetric cylinder and plate were adopted the calculational evaluations for taking into account the strain rate effect of material properties, effect of artificial viscosity and effect of time integration step size were carried out. (3) Evaluation of impact analysis algorithm of computer programs were conducted and the requirements for computer programs to be developed in future and an index for further studies have been clarified. (author)

  14. 8 x 8 fuel surveillance program at Monticello site - end of Cycle 6: fourth post-irradiation inspection, October 1978

    International Nuclear Information System (INIS)

    Skarshaug, N.H.

    1980-09-01

    A fuel surveillance program for a lead 8 x 8 reload fuel assembly was implemented at the Monticello Nuclear Power Station in May 1974 prior to Reactor Cycle 3. Inspection results of the fourth post-irradiation inspection performed on this surveillance fuel assembly in October 1978 at EOC 6, after a bundle average exposure of 25,900 MWd/MT, are presented. The measurement techniques, results obtained and comparisons to previous measurements are discussed. The bundle and individual rods examined exhibited characteristics of normal operation and were approved for continued irradiation during Monticello operating Cycle 7

  15. Program requirements to determine and relate fuel damage and failure thresholds to anticipated conditions in pressurized water reactors

    International Nuclear Information System (INIS)

    Loyd, R.F.; Croucher, D.W.

    1980-03-01

    Anticipated transients, licensing criteria, and damage mechanisms for PWR fuel rods are reviewed. Potential mechanistic fuel rod damage limits for PWRs are discussed. An expermental program to be conducted out-of-pile and in the Engineering Test Reactor (ETR) to generate a safety data base to define mechanistic fuel damage and failure thresholds and to relate these thresholds to the thermal-hydraulic and power conditions in a PWR is proposed. The requirements for performing the tests are outlined. Analytical support requirements are defined

  16. High-uranium-loaded U3O8-Al fuel element development program [contributed by N.M. Martin, ORNL

    International Nuclear Information System (INIS)

    Martin, M.M.

    1993-01-01

    The High-Uranium-Loaded U 3 O 8 -Al Fuel Element Development Program supports Argonne National Laboratory efforts to develop high-uranium-density research and test reactor fuel to accommodate use of low-uranium enrichment. The goal is to fuel most research and test reactors with uranium of less than 20% enrichment for the purpose of lowering the potential for diversion of highly-enriched material for nonpeaceful usages. The specific objective of the program is to develop the technological and engineering data base for U 3 O 8 -Al plate-type fuel elements of maximal uranium content to the point of vendor qualification for full scale fabrication on a production basis. A program and management plan that details the organization, supporting objectives, schedule, and budget is in place and preparation for fuel and irradiation studies is under way. The current programming envisions a program of about four years duration for an estimated cost of about two million dollars. During the decades of the fifties and sixties, developments at Oak Ridge National Laboratory led to the use of U 3 O 8 -Al plate-type fuel elements in the High Flux Isotope Reactor, Oak Ridge Research Reactor, Puerto Rico Nuclear Center Reactor, and the High Flux Beam Reactor. Most of the developmental information however applies only up to a uranium concentration of about 55 wt % (about 35 vol % U 3 O 8 ). The technical issues that must be addressed to further increase the uranium loading beyond 55 wt % involve plate fabrication phenomena of voids and dogboning, fuel behavior under long irradiation, and potential for the thermite reaction between U 3 O 8 and aluminum. (author)

  17. Idaho Chemical Processing Plant spent fuel and waste management technology development program plan: 1994 Update

    International Nuclear Information System (INIS)

    1994-09-01

    The Department of Energy has received spent nuclear fuel (SNF) at the Idaho Chemical Processing Plant (ICPP) for interim storage since 1951 and reprocessing since 1953. Until April 1992, the major activity of the ICPP was the reprocessing of SNF to recover fissile uranium and the management of the resulting high-level wastes (HLW). In 1992, DOE chose to discontinue reprocessing SNF for uranium recovery and shifted its focus toward the continued safe management and disposition of SNF and radioactive wastes accumulated through reprocessing activities. Currently, 1.8 million gallons of radioactive liquid wastes (1.5 million gallons of radioactive sodium-bearing liquid wastes and 0.3 million gallons of high-level liquid waste), 3,800 cubic meters of calcine waste, and 289 metric tons heavy metal of SNF are in inventory at the ICPP. Disposal of SNF and high-level waste (HLW) is planned for a repository. Preparation of SNF, HLW, and other radioactive wastes for disposal may include mechanical, physical, and/or chemical processes. This plan outlines the program strategy of the ICPP spent Fuel and Waste Management Technology Development Program (SF ampersand WMTDP) to develop and demonstrate the technology required to ensure that SNF and radioactive waste will be properly stored and prepared for final disposal in accordance with regulatory drivers. This Plan presents a brief summary of each of the major elements of the SF ampersand WMTDP; identifies key program assumptions and their bases; and outlines the key activities and decisions that must be completed to identify, develop, demonstrate, and implement a process(es) that will properly prepare the SNF and radioactive wastes stored at the ICPP for safe and efficient interim storage and final disposal

  18. Run-Beyond-Cladding-Breach (RBCB) test results for the Integral Fast Reactor (IFR) metallic fuels program

    International Nuclear Information System (INIS)

    Batte, G.L.; Hoffman, G.L.

    1990-01-01

    In 1984 Argonne National Laboratory (ANL) began an aggressive program of research and development based on the concept of a closed system for fast-reactor power generation and on-site fuel reprocessing, exclusively designed around the use of metallic fuel. This is the Integral Fast Reactor (IFR). Although the Experimental Breeder Reactor-II (EBR-II) has used metallic fuel since its creation 25 yeas ago, in 1985 ANL began a study of the characteristics and behavior of an advanced-design metallic fuel based on uranium-zirconium (U-Zr) and uranium-plutonium-zirconium (U-Pu-Zr) alloys. During the past five years several areas were addressed concerning the performance of this fuel system. In all instances of testing the metallic fuel has demonstrated its ability to perform reliably to high burnups under varying design conditions. This paper will present one area of testing which concerns the fuel system's performance under breach conditions. It is the purpose of this paper to document the observed post-breach behavior of this advanced-design metallic fuel. 2 figs., 1 tab

  19. Vehicle Technologies and Fuel Cell Technologies Office Research and Development Programs: Prospective Benefits Assessment Report for Fiscal Year 2018

    Energy Technology Data Exchange (ETDEWEB)

    Stephens, T. S. [Argonne National Lab. (ANL), Argonne, IL (United States); Birky, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Gohlke, David [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-11-01

    Under a diverse set of programs, the Vehicle Technologies and Fuel Cell Technologies Offices of the U.S. Department of Energy’s Office of Energy Efficiency and Renewable Energy invest in early-stage research of advanced batteries and electrification, engines and fuels, materials, and energy-efficient mobility systems; hydrogen production, delivery, and storage; and fuel cell technologies. This report documents the estimated benefits of successful development and implementation of advanced vehicle technologies. It presents a comparison of a scenario with completely successful implementation of Vehicle Technologies Office (VTO) and Fuel Cell Technologies Office (FCTO) technologies (the Program Success case) to a future in which there is no contribution after Fiscal Year 2017 by the VTO or FCTO to these technologies (the No Program case). Benefits were attributed to individual program technology areas, which included FCTO research and development and the VTO programs of electrification, advanced combustion engines and fuels, and materials technology. Projections for the Program Success case indicate that by 2035, the average fuel economy of on-road, light-duty vehicle stock could be 24% to 30% higher than in the No Program case, while fuel economy for on-road medium- and heavy-duty vehicle stock could be as much as 13% higher. The resulting petroleum savings in 2035 were estimated to be as high as 1.9 million barrels of oil per day, and reductions in greenhouse gas emissions were estimated to be as high as 320 million metric tons of carbon dioxide equivalent per year. Projections of light-duty vehicle adoption indicate that although advanced-technology vehicles may be somewhat more expensive to purchase, the fuel savings result in a net reduction of consumer cost. In 2035, reductions in annual fuel expenditures for vehicles (both light- and heavy-duty) are projected to range from $86 billion to $109 billion (2015$), while the projected increase in new vehicle

  20. Gas-Cooled Reactor Programs annual progress report for period ending December 31, 1973. [HTGR fuel reprocessing, fuel fabrication, fuel irradiation, core materials, and fission product distribution; GCFR fuel irradiation and steam generator modeling

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Coobs, J.H.; Lotts, A.L.

    1976-04-01

    Progress is summarized in studies relating to HTGR fuel reprocessing, refabrication, and recycle; HTGR fuel materials development and performance testing; HTGR PCRV development; HTGR materials investigations; HTGR fuel chemistry; HTGR safety studies; and GCFR irradiation experiments and steam generator modeling.

  1. Chemical Engineering Division fuel cycle programs. Progress report, January--March 1978

    Energy Technology Data Exchange (ETDEWEB)

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

    1979-04-01

    Fuel cycle studies reported for this period include studies of advanced solvent extraction techniques focussed on the development of centrifugal contactors for use in Purex processes. Miniature single-stage and eight-stage centrifugal contactors are being employed in studies of contactor performance and the kinetics of extraction. A 9-cm-ID centrifugal contactor has been completed, and fabrication drawings are being prepared for a plant-scale contactor. In other work, tricaprylmethyl-ammonium nitrate and di-n-amyl n-amylphosphonate are being evaluated as extractants in the Thorex process. Literature on the dispersion of liquids by explosions is being reviewed. A process was developed for extracting TBP degradation products from TBP-Na/sub 2/CO/sub 3/ scrub solutions while the actinides remain with the raffinate. In the program on pyrochemical and dry processing of nuclear fuel, the literature is being reviewed for acceptable materials for containment vessels, decladding methods are being evaluated, salt transport processes are being studied, a candidate flow sheet (based upon the Dow Aluminum Pyrometallurgical process) for reprocessing spent uranium metal fuel was prepared, work was begun on the use of molten salts for reprocessing actinide oxides, and the reprocessing of (Th,U)O/sub 2/ solid solution in a KCl-LiCl salt containing ThCl/sub 4/ and thorium chips was studied. Work on the encapsulation of solidified radioactive waste in a metal matrix includes study of (1) chemical interactions between simulated waste forms and matrix metals, (2) the leach rates of simulated encapsulated waste forms, and (3) the corrosion of candidate matrix metals and canister materials in brine solutions.Work to establish criteria for the handling of waste cladding hulls is continuing. The transport properties of nuclear waste in geologic media are being studied to estimate leaching of radionuclides from deep repositories by groundwater.

  2. Program description for the program Fuel program conversion 2011-07-01 - 2015-06-30; Programbeskrivning foer programmet Braensleprogrammet omvandling 2011-07-01 - 2015-06-30

    Energy Technology Data Exchange (ETDEWEB)

    2011-04-15

    The Fuel program conversion is included as one of three programs in a cohesive commitment to increased, sustainable and efficient production and use of indigenous and renewable fuels that are implemented by the Swedish Energy Agency during the period July 1, 2011 - June 30, 2015. The program focuses on increasing knowledge and development of technologies for processing of the fuel feedstock, and conversion to primarily heat and electricity in plants up to about 10 MW. Issues of particular concern is to strengthen competitiveness, efficiency and sustainability in the chain from raw material to end use, and increasing raw material base by developing knowledge and technologies for new materials that are current. The program includes basic and applied research, development and demonstration. Within the sub-area Processing efforts are prioritized which streamline and reduce costs in different types of processing operations. Processing here means not just technology for pelleting and briquetting, but also sub-processes that prepare, dry, handle or process raw materials in other ways. Within the sub-area Conversion to heat and electricity in the range <10 MW, the focus is principally, on one hand increasing the knowledge of the interaction between fuel and combustion plant, and on the other hand to implement results of research in collaboration with users and equipment suppliers. Furthermore, great importance is given to increasing the competitiveness of local heating in the transition from oil to renewable heat supply and the use of new fuel materials and fuels. Questions about the fuel supply from forestry and agriculture, including issues of sustainability, environment and resource use are treated in the two parallel running programs, the Fuel program supply and the Fuel program sustainability

  3. Canadian power reactor fuel

    International Nuclear Information System (INIS)

    Page, R.D.

    1976-03-01

    The following subjects are covered: the basic CANDU fuel design, the history of the bundle design, the significant differences between CANDU and LWR fuel, bundle manufacture, fissile and structural materials and coolants used in the CANDU fuel program, fuel and material behaviour, and performance under irradiation, fuel physics and management, booster rods and reactivity mechanisms, fuel procurement, organization and industry, and fuel costs. (author)

  4. Current status of spent fuel disposal program in Taiwan, Republic of China

    International Nuclear Information System (INIS)

    Soong, K.L.; Liu, S.J.

    1989-01-01

    In the year of 1988, Taiwan has completed a two-year preliminary study and began in late 1988 a second term for the final disposal of spent fuel program. The research conducted in the first phase was mainly concentrated on the reviews of international studies and domestic geological literatures, set-up of siting criteria, and pertinent engineering analysis of a repository. Taiwan is an area of complex and unstable geological structures, abundant ground water and high density of population. Favorable host rocks under consideration for permanent waste disposal are thick shales and mudstones, metamorphosed rocks of mesozoic basement, solidified quartzite and mesozoic granitic gneiss. The analysis of heat transfer for hypothetical case studies indicate that the temperature rises in the repository systems would be well below the proposed maximum admissible temperatures. The repository size would be less than 2 km 2 when appropriate period for surface storage of spent fuel is satisfied. A probabilistic risk analysis also demonstrates that the presence of faulting and other tectonic instabilities characterize Taiwan's disadvantages for geological disposal of nuclear waste

  5. Cost Effectiveness of a Weight Management Program Implemented in the Worksite: Translation of Fuel Your Life.

    Science.gov (United States)

    Corso, Phaedra S; Ingels, Justin B; Padilla, Heather M; Zuercher, Heather; DeJoy, David M; Vandenberg, Robert J; Wilson, Mark G

    2018-04-18

    Conduct a cost-effectiveness analysis of the Fuel Your Life (FYL) program dissemination. Employees were recruited from three workplaces randomly assigned to one of the conditions: telephone coaching, small group coaching, and self-study. Costs were collected prospectively during the efficacy trial. The main outcome measures of interest were weight loss and quality-adjusted life years (QALYs). The phone condition was most costly ($601-$589/employee) and the self-study condition was least costly ($145-$143/employee). For weight loss, delivering FYL through the small group condition was no more effective, yet more expensive, than the self-study delivery. For QALYs, the group delivery of FYL was in an acceptable cost-effectiveness range ($22,400/QALY) relative to self-study (95% CI: $10,600/QALY - dominated). Prevention programs require adaptation at the local level and significantly affect the cost, effectiveness, and cost effectiveness of the program.This is an open access article distributed under the terms of the Creative Commons Attribution-Non Commercial-No Derivatives License 4.0 (CCBY-NC-ND), where it is permissible to download and share the work provided it is properly cited. The work cannot be changed in any way or used commercially without permission from the journal. http://creativecommons.org/licenses/by-nc-nd/4.0.

  6. A data processing program for transient sodium boiling and fuel failure propagation tests, (2)

    International Nuclear Information System (INIS)

    Hasebe, Takeshi; Isozaki, Tadashi; Satoh, Akihiro; Yamaguchi, Katsuhisa; Haga, Kazuo.

    1983-01-01

    Transient Sodium Boiling Tests and Fuel Failure Propagation Tests are being conducted with the out-of-pile test facility, SIENA, in the Core Safety Section of O-arai Engineering Center. The experimental data are recorded using a digital data acquisition system controlled by a HP-1000E computer. The SICILIAN (Speedy Illustration Code for Inspection Line Anomaly) code was developed to obtain quick graphic outputs of data recorded in the magnetic tapes. The program is written in BASIC and Assembler languages and uses a data processing system composed of a desktop computer HP 9845B, a magnetic tape system, a magnetic disc and an eightcolor plotter. The SICILIAN code enables us to get graphic outputs soon after a run. These outputs are very helpful to inspect anomaly in the instrument circuit and to check the experimental conditions of coming runs. (author)

  7. Optimization of fuel-cell tram operation based on two dimension dynamic programming

    Science.gov (United States)

    Zhang, Wenbin; Lu, Xuecheng; Zhao, Jingsong; Li, Jianqiu

    2018-02-01

    This paper proposes an optimal control strategy based on the two-dimension dynamic programming (2DDP) algorithm targeting at minimizing operation energy consumption for a fuel-cell tram. The energy consumption model with the tram dynamics is firstly deduced. Optimal control problem are analyzed and the 2DDP strategy is applied to solve the problem. The optimal tram speed profiles are obtained for each interstation which consist of three stages: accelerate to the set speed with the maximum traction power, dynamically adjust to maintain a uniform speed and decelerate to zero speed with the maximum braking power at a suitable timing. The optimal control curves of all the interstations are connected with the parking time to form the optimal control method of the whole line. The optimized speed profiles are also simplified for drivers to follow.

  8. Comparison of structural computer programs used for the analysis of spent fuel shipping casks

    International Nuclear Information System (INIS)

    Friley, J.R.

    1984-09-01

    Several structural analysis computer programs were selected and used in analyses relevant to the hypothetical impact requirements for spent fuel shipping cask designs. The objective of the study was to evaluate the computer codes by performing a series of analyses and comparing results. The code evaluation efforts treated end and side impact situations only. As a result, the models were either one or two dimensional. Both clad lead and solid wall construction types were considered. For clad lead models, frictionless sliding between the lead and cladding was assumed. General agreement was achieved between the codes for problems involving non-clad models. For clad models, agreement between the codes was poor. This work was sponsored by the Department of Energy through the Transportation Technology Center at Sandia National Laboratories, Albuquerque, New Mexico. 16 references, 18 figures, 9 tables

  9. Nuclear fuel waste management program geotechnical studies of Eye-Dashwa Lakes research area rock properties

    International Nuclear Information System (INIS)

    Chernis, P.J.; Robertson, P.B.

    1992-05-01

    The Eye-Dashwa Lakes pluton near Atikokan Ontario has been used as a study area for the Canadian nuclear fuel waste management research program. The pluton consists predominately of granite. Fractures formed during cooling of the pluton were filled with a succession of different materials at different times. Measurements of a series of geophysical and geotechnical properties of rock samples are published here in this report, including especially microcrack and pore structures. An indication has been found that a larger proportion of the porosity of Whiteshell and Atikokan samples is contained in connecting pores, compared to other rocks. This may seem surprising in view of the finding that approximately 70% of the effective porosity of Atikokan samples is contained in pockets

  10. Characterization Program Management Plan for Hanford K Basin Spent Nuclear Fuel (SNF) (OCRWM)

    International Nuclear Information System (INIS)

    BAKER, R.B.; TRIMBLE, D.J.

    2000-01-01

    The management plan developed to characterize the K Basin spent nuclear fuel (SNF) and sludge was originally developed for Westinghouse Hanford Company and Pacific Northwest National Laboratory to work together on a program to provide characterization data to support removal, conditioning, and subsequent dry storage of the SNF stored at the Hanford K Basins. The plan also addressed necessary characterization for the removal, transport, and storage of the sludge from the Hanford K Basins. This plan was revised in 1999 (i.e., Revision 2) to incorporate actions necessary to respond to the deficiencies revealed as the result of Quality Assurance surveillances and audits in 1999 with respect to the fuel characterization activities. Revision 3 to this Program Management Plan responds to a Worker Assessment resolution determined in Fical Year 2000. This revision includes an update to current organizational structures and other revisions needed to keep this management plan consistent with the current project scope. The plan continues to address both the SNF and the sludge accumulated at K Basins. Most activities for the characterization of the SNF have been completed. Data validation, Office of Civilian Radioactive Waste Management (OCRWM) document reviews, and OCRWM data qualification are the remaining SNF characterization activities. The transport and storage of K Basin sludge are affected by recent path forward revisions. These revisions require additional laboratory analyses of the sludge to complete the acquisition of required supporting engineering data. Hence, this revision of the management plan provides the overall work control for these remaining SNF and sludge characterization activities given the current organizational structure of the SNF Project

  11. Development of the Simulation Program for the In-Vessel Fuel Handling System of Double Rotating Plug Type

    International Nuclear Information System (INIS)

    Kim, S. H.; Kim, J. B.

    2011-01-01

    In-vessel fuel handling machines are the main equipment of the in-vessel fuel handling system, which can move the core assembly inside the reactor vessel along with the rotating plug during refueling. The in vessel fuel handling machines for an advanced sodium cooled fast reactor(SFR) demonstration plant are composed of a direct lift machine(DM) and a fixed arm machine(FM). These machines should be able to access all areas above the reactor core by means of the rotating combination of double rotating plugs. Thus, in the in vessel fuel handling system of the double rotating plug type, it is necessary to decide the rotating plug size and evaluate the accessibility of in-vessel fuel handling machines in given core configuration. In this study, the simulation program based on LABVIEW which can effectively perform the arrangement design of the in vessel fuel handling system and simulate the rotating plug motion was developed. Fig. 1 shows the flow chart of the simulation program

  12. Public involvement: the critical path in siting controversial facilities. Proceedings of the Nuclear Energy Low-Level Waste Mangement Program conference

    International Nuclear Information System (INIS)

    1986-01-01

    The purpose of the conference was to: exchange information among those responsible for, or interested in, the development of new low-level waste disposal facilities; acquaint participants with past experiences of states and organizations in enfranchising the public in the siting of controversial facilities; and discuss various mechanisms and techniques for effectively involving the public in decision-making processes. The conference addressed four major topics: lessons from past experiences; mechanisms and techniques for public involvement, conflict resolution, and working constructively with the media. A series of presentations on each topic was followed by questions and discussion among presenters and conference participants. Several key points emerged as the conference proceeded

  13. Low NO sub x heavy fuel combustor concept program. Phase 1A: Combustion technology generation coal gas fuels

    Science.gov (United States)

    Sherlock, T. P.

    1982-01-01

    Combustion tests of two scaled burners using actual coal gas from a 25 ton/day fluidized bed coal gasifier are described. The two combustor configurations studied were a ceramic lined, staged rich/lean burner and an integral, all metal multiannual swirl burner (MASB). The tests were conducted over a range of temperature and pressures representative of current industrial combustion turbine inlet conditions. Tests on the rich lean burner were conducted at three levels of product gas heating values: 104, 197 and 254 btu/scf. Corresponding levels of NOx emissions were 5, 20 and 70 ppmv. Nitrogen was added to the fuel in the form of ammonia, and conversion efficiencies of fuel nitrogen to NOx were on the order of 4 percent to 12 percent, which is somewhat lower than the 14 percent to 18 percent conversion efficiency when src-2 liquid fuel was used. The MASB was tested only on medium btu gas (220 to 270 btu/scf), and produced approximately 80 ppmv NOx at rated engine conditions. Both burners operated similarly on actual coal gas and erbs fuel, and all heating values tested can be successfully burned in current machines.

  14. Comparison of national programs and regulations for the management of spent fuel and disposal of high-level waste in seven countries

    International Nuclear Information System (INIS)

    Numark, N.J.; Mattson, R.J.; Gaunt, J.

    1986-01-01

    This paper describes programs and regulatory requirements affecting the management of spent fuel and disposal of high-level radioactive waste in seven nations with large nuclear power programs. The comparison is intended to illustrate that the range of spent fuel management options is influenced by certain technical and political constraints. It begins by providing overall nuclear fuel cycle facts for each country, including nuclear generating capacities, rates of spent fuel discharge, and policies on spent fuel reprocessing. Spent fuel storage techniques and reprocessing activities are compared in light of constraints such as fuel type. Waste disposal investigations are described, including a summary of the status of regulatory developments affecting repository siting and disposal. A timeline is provided to illustrate the principle milestones in spent fuel management and waste disposal in each country. Finally, policies linking nuclear power licensing and development to nuclear waste management milestones and RandD progress are discussed

  15. Program ELM: A tool for rapid thermal-hydraulic analysis of solid-core nuclear rocket fuel elements

    International Nuclear Information System (INIS)

    Walton, J.T.

    1992-11-01

    This report reviews the state of the art of thermal-hydraulic analysis codes and presents a new code, Program ELM, for analysis of fuel elements. ELM is a concise computational tool for modeling the steady-state thermal-hydraulics of propellant flow through fuel element coolant channels in a nuclear thermal rocket reactor with axial coolant passages. The program was developed as a tool to swiftly evaluate various heat transfer coefficient and friction factor correlations generated for turbulent pipe flow with heat addition which have been used in previous programs. Thus, a consistent comparison of these correlations was performed, as well as a comparison with data from the NRX reactor experiments from the Nuclear Engine for Rocket Vehicle Applications (NERVA) project. This report describes the ELM Program algorithm, input/output, and validation efforts and provides a listing of the code

  16. Development of whole core thermal-hydraulic analysis program ACT. 4. Simplified fuel assembly model and parallelization by MPI

    International Nuclear Information System (INIS)

    Ohshima, Hiroyuki

    2001-10-01

    A whole core thermal-hydraulic analysis program ACT is being developed for the purpose of evaluating detailed in-core thermal hydraulic phenomena of fast reactors including the effect of the flow between wrapper-tube walls (inter-wrapper flow) under various reactor operation conditions. As appropriate boundary conditions in addition to a detailed modeling of the core are essential for accurate simulations of in-core thermal hydraulics, ACT consists of not only fuel assembly and inter-wrapper flow analysis modules but also a heat transport system analysis module that gives response of the plant dynamics to the core model. This report describes incorporation of a simplified model to the fuel assembly analysis module and program parallelization by a message passing method toward large-scale simulations. ACT has a fuel assembly analysis module which can simulate a whole fuel pin bundle in each fuel assembly of the core and, however, it may take much CPU time for a large-scale core simulation. Therefore, a simplified fuel assembly model that is thermal-hydraulically equivalent to the detailed one has been incorporated in order to save the simulation time and resources. This simplified model is applied to several parts of fuel assemblies in a core where the detailed simulation results are not required. With regard to the program parallelization, the calculation load and the data flow of ACT were analyzed and the optimum parallelization has been done including the improvement of the numerical simulation algorithm of ACT. Message Passing Interface (MPI) is applied to data communication between processes and synchronization in parallel calculations. Parallelized ACT was verified through a comparison simulation with the original one. In addition to the above works, input manuals of the core analysis module and the heat transport system analysis module have been prepared. (author)

  17. RISKIND: A computer program for calculating radiological consequences and health risks from transportation of spent nuclear fuel

    International Nuclear Information System (INIS)

    Yuan, Y.C.; Chen, S.Y.; LePoire, D.J.

    1993-02-01

    This report presents the technical details of RISIUND, a computer code designed to estimate potential radiological consequences and health risks to individuals and the collective population from exposures associated with the transportation of spent nuclear fuel. RISKIND is a user-friendly, semiinteractive program that can be run on an IBM or equivalent personal computer. The program language is FORTRAN-77. Several models are included in RISKIND that have been tailored to calculate the exposure to individuals under various incident-free and accident conditions. The incidentfree models assess exposures from both gamma and neutron radiation and can account for different cask designs. The accident models include accidental release, atmospheric transport, and the environmental pathways of radionuclides from spent fuels; these models also assess health risks to individuals and the collective population. The models are supported by databases that are specific to spent nuclear fuels and include a radionudide inventory and dose conversion factors

  18. RISKIND: A computer program for calculating radiological consequences and health risks from transportation of spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Yuan, Y.C. [Square Y, Orchard Park, NY (United States); Chen, S.Y.; LePoire, D.J. [Argonne National Lab., IL (United States). Environmental Assessment and Information Sciences Div.; Rothman, R. [USDOE Idaho Field Office, Idaho Falls, ID (United States)

    1993-02-01

    This report presents the technical details of RISIUND, a computer code designed to estimate potential radiological consequences and health risks to individuals and the collective population from exposures associated with the transportation of spent nuclear fuel. RISKIND is a user-friendly, semiinteractive program that can be run on an IBM or equivalent personal computer. The program language is FORTRAN-77. Several models are included in RISKIND that have been tailored to calculate the exposure to individuals under various incident-free and accident conditions. The incidentfree models assess exposures from both gamma and neutron radiation and can account for different cask designs. The accident models include accidental release, atmospheric transport, and the environmental pathways of radionuclides from spent fuels; these models also assess health risks to individuals and the collective population. The models are supported by databases that are specific to spent nuclear fuels and include a radionudide inventory and dose conversion factors.

  19. Computer model for refinery operations with emphasis on jet fuel production. Volume 3: Detailed systems and programming documentation

    Science.gov (United States)

    Dunbar, D. N.; Tunnah, B. G.

    1978-01-01

    The FORTRAN computing program predicts flow streams and material, energy, and economic balances of a typical petroleum refinery, with particular emphasis on production of aviation turbine fuels of varying end point and hydrogen content specifications. The program has a provision for shale oil and coal oil in addition to petroleum crudes. A case study feature permits dependent cases to be run for parametric or optimization studies by input of only the variables which are changed from the base case.

  20. Annual Radiological Environmental Monitoring Program Report for the Three Mile Island, Unit 2, Independent Spent Fuel Storage Installation

    International Nuclear Information System (INIS)

    Hall, Gregory Graham

    2002-01-01

    This report presents the results of the 2001 Radiological Environmental Monitoring Program conducted in accordance with 10 CFR 72.44 for the Three Mile Island, Unit 2, Independent Spent Fuel Storage Installation. A description of the facility and the monitoring program is provided. The results of monitoring the two predominant radiation exposure pathways, potential airborne radioactivity releases and direct radiation exposure, indicate the facility operation has not contributed to any increase in the estimated maximum potential dose commitment to the general public.

  1. Biomass as a fuel and a profitable investment: the Euro-ASEAN COGEN program

    International Nuclear Information System (INIS)

    Menu, J.-F.; Schenkel, Y.; Guillaume, M.

    1997-01-01

    The COGEN Program (''Cogen'') is an economic cooperation program between the European Commission and ASEAN (Association of Southeast Asian Nations). A pioneering initiative in the field of biomass energy. Cogen is coordinated by and from AIT (Asian Institute of Technology, Bangkok, Thailand). Its main objective is to accelerate the implementation of proven technologies generating heat and/or power from wood and agro-residues through partnerships between European and ASEAN companies. ASEAN now offers the biggest potential for energy solutions, including waste-based fuels. Within Cogen, a number of demonstration projects have been implemented in different ASEAN industries. These projects have generated over 100 million US dollars in direct investment and represent showcases of proven technology in biomass energy equipment around the region. Some biomass energy projects have been highly profitable. The success of Cogen can also be explained by an emphasis on market intelligence, i.e., information sources, channels and business opportunities rarely achieved in public-private initiatives. (author)

  2. Spent Nuclear Fuel Project document control and Records Management Program Description

    International Nuclear Information System (INIS)

    MARTIN, B.M.

    2000-01-01

    The Spent Nuclear Fuel (SNF) Project document control and records management program, as defined within this document, is based on a broad spectrum of regulatory requirements, Department of Energy (DOE) and Project Hanford and SNF Project-specific direction and guidance. The SNF Project Execution Plan, HNF-3552, requires the control of documents and management of records under the auspices of configuration control, conduct of operations, training, quality assurance, work control, records management, data management, engineering and design control, operational readiness review, and project management and turnover. Implementation of the controls, systems, and processes necessary to ensure compliance with applicable requirements is facilitated through plans, directives, and procedures within the Project Hanford Management System (PHMS) and the SNF Project internal technical and administrative procedures systems. The documents cited within this document are those which directly establish or define the SNF Project document control and records management program. There are many peripheral documents that establish requirements and provide direction pertinent to managing specific types of documents that, for the sake of brevity and clarity, are not cited within this document

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

    Energy Technology Data Exchange (ETDEWEB)

    2012-09-01

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

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

    International Nuclear Information System (INIS)

    Baker, R.D.

    1978-12-01

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

  5. Fact Sheet: Alternative Low-Sulfur Diesel Fuel Transition Program for Alaska

    Science.gov (United States)

    This fact sheet summarizes EPA's final rule modifying the diesel fuel regulations to apply an effective date of 6-1-2010 for 15 ppm sulfur requirements for highway, nonroad, locomotive and marine diesel fuel produced/imported for, distributed

  6. Experimental program on fuel rod behaviour under off-normal conditions

    International Nuclear Information System (INIS)

    Languille, A.; Cecchi, P.

    1985-01-01

    During LMFBR plant operation, fuel developments are primarily concerned with the fuel pin irradiation behaviour under steady-state conditions up to high burn-up levels. But additional studies under off-normal conditions are necessary in order to assess fuel pin performance and to define operational limits. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-01-01

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

  8. Chemical engineering division fuel cycle programs. Progress report, January--September 1977

    International Nuclear Information System (INIS)

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

    1977-01-01

    Fuel-cycle studies reported for this period include pyrochemical separation of plutonium and americium oxides from contaminated materials of construction such as steel. When slag and actinide-contaminated metal in the same process vessel are heated until liquefied, the actinides are partitioned to a high degree into the slags. Also, studies of advanced solvent extraction techniques are focused on the development of centrifugal contactors for use in Purex processes. A miniature contactor is to be used for performance studies applicable to larger units. In other work, literature on the process chemistry of zirconium and ruthenium has been reviewed to aid in improving the process when short-residence-time contactors are used. In addition, a review of information on the dispersion of reagents and products during accidents in fuel reprocessing facilities has been initiated to develop systematic data useful in identifying source terms. A review and evaluation of the encapsulation of high-level waste in a metal matrix are continuing. The data will be used to identify the state of the art and the importance of selected features of this process. In other work, criteria for the handling of hulls are being developed on the basis of past work on the pyrophoricity of zirconium alloys and related criteria from several sources. These suggested criteria will be assembled with the necessary technical rationalization into a package for review by interested parties. Other work consists of a brief program to explore the disposal of noble gas fission products by deep-well injection and laboratory-scale experiments to study the migratory characteristics of nuclear waste confined in geologic formations. 28 figures, 26 tables

  9. Program of enhancing the Korea-USA cooperation research for the development of proliferation resistant fuel cycle technology

    International Nuclear Information System (INIS)

    Yang, Myung Seung; Ahn, D. H.; Ko, W. I.

    2007-03-01

    The objective of the Program is to develop the fuel cycle technology of GEN-IV SFR (Sodium Fast Reactor) system through the Korea-USA cooperation research in order to improve the efficiency of the technology development and to increase the transparency of the research. Since the pyroprocessing research by using actual spent nuclear fuel can not be performed in Korea at present, the active demonstration research will be performed by using the USA national research facilities under the Korea-USA cooperation. Moreover, the development of safeguards technology and the methodology for the evaluation of the proliferation resistance will also be performed under the cooperation. The current cooperation national laboratories of the safeguards and pyroprocessing technology development are LANL (Los Alamos National Lab.) and INL (Idaho National Lab.), respectively. Practical research experience and technical data for the pyroprocessing technology can be achieved through the demonstration of the inactive research results, which was performed in Korea, by using actual spent nuclear fuel. The scope of the cooperation study encompass the electrolytic reduction of oxide spent fuel, electrorefining, liquid cadmium cathode process, TRU fuel fabrication, fuel performance evaluation and related safeguards technology development

  10. A dynamic programming approach for modeling low-carbon fuel technology adoption considering learning-by-doing effect

    International Nuclear Information System (INIS)

    Chen, Yuche; Zhang, Yunteng; Fan, Yueyue; Hu, Kejia; Zhao, Jianyou

    2017-01-01

    Highlights: • Dynamic programming method is used in transportation fuel portfolio planning. • The learning effect in new fuel technology is endogenously modeled through an experience curve. • Cellulosic biofuels play critical role in de-carbonization transport sector in near term. • The initial 3–4 billion gallons production is critical to bring down cellulosic biofuels’ cost. • Large penetration of Zero Emission Vehicles will discourage development of cellulosic biofuels. - Abstract: Promoting the adoption of low-carbon technologies in the transportation fuel portfolio is an effective strategy to mitigate greenhouse gas emissions from the transportation sector worldwide. However, as one of the most promising low-carbon fuels, cellulosic biofuel has not fully entered commercial production. Governments could provide guidance in developing cellulosic biofuel technologies, but no systematic approach has been proposed yet. We establish a dynamic programming framework for investigating time-dependent and adaptive decision-making processes to develop advanced fuel technologies. The learning-by-doing effect inherited in the technology development process is included in the framework. The proposed framework is applied in a case study to explore the most economical pathway for California to develop a solid cellulosic biofuel industry under its Low Carbon Fuel Standard. Our results show that cellulosic biofuel technology is playing a critical role in guaranteeing California’s 10% greenhouse gas emission reduction by 2020. Three to four billion gallons of cumulative production are needed to ensure that cellulosic biofuel is cost-competitive with petroleum-based fuels or conventional biofuels. Zero emission vehicle promoting policies will discourage the development of cellulosic biofuel. The proposed framework, with small adjustments, can also be applied to study new technology development in other energy sectors.

  11. Mathematical simulation of stressed-deformed state in rod cylindrical fuel elemnts KONDOR program

    International Nuclear Information System (INIS)

    Khmelevskij, M.Ya.; Malakhova, E.I.; Dolmatov, P.S.

    1987-01-01

    A mathematical model for numerical computation of stressed-deformed stae in a rod cylindrical fuel element is developed. The model is based on preliminary discretization of the design scheme and linearization of radial parameters as radius functions. The formulation generality enables to calculate strength parameter kinetics in any circular cylindrical fuel element (e.g. annular fuel element; solid or tubular core; ceramic, metallic or dispersion fuel) for arbitrary transient operating conditions and taking into account all possible loading factors. The method is realized in the KONDOR programm (FORTRAN, ES-1061 computer). An example illustrating computation of stress kinetics in a fast reactor fuel element during transient operation is given

  12. 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.'

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

    International Nuclear Information System (INIS)

    Salmon, R.; Kee, C.W.

    1977-06-01

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

  14. CASKETSS-DYNA2D: a nonlinear impact analysis computer program for nuclear fuel transport casks in two dimensional geometries

    International Nuclear Information System (INIS)

    Ikushima, Takeshi

    1988-10-01

    A nonlinear impact analysis computer program DYNA2D, which was developed by Hallquist, has been introduced from Lawrence Livermore National Laboratory for the purpose of using impact analysis of nuclear fuel transport casks. DYNA2D has been built in CASKETSS code system (CASKETSS means a modular code system for CASK Evaluation code system for Thermal and Structural Safety). Main features of DYNA2D are as follows; (1) This program has been programmed to provide near optimal speed on vector processing computers. (2) An explicit time integration method is used for fast calculation. (3) Many material models are available in the program. (4) A contact-impact algorithm permits gap and sliding along structural interfaces. (5) A rezoner has been embedded in the program. (6) The graphic program for representations of calculation is provided. In the paper, brief illustration of calculation method, input data and sample calculations are presented. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2013-10-01

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

  16. Ruthenium release at high temperature from irradiated PWR fuels in various oxidising conditions. Main findings from the VERCORS program

    International Nuclear Information System (INIS)

    Ducros, G.; Pontillon, Y.; Malgouyres, P.P.; Taylor, P.; Dutheillet, Y.

    2005-01-01

    Fission product release and transport in case of PWR severe accident is a major topic in reactor safety assessment due to the potential radiological consequences for surrounding populations and the environment. In this context, the Institute for Radiological Protection and Safety (IRSN) and Electricite de France (EDF) have supported the VERCORS analytical test program which was performed by the ''Commissariat a l'Energie Atomique'' (CEA). It is usually considered as complementary to the PHEBUS FP in-pile integral experimental program. 25 annealing tests were performed between 1983 and 2002 on irradiated PWR fuels under various conditions of temperature and atmospheres (oxidising or reducing conditions).The influence of the nature of the fuel (UO 2 versus MOX, burn-up) and the fuel morphology (initially intact or fragmented fuels) have also been investigated. These led to an extended data base allowing on the one hand to study mechanisms which promote fission products release, and on the other hand to enhance models implemented in severe accident codes. Among all the fission products investigated, ruthenium is of specific concern because of its high radiological effects due essentially to the combination of both its short and long half-life isotopes (i.e. 103 Ru and 106 Ru respectively), but also by its ability to generate volatile gaseous oxides (RuO 3 , RuO 4 ) in very oxidising conditions, in particular in the case of air ingress accidents. Important uncertainties still remain on the release and transport of this element in such situations, and investigations on this open issue are notably carried out in the SARNET European framework. The present communication gives a general overview of the VERCORS program and presents more deeply the main findings concerning the ruthenium release. Its global behaviour is analysed on the basis of several comparative tests: same UO 2 sample (35 and 50 GWd/t) under hydrogen or steam conditions, similar MOX sample (40 GWd/t) under

  17. Low NO subx heavy fuel combustor concept program. Phase 1A: Coal gas addendum

    Science.gov (United States)

    Rosfjord, T.; Sederquist, R.

    1982-01-01

    The performance and emissions from a rich-lean combustor fired on simulated coal gas fuels were investigated using a 12.7-cm diameter axially-staged burner originally designed for operation with high heating value liquid fuels. A simple, tubular fuel injector was substituted for the liquid fuel nozzle; no other combustor modifications were made. Four test fuels were studied including three chemically bound nitrogen-free gas mixtures with higher heating values of 88, 227, and 308 kj/mol (103, 258 and 349 Btu/scf), and a 227 kj/mol (258 Btu/scf) heating value doped with ammonia to produce a fuel nitrogen content of 0.5% (wt). Stable, ultra-low nitrogen oxide, smoke-free combustion was attained for the nitrogen-free fuels. Results with the doped fuel indicated that less than 5% conversion of NH3 to nitrogen oxide levels below Environmental Protection Agency limits could be achieved. In some instances, excessive CO levels were encountered. It is shown that use of a burner design employing a less fuel-rich primary zone than that found optimum for liquid fuels would yield more acceptable CO emissions.

  18. Natural Resource Canada`s fuel cell R and D program

    Energy Technology Data Exchange (ETDEWEB)

    Hammerli, M; Beck, N R [Natural Resources Canada, Ottawa, ON (Canada)

    1998-05-01

    The rationale for focusing fuel cell technology on the Ballard Proton exchange Membrane (PEM) system is provided. As well, research into other fuel cell types supported by Natural Resources Canada are discussed. Fuel cells are electrochemical devices that convert a fuel and an oxidant directly into electricity. Five fuel cell technologies use hydrogen as the fuel: (1) the alkaline fuel cell (AFC), (2) the proton exchange membrane fuel cell (PEMFC), (3) the phosphoric acid fuel cell (PAFC), (4) the molten carbonate fuel cell (MCFC), and (5) the solid oxide fuel cell (SOFC). The PEMFC is suitable for transportation applications because it does not contain a liquid electrolyte and it operates at about 80 degrees C. Trials on municipal bus systems are currently underway in Vancouver and Chicago. PEMFC stacks are supplied by Ballard Power Systems of Burnaby, BC, a recognized world leader in PEMFC technology. Daimler-Benz is demonstrating the methanol reformer on its NECAR-3, powered with a Ballard PEMFC. Ballard is also designing and producing two prototype fuel cell engines for the Ford Motor Company which will integrate them into its P2000 prototype vehicle platform. The Ballard technology is also suitable for distributed power generation up to about five MW, as well as for cogeneration, when fuelled with natural gas. Stuart Energy Systems (SES) has developed an advanced UNICELL-CLUSTER{sup T}M, which permits a direct coupling of the PV array to the electrolyser, a project which demonstrates the use of solar-electrolytic hydrogen production. SES is also designing a refuelling system for the BC Transit System in Vancouver for refuelling their three Zero Emission urban transit buses powered by Ballard fuel cell engines.

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

    International Nuclear Information System (INIS)

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

    1982-03-01

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

  20. Status report: Nuclear fuel operating experience in implementing the program for power generation increase at VVER NPPs of JSC concern Rosenergoatom

    International Nuclear Information System (INIS)

    Ryabinin, Y.

    2015-01-01

    The power uprate program of operating WWER-1000 plants was performed by Rosenergoatom using FA-2M and FAA-PLUS for 18-month fuel cycles. Their operation was justified at 104% of the rated power, and extension to 18-month fuel cycles was carried out at WWER-1000 units (except for Kalinin NPP-1). The analysis of actual performance data confirmed the efficiency of the actions implemented, and issues addressed related to the introduction of new fuel type, extended fuel cycles and spent nuclear fuel storage and removal

  1. The role of long-term geologic changes in the regulation of the Canadian nuclear fuel waste management program

    International Nuclear Information System (INIS)

    Flavelle, P.

    1996-01-01

    It is recognized that the geosphere is a dynamic system over the long time frames of nuclear fuel waste disposal. This paper describes how consideration of a dynamic geosphere has impacted upon the evolving regulatory environment in Canada, and how the approach taken to comply with the regulatory requirements can affect the evaluation of long-term geologic changes. AECB staff opinion is that if the maximum possible effect of geologic changes can be demonstrated to have negligible impact on the safety of a nuclear fuel waste repository, then further consideration of a dynamic geosphere is unnecessary for the current review of the Canadian Nuclear Fuel Waste Management Program. (authors). 7 refs., 4 figs

  2. Spent nuclear fuels project: FY 1995 multi-year program plan, WBS {number_sign}1.4

    Energy Technology Data Exchange (ETDEWEB)

    Denning, J.L.

    1994-09-01

    The mission of the Spent Nuclear Fuel (SNF) program is to safely, reliably, and efficiently manage, condition, transport, and store Department of Energy (DOE)-owned SNF, so that it meets acceptance criteria for disposal in a permanent repository. The Hanford Site Spent Nuclear Fuel strategic plan for accomplishing the project mission is: Establish near-term safe storage in the 105-K Basins; Complete national Environmental Policy Act (NEPA) process to obtain a decision on how and where spent nuclear fuel will be managed on the site; Define and establish alternative interim storage on site or transport off site to support implementation of the NEPA decision; and Define and establish a waste package qualified for final disposition. This report contains descriptions of the following: Work Breakdown Structure; WBS Dictionary; Responsibility Assignment Matrix; Program Logic Diagrams; Program Master Baseline Schedule; Program Performance Baseline Schedule; Milestone List; Milestone Description Sheets; Cost Baseline Summary by Year; Basis of Estimate; Waste Type Data; Planned Staffing; and Fiscal Year Work Plan.

  3. Spent nuclear fuels project: FY 1995 multi-year program plan, WBS number-sign 1.4

    International Nuclear Information System (INIS)

    Denning, J.L.

    1994-09-01

    The mission of the Spent Nuclear Fuel (SNF) program is to safely, reliably, and efficiently manage, condition, transport, and store Department of Energy (DOE)-owned SNF, so that it meets acceptance criteria for disposal in a permanent repository. The Hanford Site Spent Nuclear Fuel strategic plan for accomplishing the project mission is: Establish near-term safe storage in the 105-K Basins; Complete national Environmental Policy Act (NEPA) process to obtain a decision on how and where spent nuclear fuel will be managed on the site; Define and establish alternative interim storage on site or transport off site to support implementation of the NEPA decision; and Define and establish a waste package qualified for final disposition. This report contains descriptions of the following: Work Breakdown Structure; WBS Dictionary; Responsibility Assignment Matrix; Program Logic Diagrams; Program Master Baseline Schedule; Program Performance Baseline Schedule; Milestone List; Milestone Description Sheets; Cost Baseline Summary by Year; Basis of Estimate; Waste Type Data; Planned Staffing; and Fiscal Year Work Plan

  4. The computer program ELCOM in the planning and structural analysis of PWR fuel elements: an example

    International Nuclear Information System (INIS)

    Silva Macedo, L.V. da

    1990-01-01

    Is's presented some results obtained with the ELCOM computer code, such as deflections, moments and natural frequencies, used in the design and structural analysis of PWR fuels assemblies. It's studied the behavior of these results varying the number of spacer grids, the rigidity of the joint between the fuel pin and the spacer grid, and the fuel assembly's boundary condition, considered in the analysis, in it's mounting into the core (if clamped-clamped, clamped-hinged or hinged-hinged). (author)

  5. Oak Ridge TNS program: study of fueling techniques in support of TNS development

    International Nuclear Information System (INIS)

    Simpson, W.

    1979-12-01

    The objective of this study was to survey fueling techniques and determine which approaches were compatible with the TNS requirements. Specifically, the following tasks were undertaken: survey of existing fueling concepts for use in TNS, determination of available physical properties for D-T fuel pellets, performance of preliminary load analysis of selected pellet acceleration machines, preparation of conceptual designs, and recommendations for follow-up work

  6. Experimental program to determine maximum temperatures for dry storage of spent fuel

    International Nuclear Information System (INIS)

    Knox, C.A.; Gilbert, E.R.; White, G.D.

    1985-02-01

    Although air is used as a cover gas in some dry storage facilities, other facilities use inert cover gases which must be monitored to assure inertness of the atmosphere. Thus qualifying air as a cover gas is attractive for the dry storage of spent fuels. At sufficiently high temperatures, air can react with spent fuel (UO 2 ) at the site of cladding breaches that formed during reactor irradiation or during dry storage. The reaction rate is temperature dependent; hence the rates can be maintained at acceptable levels if temperatures are low. Tests with spent fuel are being conducted at Pacific Northwest Laboratory (PNL) to determine the allowable temperatures for storage of spent fuel in air. Tests performed with nonirradiated UO 2 pellets indicated that moisture, surface condition, gamma radiation, gadolinia content of the fuel pellet, and temperature are important variables. Tests were then initiated on spent fuel to develop design data under simulated dry storage conditions. Tests have been conducted at 200 and 230 0 C on spent fuel in air and 275 0 C in moist nitrogen. The results for nonirradiated UO 2 and published data for irradiated fuel indicate that above 230 0 C, oxidation rates are unacceptably high for extended storage in air. The tests with spent fuel will be continued for approximately three years to enable reliable extrapolations to be made for extended storage in air and inert gases with oxidizing constituents. 6 refs., 6 figs., 3 tabs

  7. GRSIS program to predict fission gas release and swelling behavior of metallic fast reactor fuel

    International Nuclear Information System (INIS)

    Lee, Chan Bock; Lee, Byung Ho; Nam, Cheol; Sohn, Dong Seong

    1999-03-01

    A mechanistic model of fission gas release and swelling for the U-(Pu)-Zr metallic fuel in the fast reactor, GRSIS (Gas Release and Swelling in ISotropic fuel matrix) was developed. Fission gas bubbles are assumed to nucleate isotropically from the gas atoms in the metallic fuel matrix since they can nucleate at both the grain boundaries and the phase boundaries which are randomly distributed inside the grain. Bubbles can grow to larger size by gas diffusion and coalition with other bubbles so that they are classified as three classes depending upon their sizes. When bubble swelling reaches the threshold value, bubbles become interconnected each other to make the open channel to the external free space, that is, the open bubbles and then fission gases inside the interconnected open bubbles are released instantaneously. During the irradiation, fission gases are released through the open bubbles. GRSIS model can take into account the fuel gap closure by fuel bubble swelling. When the fuel gap is closed by fuel swelling, the contact pressure between fuel and cladding in relation to the bubble swelling and temperature is calculated. GRSIS model was validated by comparison with the irradiation test results of U-(Pu)-Zr fuels in ANL as well as the parametric studies of the key variable in the model. (author). 13 refs., 1 tab., 22 figs

  8. GRSIS program to predict fission gas release and swelling behavior of metallic fast reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Bock; Lee, Byung Ho; Nam, Cheol; Sohn, Dong Seong

    1999-03-01

    A mechanistic model of fission gas release and swelling for the U-(Pu)-Zr metallic fuel in the fast reactor, GRSIS (Gas Release and Swelling in ISotropic fuel matrix) was developed. Fission gas bubbles are assumed to nucleate isotropically from the gas atoms in the metallic fuel matrix since they can nucleate at both the grain boundaries and the phase boundaries which are randomly distributed inside the grain. Bubbles can grow to larger size by gas diffusion and coalition with other bubbles so that they are classified as three classes depending upon their sizes. When bubble swelling reaches the threshold value, bubbles become interconnected each other to make the open channel to the external free space, that is, the open bubbles and then fission gases inside the interconnected open bubbles are released instantaneously. During the irradiation, fission gases are released through the open bubbles. GRSIS model can take into account the fuel gap closure by fuel bubble swelling. When the fuel gap is closed by fuel swelling, the contact pressure between fuel and cladding in relation to the bubble swelling and temperature is calculated. GRSIS model was validated by comparison with the irradiation test results of U-(Pu)-Zr fuels in ANL as well as the parametric studies of the key variable in the model. (author). 13 refs., 1 tab., 22 figs.

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

    International Nuclear Information System (INIS)

    1981-06-01

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

  10. An overview of the Canadian program to investigate vibration and fretting in nuclear fuel assemblies

    International Nuclear Information System (INIS)

    Oldaker, I.E.; Lane, A.D.; Forrest, C.F.

    The development of a model that would allow the fuel designer to predict the occurrence of fretting could materially reduce the amount of development testing of a new fuel design. To achieve this, we are working in several areas: to identify and measure the phenomena that excite fuel to vibrate, and to study their relation to reactor design features; to predict the vibratory response of a fuel assembly as a function of its design and environment, and; to study the relationship between vibration and fretting to determine when vibration results in fretting. (author)

  11. Subchannel analysis program for boiling water reactor fuel bundles based on five conservation equations of two-phase flow

    International Nuclear Information System (INIS)

    Bessho, Y.; Uchikawa, S.

    1985-01-01

    A subchannel analysis program, MENUETT, is developed for evaluation of thermal-hydraulic characteristics in boiling water reactor fuel bundles. This program is based on five conservation equations of two-phase flow with the drift-flux correlation. The cross flows are calculated separately for liquid and vapor phases from the lateral momentum conservation equation. The effects of turbulent mixing and void drift are accounted for in the program. The conservation equations are implicitly differentiated with the convective terms by the donor-cell method, and are solved iteratively in the axial and lateral directions. Data of the 3 X 3 rod bundle experiments are used for program verification. The lateral distributions of equilibrium quality and mass flow rate at the bundle exit calculated by the program compare satisfactorily with the experimental results

  12. Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs draft environmental impact statement

    International Nuclear Information System (INIS)

    1994-06-01

    The US Department of Energy (DOE) has prepared this report to assist its management in making two decisions. The first decision, which is programmatic, is to determine the management program for DOE spent nuclear fuel. The second decision is on the future direction of environmental restoration, waste management, and spent nuclear fuel management activities at the Idaho National Engineering Laboratory. Volume 1 of the EIS, which supports the programmatic decision, considers the effects of spent nuclear fuel management on the quality of the human and natural environment for planning years 1995 through 2035. DOE has derived the information and analysis results in Volume 1 from several site-specific appendixes. Volume 2 of the EIS, which supports the INEL-specific decision, describes environmental impacts for various environmental restoration, waste management, and spent nuclear fuel management alternatives for planning years 1995 through 2005. This Appendix B to Volume 1 considers the impacts on the INEL environment of the implementation of various DOE-wide spent nuclear fuel management alternatives. The Naval Nuclear Propulsion Program, which is a joint Navy/DOE program, is responsible for spent naval nuclear fuel examination at the INEL. For this appendix, naval fuel that has been examined at the Naval Reactors Facility and turned over to DOE for storage is termed naval-type fuel. This appendix evaluates the management of DOE spent nuclear fuel including naval-type fuel

  13. Chemical Engineering Division Fuel Cycle Programs. Quarterly progress report, January-March 1979

    Energy Technology Data Exchange (ETDEWEB)

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

    1980-01-01

    In the program on pyrochemical and dry processing methods (PDPM) for nuclear fuel, corrosion testing of refractory metals and alloys, graphite, and SiC in PDPM environments was done. A tungsten-metallized Al/sub 2/O/sub 3/-3% Y/sub 2/O/sub 3/ crucible was successfully fabricated. Tungsten microstructure of a plasma-sprayed tungsten crucible was stabilized by nickel infiltration and heat treatment. Solubility measurements of Th in Cd and Cd-Mg alloys were continued, as were experiments to study the reduction of high-fired ThO/sub 2/. Work on the fused salt electrolysis of CaO also was continued. The method of coprocessing of U and Pu by a salt transport process was modified. Tungsten-coated molybdenum crucibles were fabricated. The proliferation resistance of chloride volatility processing of thorium-based fuels is being evaluated by studying the behavior of fission product elements during chlorination of U and Th. Thermodynamic analysis of the phase relationships in the U-Pu-Zn system was initiated. The Pyro-Civex reprocessing method is being reviewed. Reactivity of UO/sub 2/ and PuO/sub 2/ with molten equimolar NaNO/sub 3/-KNO/sub 3/ is being studied along with the behavior of selected fission product elements. Work was continued on the reprocessing of actinide oxides by extracting the actinides from a bismuth solution. Rate of dissolution of UO/sub 2/ microspheres in LiCl/AlCl/sub 3/ was measured. Nitriding rates of Th and U dissolved in molten tin were measured. In work on the encapsulation of radioactive waste in metal, leach rates of a simulated waste glass were studied. Rates of dissolution of metals (potential barrier materials) in aqueous media are being studied. In work on the transport properties of nuclear waste in geologic media, the adsorption of iodate by hematite as a function of pH and iodate concentration was measured. The migration behavior of cesium in limestone was studied in relation to the cesium concentration and pH of simulated groundwater

  14. Analysis of the behavior under irradiation of high burnup nuclear fuels with the computer programs FRAPCON and FRAPTRAN

    Energy Technology Data Exchange (ETDEWEB)

    Reis, Regis; Silva, Antonio Teixeira e, E-mail: teixeira@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2017-11-01

    The objective of this paper is to verify the validity and accuracy of the results provided by computer programs FRAPCON-3.4a and FRAPTRAN-1.4, used in the simulation process of the irradiation behavior of Pressurized Water Reactors (PWR) fuel rods, in steady-state and transient operational conditions at high burnup. To achieve this goal, the results provided by these computer simulations are compared with experimental data available in the database FUMEX III. Through the results, it was found that the computer programs used have a good ability to predict the operational behavior of PWR fuel rods in high burnup steady-state conditions and under the influence of Reactivity Initiated Accident (RIA). (author)

  15. Vehicle Design Evaluation Program (VDEP). A computer program for weight sizing, economic, performance and mission analysis of fuel-conservative aircraft, multibodied aircraft and large cargo aircraft using both JP and alternative fuels

    Science.gov (United States)

    Oman, B. H.

    1977-01-01

    The NASA Langley Research Center vehicle design evaluation program (VDEP-2) was expanded by (1) incorporating into the program a capability to conduct preliminary design studies on subsonic commercial transport type aircraft using both JP and such alternate fuels as hydrogen and methane;(2) incorporating an aircraft detailed mission and performance analysis capability; and (3) developing and incorporating an external loads analysis capability. The resulting computer program (VDEP-3) provides a preliminary design tool that enables the user to perform integrated sizing, structural analysis, and cost studies on subsonic commercial transport aircraft. Both versions of the VDEP-3 Program which are designated preliminary Analysis VDEP-3 and detailed Analysis VDEP utilize the same vehicle sizing subprogram which includes a detailed mission analysis capability, as well as a geometry and weight analysis for multibodied configurations.

  16. End-of-life destructive examination of light water breeder reactor fuel rods (LWBR Development Program)

    International Nuclear Information System (INIS)

    Richardson, K.D.

    1987-10-01

    Destructive examination of 12 representative Light Water Breeder Reactor fuel rods was performed following successful operation in the Shippingport Atomic Power Station for 29,047 effective full power hours, about five years. Light Water Breeder Reactor fuel rods were unique in that the thorium oxide and uranium-233 oxide fuel was contained within Zircaloy-4 cladding. Destructive examinations included analysis of released fission gas; chemical analysis of the fuel to determine depletion, iodine, and cesium levels; chemical analysis of the cladding to determine hydrogen, iodine, and cesium levels; metallographic examination of the cladding, fuel, and other rod components to determine microstructural features and cladding corrosion features; and tensile testing of the irradiated cladding to determine mechanical strength. The examinations confirmed that Light Water Breeder Reactor fuel rod performance was excellent. No evidence of fuel rod failure was observed, and the fuel operating temperature was low (below 2580 0 F at which an increased percentage of fission gas is released). 21 refs., 80 figs., 20 tabs

  17. Technical Approach and Plan for Transitioning Spent Nuclear Fuel (SNF) Project Facilities to the Environmental Restoration Program

    International Nuclear Information System (INIS)

    SKELLY, W.A.

    1999-01-01

    This document describes the approach and process in which the 100-K Area Facilities are to be deactivated and transitioned over to the Environmental Restoration Program after spent nuclear fuel has been removed from the K Basins. It describes the Transition Project's scope and objectives, work breakdown structure, activity planning, estimated cost, and schedule. This report will be utilized as a planning document for project management and control and to communicate details of project content and integration

  18. Application of the successive linear programming technique to the optimum design of a high flux reactor using LEU fuel

    International Nuclear Information System (INIS)

    Mo, S.C.

    1991-01-01

    The successive linear programming technique is applied to obtain the optimum thermal flux in the reflector region of a high flux reactor using LEU fuel. The design variables are the reactor power, core radius and coolant channel thickness. The constraints are the cycle length, average heat flux and peak/average power density ratio. The characteristics of the optimum solutions with various constraints are discussed

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

    International Nuclear Information System (INIS)

    Shemilt, L.W.

    1993-03-01

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

  20. Nondestructive assay of spent fuel rods from a Light Water Breeder Reactor (LWBR Development Program)

    International Nuclear Information System (INIS)

    Tessler, G.; Beaudoin, B.R.; Beggs, W.J.; Freeman, L.B.; Schick, W.C. Jr.

    1987-09-01

    A gauge, called the Production Irradiated Fuel Assay Gauge (PIFAG), has been developed and utilized to measure, nondestructively, the fissile fuel content of spent fuel rods from the Light Water Breeder Reactor (LWBR) core. The PIFAG was in operation from November 1983 to May 1987. During this period, assay data were obtained for two irradiated test rods used for initial qualification of the gauge and 524 spent LWBR core rods. A review of PIFAG operations is given, including hot cell operations, calibration, assay operations, and methods used to monitor the data quality and verify the precision and accuracy of the data. The analytical model used to determine the core rod fissile fuel content from the data and the results for the 524 LWBR spent fuel rods are given