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Sample records for argonne research reactor

  1. Development of a Monolithic Research Reactor Fuel Type at Argonne National Laboratory

    SciTech Connect

    Clark, C.R.; Briggs, R.J.

    2004-10-06

    The Reduced Enrichment for Research and Test Reactors (RERTR) program has been tasked with the conversion of research reactors from highly enriched to low-enriched uranium (LEU). To convert several high power reactors, monolithic fuel, a new fuel type, is being developed. This fuel type replaces the standard fuel dispersion with a fuel alloy foil, which allows for fuel densities far in excess of that found in dispersion fuel. The single-piece fuel foil also contains a significantly lower interface area between the fuel and the aluminum in the plate than the standard fuel type, limiting the amount of detrimental fuel-aluminum interaction that can occur. Implementation of monolithic fuel is dependant on the development of a suitable fabrication method as traditional roll-bonding techniques are inadequate.

  2. Computational chemistry in Argonne`s Reactor Analysis Division

    SciTech Connect

    Gelbard, E.; Agrawal, R.; Fanning, T.

    1997-08-01

    Roughly 3 years ago work on Argonne`s Integral Fast Reactor ({open_quotes}IFR{close_quotes}) was terminated and at that time, ANL funding was redirected to a number of alternative programs. One such alternative was waste management and, since disposal of spent fuel from ANL`s EBR-II reactor presents some special problems, this seemed an appropriate area for ANL work. Methods for the treatment and disposal of spent fuel (particularly from EBR-II but also from other sources) are now under very active investigation at ANL. The very large waste form development program is mainly experimental at this point, but within the Reactor Analysis ({open_quotes}RA{close_quotes}) Division a small computational chemistry program is underway, designed to supplement the experimental program. One of the most popular proposals for the treatment of much of our high-level wastes is vitrification. As noted below, this approach has serious drawbacks for EBR-II spent fuel. ANL has proposed, instead, that spent fuel first be pretreated by a special metallurgical process which produces, as waste, chloride salts of the various fission products; these salts would then be adsorbed in zeolite A, which is subsequently bonded with glass to produce a waste form suitable for disposal. So far it has been the main mission of RA`s computational chemistry program to study the process by which leaching occurs when the glass-bonded zeolite waste form is exposed to water. It is the purpose of this paper to describe RA`s computational chemistry program, to discuss the computational techniques involved in such a program, and in general to familiarize the M. and C. Division with a computational area which is probably unfamiliar to most of its member. 11 refs., 2 figs.

  3. Argonne Liquid-Metal Advanced Burner Reactor : components and in-vessel system thermal-hydraulic research and testing experience - pathway forward.

    SciTech Connect

    Kasza, K.; Grandy, C.; Chang, Y.; Khalil, H.; Nuclear Engineering Division

    2007-06-30

    This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition from normal high

  4. Argonne National Laboratory Research Highlights 1988

    SciTech Connect

    Not Available

    1988-01-01

    The research and development highlights are summarized. The world's brightest source of X-rays could revolutionize materials research. Test of a prototype insertion device, a key in achieving brilliant X-ray beams, have given the first glimpse of the machine's power. Superconductivity research focuses on the new materials' structure, economics and applications. Other physical science programs advance knowledge of material structures and properties, nuclear physics, molecular structure, and the chemistry and structure of coal. New programming approaches make advanced computers more useful. Innovative approaches to fighting cancer are being developed. More experiments confirm the passive safety of Argonne's Integral Fast Reactor concept. Device simplifies nuclear-waste processing. Advanced fuel cell could provide better mileage, more power than internal combustion engine. New instruments find leaks in underground pipe, measure sodium impurities in molten liquids, detect flaws in ceramics. New antibody findings may explain ability to fight many diseases. Cadmium in cigarettes linked to bone loss in women. Programs fight deforestation in Nepal. New technology could reduce acid rain, mitigate greenhouse effect, enhance oil recovery. Innovative approaches transfer Argonne-developed technology to private industry. Each year Argonne educational programs reach some 1200 students.

  5. Chemical research at Argonne National Laboratory

    SciTech Connect

    1997-04-01

    Argonne National Laboratory is a research and development laboratory located 25 miles southwest of Chicago, Illinois. It has more than 200 programs in basic and applied sciences and an Industrial Technology Development Center to help move its technologies to the industrial sector. At Argonne, basic energy research is supported by applied research in diverse areas such as biology and biomedicine, energy conservation, fossil and nuclear fuels, environmental science, and parallel computer architectures. These capabilities translate into technological expertise in energy production and use, advanced materials and manufacturing processes, and waste minimization and environmental remediation, which can be shared with the industrial sector. The Laboratory`s technologies can be applied to help companies design products, substitute materials, devise innovative industrial processes, develop advanced quality control systems and instrumentation, and address environmental concerns. The latest techniques and facilities, including those involving modeling, simulation, and high-performance computing, are available to industry and academia. At Argonne, there are opportunities for industry to carry out cooperative research, license inventions, exchange technical personnel, use unique research facilities, and attend conferences and workshops. Technology transfer is one of the Laboratory`s major missions. High priority is given to strengthening U.S. technological competitiveness through research and development partnerships with industry that capitalize on Argonne`s expertise and facilities. The Laboratory is one of three DOE superconductivity technology centers, focusing on manufacturing technology for high-temperature superconducting wires, motors, bearings, and connecting leads. Argonne National Laboratory is operated by the University of Chicago for the U.S. Department of Energy.

  6. Argonne's Magellan Cloud Computing Research Project

    ScienceCinema

    Beckman, Pete

    2016-07-12

    Pete Beckman, head of Argonne's Leadership Computing Facility (ALCF), discusses the Department of Energy's new $32-million Magellan project, which designed to test how cloud computing can be used for scientific research. More information: http://www.anl.gov/Media_Center/News/2009/news091014a.html

  7. Argonne's Magellan Cloud Computing Research Project

    SciTech Connect

    Beckman, Pete

    2009-01-01

    Pete Beckman, head of Argonne's Leadership Computing Facility (ALCF), discusses the Department of Energy's new $32-million Magellan project, which designed to test how cloud computing can be used for scientific research. More information: http://www.anl.gov/Media_Center/News/2009/news091014a.html

  8. Stirling engine research at Argonne National Laboratory

    SciTech Connect

    Holtz, R.E.; Daley, J.G.; Roach, P.D.

    1986-06-01

    Stirling engine research at Argonne National Laboratory has been focused at (1) development of mathematical models and analytical tools for predicting component and engine performance, and (2) experimental research into fundamental heat transfer and fluid flow phenomena occurring in Stirling cycle devices. A result of the analytical effort has been the formation of a computer library specifically for Stirling engine researchers and developers. The library contains properties of structural materials commonly used, thermophysical properties of several working fluids, correlations for heat transfer calculations and general specifications of mechanical arrangements (including various drive mechanisms) that can be utilized to model a particular engine. The library also contains alternative modules to perform analysis at different levels of sophistication, including design optimization. A reversing flow heat transfer facility is operating at Argonne to provide data at prototypic Stirling engine operating conditions under controlled laboratory conditions. This information is needed to validate analytical models.

  9. Advanced reactor development: The LMR integral fast reactor program at Argonne

    SciTech Connect

    Till, C.E.

    1990-01-01

    Reactor technology for the 21st Century must develop with characteristics that can now be seen to be important for the future, quite different from the things when the fundamental materials and design choices for present reactors were made in the 1950s. Argonne National Laboratory, since 1984, has been developing the Integral Fast Reactor (IFR). This paper will describe the way in which this new reactor concept came about; the technical, public acceptance, and environmental issues that are addressed by the IFR; the technical progress that has been made; and our expectations for this program in the near term. 3 figs.

  10. Argonne ARPA-E Battery Research

    ScienceCinema

    Amine, Khalil; Sinkula, Michael

    2016-07-12

    Argonne National Laboratory and Envia Systems annouced a licensing agreement for Argonne's patented electrode material technology. Envia plans to commercialize these materials for use in energy storage devices for the next generation of electric, plug-in and hybrid electric vehicles. General Motors Company, LG Chem, BASF and Toda Kyoga have also licensed this suite of Argonne's technologies. For more information visit us at http://www.anl.gov

  11. Verification Survey of the Building 315 Zero Power Reactor-6 Facility, Argonne National Laboratory-East, Argonne, Illinois

    SciTech Connect

    W. C. Adams

    2007-05-25

    Oak Ridge Institute for Science and Education (ORISE) conducted independent verification radiological survey activities at Argonne National Laboratory’s Building 315, Zero Power Reactor-6 facility in Argonne, Illinois. Independent verification survey activities included document and data reviews, alpha plus beta and gamma surface scans, alpha and beta surface activity measurements, and instrumentation comparisons. An interim letter report and a draft report, documenting the verification survey findings, were submitted to the DOE on November 8, 2006 and February 22, 2007, respectively (ORISE 2006b and 2007).

  12. Battery research at Argonne National Laboratory

    SciTech Connect

    Thackeray, M.M.

    1997-10-01

    Argonne National Laboratory (ANL) has, for many years, been engaged in battery-related R and D programs for DOE and the transportation industry. In particular, from 1973 to 1995, ANL played a pioneering role in the technological development of the high-temperature (400 C) lithium-iron disulfide battery. With the emphasis of battery research moving away from high temperature systems toward ambient temperature lithium-based systems for the longer term, ANL has redirected its efforts toward the development of a lithium-polymer battery (60--80 C operation) and room temperature systems based on lithium-ion technologies. ANL`s lithium-polymer battery program is supported by the US Advanced Battery Consortium (USABC), 3M and Hydro-Quebec, and the lithium-ion battery R and D efforts by US industry and by DOE.

  13. Research reactors

    SciTech Connect

    Tonneson, L.C.; Fox, G.J.

    1996-04-01

    There are currently 284 research reactors in operation, and 12 under construction around the world. Of the operating reactors, nearly two-thirds are used exclusively for research, and the rest for a variety of purposes, including training, testing, and critical assembly. For more than 50 years, research reactor programs have contributed greatly to the scientific and educational communities. Today, six of the world`s research reactors are being shut down, three of which are in the USA. With government budget constraints and the growing proliferation concerns surrounding the use of highly enriched uranium in some of these reactors, the future of nuclear research could be impacted.

  14. "Ask Argonne" - Edwin Campos, Research Meteorologist, Part 2

    ScienceCinema

    Edwin Campos

    2016-07-12

    Argonne's Edwin Campos has for the last two decades studied weather, and in particular, clouds. His research can help make solar power a more viable option for the U.S. and the world. In this video, Dr. Campos answers questions that were submitted by the public in response to his introductory video: http://www.youtube.com/watch?v=pfdoHz.... We will be posting a new "Ask Argonne" video every other month, on various topics. Keep an eye out for your next opportunity to submit a question and see if it gets answered - and if you get a shout-out on camera.

  15. "Ask Argonne" - Edwin Campos, Research Meteorologist, Part 2

    SciTech Connect

    Edwin Campos

    2013-05-23

    Argonne's Edwin Campos has for the last two decades studied weather, and in particular, clouds. His research can help make solar power a more viable option for the U.S. and the world. In this video, Dr. Campos answers questions that were submitted by the public in response to his introductory video: http://www.youtube.com/watch?v=pfdoHz.... We will be posting a new "Ask Argonne" video every other month, on various topics. Keep an eye out for your next opportunity to submit a question and see if it gets answered - and if you get a shout-out on camera.

  16. An evaluation of alternative reactor vessel cutting technologies for the experimental boiling water reactor at Argonne National Laboratory

    SciTech Connect

    Boing, L.E.; Henley, D.R. ); Manion, W.J.; Gordon, J.W. )

    1989-12-01

    Metal cutting techniques that can be used to segment the reactor pressure vessel of the Experimental Boiling Water Reactor (EBWR) at Argonne National Laboratory (ANL) have been evaluated by Nuclear Energy Services. Twelve cutting technologies are described in terms of their ability to perform the required task, their performance characteristics, environmental and radiological impacts, and cost and schedule considerations. Specific recommendations regarding which technology should ultimately be used by ANL are included. The selection of a cutting method was the responsibility of the decommissioning staff at ANL, who included a relative weighting of the parameters described in this document in their evaluation process. 73 refs., 26 figs., 69 tabs.

  17. Frontiers: Research highlights 1946-1996 [50th Anniversary Edition. Argonne National Laboratory

    SciTech Connect

    1996-12-31

    This special edition of 'Frontiers' commemorates Argonne National Laboratory's 50th anniversary of service to science and society. America's first national laboratory, Argonne has been in the forefront of U.S. scientific and technological research from its beginning. Past accomplishments, current research, and future plans are highlighted.

  18. Frontiers: Research Highlights 1946-1996 [50th Anniversary Edition. Argonne National Laboratory

    DOE R&D Accomplishments Database

    1996-01-01

    This special edition of 'Frontiers' commemorates Argonne National Laboratory's 50th anniversary of service to science and society. America's first national laboratory, Argonne has been in the forefront of U.S. scientific and technological research from its beginning. Past accomplishments, current research, and future plans are highlighted.

  19. Decontamination and decommissioning of the Argonne Thermal Source Reactor at Argonne National Laboratory - East project final report.

    SciTech Connect

    Fellhauer, C.; Garlock, G.; Mathiesen, J.

    1998-12-02

    The ATSR D&D Project was directed toward the following goals: (1) Removal of radioactive and hazardous materials associated with the ATSR Reactor facility; (2) Decontamination of the ATSR Reactor facility to unrestricted use levels; and (3)Documentation of all project activities affecting quality (i.e., waste packaging, instrument calibration, audit results, and personnel exposure). These goals had been set in order to eliminate the radiological and hazardous safety concerns inherent in the ATSR Reactor facility and to allow, upon completion of the project, unescorted and unmonitored access to the area. The reactor aluminum, reactor lead, graphite piles in room E-111, and the contaminated concrete in room E-102 were the primary areas of concern. NES, Incorporated (Danbury, CT) characterized the ATSR Reactor facility from January to March 1998. The characterization identified a total of thirteen radionuclides, with a total activity of 64.84 mCi (2.4 GBq). The primary radionuclides of concern were Co{sup 60}, Eu{sup 152}, Cs{sup 137}, and U{sup 238}. No additional radionuclides were identified during the D&D of the facility. The highest dose rates observed during the project were associated with the reactor tank and shield tank. Contact radiation levels of 30 mrem/hr (0.3 mSv/hr) were measured on reactor internals during dismantlement of the reactor. A level of 3 mrem/hr (0.03 mSv/hr) was observed in a small area (hot spot) in room E-102. DOE Order 5480.2A establishes the maximum whole body exposure for occupational workers at 5 rem/yr (50 mSv/yr); the administrative limit at ANL-E is 1 rem/yr (10 mSv/yr).

  20. Environmental Assessment and FONSI Proposed Decontamination and Disassembly of the Argonne Thermal Source Reactor (ATSR) at Argonne National Laboratory

    SciTech Connect

    N /A

    1998-07-15

    The purpose of this project is to protect human health and the environment from risks associated with the contaminated surplus ATSR. The proposed action is needed because the ATSR, a former experimental reactor, contains residual radioactivity and hazardous materials.

  1. "Ask Argonne" - Edwin Campos, Research Meteorologist, Part 1

    SciTech Connect

    Edwin Campos

    2013-05-08

    Dr. Edwin Campos is a Research Meteorologist at Argonne National Laboratory. For the last two decades, he has studied weather, and in particular, clouds. Clouds are one of the most uncertain variables in climate predictions and are often related to transportation hazards. Clouds can also impact world-class sporting events like the Olympics. You may have questions about the role of clouds, or weather, on our daily lives. How is severe weather monitored for airports? What is the impact of clouds and wind on the generation of electricity? One of the projects Edwin is working on is short-term forecasting as it relates to solar electricity. For this, Edwin's team is partnering with industry and academia to study new ways of forecasting clouds, delivering technologies that will allow the incorporation of more solar power into the electric grid. Post a question for Edwin as a comment below, and it might get answered in the follow-up video we'll post in the next few weeks.

  2. "Ask Argonne" - Edwin Campos, Research Meteorologist, Part 1

    ScienceCinema

    Edwin Campos

    2016-07-12

    Dr. Edwin Campos is a Research Meteorologist at Argonne National Laboratory. For the last two decades, he has studied weather, and in particular, clouds. Clouds are one of the most uncertain variables in climate predictions and are often related to transportation hazards. Clouds can also impact world-class sporting events like the Olympics. You may have questions about the role of clouds, or weather, on our daily lives. How is severe weather monitored for airports? What is the impact of clouds and wind on the generation of electricity? One of the projects Edwin is working on is short-term forecasting as it relates to solar electricity. For this, Edwin's team is partnering with industry and academia to study new ways of forecasting clouds, delivering technologies that will allow the incorporation of more solar power into the electric grid. Post a question for Edwin as a comment below, and it might get answered in the follow-up video we'll post in the next few weeks.

  3. Light-water-reactor safety research program. Quarterly progress report, January-March 1980

    SciTech Connect

    Massey, W.E.; Kyger, J.A.

    1980-08-01

    This progress report summarizes the Argonne National Laboratory work performed during January, February, and March 1980 on water-reactor-safety problems. The research and development area covered is Transient Fuel Response and Fission-Product Release.

  4. Applied mathematical sciences research at Argonne, April 1, 1981-March 31, 1982

    SciTech Connect

    Pieper, G.W.

    1982-01-01

    This report reviews the research activities in Applied Mathematical Sciences at Argonne National Laboratory for the period April 1, 1981, through March 31, 1982. The body of the report discusses various projects carried out in three major areas of research: applied analysis, computational mathematics, and software engineering. Information on section staff, visitors, workshops, and seminars is found in the appendices.

  5. Argonne National Laboratory: Laboratory Directed Research and Development FY 1993 program activities. Annual report

    SciTech Connect

    1993-12-23

    The purposes of Argonne`s Laboratory Directed Research and Development (LDRD) Program are to encourage the development of novel concepts, enhance the Laboratory`s R&D capabilities, and further the development of its strategic initiatives. Projects are selected from proposals for creative and innovative R&D studies which are not yet eligible for timely support through normal programmatic channels. Among the aims of the projects supported by the Program are establishment of engineering ``proof-of-principle`` assessment of design feasibility for prospective facilities; development of an instrumental prototype, method, or system; or discovery in fundamental science. Several of these projects are closely associated with major strategic thrusts of the Laboratory as described in Argonne`s Five Year Institutional Plan, although the scientific implications of the achieved results extend well beyond Laboratory plans and objectives. The projects supported by the Program are distributed across the major programmatic areas at Argonne as indicated in the Laboratory LDRD Plan for FY 1993.

  6. Argonne National Laboratory, High Energy Physics Division, semiannual report of research activities, July 1, 1989--December 31, 1989

    SciTech Connect

    Not Available

    1989-01-01

    This report discusses research being conducted at the Argonne National Laboratory in the following areas: Experimental High Energy Physics; Theoretical High Energy Physics; Experimental Facilities Research; Accelerator Research and Development; and SSC Detector Research and Development.

  7. Reduced enrichment for research and test reactors: Proceedings

    SciTech Connect

    Not Available

    1993-07-01

    The 15th annual Reduced Enrichment for Research and Test Reactors (RERTR) international meeting was organized by Ris{o} National Laboratory in cooperation with the International Atomic Energy Agency and Argonne National Laboratory. The topics of the meeting were the following: National Programs, Fuel Fabrication, Licensing Aspects, States of Conversion, Fuel Testing, and Fuel Cycle. Individual papers have been cataloged separately.

  8. Argonne National Laboratory High Energy Physics Division semiannual report of research activities, January 1, 1989--June 30, 1989

    SciTech Connect

    Not Available

    1989-01-01

    This paper discuss the following areas on High Energy Physics at Argonne National Laboratory: experimental program; theory program; experimental facilities research; accelerator research and development; and SSC detector research and development.

  9. Argonne National Laboratory annual report of Laboratory Directed Research and Development Program Activities FY 2009.

    SciTech Connect

    Office of the Director

    2010-04-09

    I am pleased to submit Argonne National Laboratory's Annual Report on its Laboratory Directed Research and Development (LDRD) activities for fiscal year 2009. Fiscal year 2009 saw a heightened focus by DOE and the nation on the need to develop new sources of energy. Argonne scientists are investigating many different sources of energy, including nuclear, solar, and biofuels, as well as ways to store, use, and transmit energy more safely, cleanly, and efficiently. DOE selected Argonne as the site for two new Energy Frontier Research Centers (EFRCs) - the Institute for Atom-Efficient Chemical Transformations and the Center for Electrical Energy Storage - and funded two other EFRCs to which Argonne is a major partner. The award of at least two of the EFRCs can be directly linked to early LDRD-funded efforts. LDRD has historically seeded important programs and facilities at the lab. Two of these facilities, the Advanced Photon Source and the Center for Nanoscale Materials, are now vital contributors to today's LDRD Program. New and enhanced capabilities, many of which relied on LDRD in their early stages, now help the laboratory pursue its evolving strategic goals. LDRD has, since its inception, been an invaluable resource for positioning the Laboratory to anticipate, and thus be prepared to contribute to, the future science and technology needs of DOE and the nation. During times of change, LDRD becomes all the more vital for facilitating the necessary adjustments while maintaining and enhancing the capabilities of our staff and facilities. Although I am new to the role of Laboratory Director, my immediate prior service as Deputy Laboratory Director for Programs afforded me continuous involvement in the LDRD program and its management. Therefore, I can attest that Argonne's program adhered closely to the requirements of DOE Order 413.2b and associated guidelines governing LDRD. Our LDRD program management continually strives to be more efficient. In addition to

  10. Argonne's 2012 Earth Day Event

    ScienceCinema

    None

    2016-07-12

    Argonne's 2012 Earth Day event drew crowds from across the laboratory. Argonne and U.S. Department of Energy employees toured booths and interactive displays set up by Argonne programs and clubs. Several of Argonne's partners participated, including U.S. Department of Energy, University of Chicago, Abri Credit Union, DuPage County Forest Preserve, DuPage Water Commission, PACE and Morton Arboretum. Argonne scientists and engineers also participated in a poster session, discussing their clean energy research.

  11. Argonne's 2012 Earth Day Event

    SciTech Connect

    2012-01-01

    Argonne's 2012 Earth Day event drew crowds from across the laboratory. Argonne and U.S. Department of Energy employees toured booths and interactive displays set up by Argonne programs and clubs. Several of Argonne's partners participated, including U.S. Department of Energy, University of Chicago, Abri Credit Union, DuPage County Forest Preserve, DuPage Water Commission, PACE and Morton Arboretum. Argonne scientists and engineers also participated in a poster session, discussing their clean energy research.

  12. Argonne nuclear pioneer: Leonard Koch

    SciTech Connect

    Koch, Leonard

    2012-01-01

    Leonard Koch joined Argonne National Laboratory in 1948. He helped design and build Experimental Breeder Reactor-1 (EBR-1), the first reactor to generate useable amounts of electricity from nuclear energy.

  13. Surveys of research in the Chemistry Division, Argonne National Laboratory

    SciTech Connect

    Grazis, B.M.

    1992-11-01

    Research reports are presented on reactive intermediates in condensed phase (radiation chemistry, photochemistry), electron transfer and energy conversion, photosynthesis and solar energy conversion, metal cluster chemistry, chemical dynamics in gas phase, photoionization-photoelectrons, characterization and reactivity of coal and coal macerals, premium coal sample program, chemical separations, heavy elements coordination chemistry, heavy elements photophysics/photochemistry, f-electron interactions, radiation chemistry of high-level wastes (gas generation in waste tanks), ultrafast molecular electronic devices, and nuclear medicine. Separate abstracts have been prepared. Accelerator activites and computer system/network services are also reported.

  14. Surveys of research in the Chemistry Division, Argonne National Laboratory

    SciTech Connect

    Grazis, B.M.

    1992-01-01

    Research reports are presented on reactive intermediates in condensed phase (radiation chemistry, photochemistry), electron transfer and energy conversion, photosynthesis and solar energy conversion, metal cluster chemistry, chemical dynamics in gas phase, photoionization-photoelectrons, characterization and reactivity of coal and coal macerals, premium coal sample program, chemical separations, heavy elements coordination chemistry, heavy elements photophysics/photochemistry, f-electron interactions, radiation chemistry of high-level wastes (gas generation in waste tanks), ultrafast molecular electronic devices, and nuclear medicine. Separate abstracts have been prepared. Accelerator activites and computer system/network services are also reported.

  15. Argonne National Laboratory Annual Report of Laboratory Directed Research and Development Program Activities for FY 1994

    SciTech Connect

    1995-02-25

    The purposes of Argonne's Laboratory Directed Research and Development (LDRD) Program are to encourage the development of novel concepts, enhance the Laboratory's R and D capabilities, and further the development of its strategic initiatives. Projects are selected from proposals for creative and innovative R and D studies which are not yet eligible for timely support through normal programmatic channels. Among the aims of the projects supported by the Program are establishment of engineering proof-of-principle; assessment of design feasibility for prospective facilities; development of an instrumental prototype, method, or system; or discovery in fundamental science. Several of these projects are closely associated with major strategic thrusts of the Laboratory as described in Argonne's Five-Year Institutional Plan, although the scientific implications of the achieved results extend well beyond Laboratory plans and objectives. The projects supported by the Program are distributed across the major programmatic areas at Argonne as indicated in the Laboratory's LDRD Plan for FY 1994. Project summaries of research in the following areas are included: (1) Advanced Accelerator and Detector Technology; (2) X-ray Techniques for Research in Biological and Physical Science; (3) Nuclear Technology; (4) Materials Science and Technology; (5) Computational Science and Technology; (6) Biological Sciences; (7) Environmental Sciences: (8) Environmental Control and Waste Management Technology; and (9) Novel Concepts in Other Areas.

  16. Argonne National Laboratory Annual Report of Laboratory Directed Research and Development program activities FY 2011.

    SciTech Connect

    2012-04-25

    As a national laboratory Argonne concentrates on scientific and technological challenges that can only be addressed through a sustained, interdisciplinary focus at a national scale. Argonne's eight major initiatives, as enumerated in its strategic plan, are Hard X-ray Sciences, Leadership Computing, Materials and Molecular Design and Discovery, Energy Storage, Alternative Energy and Efficiency, Nuclear Energy, Biological and Environmental Systems, and National Security. The purposes of Argonne's Laboratory Directed Research and Development (LDRD) Program are to encourage the development of novel technical concepts, enhance the Laboratory's research and development (R and D) capabilities, and pursue its strategic goals. projects are selected from proposals for creative and innovative R and D studies that require advance exploration before they are considered to be sufficiently developed to obtain support through normal programmatic channels. Among the aims of the projects supported by the LDRD Program are the following: establishment of engineering proof of principle, assessment of design feasibility for prospective facilities, development of instrumentation or computational methods or systems, and discoveries in fundamental science and exploratory development.

  17. Argonne National Laboratory Annual Report of Laboratory Directed Research and Development program activities FY 2010.

    SciTech Connect

    2012-04-25

    As a national laboratory Argonne concentrates on scientific and technological challenges that can only be addressed through a sustained, interdisciplinary focus at a national scale. Argonne's eight major initiatives, as enumerated in its strategic plan, are Hard X-ray Sciences, Leadership Computing, Materials and Molecular Design and Discovery, Energy Storage, Alternative Energy and Efficiency, Nuclear Energy, Biological and Environmental Systems, and National Security. The purposes of Argonne's Laboratory Directed Research and Development (LDRD) Program are to encourage the development of novel technical concepts, enhance the Laboratory's research and development (R and D) capabilities, and pursue its strategic goals. projects are selected from proposals for creative and innovative R and D studies that require advance exploration before they are considered to be sufficiently developed to obtain support through normal programmatic channels. Among the aims of the projects supported by the LDRD Program are the following: establishment of engineering proof of principle, assessment of design feasibility for prospective facilities, development of instrumentation or computational methods or systems, and discoveries in fundamental science and exploratory development.

  18. Reactor Safety Research Programs

    SciTech Connect

    Edler, S. K.

    1981-07-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from January 1 through March 31, 1981, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipeto- pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-ofcoolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  19. DOE/EA-1519: Environmental Assessment for the Proposed Decontamination and Decommissioning of the Zero Power Reactors (Building 315) at Argonne National Laboratory (April 2005)

    SciTech Connect

    N /A

    2005-04-30

    The U.S. Department of Energy (DOE) is proposing to decontaminate and decommission the Zero Power Reactor (ZPR) facilities located in Building 315 at Argonne National Laboratory (ANL) in Argonne, Illinois (Figure 1-1). The proposed action would occur in two phases: ZPR-6 would be the focus of Phase I and ZPR-9 would be the focus of Phase II. DOE has prepared this environmental assessment (EA) in accordance with the National Environmental Policy Act (NEPA), 42 U.S.C. {section} 4321 et seq., and applicable regulations (Title 40, Code of Federal Regulations [CFR] Parts 1500-1508 and 10 CFR Part 1021). This section describes the reactors and their current status.

  20. Reactor Safety Research Programs

    SciTech Connect

    Dotson, CW

    1980-08-01

    This document summarizes the work performed by Pacific Northwest laboratory from October 1 through December 31, 1979, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission. Evaluation of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibilty of determining structural graphite strength, evaluating the feasibilty of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include the loss-of-coolant accident simulation tests at the NRU reactor, Chalk River, Canada; the fuel rod deformation and post-accident coolability tests for the ESSOR Test Reactor Program, lspra, Italy; the blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and the experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  1. Argonne National Laboratory-East summary site environmental report for calendar year 2002.

    SciTech Connect

    Golchert, N. W.; Kolzow, R. G.

    2004-03-08

    Argonne performs research and development in many areas of science and technology. General fields of research at Argonne include, but are not limited to, biosciences, biotechnology, chemical engineering, chemistry, decision and information sciences, energy systems and technology, high energy physics, materials science, math and computer science, nuclear reactors, physics, and environmental science. Argonne is not, and never has been, a weapons laboratory. Several missions provide focus for Argonne scientists. Basic research helps better understand the world, and applied research helps protect and improve it. For example, the prairies of Argonne provide sites for environmental studies that provide valuable information about invader species and the food webs within ecosystems. Argonne also operates world-class research facilities, such as the Advanced Photon Source (APS), which is a national research facility funded by the U.S. Department of Energy (DOE). Scientists use high brilliance X-rays from the APS for basic and applied research in many fields. Argonne also seeks to ensure our energy future. Currently, scientists and engineers are developing cleaner and more efficient energy sources, such as fuel cells and advanced electric power generation. Argonne has spent much of its history on developing nuclear reactor technology. That research is now being applied to American and Soviet nuclear reactors to improve the safety and life of the reactors. Other Argonne research seeks to improve the way we manage our environment. For example, Argonne scientists created a new catalyst that could help carmakers eliminate 95 percent of nitrogen-oxide emitted by diesel engines by the year 2007. Research and development solutions such as these will help protect our ecosystems.

  2. Human-health effects of radium: an epidemiolgic perspective of research at Argonne National Laboratory

    SciTech Connect

    Stebbings, J.H.

    1982-01-01

    The topic of health effects of radium has recently been considerably broadened by the identification of multiple myeloma as a specific outcome of bone-seeking radionuclides, and by evidence that the incidence of breast cancer may be significantly increased by radium exposure. All soft-tissue tumors are now suspect, especially leukemias. Concepts of dose-response need to be broadened to include the concept of risk factors, or, if one prefers, of susceptible subgroups. Biological factors relating to radium uptake and retention require study, as do risk factors modifying risk of both the clasical tumors, osteosarcoma and nasal sinus/mastoid, and the more recently suspect soft-tissue tumors. The history, organization, and current research activities in epidemiology at Argonne National Laboratory are described, and findings of the last decade and a half reviewed. Plans for future research are briefly discussed.

  3. Argonne`s new Wakefield Test Facility

    SciTech Connect

    Simpson, J.D.

    1992-07-20

    The first phase of a high current, short bunch length electron beam research facility, the AWA, is near completion at Argonne. At the heart of the facility is a photocathode based electron gun and accelerating sections designed to deliver 20 MeV pulses with up to 100 nC per pulse and with pulse lengths of approximately 15 ps (fw). Using a technique similar to that originated at Argonne`s AATF facility, a separate weak probe pulse can be generated and used to diagnose wake effects produced by the intense pulses. Initial planned experiments include studies of plasma wakefields and dielectric wakefield devices, and expect to demonstrate large, useful accelerating gradients (> 100 MeV/m). Later phases of the facility will increase the drive bunch energy to more than 100 MeV to enable acceleration experiments up to the GeV range. Specifications, design details, and commissioning progress are presented.

  4. Reduced enrichment for research and test reactors: Proceedings

    SciTech Connect

    Not Available

    1988-05-01

    The international effort to develop new research reactor fuel materials and designs based on the use of low-enriched uranium, instead of highly-enriched uranium, has made much progress during the eight years since its inception. To foster direct communication and exchange of ideas among the specialist in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the ninth of a series which began in 1978. All previous meetings of this series are listed on the facing page. The focus of this meeting was on the LEU fuel demonstration which was in progress at the Oak Ridge Research (ORR) reactor, not far from where the meeting was held. The visit to the ORR, where a silicide LEU fuel with 4.8 g A/cm/sup 3/ was by then in routine use, illustrated how far work has progressed.

  5. Argonne National Laboratory, High Energy Physics Division: Semiannual report of research activities, July 1, 1986-December 31, 1986

    SciTech Connect

    Not Available

    1987-01-01

    This paper discusses the research activity of the High Energy Physics Division at the Argonne National Laboratory for the period, July 1986-December 1986. Some of the topics included in this report are: high resolution spectrometers, computational physics, spin physics, string theories, lattice gauge theory, proton decay, symmetry breaking, heavy flavor production, massive lepton pair production, collider physics, field theories, proton sources, and facility development. (LSP)

  6. Research reactor fork users manual

    SciTech Connect

    Hsue, S.T.; Menlove, H.O.; Bosler, G.E.; Dye, H.R.; Walton, G.; Halbig, J.K.; Siebelist, R.

    1993-11-01

    This manual describes the design features and operating characteristics of the research reactor fork. The system includes an ion chamber for gross gamma-ray counting, fission chambers for neutron counting, and a collimated high-resolution spectroscopy system for gamma-ray measurements. The neutron and ion chamber measurements are designed to be made underwater in spent-fuel cooling ponds. The neutron and gamma-ray detectors have been designed with high efficiencies to accommodate the relatively low emission rates of neutrons and gamma rays from low-burnup, research-type reactor fuel. This manual presents the design, performance, and test results for the system.

  7. Studies of acute and chronic radiation injury at the Biological and Medical Research Division, Argonne National Laboratory, 1970-1992: The JANUS Program Survival and Pathology Data

    SciTech Connect

    Grahn, D.; Wright, B.J.; Carnes, B.A.; Williamson, F.S.; Fox, C.

    1995-02-01

    A research reactor for exclusive use in experimental radiobiology was designed and built at Argonne National Laboratory in the 1960`s. It was located in a special addition to Building 202, which housed the Division of Biological and Medical Research. Its location assured easy access for all users to the animal facilities, and it was also near the existing gamma-irradiation facilities. The water-cooled, heterogeneous 200-kW(th) reactor, named JANUS, became the focal point for a range of radiobiological studies gathered under the rubic of {open_quotes}the JANUS program{close_quotes}. The program ran from about 1969 to 1992 and included research at all levels of biological organization, from subcellular to organism. More than a dozen moderate- to large-scale studies with the B6CF{sub 1} mouse were carried out; these focused on the late effects of whole-body exposure to gamma rays or fission neutrons, in matching exposure regimes. In broad terms, these studies collected data on survival and on the pathology observed at death. A deliberate effort was made to establish the cause of death. This archieve describes these late-effects studies and their general findings. The database includes exposure parameters, time of death, and the gross pathology and histopathology in codified form. A series of appendices describes all pathology procedures and codes, treatment or irradiation codes, and the manner in which the data can be accessed in the ORACLE database management system. A series of tables also presents summaries of the individual experiments in terms of radiation quality, sample sizes at entry, mean survival times by sex, and number of gross pathology and histopathology records.

  8. Proceedings of the 1988 International Meeting on Reduced Enrichment for Research and Test Reactors

    SciTech Connect

    Not Available

    1993-07-01

    The international effort to develop and implement new research reactor fuels utilizing low-enriched uranium, instead of highly- enriched uranium, continues to make solid progress. This effort is the cornerstone of a widely shared policy aimed at reducing, and possibly eliminating, international traffic in highly-enriched uranium and the nuclear weapon proliferation concerns associated with this traffic. To foster direct communication and exchange of ideas among the specialists in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the eleventh of a series which began 1978. Individual papers presented at the meeting have been cataloged separately.

  9. Technical basis in support of the conversion of the University of Missouri Research Reactor (MURR) core from highly-enriched to low-enriched uranium - core neutron physics

    SciTech Connect

    Stillman, J.; Feldman, E.; Foyto, L; Kutikkad, K; McKibben, J C; Peters, N.; Stevens, J.

    2012-09-01

    This report contains the results of reactor design and performance for conversion of the University of Missouri Research Reactor (MURR) from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL) and the MURR Facility. The core conversion to LEU is being performed with financial support of the U. S. government.

  10. Green Supercomputing at Argonne

    SciTech Connect

    Pete Beckman

    2009-11-18

    Pete Beckman, head of Argonne's Leadership Computing Facility (ALCF) talks about Argonne National Laboratory's green supercomputing—everything from designing algorithms to use fewer kilowatts per operation to using cold Chicago winter air to cool the machine more efficiently.

  11. Fuel elements of research reactors in China

    SciTech Connect

    Yongmao, Z.; Dianshan, C.; Guofang, Q.

    1988-01-01

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

  12. Integration of improved decontamination and characterization technologies in the decommissioning of the CP-5 research reactor

    SciTech Connect

    Bhattacharyya, S. K.; Boing, L. E.

    2000-02-17

    The aging of research reactors worldwide has resulted in a heightened awareness in the international technical decommissioning community of the timeliness to review and address the needs of these research institutes in planning for and eventually performing the decommissioning of these facilities. By using the reactors already undergoing decommissioning as test beds for evaluating enhanced or new/innovative technologies for decommissioning, it is possible that new techniques could be made available for those future research reactor decommissioning projects. Potentially, the new technologies will result in: reduced radiation doses to the work force, larger safety margins in performing decommissioning and cost and schedule savings to the research institutes in performing the decommissioning of these facilities. Testing of these enhanced technologies for decontamination, dismantling, characterization, remote operations and worker protection are critical to furthering advancements in the technical specialty of decommissioning. Furthermore, regulatory acceptance and routine utilization for future research reactor decommissioning will be assured by testing and developing these technologies in realistically contaminated environments prior to use in the research reactors. The decommissioning of the CP-5 Research Reactor is currently in the final phase of dismantlement. In this paper the authors present results of work performed at Argonne National Laboratory (ANL) in the development, testing and deployment of innovative and/or enhanced technologies for the decommissioning of research reactors.

  13. Research Program of a Super Fast Reactor

    SciTech Connect

    Oka, Yoshiaki; Ishiwatari, Yuki; Liu, Jie; Terai, Takayuki; Nagasaki, Shinya; Muroya, Yusa; Abe, Hiroaki; Akiba, Masato; Akimoto, Hajime; Okumura, Keisuke; Akasaka, Naoaki; GOTO, Shoji

    2006-07-01

    Research program of a supercritical-pressure light water cooled fast reactor (Super Fast Reactor) is funded by MEXT (Ministry of Education, Culture, Sports, Science and Technology) in December 2005 as one of the research programs of Japanese NERI (Nuclear Energy Research Initiative). It consists of three programs. (1) development of Super Fast Reactor concept; (2) thermal-hydraulic experiments; (3) material developments. The purpose of the concept development is to pursue the advantage of high power density of fast reactor over thermal reactors to achieve economic competitiveness of fast reactor for its deployment without waiting for exhausting uranium resources. Design goal is not breeding, but maximizing reactor power by using plutonium from spent LWR fuel. MOX will be the fuel of the Super Fast Reactor. Thermal-hydraulic experiments will be conducted with HCFC22 (Hydro chlorofluorocarbons) heat transfer loop of Kyushu University and supercritical water loop at JAEA. Heat transfer data including effect of grid spacers will be taken. The critical flow and condensation of supercritical fluid will be studied. The materials research includes the development and testing of austenitic stainless steel cladding from the experience of PNC1520 for LMFBR. Material for thermal insulation will be tested. SCWR (Supercritical-Water Cooled Reactor) of GIF (Generation-4 International Forum) includes both thermal and fast reactors. The research of the Super Fast Reactor will enhance SCWR research and the data base. The research period will be until March 2010. (authors)

  14. United States Domestic Research Reactor Infrastrucutre TRIGA Reactor Fuel Support

    SciTech Connect

    Douglas Morrell

    2011-03-01

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

  15. Argonne's atlas control system upgrade.

    SciTech Connect

    Munson, F.; Quock, D.; Chapin, B.; Figueroa, J.

    1999-09-27

    The ATLAS facility (Argonne Tandem-Linac Accelerator System) is located at the Argonne National Laboratory. The facility is a tool used in nuclear and atomic physics research, which focuses primarily on heavy-ion physics. The accelerator as well as its control system are evolutionary in nature, and consequently, continue to advance. In 1998 the most recent project to upgrade the ATLAS control system was completed. This paper briefly reviews the upgrade, and summarizes the configuration and features of the resulting control system.

  16. Mira: Argonne's 10-petaflops supercomputer

    ScienceCinema

    Papka, Michael; Coghlan, Susan; Isaacs, Eric; Peters, Mark; Messina, Paul

    2016-07-12

    Mira, Argonne's petascale IBM Blue Gene/Q system, ushers in a new era of scientific supercomputing at the Argonne Leadership Computing Facility. An engineering marvel, the 10-petaflops supercomputer is capable of carrying out 10 quadrillion calculations per second. As a machine for open science, any researcher with a question that requires large-scale computing resources can submit a proposal for time on Mira, typically in allocations of millions of core-hours, to run programs for their experiments. This adds up to billions of hours of computing time per year.

  17. Applied Mathematical Sciences research at Argonne, October 1, 1978-March 31, 1980

    SciTech Connect

    Pieper, G. W.

    1980-01-01

    This report reviews the research activities of the Applied Mathematical Sciences Section for the period October 1, 1978, through March 31, 1980. The body of the report discusses various projects carried out in four major areas of research: applied analysis, computational mathematics, software engineering, and software clinics. Information on section staff, visitors, workshops, and seminars is found in the appendices. Descriptions of individual research topics are very brief.

  18. Green Supercomputing at Argonne

    ScienceCinema

    Beckman, Pete

    2016-07-12

    Pete Beckman, head of Argonne's Leadership Computing Facility (ALCF) talks about Argonne National Laboratory's green supercomputing—everything from designing algorithms to use fewer kilowatts per operation to using cold Chicago winter air to cool the machine more efficiently. Argonne was recognized for green computing in the 2009 HPCwire Readers Choice Awards. More at http://www.anl.gov/Media_Center/News/2009/news091117.html Read more about the Argonne Leadership Computing Facility at http://www.alcf.anl.gov/

  19. Reactivity Transients in Nuclear Research Reactors

    SciTech Connect

    2015-01-01

    Version 01 AIREMOD-RR is a point kinetics code which can simulate fast transients in nuclear research reactor cores. It can also be used for theoretical reactor dynamics studies. It is used for research reactor kinetic analysis and provides a point neutron kinetic capability. The thermal hydraulic behavior is governed by a one-dimensional heat balance equation. The calculations are restricted to a single equivalent unit cell which consists of fuel, clad and coolant.

  20. Utilisation of British University Research Reactors.

    ERIC Educational Resources Information Center

    Duncton, P. J.; And Others

    British experience relating to the employment of university research reactors and subcritical assemblies in the education of nuclear scientists and technologists, in the training of reactor operators and for fundamental pure and applied research in this field is reviewed. The facilities available in a number of British universities and the uses…

  1. Research reactor job analysis - A project description

    SciTech Connect

    Yoder, John

    1988-07-01

    Addressing the need of the improved training in nuclear industry, nuclear utilities established training program guidelines based on Performance-Based Training (PBT) concepts. The comparison of commercial nuclear power facilities with research and test reactors owned by the U.S. Department of Energy (DOE), made in an independent review of personnel selection, training, and qualification requirements for DOE-owned reactors pointed out that the complexity of the most critical tasks in research reactors is less than that in power reactors. The U.S. Department of Energy (DOE) started a project by commissioning Oak Ridge Associated Universities (ORAU) to conduct a job analysis survey of representative research reactor facilities. The output of the project consists of two publications: Volume 1 - Research Reactor Job Analysis: Overview, which contains an Introduction, Project Description, Project Methodology,, and. An Overview of Performance-Based Training (PBT); and Volume 2 - Research Reactor Job Analysis: Implementation, which contains Guidelines for Application of Preliminary Task Lists and Preliminary Task Lists for Reactor Operators and Supervisory Reactor Operators.

  2. Green Supercomputing at Argonne

    ScienceCinema

    Pete Beckman

    2016-07-12

    Pete Beckman, head of Argonne's Leadership Computing Facility (ALCF) talks about Argonne National Laboratory's green supercomputing—everything from designing algorithms to use fewer kilowatts per operation to using cold Chicago winter air to cool the machine more efficiently.

  3. The Advanced Photon Source: A national synchrotron radiation research facility at Argonne National Laboratory

    SciTech Connect

    1995-10-01

    The vision of the APS sprang from prospective users, whose unflagging support the project has enjoyed throughout the decade it has taken to make this facility a reality. Perhaps the most extraordinary aspect of synchrotron radiation research, is the extensive and diverse scientific makeup of the user community. From this primordial soup of scientists exchanging ideas and information, come the collaborative and interdisciplinary accomplishments that no individual alone could produce. So, unlike the solitary Roentgen, scientists are engaged in a collective and dynamic enterprise with the potential to see and understand the structures of the most complex materials that nature or man can produce--and which underlie virtually all modern technologies. This booklet provides scientists and laymen alike with a sense of both the extraordinary history of x-rays and the knowledge they have produced, as well as the potential for future discovery contained in the APS--a source a million million times brighter than the Roentgen tube.

  4. Advanced Reciprocating Engine Systems (ARES) Research at Argonne National Laboratory. A Report

    SciTech Connect

    Gupta, Sreenath; Biruduganti, Muni; Bihari, Bipin; Sekar, Raj

    2014-08-01

    The goals of these experiments were to determine the potential of employing spectral measurements to deduce combustion metrics such as HRR, combustion temperatures, and equivalence ratios in a natural gas-fired reciprocating engine. A laser-ignited, natural gas-fired single-cylinder research engine was operated at various equivalence ratios between 0.6 and 1.0, while varying the EGR levels between 0% and maximum to thereby ensure steady combustion. Crank angle-resolved spectral signatures were collected over 266-795 nm, encompassing chemiluminescence emissions from OH*, CH*, and predominantly by CO2* species. Further, laser-induced gas breakdown spectra were recorded under various engine operating conditions.

  5. Proceedings of the 1990 International Meeting on Reduced Enrichment for Research and Test Reactors

    SciTech Connect

    Not Available

    1993-07-01

    The global effort to reduce, and possibly, eliminate the international traffic in highly-enriched uranium caused by its use in research reactors requires extensive cooperation and free exchange of information among all participants. To foster this free exchange of information, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the thirteenth of a series which began in 1978. The common effort brought together, past, a large number of specialists from many countries. On hundred twenty-three participants from 26 countries, including scientists, reactor operators, and personnel from commercial fuel suppliers, research centers, and government organizations, convened in Newport, Rhode Island to discuss their results, their activities, and their plans relative to converting research reactors to low-enriched fuels. As more and more reactors convert to the use of low-enriched uranium, the emphasis of our effort has begun to shift from research and development to tasks more directly related to implementation of the new fuels and technologies that have been developed, and to refinements of those fuels and technologies. It is appropriate, for this reason, that the emphasis of this meeting was placed on safety and on conversion experiences. This individual papers in this report have been cataloged separately.

  6. Status of reduced enrichment programs for research reactors in Japan

    SciTech Connect

    Kanda, Keiji; Nishihara, Hedeaki; Shirai, Eiji; Oyamada, Rokuro; Sanokawa, Konomo

    1997-08-01

    The reduced enrichment programs for the JRR-2, JRR-3, JRR-4 and JMTR of Japan Atomic Energy Research Institute (JAERI), and the KUR of Kyoto University Research Reactor Institute (KURRI) have been partially completed and are mostly still in progress under the Joint Study Programs with Argonne National Laboratory (ANL). The JMTR and JRR-2 have been already converted to use MEU aluminide fuels in 1986 and 1987, respectively. The operation of the upgraded JRR-3(JRR-3M) has started in March 1990 with the LEU aluminide fuels. Since May 1992, the two elements have been inserted in the KUR. The safety review application for the full core conversion to use LEU silicide in the JMTR was approved in February 1992 and the conversion has been done in January 1994. The Japanese Government approved a cancellation of the KUHFR Project in February 1991, and in April 1994 the U.S. Government gave an approval to utilize HEU in the KUR instead of the KUHFR. Therefore, the KUR will be operated with HEU fuel until 2001. Since March 1994, Kyoto University is continuing negotiation with UKAEA Dounreay on spent fuel reprocessing and blending down of recovered uranium, in addition to that with USDOE.

  7. Fuel elements of research reactor CM

    SciTech Connect

    Kozlov, A.V.; Morozov, A.V.; Vatulin, A.V.; Ershov, S.A.

    2013-07-01

    In 1961 the CM research reactor was commissioned at the Research Institute of Atomic Reactors (Dimitrovgrad, Russia), it was intended to carry on investigations and the production of transuranium nuclides. The reactor is of a tank type. Original fuel assembly contained plate fuels that were spaced with vanes and corrugated bands. Nickel was used as a cladding material, fuel meat was produced from UO{sub 2} + electrolytic nickel composition. Fuel plates have been replaced by self-spacing cross-shaped dispersion fuels clad in stainless steel. In 2005 the reactor was updated. The purpose of this updating was to increase the quantity of irradiation channels in the reactor core and to improve the neutron balance. The updating was implemented at the expense of 20 % reduction in the quantity of fuel elements in the core which released a space for extra channels and decreased the mass of structural materials in the core. The updated reactor is loaded with modified standard fuel elements with 20 % higher uranium masses. At the same time stainless steel in fuel assembly shrouds was substituted by zirconium alloy. Today in progress are investigations and work to promote the second stage of reactor updating that involve developments of cross-shaped fuel elements having low neutron absorption matrix materials. This article gives an historical account of the design and main technical changes that occurred for the CM reactor since its commissioning.

  8. Gaseous fuel nuclear reactor research

    NASA Technical Reports Server (NTRS)

    Schwenk, F. C.; Thom, K.

    1975-01-01

    Gaseous-fuel nuclear reactors are described; their distinguishing feature is the use of fissile fuels in a gaseous or plasma state, thereby breaking the barrier of temperature imposed by solid-fuel elements. This property creates a reactor heat source that may be able to heat the propellant of a rocket engine to 10,000 or 20,000 K. At this temperature level, gas-core reactors would provide the breakthrough in propulsion needed to open the entire solar system to manned and unmanned spacecraft. The possibility of fuel recycling makes possible efficiencies of up to 65% and nuclear safety at reduced cost, as well as high-thrust propulsion capabilities with specific impulse up to 5000 sec.

  9. Supply of enriched uranium for research reactors

    SciTech Connect

    Mueller, H.

    1997-08-01

    Since the RERTR-meeting In Newport/USA in 1990 the author delivered a series of papers in connection with the fuel cycle for research reactors dealing with its front-end. In these papers the author underlined the need for unified specifications for enriched uranium metal suitable for the production of fuel elements and made proposals with regard to the re-use of in Europe reprocessed highly enriched uranium. With regard to the fuel cycle of research reactors the research reactor community was since 1989 more concentrating on the problems of its back-end since the USA stopped the acceptance of spent research reactor fuel on December 31, 1988. Now, since it is apparent that these back-end problem have been solved by AEA`s ability to reprocess and the preparedness of the USA to again accept physically spent research reactor fuel the author is focusing with this paper again on the front-end of the fuel cycle on the question whether there is at all a safe supply of low and high enriched uranium for research reactors in the future.

  10. Argonne's new Wakefield Test Facility

    SciTech Connect

    Simpson, J.D.

    1992-07-20

    The first phase of a high current, short bunch length electron beam research facility, the AWA, is near completion at Argonne. At the heart of the facility is a photocathode based electron gun and accelerating sections designed to deliver 20 MeV pulses with up to 100 nC per pulse and with pulse lengths of approximately 15 ps (fw). Using a technique similar to that originated at Argonne's AATF facility, a separate weak probe pulse can be generated and used to diagnose wake effects produced by the intense pulses. Initial planned experiments include studies of plasma wakefields and dielectric wakefield devices, and expect to demonstrate large, useful accelerating gradients (> 100 MeV/m). Later phases of the facility will increase the drive bunch energy to more than 100 MeV to enable acceleration experiments up to the GeV range. Specifications, design details, and commissioning progress are presented.

  11. Argonne-West facility requirements for a radioactive waste treatment demonstration

    SciTech Connect

    Dwight, C.C.; Felicione, F.S.; Black, D.B.; Kelso, R.B.; McClellan, G.C.

    1995-03-01

    At Argonne National Laboratory-West (ANL-W), near Idaho Falls, Idaho, facilities that were originally constructed to support the development of liquid-metal reactor technology are being used and/or modified to meet the environmental and waste management research needs of DOE. One example is the use of an Argonne-West facility to conduct a radioactive waste treatment demonstration through a cooperative project with Science Applications International Corporation (SAIC) and Lockheed Idaho Technologies Company. The Plasma Hearth Process (PBP) project will utilize commercially-adapted plasma arc technology to demonstrate treatment of actual mixed waste. The demonstration on radioactive waste will be conducted at Argonne`s Transient Reactor Test Facility (TREAT). Utilization of an existing facility for a new and different application presents a unique set of issues in meeting applicable federal state, and local requirements as well as the additional constraints imposed by DOE Orders and ANL-W site requirements. This paper briefly describes the PHP radioactive demonstrations relevant to the interfaces with the TREAT facility. Safety, environmental design, and operational considerations pertinent to the PHP radioactive demonstration are specifically addressed herein. The personnel equipment, and facility interfaces associated with a radioactive waste treatment demonstration are an important aspect of the demonstration effort. Areas requiring significant effort in preparation for the PBP Project being conducted at the TREAT facility include confinement design, waste handling features, and sampling and analysis considerations. Information about the facility in which a radioactive demonstration will be conducted, specifically Argonne`s TREAT facility in the case of PHP, may be of interest to other organizations involved in developing and demonstrating technologies for mixed waste treatment.

  12. Strengthening IAEA Safeguards for Research Reactors

    SciTech Connect

    Reid, Bruce D.; Anzelon, George A.; Budlong-Sylvester, Kory

    2016-09-01

    During their December 10-11, 2013, workshop in Grenoble France, which focused on the history and future of safeguarding research reactors, the United States, France and the United Kingdom (UK) agreed to conduct a joint study exploring ways to strengthen the IAEA’s safeguards approach for declared research reactors. This decision was prompted by concerns about: 1) historical cases of non-compliance involving misuse (including the use of non-nuclear materials for production of neutron generators for weapons) and diversion that were discovered, in many cases, long after the violations took place and as part of broader pattern of undeclared activities in half a dozen countries; 2) the fact that, under the Safeguards Criteria, the IAEA inspects some reactors (e.g., those with power levels under 25 MWt) less than once per year; 3) the long-standing precedent of States using heavy water research reactors (HWRR) to produce plutonium for weapons programs; 4) the use of HEU fuel in some research reactors; and 5) various technical characteristics common to some types of research reactors that could provide an opportunity for potential proliferators to misuse the facility or divert material with low probability of detection by the IAEA. In some research reactors it is difficult to detect diversion or undeclared irradiation. In addition, infrastructure associated with research reactors could pose a safeguards challenge. To strengthen the effectiveness of safeguards at the State level, this paper advocates that the IAEA consider ways to focus additional attention and broaden its safeguards toolbox for research reactors. This increase in focus on the research reactors could begin with the recognition that the research reactor (of any size) could be a common path element on a large number of technically plausible pathways that must be considered when performing acquisition pathway analysis (APA) for developing a State Level Approach (SLA) and Annual Implementation Plan (AIP). To

  13. Reactor operations: Brookhaven Medical Research Reactor, Brookhaven High Flux Beam Reactor. Informal report, June 1995

    SciTech Connect

    1995-06-01

    Part one of this report gives the operating history of the Brookhaven Medical Research Reactor for the month of June. Also included are the BMRR technical safety surveillance requirements record and the summary of BMRR irradiations for the month. Part two gives the operating histories of the Brookhaven High Flux Beam Reactor and the Cold Neutron Facility at HFBR for June. Also included are the HFBR technical safety surveillance requirements record and the summary of HFBR irradiations for the month.

  14. Corrosion Minimization for Research Reactor Fuel

    SciTech Connect

    Eric Shaber; Gerard Hofman

    2005-06-01

    Existing university research reactors are being converted to use low-enriched uranium fue to eliminate the use of highly-enriched uranium. These conversions require increases in fuel loading that will result in the use of elements with more fuel plates, resulting in a net decrease in the water annulus between fuel plates. The proposed decrease in the water annulus raises questions about the requirements and stability of the surface hydroxide on the aluminum fuel cladding and the potential for runaway corrosion resulting in fuel over-temperature incidents. The Nuclear Regulatory Commission (NRC), as regulator for these university reactors, must ensure that proposed fuel modifications will not result in any increased risk or hazard to the reactor operators or the public. This document reviews the characteristics and behavior of aluminum hydroxides, analyzes the drivers for fuel plate corrosion, reviews relevant historical incidents, and provides recommendations on fuel design, surface treatment, and reactor operational practices to avoid corrosion issues.

  15. Reactor pulse repeatability studies at the annular core research reactor

    SciTech Connect

    DePriest, K.R.; Trinh, T.Q.; Luker, S. M.

    2011-07-01

    The Annular Core Research Reactor (ACRR) at Sandia National Laboratories is a water-moderated pool-type reactor designed for testing many types of objects in the pulse and steady-state mode of operations. Personnel at Sandia began working to improve the repeatability of pulse operations for experimenters in the facility. The ACRR has a unique UO{sub 2}-BeO fuel that makes the task of producing repeatable pulses difficult with the current operating procedure. The ACRR produces a significant quantity of photoneutrons through the {sup 9}Be({gamma}, n){sup 8}Be reaction in the fuel elements. The photoneutrons are the result of the gammas produced during fission and in fission product decay, so their production is very much dependent on the reactor power history and changes throughout the day/week of experiments in the facility. Because the photoneutrons interfere with the delayed-critical measurements required for accurate pulse reactivity prediction, a new operating procedure was created. The photoneutron effects at delayed critical are minimized when using the modified procedure. In addition, the pulse element removal time is standardized for all pulse operations with the modified procedure, and this produces less variation in reactivity removal times. (authors)

  16. Trends in fusion reactor safety research

    SciTech Connect

    Herring, J.S.; Holland, D.F.; Piet, S.J.

    1991-01-01

    Fusion has the potential to be an attractive energy source. From the safety and environmental perspective, fusion must avoid concerns about catastrophic accidents and unsolvable waste disposal. In addition, fusion must achieve an acceptable level of risk from operational accidents that result in public exposure and economic loss. Finally, fusion reactors must control routine radioactive effluent, particularly tritium. Major progress in achieving this potential rests on development of low-activation materials or alternative fuels. The safety and performance of various material choices and fuels for commercial fusion reactors can be investigated relatively inexpensively through reactor design studies. These studies bring together experts in a wide range of backgrounds and force the group to either agree on a reactor design or identify areas for further study. Fusion reactors will be complex with distributed radioactive inventories. The next generation of experiments will be critical in demonstrating that acceptable levels of safe operation can be achieved. These machines will use materials which are available today and for which a large database exists (e.g. for 316 stainless steel). Researchers have developed a good understanding of the risks associated with operation of these devices. Specifically, consequences from coolant system failures, loss of vacuum events, tritium releases, and liquid metal reactions have been studied. Recent studies go beyond next step designs and investigate commercial reactor concerns including tritium release and liquid metal reactions. 18 refs.

  17. Nuclear reactors for research and radioisotope production in Argentina

    SciTech Connect

    Duran, H.H.

    1981-01-01

    In Argentina, the construction, operation, and use of research and radioisotope production reactors is and has been an important method of personnel preparation for the nuclear power program. Moreover, it is a very suitable means for technology transfer to countries developing their own nuclear programs. At present, the following research reactors are in operation in Argentina: Argentine Reactor 0 (RA-0); Argentine Reactor 1 (RA-1); Argentine Reactor 2 (RA-2); Argentine Reactor 3 (RA-3); Argentine Reactor 4 (RA-4). The Argentine Reactor 6 (RA-6), under construction, should reach criticality in 1981.

  18. Preliminary study on new configuration with LEU fuel assemblies for the Dalat nuclear research reactor

    SciTech Connect

    Van Lam Pham; Vinh Vinh Le; Ton Nghiem Huynh; Ba Vien Luong; Kien Cuong Nguyen

    2008-07-15

    The fuel conversion of the Dalat Nuclear Research Reactor (DNRR) is being realized. The DNRR is a pool type research reactor which was reconstructed from the 250 kW TRIGA- MARK II reactor. The reconstructed reactor attained its nominal power of 500 kW in February 1984. According to the results of design and safety analyses performed by the joint study between RERTR Program at Argonne National Laboratory (ANL) and Vietnam Atomic Energy Commission (VAEC) the mixed core of irradiated HEU and new LEU WWR-M2 fuel assemblies will be created soon. This paper presents the results of preliminary study on new configuration with only LEU fuel assemblies for the DNRR. The codes MCNP, REBUS and VARI3D are used to calculate neutron flux performance in irradiation positions and kinetics parameters. The idea of change of Beryllium rod reloading enables to get working configuration assured shutdown margin, thermal-hydraulic safety and increase in thermal neutron flux in neutron trap at the center of DNRR active core. (author)

  19. Dismantling the nuclear research reactor Thetis

    SciTech Connect

    Michiels, P.

    2013-07-01

    The research reactor Thetis, in service since 1967 and stopped in 2003, is part of the laboratories of the institution of nuclear science of the University of Ghent. The reactor, of the pool-type, was used as a neutron-source for the production of radio-isotopes and for activation analyses. The reactor is situated in a water pool with inner diameter of 3 m. and a depth of 7.5 m. The reactor core is situated 5.3 m under water level. Besides the reactor, the pool contains pneumatic loops, handling tools, graphite blocks for neutron moderation and other experimental equipment. The building houses storage rooms for fissile material and sources, a pneumatic circuit for transportation of samples, primary and secondary cooling circuits, water cleaning resin circuits, a ventilation system and other necessary devices. Because of the experimental character of the reactor, laboratories with glove boxes and other tools were needed and are included in the dismantling program. The building is in 3 levels with a crawl-space. The ground-floor contains the ventilation installation, the purification circuits with tanks, cooling circuits and pneumatic transport system. On the first floor, around the reactor hall, the control-room, visiting area, end-station for pneumatic transport, waste-storage room, fuel storage room and the labs are located. The second floor contains a few laboratories and end stations of the two high speed transfer tubes. The lowest level of the pool is situated under ground level. The reactor has been operated at a power of 150 kW and had a max operating power of 250 kW. Belgoprocess has been selected to decommission the reactor, the labs, storage halls and associated circuits to free release the building for conventional reuse and for the removal of all its internals as legal defined. Besides the dose-rate risk and contamination risk, there is also an asbestos risk of contamination. During construction of the installation, asbestos-containing materials were

  20. Running WASP at Argonne

    SciTech Connect

    Huber, C.C.

    1981-01-01

    The WASP model was initially implemented at Argonne for the International Training course on Electric System Planning being conducted at Argonne. This implementation was done with special consideration to course participants who are unfamiliar with WASP and with the computer system they use during the course. Cataloged Procedures were developed for this purpose. The procedures simplify using WASP and enable participants to quickly start using WASP with a minimum of training. Within the procedures, features were added that enhance WASP. These features include a formatted printout of WASP input data and a historical log of all runs and inut data used. For the RENAME step, an alternate method is presented, with special comment concerning the WASP3 release.

  1. A neutronic feasibility study for LEU conversion of the Brookhaven Medical Research Reactor (BMRR).

    SciTech Connect

    Hanan, N. A.

    1998-01-14

    A neutronic feasibility study for converting the Brookhaven Medical Research Reactor from HEU to LEU fuel was performed at Argonne National Laboratory in cooperation with Brookhaven National Laboratory. Two possible LEU cores were identified that would provide nearly the same neutron flux and spectrum as the present HEU core at irradiation facilities that are used for Boron Neutron Capture Therapy and for animal research. One core has 17 and the other has 18 LEU MTR-type fuel assemblies with uranium densities of 2.5g U/cm{sup 3} or less in the fuel meat. This LEU fuel is fully-qualified for routine use. Thermal hydraulics and safety analyses need to be performed to complete the feasibility study.

  2. Neutron scattering at Australia's replacement research reactor

    NASA Astrophysics Data System (ADS)

    Robinson, R. A.; Kennedy, S. J.

    2002-01-01

    On August 25 1999, the Australian government gave final approval to build a research reactor to replace the existing HIFAR reactor at Lucas Heights. The replacement reactor, which will commence operation in 2005, will be multipurpose in function, with capabilities for both neutron-beam research and radioisotope production. Regarding beams, cold and thermal neutron sources are to be installed and the intent is to use supermirror guides, with coatings with critical angles up to 3 times that of natural Ni, to transport cold and thermal neutron beams into a large modern guide hall. The reactor and all the associated infrastructure, with the exception of the neutron beam instruments, is to be built by INVAP, SE and subcontractors in a turnkey contract. The goal is to have at least eight leading-edge neutron-beam instruments ready in 2005, and they will be developed by ANSTO and other contracted organisations, in consultation with the Australian user community and interested overseas parties. A review of the planned scientific capabilities, a description of the facility and a status report on the activities so far is given.

  3. Measuring the productivity of university research reactors

    SciTech Connect

    Voth, M.H.

    1989-11-01

    University Research Reactors (URRs) on 33 campuses in the United States provide valuable contributions to academic instruction and research programs. In most cases, there are no alternative diagnostic techniques to supplant the need for a reactor and associated facilities. Since URRs constitute a major financial commitment, it is important that they be operated in a productive manner. Productivity may be defined as the sum of new knowledge generated, existing knowledge transferred to others, and analytical services provided to assist in the generation of new knowledge; another definition of productivity is this sum expressed as a function of the cost incurred. In either case, a consistent measurement is difficult and more qualitative than quantitative. A uniform reporting system has been proposed that defines simplified categories through which meaningful comparisons can be performed.

  4. Neutron scattering at the OPAL research reactor

    NASA Astrophysics Data System (ADS)

    McIntyre, Garry J.; Holden, Peter J.

    2016-09-01

    The current suite of 14 neutron scattering instruments at the multipurpose OPAL research reactor is described. All instruments have been constructed following best practice, using state-of-the-art components and in close consultation with the regional user base. First results from the most recently commissioned instruments match their design performance parameters. Selected recent scientific highlights illustrate some unique combinations of instrumentation and the regional flavour of topical applications.

  5. Results from the Argonne, Los Alamos, JAERI collaboration

    SciTech Connect

    Meadows, J.; Smith, D.; Greenwood, L.; Haight, R.; Ikeda, Y.; Konno, C.

    1993-07-01

    Four sample packets containing elemental Ti, Fe, Ni, Cu, Nb, Ag, Eu, Tb and Hf have been irradiated in three distinct accelerator neutron fields, at Argonne National Laboratory and Los Alamos National Laboratory, USA, and Japan Atomic Energy Research Institute, Tokai, Japan. The acquired experimental data include differential cross sections and integral cross sections for the continuum neutron spectrum produced by 7-MeV deuterons incident on thick Be-metal target. The U-238(n,f) cross section was also measured at 10.3 MeV as a consistency check on the experimental technique. This the third progress report on a project which has been carried out under the auspices of an IAEA Coordinated Research Program entitled ``Activation Cross Sections for the Generation Of Long-lived Radionuclides of Importance in Fusion Reactor Technology``. The present report provides the latest results from this work. Comparison is made between the 14.7-MeV cross-section values obtained from the separate investigations at Argonne and JAERI. Generally, good agreement observed within the experimental errors when consistent sample parameters, radioactivity decay data and reference cross values are employed. A comparison is also made between the experimental results and those derived from calculations using a nuclear model. Experimental neutron information on the Be(d,n) neutron spectrum was incorporated in the comparisons for the integral results. The agreement is satisfactory considering the various uncertainties that are involved.

  6. Change in argonne national laboratory: a case study.

    PubMed

    Mozley, A

    1971-10-01

    Despite traditional opposition to change within an institution and the known reluctance of an "old guard" to accept new managerial policies and techniques, the reactions suggested in this study go well beyond the level of a basic resistance to change. The response, indeed, drawn from a random sampling of Laboratory scientific and engineering personnel, comes close to what Philip Handler has recently described as a run on the scientific bank in a period of depression (1, p. 146). It appears that Argonne's apprehension stems less from the financial cuts that have reduced staff and diminished programs by an annual 10 percent across the last 3 fiscal years than from the administrative and conceptual changes that have stamped the institution since 1966. Administratively, the advent of the AUA has not forged a sense of collaborative effort implicit in the founding negotiations or contributed noticeably to increasing standards of excellence at Argonne. The AUA has, in fact, yet to exercise the constructive powers vested in them by the contract of reviewing and formulating long-term policy on the research and reactor side. Additionally, the University of Chicago, once the single operator, appears to have forfeited some of the trust and understanding that characterized the Laboratory's attitude to it in former years. In a period of complex and sensitive management the present directorate at Argonne is seriously dissociated from a responsible spectrum of opinion within the Laboratory. The crux of discontent among the creative scientific and engineering community appears to lie in a developed sense of being overadministered. In contrast to earlier periods, Argonne's professional staff feels a critical need for a voice in the formulation of Laboratory programs and policy. The Argonne senate could supply this mechanism. Slow to rally, their present concern springs from a firm conviction that the Laboratory is "withering on the vine." By contrast, the Laboratory director Powers

  7. Argonne Director Eric Isaacs talks about ARRA funding at Argonne

    ScienceCinema

    Isaacs, Eric

    2016-07-12

    Argonne is set to receive over $150 million in stimulus funds. Director Eric Isaacs describes how these funds will be put to good use—hiring employees and contractors, cleaning up the nuclear footprint, and investing in technologies for America's future. More info on Argonne and ARRA here: http://www.anl.gov/recovery/index.html

  8. Argonne Director Eric Isaacs talks about ARRA funding at Argonne

    SciTech Connect

    Isaacs, Eric

    2009-01-01

    Argonne is set to receive over $150 million in stimulus funds. Director Eric Isaacs describes how these funds will be put to good use—hiring employees and contractors, cleaning up the nuclear footprint, and investing in technologies for America's future. More info on Argonne and ARRA here: http://www.anl.gov/recovery/index.html

  9. Upgrade of the Dow TRIGA research reactor

    SciTech Connect

    Kocher, C.W.

    1991-11-01

    Useful operation of the Dow TRIGA{sup a} research reactor over a period of >20 years has led to a commitment to upgrades enabling another two decades of use with increased capabilities. The reactor utilization program and the upgrades are described in this paper. These included requesting a 20-yr license instead of the 10-yr license, which had been used previously; changing the license to allow operation at power levels of up to 300 kW, which provided improved analytical sensitivity; adding fuel elements to the core, which allowed better performance at the higher power levels; renovating the laboratories, which included consolidating the radioactive materials handling areas and improving the sample preparation areas; installing new shielding, detectors, computers, and sample-handling robots for greater productivity and sensitivity; replacing the 1967-1974 era control console and renovating the control rod drives to provide greater safety, reliability, and maintenance capabilities; and identifying, training, and licensing more senior reactor operators to allow the staff to continue operating and improving this system well past the turn of the century.

  10. Research reactor de-fueling and fuel shipment

    SciTech Connect

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

    1998-08-01

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

  11. Flow Induced Vibration Program at Argonne National Laboratory

    SciTech Connect

    Not Available

    1984-01-01

    Argonne National Laboratory has had a Flow Induced Vibration Program since 1967; the Program currently resides in the Laboratory's Components Technology Division. Throughout its existence, the overall objective of the program has been to develop and apply new and/or improved methods of analysis and testing for the design evaluation of nuclear reactor plant components and heat exchange equipment from the standpoint of flow induced vibration. Historically, the majority of the program activities have been funded by the US Atomic Energy Commission (AEC), Energy Research and Development Administration (ERDA), and Department of Energy (DOE). Current DOE funding is from the Breeder Mechanical Component Development Division, Office of Breeder Technology Projects; Energy Conversion and Utilization Technology (ECUT) Program, Office of Energy Systems Research; and Division of Engineering, Mathematical and Geosciences, Office of Basic Energy Sciences. Testing of Clinch River Breeder Reactor upper plenum components has been funded by the Clinch River Breeder Reactor Plant (CRBRP) Project Office. Work has also been performed under contract with Foster Wheeler, General Electric, Duke Power Company, US Nuclear Regulatory Commission, and Westinghouse.

  12. Accident Analyses for Conversion of the University of Missouri Research Reactor (MURR) from Highly-Enriched to Low-Enriched Uranium

    SciTech Connect

    Stillman, J. A.; Feldman, E. E.; Wilson, E. H.; Foyto, L. P.; Kutikkad, K.; McKibben, J. C.; Peters, N. J.; Cowherd, W. M.; Rickman, B.

    2014-12-01

    This report contains the results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL), the MURR Facility, and the Nuclear Engineering Program – College of Engineering, University of Missouri-Columbia. The core conversion to LEU is being performed with financial support from the U. S. government. This report contains the results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL), the MURR Facility, and the Nuclear Engineering Program – College of Engineering, University of Missouri-Columbia. The core conversion to LEU is being performed with financial support from the U. S. government. In the framework of non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context most research and test reactors, both domestic and international, have started a program of conversion to the use of LEU fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (U-Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MURR. This report presents the results of a study of core behavior under a set of accident conditions for MURR cores fueled with HEU U-Alx dispersion fuel or LEU monolithic U-Mo alloy fuel with 10 wt% Mo

  13. The LEU conversion status of U.S. Research Reactors.

    SciTech Connect

    Matos, J. E.

    1997-11-14

    This paper summarizes the conversion status of US research and test reactors and estimates uranium densities needed to convert reactors with power levels 21 MW from HEU ({ge} 20% U-235) to LEU (<20% U-235) fuels. Detailed conversion studies for each reactor need to be completed in order to establish the feasibility of using LEU fuels.

  14. Argonne National Laboratory institutional plan FY 2001--FY 2006.

    SciTech Connect

    Beggs, S.D.

    2000-12-07

    This Institutional Plan describes what Argonne management regards as the optimal future development of Laboratory activities. The document outlines the development of both research programs and support operations in the context of the nation's R and D priorities, the missions of the Department of Energy (DOE) and Argonne, and expected resource constraints. The Draft Institutional Plan is the product of many discussions between DOE and Argonne program managers, and it also reflects programmatic priorities developed during Argonne's summer strategic planning process. That process serves additionally to identify new areas of strategic value to DOE and Argonne, to which Laboratory Directed Research and Development funds may be applied. The Draft Plan is provided to the Department before Argonne's On-Site Review. Issuance of the final Institutional Plan in the fall, after further comment and discussion, marks the culmination of the Laboratory's annual planning cycle. Chapter II of this Institutional Plan describes Argonne's missions and roles within the DOE laboratory system, its underlying core competencies in science and technology, and six broad planning objectives whose achievement is considered critical to the future of the Laboratory. Chapter III presents the Laboratory's ''Science and Technology Strategic Plan,'' which summarizes key features of the external environment, presents Argonne's vision, and describes how Argonne's strategic goals and objectives support DOE's four business lines. The balance of Chapter III comprises strategic plans for 23 areas of science and technology at Argonne, grouped according to the four DOE business lines. The Laboratory's 14 major initiatives, presented in Chapter IV, propose important advances in key areas of fundamental science and technology development. The ''Operations and Infrastructure Strategic Plan'' in Chapter V includes strategic plans for human resources; environmental protection, safety, and health; site and

  15. The ATLAS Facility at Argonne National Laboratory

    SciTech Connect

    1997-07-01

    The Argonne Tandem Linac Accelerator System (ATLAS) is a superconducting low-energy heavy ion accelerator. Its primary purpose is to provide beams for research in nuclear structure physics. This report begins with a brief history of ATLAS and then describes the current design of the facility. Also summarized are the experimental equipment and research programs. It concludes with a proposal for turning ATLAS into a radioactive beam facility.

  16. Health physics research reactor reference dosimetry

    SciTech Connect

    Sims, C.S.; Ragan, G.E.

    1987-06-01

    Reference neutron dosimetry is developed for the Health Physics Research Reactor (HPRR) in the new operational configuration directly above its storage pit. This operational change was physically made early in CY 1985. The new reference dosimetry considered in this document is referred to as the 1986 HPRR reference dosimetry and it replaces any and all HPRR reference documents or papers issued prior to 1986. Reference dosimetry is developed for the unshielded HPRR as well as for the reactor with each of five different shield types and configurations. The reference dosimetry is presented in terms of three different dose and six different dose equivalent reporting conventions. These reporting conventions cover most of those in current use by dosimetrists worldwide. In addition to the reference neutron dosimetry, this document contains other useful dosimetry-related data for the HPRR in its new configuration. These data include dose-distance measurements and calculations, gamma dose measurements, neutron-to-gamma ratios, ''9-to-3 inch'' ratios, threshold detector unit measurements, 56-group neutron energy spectra, sulfur fluence measurements, and details concerning HPRR shields. 26 refs., 11 figs., 31 tabs.

  17. An Account of Oak Ridge National Laboratory's Thirteen Research Reactors

    SciTech Connect

    Rosenthal, Murray Wilford

    2009-08-01

    The Oak Ridge National Laboratory has built and operated 13 nuclear reactors in its 66-year history. The first was the graphite reactor, the world's first operational nuclear reactor, which served as a plutonium production pilot plant during World War II. It was followed by two aqueous-homogeneous reactors and two red-hot molten-salt reactors that were parts of power-reactor development programs and by eight others designed for research and radioisotope production. One of the eight was an all-metal fast burst reactor used for health physics studies. All of the others were light-water cooled and moderated, including the famous swimming-pool reactor that was copied dozens of times around the world. Two of the reactors were hoisted 200 feet into the air to study the shielding needs of proposed nuclear-powered aircraft. The final reactor, and the only one still operating today, is the High Flux Isotope Reactor (HFIR) that was built particularly for the production of californium and other heavy elements. With the world's highest flux and recent upgrades that include the addition of a cold neutron source, the 44-year-old HFIR continues to be a valuable tool for research and isotope production, attracting some 500 scientific visitors and guests to Oak Ridge each year. This report describes all of the reactors and their histories.

  18. REACTOR PHYSICS MODELING OF SPENT NUCLEAR RESEARCH REACTOR FUEL FOR SNM ATTRIBUTION AND NUCLEAR FORENSICS

    SciTech Connect

    Sternat, M.; Beals, D.; Webb, R.; Nichols, T.

    2010-06-09

    Nuclear research reactors are the least safeguarded type of reactor; in some cases this may be attributed to low risk and in most cases it is due to difficulty from dynamic operation. Research reactors vary greatly in size, fuel type, enrichment, power and burnup providing a significant challenge to any standardized safeguard system. If a whole fuel assembly was interdicted, based on geometry and other traditional forensics work, one could identify the material's origin fairly accurately. If the material has been dispersed or reprocessed, in-depth reactor physics models may be used to help with the identification. Should there be a need to attribute research reactor fuel material, the Savannah River National Laboratory would perform radiochemical analysis of samples of the material as well as other non-destructive measurements. In depth reactor physics modeling would then be performed to compare to these measured results in an attempt to associate the measured results with various reactor parameters. Several reactor physics codes are being used and considered for this purpose, including: MONTEBURNS/ORIGEN/MCNP5, CINDER/MCNPX and WIMS. In attempt to identify reactor characteristics, such as time since shutdown, burnup, or power, various isotopes are used. Complexities arise when the inherent assumptions embedded in different reactor physics codes handle the isotopes differently and may quantify them to different levels of accuracy. A technical approach to modeling spent research reactor fuel begins at the assembly level upon acquiring detailed information of the reactor to be modeled. A single assembly is run using periodic boundary conditions to simulate an infinite lattice which may be repeatedly burned to produce input fuel isotopic vectors of various burnups for a core level model. A core level model will then be constructed using the assembly level results as inputs for the specific fuel shuffling pattern in an attempt to establish an equilibrium cycle. The

  19. REACTOR PHYSICS MODELING OF SPENT RESEARCH REACTOR FUEL FOR TECHNICAL NUCLEAR FORENSICS

    SciTech Connect

    Nichols, T.; Beals, D.; Sternat, M.

    2011-07-18

    Technical nuclear forensics (TNF) refers to the collection, analysis and evaluation of pre- and post-detonation radiological or nuclear materials, devices, and/or debris. TNF is an integral component, complementing traditional forensics and investigative work, to help enable the attribution of discovered radiological or nuclear material. Research is needed to improve the capabilities of TNF. One research area of interest is determining the isotopic signatures of research reactors. Research reactors are a potential source of both radiological and nuclear material. Research reactors are often the least safeguarded type of reactor; they vary greatly in size, fuel type, enrichment, power, and burn-up. Many research reactors are fueled with highly-enriched uranium (HEU), up to {approx}93% {sup 235}U, which could potentially be used as weapons material. All of them have significant amounts of radiological material with which a radioactive dispersal device (RDD) could be built. Therefore, the ability to attribute if material originated from or was produced in a specific research reactor is an important tool in providing for the security of the United States. Currently there are approximately 237 operating research reactors worldwide, another 12 are in temporary shutdown and 224 research reactors are reported as shut down. Little is currently known about the isotopic signatures of spent research reactor fuel. An effort is underway at Savannah River National Laboratory (SRNL) to analyze spent research reactor fuel to determine these signatures. Computer models, using reactor physics codes, are being compared to the measured analytes in the spent fuel. This allows for improving the reactor physics codes in modeling research reactors for the purpose of nuclear forensics. Currently the Oak Ridge Research reactor (ORR) is being modeled and fuel samples are being analyzed for comparison. Samples of an ORR spent fuel assembly were taken by SRNL for analytical and radiochemical

  20. Commercialization plan for Argonne`s lubricious coatings

    SciTech Connect

    Skackson, R.H.

    1991-03-14

    This report summarizes an interview conducted with Dr. David S. Grummon, a specialist in ion beam assisted deposition (IBAD) and coatings at Michigan State University, to define the concept and components of an ion beam assisted deposition (IBAD) system; summarizes findings of a literature search conducted to identify prior and current efforts in developing lubricious coatings; and reports on the market potential for Argonne`s IBAD lubricious coating process.

  1. Argonne National Laboratory-East evolution of solid waste management

    SciTech Connect

    Trychta, K.; McHenry, J.; Thuot, J.

    1996-07-01

    The purpose of this report is to provide the reader with a basic understanding of Argonne National Laboratory`s current general refuse disposal and material recycling programs, how they were developed, and where they are going. In order to better understand the current situation, a brief description of the facilities past practices is explained. ANL is a multi-program research and development center owned by DOE and operated by the University of Chicago. Argonne`s primary facilities are on a 1,700 acre site, 27 miles southwest of Chicago. Fifty-seven major buildings house approximately 4,500 employees at the site.

  2. Argonne's Vilas Pol on NOVA!

    SciTech Connect

    2011-01-01

    Can innovations in materials science help clean up our world? Argonne's material scientist Vilas Pol guest starred on NOVA's "Making Stuff: Cleaner," where David Pogue explored the rapidly-developing science and business of clean energy.

  3. Argonne's Vilas Pol on NOVA!

    ScienceCinema

    None

    2016-07-12

    Can innovations in materials science help clean up our world? Argonne's material scientist Vilas Pol guest starred on NOVA's "Making Stuff: Cleaner," where David Pogue explored the rapidly-developing science and business of clean energy.

  4. Argonne Wakefield Accelerator Update `92

    SciTech Connect

    Rosing, M.; Balka, L.; Chojnacki, E.; Gai, W.; Ho, C.; Konecny, R.; Power, J.; Schoessow, P.; Simpson, J.

    1992-09-01

    The Argonne Wakefield Accelerator (AWA) is an experiment designed to test various ideas related to wakefield technology. Construction is now underway for a 100 nC electron beam in December of 1992. This report updates this progress.

  5. Argonne Wakefield Accelerator Update '92

    SciTech Connect

    Rosing, M.; Balka, L.; Chojnacki, E.; Gai, W.; Ho, C.; Konecny, R.; Power, J.; Schoessow, P.; Simpson, J.

    1992-01-01

    The Argonne Wakefield Accelerator (AWA) is an experiment designed to test various ideas related to wakefield technology. Construction is now underway for a 100 nC electron beam in December of 1992. This report updates this progress.

  6. Dynamic simulation platform to verify the performance of the reactor regulating system for a research reactor

    SciTech Connect

    2015-07-01

    Digital instrumentation and controls system technique is being introduced in new constructed research reactor or life extension of older research reactor. Digital systems are easy to change and optimize but the validated process for them is required. Also, to reduce project risk or cost, we have to make it sure that configuration and control functions is right before the commissioning phase on research reactor. For this purpose, simulators have been widely used in developing control systems in automotive and aerospace industries. In these literatures, however, very few of these can be found regarding test on the control system of research reactor with simulator. Therefore, this paper proposes a simulation platform to verify the performance of RRS (Reactor Regulating System) for research reactor. This simulation platform consists of the reactor simulation model and the interface module. This simulation platform is applied to I and C upgrade project of TRIGA reactor, and many problems of RRS configuration were found and solved. And it proved that the dynamic performance testing based on simulator enables significant time saving and improves economics and quality for RRS in the system test phase. (authors)

  7. Photoneutron effects on pulse reactor kinetics for the Annular Core Research Reactor (ACRR).

    SciTech Connect

    Parma, Edward J., Jr.

    2009-06-01

    The Annular Core Research Reactor (ACRR) is a swimming-pool type pulsed reactor that maintains an epithermal neutron flux and a nine-inch diameter central dry cavity. One of its uses is neutron and gamma-ray irradiation damage studies on electronic components under transient reactor power conditions. In analyzing the experimental results, careful attention must be paid to the kinetics associated with the reactor to ensure that the transient behavior of the electronic device is understood. Since the ACRR fuel maintains a substantial amount of beryllium, copious quantities of photoneutrons are produced that can significantly alter the expected behavior of the reactor power, especially following a reactor pulse. In order to understand these photoneutron effects on the reactor kinetics, the KIFLE transient reactor-analysis code was modified to include the photoneutron groups associated with the beryllium. The time-dependent behavior of the reactor power was analyzed for small and large pulses, assuming several initial conditions including following several pulses during the day, and following a long steady-state power run. The results indicate that, for these types of initial conditions, the photoneutron contribution to the reactor pulse energy can have a few to tens of percent effect.

  8. Aqueous processing of U-10Mo scrap for high performance research reactor fuel

    NASA Astrophysics Data System (ADS)

    Youker, Amanda J.; Stepinski, Dominique C.; Maggos, Laura E.; Bakel, Allen J.; Vandegrift, George F.

    2012-08-01

    The Global Threat Reduction Initiative (GTRI) Conversion program, which is part of the US government's National Nuclear Security Administration (NNSA), supports the conversion of civilian use of highly enriched uranium (HEU) to low enriched uranium (LEU) for reactor fuel and targets. The reason for conversion is to eliminate the use of any material that may pose a threat to the United States or other foreign countries. High performance research reactors (HPRRs) cannot make the conversion to a standard LEU fuel because they require a more dense fuel to meet their performance requirements. As a result, a more dense fuel consisting of a monolithic uranium-molybdenum alloy containing 10% (w/w) Mo with Al cladding and a Zr bonding-layer is being considered. Significant losses are expected in the fabrication of this fuel, so a means to recycle the scrap pieces is needed. Argonne National Laboratory has developed an aqueous-processing flowsheet for scrap recovery in the fuel fabrication process for high-density LEU-monolithic fuel based on data found in the literature. Experiments have been performed to investigate dissolution conditions for solutions containing approximately 20 g-U/L and 50 g-U/L with and without Fe(NO3)3. HNO3 and HF concentrations have been optimized for timely dissolution of the fuel scrap and prevention of the formation of the U-Zr2 intermetallic, explosive complex, while meeting the requirements needed for further processing.

  9. Loss-of-Flow and Loss-of-Pressure Simulations of the BR2 Research Reactor with HEU and LEU Fuel

    SciTech Connect

    Licht, J.; Bergeron, A.; Dionne, B.; Sikik, E.; Van den Branden, G.; Koonen, E.

    2016-01-01

    Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The reactor core of BR2 is located inside a pressure vessel that contains 79 channels in a hyperboloid configuration. The core configuration is highly variable as each channel can contain a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Because of this variability, a representative core configuration, based on current reactor use, has been defined for the fuel conversion analyses. The code RELAP5/Mod 3.3 was used to perform the transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. The input model has been modernized relative to that historically used at BR2 taking into account the best modeling practices developed by Argonne National Laboratory (ANL) and BR2 engineers.

  10. Korea Research Reactor -1 & 2 Decommissioning Project in Korea

    SciTech Connect

    Park, S. K.; Chung, U. S.; Jung, K. J.; Park, J. H.

    2003-02-24

    Korea Research Reactor 1 (KRR-1), the first research reactor in Korea, has been operated since 1962, and the second one, Korea Research Reactor 2 (KRR-2) since 1972. The operation of both of them was phased out in 1995 due to their lifetime and operation of the new and more powerful research reactor, HANARO (High-flux Advanced Neutron Application Reactor; 30MW). Both are TRIGA Pool type reactors in which the cores are small self-contained units sitting in tanks filled with cooling water. The KRR-1 is a TRIGA Mark II, which could operate at a level of up to 250 kW. The second one, the KRR-2 is a TRIGA Mark III, which could operate at a level of up 2,000 kW. The decontamination and decommissioning (D & D) project of these two research reactors, the first D & D project in Korea, was started in January 1997 and will be completed to stage 3 by 2008. The aim of this decommissioning program is to decommission the KRR-1 & 2 reactors and to decontaminate the residual building structure s and the site to release them as unrestricted areas. KAERI (Korea Atomic Energy Research Institute) submitted the decommissioning plan and the environmental impact assessment reports to the Ministry of Science and Technology (MOST) for the license in December 1998, and was approved in November 2000.

  11. Argonne OutLoud presents: The Solar Energy Challenge

    ScienceCinema

    Seth Darling

    2016-07-12

    To better understand the current and future role of solar energy, Argonne's Seth Darling framed the global energy supply and demand outlook over the next 40 years while examining potential energy sources from a feasibility and sustainability perspective. He also discussed the promise and challenges of solar energy while providing a broad overview of related research taking place at Argonne as well as his group's work on organic solar cells.

  12. Sodium fast reactor safety and licensing research plan. Volume II.

    SciTech Connect

    Ludewig, H.; Powers, D. A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A.; Phillips, J.; Zeyen, R.; Clement, B.; Garner, Frank; Walters, Leon; Wright, Steve; Ott, Larry J.; Suo-Anttila, Ahti Jorma; Denning, Richard; Ohshima, Hiroyuki; Ohno, S.; Miyhara, S.; Yacout, Abdellatif; Farmer, M.; Wade, D.; Grandy, C.; Schmidt, R.; Cahalen, J.; Olivier, Tara Jean; Budnitz, R.; Tobita, Yoshiharu; Serre, Frederic; Natesan, Ken; Carbajo, Juan J.; Jeong, Hae-Yong; Wigeland, Roald; Corradini, Michael; Thomas, Justin; Wei, Tom; Sofu, Tanju; Flanagan, George F.; Bari, R.; Porter D.; Lambert, J.; Hayes, S.; Sackett, J.; Denman, Matthew R.

    2012-05-01

    Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

  13. Research and development on the application of advanced control technologies to advanced nuclear reactor systems: A US national perspective

    SciTech Connect

    White, J.D.; Monson, L.R.; Carrol, D.G.; Dayal, Y.; Argonne National Lab., IL; General Electric Co., San Jose, CA )

    1989-01-01

    Control system designs for nuclear power plants are becoming more advanced through the use of digital technology and automation. This evolution is taking place because of: (1) the limitations in analog based control system performance and maintenance and availability and (2) the promise of significant improvement in plant operation and availability due to advances in digital and other control technologies. Digital retrofits of control systems in US nuclear plants are occurring now. Designs of control and protection systems for advanced LWRs are based on digital technology. The use of small inexpensive, fast, large-capacity computers in these designs is the first step of an evolutionary process described in this paper. Under the sponsorship of the US Department of Energy (DOE), Oak Ridge National Laboratory, Argonne National Laboratory, GE Nuclear Energy and several universities are performing research and development in the application of advances in control theory, software engineering, advanced computer architectures, artificial intelligence, and man-machine interface analysis to control system design. The target plant concept for the work described in this paper is the Power Reactor Inherently Safe Module reactor (PRISM), an advanced modular liquid metal reactor concept. This and other reactor designs which provide strong passive responses to operational upsets or accidents afford good opportunities to apply these advances in control technology. 18 refs., 5 figs.

  14. 78 FR 58575 - Review of Experiments for Research Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-09-24

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Review of Experiments for Research Reactors AGENCY: Nuclear Regulatory Commission. ACTION..., by email at Alexander.Adams@nrc.gov , Office of Nuclear Reactor Regulation, U.S. Nuclear...

  15. Oak Ridge Research reactor shutdown maintenance and surveillance

    SciTech Connect

    Coleman, G.H.; Laughlin, D.L.

    1991-05-01

    The Department of Energy ordered the Oak Ridge Research Reactor to be placed in permanent shutdown on July 14, 1987. The paper outlines routine maintenance activities and surveillance tests performed April through September, 1990, on the reactor instrumentation and controls, process system, and the gaseous waste filter system. Preparations are being made to transfer the facility to the Remedial Action Program. 6 tabs. (MHB)

  16. DISMANTLING OF THE REACTOR BLOCK OF THE FRJ-1 RESEARCH REACTOR (MERLIN)

    SciTech Connect

    Stahn, B.; Matela, K.; Zehbe, C.; Poeppinghaus, J.; Cremer, J.

    2003-02-27

    This report describes the past procedure in dismantling the reactor block of the FRJ-1 research reactor (MERLIN). Furthermore, it gives an outlook on future activities up to the final removal of the reactor block. MERLIN is an abbreviation for Medium Energy Research Light Water Moderated Industrial Nuclear Reactor. The FRJ-1 (MERLIN) was shut down in 1985 and the fuel elements removed from the facility. After dismantling the coolant loops and removing the reactor tank internals with subsequent draining of the reactor tank water, the first activities for dismantling the reactor block were carried out in summer 2001. The relevant license was granted in late July 2001 by the licensing authority specifying 8 incidental provisions. After dismantling the reactor extension (gates of the thermal columns and steel platforms surrounding the reactor block), a heavy-load platform including a casing around the reactor block was constructed. Two ventilation systems with a volume flow of 10,000 and 2 ,000 m3/h will, moreover, serve to avoid a spread of contamination. The reactor block will be dismantled in three phases divided according to upper, central and bottom sections. Dismantling the upper section started in August 2002. This section as well as the bottom section can probably be completely measured for clearance. For this reason, the activities have so far been carried out manually using mechanical and thermal techniques. The central section will probably have to be largely disposed of as radioactive waste. This is the region of the former reactor core in which the experimental devices are also integrated. Most of this work will probably have to be carried out by remote handling. More than 80 % of the dismantled materials of the reactor block can probably be measured for clearance. For this purpose, a clearance measurement device was taken into operation in the FRJ-1. On this occasion, the limits of clearance measurement have become evident. For concrete, which constitutes

  17. A Potential NASA Research Reactor to Support NTR Development

    NASA Technical Reports Server (NTRS)

    Eades, Michael; Gerrish, Harold; Hardin, Leroy

    2013-01-01

    In support of efforts for research into the design and development of a man rated Nuclear Thermal Rocket (NTR) engine, the National Aeronautics and Space Administration (NASA), Marshall Space Flight Center (MSFC), is evaluating the potential for building a Nuclear Regulatory Commission (NRC) licensed research reactor. The proposed reactor would be licensed by NASA and operated jointly by NASA and university partners. The purpose of this reactor would be to perform further research into the technologies and systems needed for a successful NTR project and promote nuclear training and education.

  18. Prediction of stainless steel activation in experimental breeder reactor 2 (EBR-II) reflector and blanket subassemblies

    SciTech Connect

    Bunde, K.A.

    1996-12-31

    Stainless steel structural components in nuclear reactors become radioactive wastes when no longer useful. Prior to disposal, certain physical attributes must be analyzed. These attributes include structural integrity, chemical stability, and the radioactive material content among others. The focus of this work is the estimation of the radioactive material content of stainless steel wastes from a research reactor operated by Argonne National Laboratory.

  19. The big and little of fifty years of Moessbauer spectroscopy at Argonne.

    SciTech Connect

    Westfall, C.

    2005-09-20

    the $50 million Zero Gradient Synchrotron (ZGS) and the $30 million Experimental Breeder Reactor (EBR) II. Starting in the mid-1990s, Argonne physicists expanded their exploration of the properties of matter by employing a new type of Moessbauer spectroscopy--this time using synchrotron light sources such as Argonne's Advanced Photon Source (APS), which at $1 billion was the most expensive U.S. accelerator project of its time. Traditional Moessbauer spectroscopy looks superficially like prototypical ''Little Science'' and Moessbauer spectroscopy using synchrotrons looks like prototypical ''Big Science''. In addition, the growth from small to larger scale research seems to follow the pattern familiar from high energy physics even though the wide range of science performed using Moessbauer spectroscopy did not include high energy physics. But is the story of Moessbauer spectroscopy really like the tale told by high energy physicists and often echoed by historians? What do U.S. national laboratories, the ''Home'' of Big Science, have to offer small-scale research? And what does the story of the 50-year development of Moessbauer spectroscopy at Argonne tell us about how knowledge is produced at large laboratories? In a recent analysis of the development of relativistic heavy ion science at Lawrence Berkeley Laboratory I questioned whether it was wise for historians to speak in terms of ''Big Science'', pointing out at that Lawrence Berkeley Laboratory hosted large-scale projects at three scales, the grand scale of the Bevatron, the modest scale of the HILAC, and the mezzo scale of the combined machine, the Bevalac. I argue that using the term ''Big Science'', which was coined by participants, leads to a misleading preoccupation with the largest projects and the tendency to see the history of physics as the history of high energy physics. My aim here is to provide an additional corrective to such views as well as further information about the web of connections that allows

  20. Argonne explains nuclear recycling in 4 minutes

    SciTech Connect

    2012-01-01

    Currently, when using nuclear energy only about five percent of the uranium used in a fuel rod gets fissioned for energy; after that, the rods are taken out of the reactor and put into permanent storage. There is a way, however, to use almost all of the uranium in a fuel rod. Recycling used nuclear fuel could produce hundreds of years of energy from just the uranium we've already mined, all of it carbon-free. Problems with older technology put a halt to recycling used nuclear fuel in the United States, but new techniques developed by scientists at Argonne National Laboratory address many of those issues. For more information, visit http://www.anl.gov/energy/nuclear-energy.

  1. Argonne Plasma Engineering Experiment (APEX) Tokamak

    SciTech Connect

    Norem, J.H.; Balka, L.J.; Kulovitz, E.E.; Magill, S.R.; McGhee, D.G.; Moretti, A.; Praeg, W.F.

    1981-03-01

    The Argonne Plasma Engineering Experiment (APEX) Tokamak was designed to provide hot plasmas for reactor-relevant experiments with rf heating (current drive) and plasma wall experiments, principally in-situ low-Z wall coating and maintenance. The device, sized to produce energetic plasmas at minimum cost, is small (R = 51 cm, r = 15 cm) but capable of high currents (100 kA) and long pulse durations (100 ms). A design using an iron central core with no return legs, pure tension tapewound toroidal field coils, digital radial position control, and UHV vacuum technology was used. Diagnostics include monochrometers, x-ray detectors, and a microwave interferometer and radiometer for density and temperature measurements. Stable 100 ms shots were produced with electron temperatures in the range 500 to 1000 eV. Initial results included studies of thermal desorption and recoating of wall materials.

  2. Argonne explains nuclear recycling in 4 minutes

    ScienceCinema

    None

    2016-07-12

    Currently, when using nuclear energy only about five percent of the uranium used in a fuel rod gets fissioned for energy; after that, the rods are taken out of the reactor and put into permanent storage. There is a way, however, to use almost all of the uranium in a fuel rod. Recycling used nuclear fuel could produce hundreds of years of energy from just the uranium we've already mined, all of it carbon-free. Problems with older technology put a halt to recycling used nuclear fuel in the United States, but new techniques developed by scientists at Argonne National Laboratory address many of those issues. For more information, visit http://www.anl.gov/energy/nuclear-energy.

  3. The effective management of medical isotope production in research reactors

    SciTech Connect

    Drummond, D.T. )

    1993-01-01

    During the 50-yr history of the use of radioisotopes for medical applications, research reactors have played a pivotal role in the production of many if not most of the key products. The marriage between research reactors and production operations is subject to significant challenges on two fronts. The medical applications of the radioisotope products impose some unique constraints and requirements on the production process. In addition, the mandates and priorities of a research reactor are not always congruent with the demands of a production environment. This paper briefly reviews the historical development of medical isotope production, identifies the unique challenges facing this endeavor, and discusses the management of the relationship between the isotope producer and the research reactor operator. Finally, the key elements of a successful relationship are identified.

  4. Renewing Liquid Fueled Molten Salt Reactor Research and Development

    NASA Astrophysics Data System (ADS)

    Towell, Rusty; NEXT Lab Team

    2016-09-01

    Globally there is a desperate need for affordable, safe, and clean energy on demand. More than anything else, this would raise the living conditions of those in poverty around the world. An advanced reactor that utilizes liquid fuel and molten salts is capable of meeting these needs. Although, this technology was demonstrated in the Molten Salt Reactor Experiment (MSRE) at ORNL in the 60's, little progress has been made since the program was cancelled over 40 years ago. A new research effort has been initiated to advance the technical readiness level of key reactor components. This presentation will explain the motivation and initial steps for this new research initiative.

  5. Argonne's SpEC Module

    ScienceCinema

    Harper, Jason

    2016-07-12

    Jason Harper, an electrical engineer in Argonne National Laboratory's EV-Smart Grid Interoperability Center, discusses his SpEC Module invention that will enable fast charging of electric vehicles in under 15 minutes. The module has been licensed to BTCPower.

  6. Argonne's SpEC Module

    SciTech Connect

    Harper, Jason

    2014-05-05

    Jason Harper, an electrical engineer in Argonne National Laboratory's EV-Smart Grid Interoperability Center, discusses his SpEC Module invention that will enable fast charging of electric vehicles in under 15 minutes. The module has been licensed to BTCPower.

  7. Advances by the Integral Fast Reactor Program

    SciTech Connect

    Lineberry, M.J.; Pedersen, D.R.; Walters, L.C.; Cahalan, J.E.

    1991-01-01

    The advances by the Integral Fast Reactor Program at Argonne National Laboratory are the subject of this paper. The Integral Fast Reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The advances stressed in the paper include fuel irradiation performance, improved passive safety, and the development of a prototype fuel cycle facility. 14 refs.

  8. Background radiation measurements at high power research reactors

    SciTech Connect

    Ashenfelter, J.; Yeh, M.; Balantekin, B.; Baldenegro, C. X.; Band, H. R.; Barclay, G.; Bass, C. D.; Berish, D.; Bowden, N. S.; Bryan, C. D.; Cherwinka, J. J.; Chu, R.; Classen, T.; Davee, D.; Dean, D.; Deichert, G.; Dolinski, M. J.; Dolph, J.; Dwyer, D. A.; Fan, S.; Gaison, J. K.; Galindo-Uribarri, A.; Gilje, K.; Glenn, A.; Green, M.; Han, K.; Hans, S.; Heeger, K. M.; Heffron, B.; Jaffe, D. E.; Kettell, S.; Langford, T. J.; Littlejohn, B. R.; Martinez, D.; McKeown, R. D.; Morrell, S.; Mueller, P. E.; Mumm, H. P.; Napolitano, J.; Norcini, D.; Pushin, D.; Romero, E.; Rosero, R.; Saldana, L.; Seilhan, B. S.; Sharma, R.; Stemen, N. T.; Surukuchi, P. T.; Thompson, S. J.; Varner, R. L.; Wang, W.; Watson, S. M.; White, B.; White, C.; Wilhelmi, J.; Williams, C.; Wise, T.; Yao, H.; Yen, Y. -R.; Zhang, C.; Zhang, X.

    2015-10-23

    Research reactors host a wide range of activities that make use of the intense neutron fluxes generated at these facilities. Recent interest in performing measurements with relatively low event rates, e.g. reactor antineutrino detection, at these facilities necessitates a detailed understanding of background radiation fields. Both reactor-correlated and naturally occurring background sources are potentially important, even at levels well below those of importance for typical activities. Here we describe a comprehensive series of background assessments at three high-power research reactors, including γ-ray, neutron, and muon measurements. For each facility we describe the characteristics and identify the sources of the background fields encountered. Furthermore, the general understanding gained of background production mechanisms and their relationship to facility features will prove valuable for the planning of any sensitive measurement conducted therein.

  9. Background radiation measurements at high power research reactors

    DOE PAGES

    Ashenfelter, J.; Yeh, M.; Balantekin, B.; ...

    2015-10-23

    Research reactors host a wide range of activities that make use of the intense neutron fluxes generated at these facilities. Recent interest in performing measurements with relatively low event rates, e.g. reactor antineutrino detection, at these facilities necessitates a detailed understanding of background radiation fields. Both reactor-correlated and naturally occurring background sources are potentially important, even at levels well below those of importance for typical activities. Here we describe a comprehensive series of background assessments at three high-power research reactors, including γ-ray, neutron, and muon measurements. For each facility we describe the characteristics and identify the sources of the backgroundmore » fields encountered. Furthermore, the general understanding gained of background production mechanisms and their relationship to facility features will prove valuable for the planning of any sensitive measurement conducted therein.« less

  10. Background radiation measurements at high power research reactors

    NASA Astrophysics Data System (ADS)

    Ashenfelter, J.; Balantekin, B.; Baldenegro, C. X.; Band, H. R.; Barclay, G.; Bass, C. D.; Berish, D.; Bowden, N. S.; Bryan, C. D.; Cherwinka, J. J.; Chu, R.; Classen, T.; Davee, D.; Dean, D.; Deichert, G.; Dolinski, M. J.; Dolph, J.; Dwyer, D. A.; Fan, S.; Gaison, J. K.; Galindo-Uribarri, A.; Gilje, K.; Glenn, A.; Green, M.; Han, K.; Hans, S.; Heeger, K. M.; Heffron, B.; Jaffe, D. E.; Kettell, S.; Langford, T. J.; Littlejohn, B. R.; Martinez, D.; McKeown, R. D.; Morrell, S.; Mueller, P. E.; Mumm, H. P.; Napolitano, J.; Norcini, D.; Pushin, D.; Romero, E.; Rosero, R.; Saldana, L.; Seilhan, B. S.; Sharma, R.; Stemen, N. T.; Surukuchi, P. T.; Thompson, S. J.; Varner, R. L.; Wang, W.; Watson, S. M.; White, B.; White, C.; Wilhelmi, J.; Williams, C.; Wise, T.; Yao, H.; Yeh, M.; Yen, Y.-R.; Zhang, C.; Zhang, X.

    2016-01-01

    Research reactors host a wide range of activities that make use of the intense neutron fluxes generated at these facilities. Recent interest in performing measurements with relatively low event rates, e.g. reactor antineutrino detection, at these facilities necessitates a detailed understanding of background radiation fields. Both reactor-correlated and naturally occurring background sources are potentially important, even at levels well below those of importance for typical activities. Here we describe a comprehensive series of background assessments at three high-power research reactors, including γ-ray, neutron, and muon measurements. For each facility we describe the characteristics and identify the sources of the background fields encountered. The general understanding gained of background production mechanisms and their relationship to facility features will prove valuable for the planning of any sensitive measurement conducted therein.

  11. LEU conversion status of US research reactors, September 1996

    SciTech Connect

    Matos, J.E.

    1996-10-07

    This paper summarizes the conversion status of research and test reactors in the United States from the use of fuels containing highly- enriched uranium (HEU, greater than or equal to 20%) to the use of fuels containing low-enriched uranium (LEU, < 20%). Estimates of the uranium densities required for conversion are made for reactors with power levels greater than or equal to 1 MW that are not currently involved in the LEU conversion process.

  12. Use of research reactors in multidisciplinary education at Cornell University

    SciTech Connect

    Clark, D.D. )

    1992-01-01

    Multidisciplinary aspects of nuclear science and technology form a large part of the research and teaching activities of the Nuclear Science and Engineering (NS and E) Program at Cornell, and the two reactors housed in Ward Laboratory - a 500-kW TRIGA and a 100-W critical facility (zero-power reactor (ZPR))- play a central role in those activities. Several primarily educational and multidisciplinary features of the NS and E program are described in this paper.

  13. Reactor Safety Research Programs Quarterly Report April- June 1981

    SciTech Connect

    Edler, S. K.

    1981-09-01

    This document summarizes the work performed by Pacific Northwest laboratory (PNL} from April1 through June 30, 1981, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, lspra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory {INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  14. WASP training experience at Argonne

    SciTech Connect

    Buehring, W.A.; Huber, C.C.

    1981-01-01

    Special training courses in electric-system expansion planning have been sponsored by the International Atomic Energy Agency and held at Argonne National Laboratory in each of the past four years. The nine-week courses have been attended by 89 individuals representing 34 countries. The course is intended to not only address the workings of the Wien Automatic System Planning Package (WASP), but also some fundamentals of electric-system planning; including engineering economics, load forecasting, cost estimation, and reliability analysis. Course participants have developd some innovative approaches to system planning using WASP, such as accounting for district heating and analyzing regions of stability for the optimum solution. The WASP-III model was implemented at Argonne for limited use in the 1980 course and was used exclusively in the 1981 course. Previous courses used the WASP-II model. Reactions to the course by past participants generally have been quite favorable.

  15. 75 FR 70042 - In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-11-16

    ... All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent Nuclear Fuel; Order... above, shall be submitted to the NRC to the attention of the Director, Office of Nuclear Reactor... properly marked and handled in accordance with 10 CFR 73.21. The Director, Office of Nuclear...

  16. Computational Science at the Argonne Leadership Computing Facility

    NASA Astrophysics Data System (ADS)

    Romero, Nichols

    2014-03-01

    The goal of the Argonne Leadership Computing Facility (ALCF) is to extend the frontiers of science by solving problems that require innovative approaches and the largest-scale computing systems. ALCF's most powerful computer - Mira, an IBM Blue Gene/Q system - has nearly one million cores. How does one program such systems? What software tools are available? Which scientific and engineering applications are able to utilize such levels of parallelism? This talk will address these questions and describe a sampling of projects that are using ALCF systems in their research, including ones in nanoscience, materials science, and chemistry. Finally, the ways to gain access to ALCF resources will be presented. This research used resources of the Argonne Leadership Computing Facility at Argonne National Laboratory, which is supported by the Office of Science of the U.S. Department of Energy under contract DE-AC02-06CH11357.

  17. Positron beam facility at Kyoto University Research Reactor

    NASA Astrophysics Data System (ADS)

    Xu, Q.; Sato, K.; Yoshiie, T.; Sano, T.; Kawabe, H.; Nagai, Y.; Nagumo, K.; Inoue, K.; Toyama, T.; Oshima, N.; Kinomura, A.; Shirai, Y.

    2014-04-01

    A positron beam facility is presently under construction at the Kyoto University Research Reactor (KUR), which is a light-water moderated tank-type reactor operated at a rated thermal power of 5 MW. A cadmium (Cd) - tungsten (W) source similar to that used in NEPOMUC was chosen in the KUR because Cd is very efficient at producing γ-rays when exposed to thermal neutron flux, and W is a widely used in converter and moderator materials. High-energy positrons are moderated by a W moderator with a mesh structure. Electrical lenses and a solenoid magnetic field are used to extract the moderated positrons and guide them to a platform outside of the reactor, respectively. Since Japan is an earthquake-prone country, a special attention is paid for the design of the in-pile positron source so as not to damage the reactor in the severe earthquake.

  18. A probabilistic safety analysis of incidents in nuclear research reactors.

    PubMed

    Lopes, Valdir Maciel; Agostinho Angelo Sordi, Gian Maria; Moralles, Mauricio; Filho, Tufic Madi

    2012-06-01

    This work aims to evaluate the potential risks of incidents in nuclear research reactors. For its development, two databases of the International Atomic Energy Agency (IAEA) were used: the Research Reactor Data Base (RRDB) and the Incident Report System for Research Reactor (IRSRR). For this study, the probabilistic safety analysis (PSA) was used. To obtain the result of the probability calculations for PSA, the theory and equations in the paper IAEA TECDOC-636 were used. A specific program to analyse the probabilities was developed within the main program, Scilab 5.1.1. for two distributions, Fischer and chi-square, both with the confidence level of 90 %. Using Sordi equations, the maximum admissible doses to compare with the risk limits established by the International Commission on Radiological Protection (ICRP) were obtained. All results achieved with this probability analysis led to the conclusion that the incidents which occurred had radiation doses within the stochastic effects reference interval established by the ICRP-64.

  19. Tiger team assessment of the Argonne Illinois site

    SciTech Connect

    Not Available

    1990-10-19

    This report documents the results of the Department of Energy's (DOE) Tiger Team Assessment of the Argonne Illinois Site (AIS) (including the DOE Chicago Operations Office, DOE Argonne Area Office, Argonne National Laboratory-East, and New Brunswick Laboratory) and Site A and Plot M, Argonne, Illinois, conducted from September 17 through October 19, 1990. The Tiger Team Assessment was conducted by a team comprised of professionals from DOE, contractors, consultants. The purpose of the assessment was to provide the Secretary of Energy with the status of Environment, Safety, and Health (ES H) Programs at AIS. Argonne National Laboratory-East (ANL-E) is the principal tenant at AIS. ANL-E is a multiprogram laboratory operated by the University of Chicago for DOE. The mission of ANL-E is to perform basic and applied research that supports the development of energy-related technologies. There are a significant number of ES H findings and concerns identified in the report that require prompt management attention. A significant change in culture is required before ANL-E can attain consistent and verifiable compliance with statutes, regulations and DOE Orders. ES H activities are informal, fragmented, and inconsistently implemented. Communication is seriously lacking, both vertically and horizontally. Management expectations are not known or commondated adequately, support is not consistent, and oversight is not effective.

  20. Conversion Preliminary Safety Analysis Report for the NIST Research Reactor

    SciTech Connect

    Diamond, D. J.; Baek, J. S.; Hanson, A. L.; Cheng, L-Y; Brown, N.; Cuadra, A.

    2015-01-30

    The NIST Center for Neutron Research (NCNR) is a reactor-laboratory complex providing the National Institute of Standards and Technology (NIST) and the nation with a world-class facility for the performance of neutron-based research. The heart of this facility is the NIST research reactor (aka NBSR); a heavy water moderated and cooled reactor operating at 20 MW. It is fueled with high-enriched uranium (HEU) fuel elements. A Global Threat Reduction Initiative (GTRI) program is underway to convert the reactor to low-enriched uranium (LEU) fuel. This program includes the qualification of the proposed fuel, uranium and molybdenum alloy foil clad in an aluminum alloy, and the development of the fabrication techniques. This report is a preliminary version of the Safety Analysis Report (SAR) that would be submitted to the U.S. Nuclear Regulatory Commission (NRC) for approval prior to conversion. The report follows the recommended format and content from the NRC codified in NUREG-1537, “Guidelines for Preparing and Reviewing Applications for the Licensing of Non-power Reactors,” Chapter 18, “Highly Enriched to Low-Enriched Uranium Conversions.” The emphasis in any conversion SAR is to explain the differences between the LEU and HEU cores and to show the acceptability of the new design; there is no need to repeat information regarding the current reactor that will not change upon conversion. Hence, as seen in the report, the bulk of the SAR is devoted to Chapter 4, Reactor Description, and Chapter 13, Safety Analysis.

  1. Reactor Safety Research: Semiannual report, July-December 1986

    SciTech Connect

    Not Available

    1987-11-01

    Sandia National Laboratories is conducting, under USNRC sponsorship, phenomenological research related to the safety of commercial nuclear power reactors. The research includes experiments to simulate the phenomenology of the accident conditions and the development of analytical models, verified by experiment, which can be used to predict reactor and safety systems performance and behavior under abnormal conditions. The objective of this work is to provide NRC requisite data bases and analytical methods to (1) identify and define safety issues, (2) understand the progression of risk-significant accident sequences, and (3) conduct safety assessments. The collective NRC-sponsored effort at Sandia National Laboratories is directed at enhancing the tehcnology base supporting licensing decisions.

  2. Reactor Safety Research Programs Quarterly Report April -June 1980

    SciTech Connect

    Edler, S. K.

    1980-11-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from April 1 through June 30, 1980, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission {NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining structural graphite strength, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation and postaccident coolability tests for the ESSOR Test Reactor Program, Ispra, Italy; blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  3. Reactor Safety Research Programs Quarterly Report July- September 1980

    SciTech Connect

    Edler, S. K.

    1980-12-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from April 1 through June 30, 1980, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission {NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining structural graphite strength, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation and postaccident coolability tests for the ESSOR Test Reactor Program, Ispra, Italy; blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  4. Reactor Safety Research Programs Quarterly Report October - December 1980

    SciTech Connect

    Edler, S K

    1981-04-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from October 1 through December 31, 1980, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining structural graphite strength, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NOE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation and postaccident coolability tests for the ESSOR Test Reactor Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  5. Reactor Safety Research Programs Quarterly Report January - March 1980

    SciTech Connect

    Hagen, C. M

    1980-10-01

    This document summarizes the work performed by Pacific Northwest Laboratory from January 1 through March 31, 1980, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission. Evaluation of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibilty of determining structural graphite strength, evaluating the feasibilty of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor steam generator tubes where serviceinduced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include the loss-of-coolant accident simulation tests at the NRU reactor, Chalk River, Canada; the fuel rod deformation and post-accident coolability tests for the ESSOR Test Reactor Program, Ispra, Italy; the blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and the experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  6. Argonne OutLoud: "Climate Change: Fact, Fiction and What You Can Do"

    ScienceCinema

    Sisterson, Douglas

    2016-07-12

    Research meteorologist Doug Sisterson discusses climate change and the cutting-edge research taking place at Argonne as well as collaborative research with other institutions, including the University of Chicago.

  7. Argonne OutLoud: "Climate Change: Fact, Fiction and What You Can Do"

    SciTech Connect

    Sisterson, Douglas

    2015-02-13

    Research meteorologist Doug Sisterson discusses climate change and the cutting-edge research taking place at Argonne as well as collaborative research with other institutions, including the University of Chicago.

  8. Proposed environmental remediation at Argonne National Laboratory, Argonne, Illinois

    SciTech Connect

    1997-05-01

    The Department of Energy (DOE) has prepared an Environmental Assessment evaluating proposed environmental remediation activity at Argonne National Laboratory-East (ANL-E), Argonne, Illinois. The environmental remediation work would (1) reduce, eliminate, or prevent the release of contaminants from a number of Resource Conservation and Recovery Act (RCRA) Solid Waste Management Units (SWMUs) and two radiologically contaminated sites located in areas contiguous with SWMUs, and (2) decrease the potential for exposure of the public, ANL-E employees, and wildlife to such contaminants. The actions proposed for SWMUs are required to comply with the RCRA corrective action process and corrective action requirements of the Illinois Environmental Protection Agency; the actions proposed are also required to reduce the potential for continued contaminant release. Based on the analysis in the EA, the DOE has determined that the proposed action does not constitute a major federal action significantly affecting the quality of the human environment within the meaning of the National Environmental Policy Act of 1969 (NEPA). Therefore, the preparation of an Environmental Impact Statement is not required.

  9. Environmental Survey preliminary report, Argonne National Laboratory, Argonne, Illinois

    SciTech Connect

    Not Available

    1988-11-01

    This report presents the preliminary findings of the first phase of the Environmental Survey of the United States Department of Energy's (DOE) Argonne National Laboratory (ANL), conducted June 15 through 26, 1987. The Survey is being conducted by an interdisciplinary team of environmental specialists, led and managed by the Office of Environment, Safety and Health's Office of Environmental Audit. The team includes outside experts supplied by a private contractor. The objective of the Survey is to identify environmental problems and areas of environmental risk associated with ANL. The Survey covers all environmental media and all areas of environmental regulation. It is being performed in accordance with the DOE Environmental Survey Manual. The on-site phase of the Survey involves the review of existing site environmental data, observations of the operations carried on at ANL, and interviews with site personnel. The Survey team developed a Sampling and Analysis (S A) Plan to assist in further assessing certain of the environmental problems identified during its on-site activities. The S A Plan will be executed by the Oak Ridge National Laboratory (ORNL). When completed, the S A results will be incorporated into the Argonne National Laboratory Environmental Survey findings for inclusion in the Environmental Survey Summary Report. 75 refs., 24 figs., 60 tabs.

  10. Reactor pressure vessel structural integrity research

    SciTech Connect

    Pennell, W.E.; Corwin, W.R.

    1995-04-01

    Development continues on the technology used to assess the safety of irradiation-embrittled nuclear reactor pressure vessels (RPVs) containing flaws. Fracture mechanics tests on RPV steel, coupled with detailed elastic-plastic finite-element analyses of the crack-tip stress fields, have shown that (1) constraint relaxation at the crack tip of shallows surface flaws results in increased data scatter but no increase in the lower-bound fracture toughness, (2) the nil ductility temperature (NDT) performs better than the reference temperature for nil ductility transition (RT{sub NDT}) as a normalizing parameter for shallow-flaw fracture toughness data, (3) biaxial loading can reduce the shallow-flaw fracture toughness, (4) stress-based dual-parameter fracture toughness correlations cannot predict the effect of biaxial loading on a shallow-flaw fracture toughness because in-plane stresses at the crack tip are not influenced by biaxial loading, and (5) an implicit strain-based dual-parameter fracture toughness correlation can predict the effect of biaxial loading on shallow-flaw fracture toughness. Experimental irradiation investigations have shown that (1) the irradiation-induced shift in Charpy V-notch vs temperature behavior may not be adequate to conservatively assess fracture toughness shifts due to embrittlement, and (2) the wide global variations of initial chemistry and fracture properties of a nominally uniform material within a pressure vessel may confound accurate integrity assessments that require baseline properties.

  11. Locating tritium sources in a research reactor building.

    PubMed

    Fukui, Masami

    2005-10-01

    Despite renovation of the D2O facility, tritium concentrations in the condensates of reactor room air showed tens of Bq mL before venting resumption on July 1997. This suggested the presence of tritium sources in the research reactor-containment building. An investigation was therefore initiated to locate the source and determine the distribution of tritium in the containment building. Air monitoring in the working area using a dish of water placed in the building suggested that the source of tritium was near the reactor core. Monitoring exhaust air from the two facilities (a cold neutron source and a D(2)O tank) showed high specific activity on the order of 10 Bq mL(-1), suggesting the presence of tritium in condensates near the reactor core. The major concern was whether the leakage of liquid deuterium (4 L) and heavy water (2 x 10(3) L) used as a moderator had occurred. The concentration of tritium in condensates has not increased over the past few years in either the exhaust line or working area, and the deuterium itself has not been found in the surrounding environment. The concentration of tritium measured using an ionization chamber after Ar decay was dependent on the thermal output of the research reactor, indicating that the tritium was produced by the irradiation process within shielding/moderator materials or cover gas with neutrons.

  12. Argonne National Laboratory 1985 publications

    SciTech Connect

    Kopta, J.A.; Hale, M.R.

    1987-08-01

    This report is a bibliography of scientific and technical 1985 publications of Argonne National Laboratory. Some are ANL contributions to outside organizations' reports published in 1985. This compilation, prepared by the Technical Information Services Technical Publications Section (TPB), lists all nonrestricted 1985 publications submitted to TPS by Laboratory's Divisions. The report is divided into seven parts: Journal Articles - Listed by first author, ANL Reports - Listed by report number, ANL and non-ANL Unnumbered Reports - Listed by report number, Non-ANL Numbered Reports - Listed by report number, Books and Book Chapters - Listed by first author, Conference Papers - Listed by first author, Complete Author Index.

  13. The Field Museum brings T. Rex Arm to Argonne

    ScienceCinema

    Makovicky, Peter; Soriano Hoyuelos, Carmen

    2016-12-29

    The tiny arms on the otherwise mighty Tyrannosaurus Rex are one of the biggest and most enduring mysteries in paleontology. Researchers will soon get more insight because an arm bone from the most famous T. Rex in history — the Field Museum's SUE — was recently studied at the Advanced Photon Source at Argonne National Laboratory.

  14. The Field Museum brings T. Rex Arm to Argonne

    SciTech Connect

    Makovicky, Peter; Soriano Hoyuelos, Carmen

    2016-10-21

    The tiny arms on the otherwise mighty Tyrannosaurus Rex are one of the biggest and most enduring mysteries in paleontology. Researchers will soon get more insight because an arm bone from the most famous T. Rex in history — the Field Museum's SUE — was recently studied at the Advanced Photon Source at Argonne National Laboratory.

  15. 2014 Annual Report - Argonne Leadership Computing Facility

    SciTech Connect

    Collins, James R.; Papka, Michael E.; Cerny, Beth A.; Coffey, Richard M.

    2014-01-01

    The Argonne Leadership Computing Facility provides supercomputing capabilities to the scientific and engineering community to advance fundamental discovery and understanding in a broad range of disciplines.

  16. 2015 Annual Report - Argonne Leadership Computing Facility

    SciTech Connect

    Collins, James R.; Papka, Michael E.; Cerny, Beth A.; Coffey, Richard M.

    2015-01-01

    The Argonne Leadership Computing Facility provides supercomputing capabilities to the scientific and engineering community to advance fundamental discovery and understanding in a broad range of disciplines.

  17. Fault detection system for Argentine Research Reactor instrumentation

    SciTech Connect

    Polenta, H.P. ); Bernard, J.A. ); Ray, A. )

    1993-01-20

    The design and implementation of a redundancy management scheme for the on-line detection and isolation of faulty sensors is presented. Such a device is potentially useful in reactor-powered spacecraft for enhancing the processing capabilities of the main computer. The fault detection device can be used as an integral part of intelligent instrumentation systems. The device has been built using an 8-bit microcontroller and commercially available electronic hardware. The software is completely portable. The operation of this device has been successfully demonstrated for real-time validation of sensor data on Argentina's RA-1 Research Reactor.

  18. Research reactor of the future: The advanced neutron source

    SciTech Connect

    Appleton, B.; West, C.

    1994-12-31

    Agents for cancer detection and treatment, stronger materials, better electronic gadgets, and other consumer and industrial products - these are assured benefits of a research reactor project proposed for Oak Ridge. Just as American companies have again assumed world leadership in producing semiconductor chips as well as cars and trucks, the United States is poised to retake the lead in neutron science by building and operating the $2.9 billion Advanced Neutron Source (ANS) research reactor by the start of the next century. In 1985, the neutron community, led by ORNL researchers, proposed a pioneering project, later called the ANS. Scheduled to begin operation in 2003, the ANS is seen not only as a replacement for the aging HFIR and HFBR but also as the best laboratory in the world for conducting neutron-based research.

  19. Reactor subchannel analysis -- Electric Power Research Institute perspective

    SciTech Connect

    Srikantiah, G.

    1995-12-01

    One of the basic objectives of subchannel flow simulation and analysis effort sponsored by the Electric Power Research Institute was the development of a computer code for subchannel analysis and its verification and validation for applications to reactor thermal margin evaluation under steady and transient conditions. A historical perspective is given of the development of specifications for a reactor core subchannel thermal-hydraulics analysis code for utility applications in the evaluation of reactor safety limits during normal operation and accident scenarios. The subchannel analysis capabilities of the VIPRE-01 code based on the homogeneous equilibrium with the algebraic slip model of two-phase flow are presented. The code, which received a safety evaluation report from the US Nuclear Regulatory Commission in 1986, is in wide use in the utility industry for fuel reload safety analysis, critical heat flux correlation development and testing, thermal margin analysis, and core thermal-hydraulic analysis. A considerable amount of work has been done during the past few years on the development of VIPRE-02, an advanced subchannel analysis code based on the two-fluid model of two-phase flow capable of simulating reactor cores, vessels, and internal structures. The functional specifications, development of VIPRE-02, and current applications for VIPRE-02, such as boiling water reactor mixed fuel core evaluation, are also discussed. Code is also used for PWR`s.

  20. Convective cooling in a pool-type research reactor

    SciTech Connect

    Sipaun, Susan; Usman, Shoaib

    2016-01-22

    A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U{sub 3}Si{sub 2}Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system’s performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm{sup −3}. An MSTR model consisting of 20% of MSTR’s nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s{sup −1} from the 4” pipe, and predicted pool surface temperature not exceeding 30°C.

  1. Convective cooling in a pool-type research reactor

    NASA Astrophysics Data System (ADS)

    Sipaun, Susan; Usman, Shoaib

    2016-01-01

    A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U3Si2Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system's performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm-3. An MSTR model consisting of 20% of MSTR's nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s-1 from the 4" pipe, and predicted pool surface temperature not exceeding 30°C.

  2. Yale High Energy Physics Research: Precision Studies of Reactor Antineutrinos

    SciTech Connect

    Heeger, Karsten M.

    2014-09-13

    This report presents experimental research at the intensity frontier of particle physics with particular focus on the study of reactor antineutrinos and the precision measurement of neutrino oscillations. The experimental neutrino physics group of Professor Heeger and Senior Scientist Band at Yale University has had leading responsibilities in the construction and operation of the Daya Bay Reactor Antineutrino Experiment and made critical contributions to the discovery of non-zero$\\theta_{13}$. Heeger and Band led the Daya Bay detector management team and are now overseeing the operations of the antineutrino detectors. Postdoctoral researchers and students in this group have made leading contributions to the Daya Bay analysis including the prediction of the reactor antineutrino flux and spectrum, the analysis of the oscillation signal, and the precision determination of the target mass yielding unprecedented precision in the relative detector uncertainty. Heeger's group is now leading an R\\&D effort towards a short-baseline oscillation experiment, called PROSPECT, at a US research reactor and the development of antineutrino detectors with advanced background discrimination.

  3. Awareness, Preference, Utilization, and Messaging Research for the Spallation Neutron Source and High Flux Isotope Reactor

    SciTech Connect

    Bryant, Rebecca; Kszos, Lynn A

    2011-03-01

    Oak Ridge National Laboratory (ORNL) offers the scientific community unique access to two types of world-class neutron sources at a single site - the Spallation Neutron Source (SNS) and the High Flux Isotope Reactor (HFIR). The 85-MW HFIR provides one of the highest steady-state neutron fluxes of any research reactor in the world, and the SNS is one of the world's most intense pulsed neutron beams. Management of these two resources is the responsibility of the Neutron Sciences Directorate (NScD). NScD commissioned this survey research to develop baseline information regarding awareness of and perceptions about neutron science. Specific areas of investigative interest include the following: (1) awareness levels among those in the scientific community about the two neutron sources that ORNL offers; (2) the level of understanding members of various scientific communities have regarding benefits that neutron scattering techniques offer; and (3) any perceptions that negatively impact utilization of the facilities. NScD leadership identified users of two light sources in North America - the Advanced Photon Source (APS) at Argonne National Laboratory and the National Synchrotron Light Source (NSLS) at Brookhaven National Laboratory - as key publics. Given the type of research in which these scientists engage, they would quite likely benefit from including the neutron techniques available at SNS and HFIR among their scientific investigation tools. The objective of the survey of users of APS, NSLS, SNS, and HFIR was to explore awareness of and perceptions regarding SNS and HFIR among those in selected scientific communities. Perceptions of SNS and FHIR will provide a foundation for strategic communication plan development and for developing key educational messages. The survey was conducted in two phases. The first phase included qualitative methods of (1) key stakeholder meetings; (2) online interviews with user administrators of APS and NSLS; and (3) one-on-one interviews

  4. Research reactor usage at the Idaho National Engineering Laboratory in support of university research and education

    SciTech Connect

    Woodall, D.M.; Dolan, T.J.; Stephens, A.G. )

    1990-01-01

    The Idaho National Engineering Laboratory is a US Department of Energy laboratory which has a substantial history of research and development in nuclear reactor technologies. There are a number of available nuclear reactor facilities which have been incorporated into the research and training needs of university nuclear engineering programs. This paper addresses the utilization of the Advanced Reactivity Measurement Facility (ARMF) and the Coupled Fast Reactivity Measurement Facility (CFRMF) for thesis and dissertation research in the PhD program in Nuclear Science and Engineering by the University of Idaho and Idaho State University. Other reactors at the INEL are also being used by various members of the academic community for thesis and dissertation research, as well as for research to advance the state of knowledge in innovative nuclear technologies, with the EBR-II facility playing an essential role in liquid metal breeder reactor research. 3 refs.

  5. Needs and Requirements for Future Research Reactors (ORNL Perspectives)

    SciTech Connect

    Ilas, Germina; Bryan, Chris; Gehin, Jess C.

    2016-02-10

    The High Flux Isotope Reactor (HFIR) is a vital national and international resource for neutron science research, production of radioisotopes, and materials irradiation. While HFIR is expected to continue operation for the foreseeable future, interest is growing in understanding future research reactors features, needs, and requirements. To clarify, discuss, and compile these needs from the perspective of Oak Ridge National Laboratory (ORNL) research and development (R&D) missions, a workshop, titled “Needs and Requirements for Future Research Reactors”, was held at ORNL on May 12, 2015. The workshop engaged ORNL staff that is directly involved in research using HFIR to collect valuable input on the reactor’s current and future missions. The workshop provided an interactive forum for a fruitful exchange of opinions, and included a mix of short presentations and open discussions. ORNL staff members made 15 technical presentations based on their experience and areas of expertise, and discussed those capabilities of the HFIR and future research reactors that are essential for their current and future R&D needs. The workshop was attended by approximately 60 participants from three ORNL directorates. The agenda is included in Appendix A. This document summarizes the feedback provided by workshop contributors and participants. It also includes information and insights addressing key points that originated from the dialogue started at the workshop. A general overview is provided on the design features and capabilities of high performance research reactors currently in use or under construction worldwide. Recent and ongoing design efforts in the US and internationally are briefly summarized, followed by conclusions and recommendations.

  6. 75 FR 61220 - Massachusetts Institute of Technology: Massachusetts Institute of Technology Research Reactor...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-10-04

    ... COMMISSION Massachusetts Institute of Technology: Massachusetts Institute of Technology Research Reactor... Massachusetts Institute of Technology (MIT, the licensee), which would authorize continued operation of the Massachusetts Institute of Technology Research Reactor (MITR-II, the facility), located in Cambridge,...

  7. Argonne National Laboratory's Recycling Pilot Plant

    ScienceCinema

    Spangenberger, Jeff; Jody, Sam

    2016-07-12

    Argonne has a Recycling Pilot Plant designed to save the non-metal portions of junked cars. Here, program managers demonstrate how plastic shredder residue can be recycled. (Currently these automotive leftovers are sent to landfills.) For more information, visit Argonne's Transportation Technology R&D Center Web site at http://www.transportation.anl.gov.

  8. Argonne News Brief: 70 Years of Discovery

    SciTech Connect

    2016-06-17

    Since its creation in 1946, Argonne National Laboratory has addressed the nation’s most pressing challenges in science, energy, the environment, and national security. United by a common goal – to improve the world – Argonne continues to drive the scientific and technological breakthroughs needed to ensure a sustainable future.

  9. Argonne National Laboratory's Recycling Pilot Plant

    SciTech Connect

    Spangenberger, Jeff; Jody, Sam

    2009-01-01

    Argonne has a Recycling Pilot Plant designed to save the non-metal portions of junked cars. Here, program managers demonstrate how plastic shredder residue can be recycled. (Currently these automotive leftovers are sent to landfills.) For more information, visit Argonne's Transportation Technology R&D Center Web site at http://www.transportation.anl.gov.

  10. The Argonne Voyager multimedia server

    SciTech Connect

    Disz, T.; Judson, I.; Olson, R.; Stevens, R.

    1997-07-01

    With the growing presence of multimedia-enabled systems, one will see an integration of collaborative computing concepts into the everyday environments of future scientific and technical workplaces. Desktop teleconferencing is in common use today, while more complex desktop teleconferencing technology that relies on the availability of multipoint (greater than two nodes) enabled tools is now starting to become available on PCs. A critical problem when using these collaboration tools is the inability to easily archive multistream, multipoint meetings and make the content available to others. Ideally one would like the ability to capture, record, playback, index, annotate and distribute multimedia stream data as easily as one currently handles text or still image data. While the ultimate goal is still some years away, the Argonne Voyager project is aimed at exploring and developing media server technology needed to provide a flexible virtual multipoint recording/playback capability. In this article the authors describe the motivating requirements, architecture implementation, operation, performance, and related work.

  11. Argonne National Laboratory 1986 publications

    SciTech Connect

    Kopta, J.A.; Springer, C.J.

    1987-12-01

    This report is a bibliography of scientific and technical 1986 publications of Argonne National Laboratory. Some are ANL contributions to outside organizations' reports published in 1986. This compilation, prepared by the Technical Information Services Technical Publications Section (TPS), lists all nonrestricted 1986 publications submitted to TPS by the Laboratory's Divisions. Author indexes list ANL authors only. If a first author is not an ANL employee, an asterisk in the bibliographic citation indicates the first ANL author. The report is divided into seven parts: Journal Articles -- Listed by first author; ANL Reports -- Listed by report number; ANL and non-ANL Unnumbered Reports -- Listed by report number; Non-ANL Numbered Reports -- Listed by report number; Books and Book Chapters -- Listed by first author; Conference Papers -- Listed by first author; and Complete Author Index.

  12. 75 FR 27368 - Aerotest Operations, Inc., Aerotest Radiography and Research Reactor; Notice of Consideration of...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-05-14

    ... COMMISSION Aerotest Operations, Inc., Aerotest Radiography and Research Reactor; Notice of Consideration of... INFORMATION CONTACT: Cindy Montgomery, Project Manager, Research and Test Reactors Licensing Branch, Division... Operating License No. R-98 for the Aerotest Radiography and Research Reactor (ARRR), currently held...

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

    SciTech Connect

    Stahl, D.

    1982-12-01

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

  14. A Research Reactor Concept to Support NTP Development

    NASA Technical Reports Server (NTRS)

    Eades, Michael J.; Blue, T. E.; Gerrish, Harold P.; Hardin, Leroy A.

    2014-01-01

    In support of efforts for research into the design and development of man rated Nuclear Thermal Propulsion (NTP), the National Aeronautics and Space Administration (NASA), Marshall Space Flight Center (MSFC), is evaluating the potential for building a Nuclear Regulatory Commission (NRC) licensed NTP based research reactor (NTPRR). The proposed NTPRR would be licensed by NASA and operated jointly by NASA and university partners. The purpose of the NTPRR would be used to perform further research into the technologies and systems needed for a successful NTP project and promote nuclear training and education.

  15. Neutron depth profiling at the University of Texas research reactor

    SciTech Connect

    Unlu, K.; Wehring, B.W. )

    1993-01-01

    A neutron depth profiling (NDP) facility has been developed at the University of Texas at Austin (UT) Nuclear Engineering Teaching Laboratory. The UT-NDP utilizes thermal neutrons from a tangential beam port of the 1-MW TRIGA Mark II research reactor. Aspects of the designs of the thermal neutron beam and target chamber for the UT-NDP facility are given in this paper. Also, a brief description of NDP and possible applications are included.

  16. On use of ZPR research reactors and associated instrumentation and measurement methods for reactor physics studies

    SciTech Connect

    Chauvin, J.P.; Blaise, P.; Lyoussi, A.

    2015-07-01

    The French atomic and alternative energies -CEA- is strongly involved in research and development programs concerning the use of nuclear energy as a clean and reliable source of energy and consequently is working on the present and future generation of reactors on various topics such as ageing plant management, optimization of the plutonium stockpile, waste management and innovative systems exploration. Core physics studies are an essential part of this comprehensive R and D effort. In particular, the Zero Power Reactor (ZPR) of CEA: EOLE, MINERVE and MASURCA play an important role in the validation of neutron (as well photon) physics calculation tools (codes and nuclear data). The experimental programs defined in the CEA's ZPR facilities aim at improving the calculation routes by reducing the uncertainties of the experimental databases. They also provide accurate data on innovative systems in terms of new materials (moderating and decoupling materials) and new concepts (ADS, ABWR, new MTR (e.g. JHR), GENIV) involving new fuels, absorbers and coolant materials. Conducting such interesting experimental R and D programs is based on determining and measuring main parameters of phenomena of interest to qualify calculation tools and nuclear data 'libraries'. Determining these parameters relies on the use of numerous and different experimental techniques using specific and appropriate instrumentation and detection tools. Main ZPR experimental programs at CEA, their objectives and challenges will be presented and discussed. Future development and perspectives regarding ZPR reactors and associated programs will be also presented. (authors)

  17. 77 FR 26321 - Reed College, Reed Research Nuclear Reactor, Renewed Facility Operating License No. R-112

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-05-03

    ... COMMISSION Reed College, Reed Research Nuclear Reactor, Renewed Facility Operating License No. R-112 AGENCY..., Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Rockville, MD 20852. Telephone..., Research and Test Reactors Licensing Branch, Division of Policy and Rulemaking, Office of Nuclear...

  18. Advanced Reactor Safety Research Division. Quarterly progress report, April 1-June 30, 1980

    SciTech Connect

    Romano, A.J.

    1980-01-01

    The Advanced Reactor Safety Research Programs Quarterly Progress Report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the USNRC Division of Reactor Safety Research. The projects reported each quarter are the following: HTGR safety evaluation, SSC Code Development, LMFBR Safety Experiments, and Fast Reactor Safety Code Validation.

  19. Advanced Reactor Safety Research Division. Quarterly progress report, January 1-March 31, 1980

    SciTech Connect

    Agrawal, A.K.; Cerbone, R.J.; Sastre, C.

    1980-06-01

    The Advanced Reactor Safety Research Programs quarterly progress report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the USNRC Division of Reactor Safety Research. The projects reported each quarter are the following: HTGR Safety Evaluation, SSC Code Development, LMFBR Safety Experiments, and Fast Reactor Safety Code Validation.

  20. The design and performance of the research reactor fuel counter

    SciTech Connect

    Abhold, M.E.; Hsue, S.T.; Menlove, H.O.; Walton, G.; Holt, S.

    1996-09-01

    This paper describes the design features, hardware specifications, and performance characteristics of the Research Reactor Fuel Counter (RRFC) System. The system is an active mode neutron coincidence counter intended to assay material test reactor fuel assemblies under water. The RRFC contains 12 {sup 3}He tubes, each with its own preamplifier, and a single ion chamber. The neutron counting electronics are based on the Los Alamos Portable Shift Register (PSR) and the gamma readout is a manual-range pico-ammeter of Los Alamos design. The RRFC is connected to the surface by a 20-m-long cable bundle. The PSR is controlled by a portable IBM computer running a modified version of the Los Alamos neutron coincidence counting code also called RRFC. There is a manual that describes the RRFC software.

  1. Physics of reactor safety. Quarterly report, July-September 1980. Volume III

    SciTech Connect

    Not Available

    1980-11-01

    This Quarterly progress report summarizes work done during the months of July-September 1980 in Argonne National Laboratory's Applied Physics and Components Technology Divisions for the Division of Reactor Safety Research of the US Nuclear Regulatory Commission. The work in the Applied Physics Division includes reports on reactor safety modeling and assessment by members of the Reactor Safety Appraisals Section. Work on reactor core thermal-hydraulics is performed in ANL's Components Technology Division, emphasizing 3-dimensional code development for LMFBR accidents under natural convection conditions. An executive summary is provided including a statement of the findings and recommendations of the report.

  2. The Integral Fast Reactor

    SciTech Connect

    Till, C.E.; Chang, Y.I. ); Lineberry, M.J. )

    1990-01-01

    Argonne National Laboratory, since 1984, has been developing the Integral Fast Reactor (IFR). This paper will describe the way in which this new reactor concept came about; the technical, public acceptance, and environmental issues that are addressed by the IFR; the technical progress that has been made; and our expectations for this program in the near term. 5 refs., 3 figs.

  3. Preparing for radiological assessments in the event of a tornado strike at Argonne National Lab. -East

    SciTech Connect

    Goodkind, M.E.; Klimczak, C.A.; Munyon, W.J. )

    1993-01-01

    Argonne National Laboratory-East (ANL) is a Department of Energy (DOE)-owned, contractor-operated national laboratory located 22 miles southwest of downtown Chicago on a wooded, 1700-acre site. The principal nuclear facilities at ANL include a large fast neutron source (Intense Pulse Neutron Source) in which high-energy protons strike a uranium target to produce neutrons for research studies; [sup 60]Co irradiation sources; chemical and metallurgical plutonium laboratories, some of which are currently being decommissioned; several large hot cell facilities designed for work with multicurie quantities of actinide elements and irradiated reactor fuel materials; a few small research reactors currently in different phases of being decommissioned; and a variety of research laboratories handling many different sources in various chemical and physical forms. The hazards analysis for the ANL site shows that tornado strikes are a serious threat. The site has been struck twice in the past 20 yr, receiving only minor building damage and no release of radioactivity to the environment. Although radioactive materials in general are handled in areas that provide good tornado protection, ANL is prepared to address the problems that would occur should there be a loss of control of radioactive materials due to severe building damage.

  4. Argonne Wakefield Accelerator facility upgrade.

    SciTech Connect

    Conde, M.E.; Gai, W.; Konecny, R.; Power, J.G.; Schoessow, P.; Sun, X.

    2001-07-11

    The Argonne Wakefield Accelerator has been successfully used for conducting wakefield experiments in dielectric loaded structures and plasmas. Although the initial wakefield experiments were successful, higher drive beam quality would substantially improve the wakefield accelerating gradients. For this reason they have built a new 1-1/2 cell L-band photocathode RF gun. This gun is expected to produce 10-100 nC bunches with 2-5 ps rms pulse length and normalized emittance less than 100 mm mrad. The gun will initially have a copper photocathode, which will soon be replaced by a high quantum efficiency cesium telluride one, allowing the generation of a train of high charge bunches. the beam energy at the exit of the gun cavity will be in the range 7.5-10 MeV. A standing-wave linac structure operating at the same frequency (1.3 GHz) will increase the beam energy to about 15 MeV. This beam will be used in high-gradient wakefield acceleration experiments and other high intensity electron beam applications. Traveling-wave dielectric loaded structures, operating at 7.8 and 15.6 GHz, will be excited by the propagation of single bunches or by trains of up to 32 electron bunches, reaching gradients in excess of 100 MV/m over distances of the order of 1 meter.

  5. Eastern Europe Research Reactor Initiative nuclear education and training courses - Current activities and future challenges

    SciTech Connect

    Snoj, L.; Sklenka, L.; Rataj, J.; Boeck, H.

    2012-07-01

    The Eastern Europe Research Reactor Initiative was established in January 2008 to enhance cooperation between the Research Reactors in Eastern Europe. It covers three areas of research reactor utilisation: irradiation of materials and fuel, radioisotope production, neutron beam experiments, education and training. In the field of education and training an EERRI training course was developed. The training programme has been elaborated with the purpose to assist IAEA Member States, which consider building a research reactor (RR) as a first step to develop nuclear competence and infrastructure in the Country. The major strength of the reactor is utilisation of three different research reactors and a lot of practical exercises. Due to high level of adaptability, the course can be tailored to specific needs of institutions with limited or no access to research reactors. (authors)

  6. Push technology at Argonne National Laboratory.

    SciTech Connect

    Noel, R. E.; Woell, Y. N.

    1999-04-06

    Selective dissemination of information (SDI) services, also referred to as current awareness searches, are usually provided by periodically running computer programs (personal profiles) against a cumulative database or databases. This concept of pushing relevant content to users has long been integral to librarianship. Librarians traditionally turned to information companies to implement these searches for their users in business, academia, and the science community. This paper describes how a push technology was implemented on a large scale for scientists and engineers at Argonne National Laboratory, explains some of the challenges to designers/maintainers, and identifies the positive effects that SDI seems to be having on users. Argonne purchases the Institute for Scientific Information (ISI) Current Contents data (all subject areas except Humanities), and scientists no longer need to turn to outside companies for reliable SDI service. Argonne's database and its customized services are known as ACCESS (Argonne-University of Chicago Current Contents Electronic Search Service).

  7. Independent Verification of Research Reactor Operation (Analysis of the Georgian IRT-M Reactor by the Isotope Ratio Method)

    SciTech Connect

    Cliff, John B.; Frank, Douglas P.; Gerlach, David C.; Gesh, Christopher J.; Little, Winston W.; Reid, Bruce D.; Tsiklauri, Georgi V.; Abramidze, Sh; Rostomashvili, Z.; Kiknadze, G.; Dzhavakhishvily, O.; Nabakhtiani, G.

    2010-08-11

    The U.S. Department of Energy’s Office of Nonproliferation and International Security (NA-24) develops technologies to aid in implementing international nuclear safeguards. The Isotope Ratio Method (IRM) was successfully developed in 2005 – 2007 by Pacific Northwest National Laboratory (PNNL) and the Republic of Georgia’s Andronikashvili Institute of Physics as a generic technology to verify the declared operation of water-moderated research reactors, independent of spent fuel inventory. IRM estimates the energy produced over the operating lifetime of a fission reactor by measuring the ratios of the isotopes of trace impurity elements in non-fuel reactor components.The Isotope Ratio Method is a technique for estimating the energy produced over the operating lifetime of a fission reactor by measuring the ratios of the isotopes of impurity elements in non-fuel reactor components.

  8. Present status of liquid metal research for a fusion reactor

    NASA Astrophysics Data System (ADS)

    Tabarés, Francisco L.

    2016-01-01

    Although the use of solid materials as targets of divertor plasmas in magnetic fusion research is accepted as the standard solution for the very challenging issue of power and particle handling in a fusion reactor, a generalized feeling that the present options chosen for ITER will not represent the best choice for a reactor is growing up. The problems found for tungsten, the present selection for the divertor target of ITER, in laboratory tests and in hot plasma fusion devices suggest so. Even in the absence of the strong neutron irradiation expected in a reactor, issues like surface melting, droplet ejection, surface cracking, dust generation, etc., call for alternative solutions in a long pulse, high efficient fusion energy-producing continuous machine. Fortunately enough, decades of research on plasma facing materials based on liquid metals (LMs) have produced a wealth of appealing ideas that could find practical application in the route to the realization of a commercial fusion power plant. The options presently available, although in a different degree of maturity, range from full coverage of the inner wall of the device with liquid metals, so that power and particle exhaust together with neutron shielding could be provided, to more conservative combinations of liquid metal films and conventional solid targets basically representing a sort of high performance, evaporative coating for the alleviation of the surface degradation issues found so far. In this work, an updated review of worldwide activities on LM research is presented, together with some open issues still remaining and some proposals based on simple physical considerations leading to the optimization of the most conservative alternatives.

  9. REACTOR

    DOEpatents

    Szilard, L.

    1963-09-10

    A breeder reactor is described, including a mass of fissionable material that is less than critical with respect to unmoderated neutrons and greater than critical with respect to neutrons of average energies substantially greater than thermal, a coolant selected from sodium or sodium--potassium alloys, a control liquid selected from lead or lead--bismuth alloys, and means for varying the quantity of control liquid in the reactor. (AEC)

  10. Oak Ridge National Laboratory Research Reactor Experimenters' Guide

    SciTech Connect

    Cagle, C.D.

    1982-10-01

    The Oak Ridge National Laboratory has three multipurpose research reactors which accommodate testing loops, target irradiations, and beam-type experiments. Since the experiments must share common or similar facilities and utilities, be designed and fabricated by the same groups, and meet the same safety criteria, certain standards for these have been developed. These standards deal only with those properties from which safety and economy of time and money can be maximized and do not relate to the intent of the experiment or quality of the data obtained. The necessity for, and the limitations of, the standards are discussed; and a compilation of general standards is included.

  11. Some Tooling for Manufacturing Research Reactor Fuel Plates

    SciTech Connect

    Knight, R.W.

    1999-10-03

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

  12. Research in nondestructive evaluation techniques for nuclear reactor concrete structures

    SciTech Connect

    Clayton, Dwight; Smith, Cyrus

    2014-02-18

    The purpose of the Materials Aging and Degradation (MAaD) Pathway of the Department of Energy's Light Water Reactor Sustainability (LWRS) Program is to develop the scientific basis for understanding and predicting longterm environmental degradation behavior of material in nuclear power plants and to provide data and methods to assess the performance of systems, structures, and components (SSCs) essential to safe and sustained nuclear power plant operations. The understanding of aging-related phenomena and their impacts on SSCs is expected to be a significant issue for any nuclear power plant planning for long-term operations (i.e. service beyond the initial license renewal period). Management of those phenomena and their impacts during long-term operations can be better enable by improved methods and techniques for detection, monitoring, and prediction of SSC degradation. The MAaD Pathway R and D Roadmap for Concrete, 'Light Water Reactor Sustainability Nondestructive Evaluation for Concrete Research and Development Roadmap', focused initial research efforts on understanding the recent concrete issues at nuclear power plants and identifying the availability of concrete samples for NDE techniques evaluation and testing. [1] An overview of the research performed by ORNL in these two areas is presented here.

  13. Research in nondestructive evaluation techniques for nuclear reactor concrete structures

    NASA Astrophysics Data System (ADS)

    Clayton, Dwight; Smith, Cyrus

    2014-02-01

    The purpose of the Materials Aging and Degradation (MAaD) Pathway of the Department of Energy's Light Water Reactor Sustainability (LWRS) Program is to develop the scientific basis for understanding and predicting longterm environmental degradation behavior of material in nuclear power plants and to provide data and methods to assess the performance of systems, structures, and components (SSCs) essential to safe and sustained nuclear power plant operations. The understanding of aging-related phenomena and their impacts on SSCs is expected to be a significant issue for any nuclear power plant planning for long-term operations (i.e. service beyond the initial license renewal period). Management of those phenomena and their impacts during long-term operations can be better enable by improved methods and techniques for detection, monitoring, and prediction of SSC degradation. The MAaD Pathway R&D Roadmap for Concrete, "Light Water Reactor Sustainability Nondestructive Evaluation for Concrete Research and Development Roadmap", focused initial research efforts on understanding the recent concrete issues at nuclear power plants and identifying the availability of concrete samples for NDE techniques evaluation and testing. [1] An overview of the research performed by ORNL in these two areas is presented here.

  14. Management of historical waste from research reactors: the Dutch experience

    SciTech Connect

    Van Heek, Aliki; Metz, Bert; Janssen, Bas; Groothuis, Ron

    2013-07-01

    Most radioactive waste emerges as well-defined waste streams from operating power reactors. The management of this is an on-going practice, based on comprehensive (IAEA) guidelines. A special waste category however consists of the historical waste from research reactors, mostly originating from various experiments in the early years of the nuclear era. Removal of the waste from the research site, often required by law, raises challenges: the waste packages must fulfill the acceptance criteria from the receiving storage site as well as the criteria for nuclear transports. Often the aged waste containers do not fulfill today's requirements anymore, and their contents are not well documented. Therefore removal of historical waste requires advanced characterization, sorting, sustainable repackaging and sometimes conditioning of the waste. This paper describes the Dutch experience of a historical waste removal campaign from the Petten High Flux research reactor. The reactor is still in operation, but Dutch legislation asks for central storage of all radioactive waste at the COVRA site in Vlissingen since the availability of the high- and intermediate-level waste storage facility HABOG in 2004. In order to comply with COVRA's acceptance criteria, the complex and mixed inventory of intermediate and low level waste must be characterized and conditioned, identifying the relevant nuclides and their activities. Sorting and segregation of the waste in a Hot Cell offers the possibility to reduce the environmental footprint of the historical waste, by repackaging it into different classes of intermediate and low level waste. In this way, most of the waste volume can be separated into lower level categories not needing to be stored in the HABOG, but in the less demanding LOG facility for low-level waste instead. The characterization and sorting is done on the basis of a combination of gamma scanning with high energy resolution of the closed waste canister and low

  15. Radioisotope research, production, and processing at the University of Missouri Research Reactor

    SciTech Connect

    Ehrhardt, G.J.; Ketring, A.R.; Ja, Wei; Ma, D.; Zinn, K.; Lanigan, J.

    1995-12-31

    The University of Missouri Research Reactor (MURR) is a 10 MW, light-water-cooled and moderated research reactor which first achieved criticality in 1996 and is currently the highest powered university-owned research reactor in the U.S. For many years a major supplier of reactor-produced isotopes for research and commercial purposes, in the last 15 years MURR has concentrated on development of reactor-produced beta-particle emitters for experimental use in nuclear medicine therapy of cancer and rheumatoid arthritis. MURR has played a major role in the development of bone cancer pain palliation with the agents {sup 153}Sm EDTMP and {sup 186}Re/{sup 188}Re HEDP, as well as in the use of {sup 186}Re, {sup 177}Lu, {sup 166}Ho, and {sup 105}Rh for radioimmunotherapy and receptor-agent-guided radiotherapy. MURR is also responsible for the development of therapeutic, {sup 90}Y-labeled glass microspheres for the treatment of liver tumors, a product ({sup 90}Y Therasphere{trademark}) which is currently an approved drug in Canada. MURR has also pioneered the development of {sup 188}W/{sup 188}Re and {sup 99}Mo/{sup 99m}Tc gel generators, which make the use of low specific activity {sup 188}W and {sup 99}Mo practical for such isotope generators.

  16. Proceedings of the NEACRP/IAEA Specialists meeting on the international comparison calculation of a large sodium-cooled fast breeder reactor at Argonne National Laboratory on February 7-9, 1978

    SciTech Connect

    LeSage, L.G.; McKnight, R.D.; Wade, D.C.; Freese, K.E.; Collins, P.J.

    1980-08-01

    The results of an international comparison calculation of a large (1250 MWe) LMFBR benchmark model are presented and discussed. Eight reactor configurations were calculated. Parameters included with the comparison were: eigenvalue, k/sub infinity/, neutron balance data, breeding reaction rate ratios, reactivity worths, central control rod worth, regional sodium void reactivity, core Doppler and effective delayed neutron fraction. Ten countries participated in the comparison, and sixteen solutions were contributed. The discussion focuses on the variation in parameter values, the degree of consistency among the various parameters and solutions, and the identification of unexpected results. The results are displayed and discussed both by individual participants and by groupings of participants (e.g., results from adjusted data sets versus non-adjusted data sets).

  17. IGORR-IV -- Proceedings of the fourth meeting of the International Group on Research Reactors

    SciTech Connect

    Rosenbalm, K.F.

    1995-12-31

    The International Group on Research Reactors was formed to facilitate the sharing of knowledge and experience among those institutions and individuals who are actively working to design, build, and promote new research reactors or to make significant upgrades to existing facilities. Twenty-nine papers were presented in five sessions and written versions of the papers or hard copies of the vugraphs used are published in these proceedings. The five sessions were: (1) Operating Research Reactors and Facility Upgrades; (2) Research Reactors in Design and Construction; (3) ANS Closeout Activities; (4) and (5) Research, Development, and Analysis Results.

  18. Radiation dosimetry at the BNL Medical Research Reactor

    SciTech Connect

    Holden, N.E.; Reciniello, R.N.; Greenberg, D.D.; Hu, J.P.

    1998-11-01

    The Medical Research Reactor, BMRR, at the Brookhaven National Laboratory, BNL, is a three megawatt, 3 MW, heterogeneous, tank-type, light water cooled and moderated, graphite reflected reactor, which was designed for biomedical studies, and became operational in 1959. It provides thermal and epithermal neutron beams suitable for research studies such as radiation therapy of various types of tumors. At the present time, the major program at BMRR is Boron Neutron Capture Therapy, BNCT. Modifications have been made to the BMRR to significantly increase the available epithermal neutron flux density to a patient in clinical trials of BNCT. The data indicate that the flux density and dose rate are concentrated in the center of the beam, the patient absorbs neutrons rather than gamma radiation and as noted previously even with the increasing flux values, gamma-ray dose received by the attending personnel has remained minimal. Flux densities in the center of the thermal port and epithermal port beams have been characterized with an agreement between the measurements and the calculations.

  19. REACTOR

    DOEpatents

    Roman, W.G.

    1961-06-27

    A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

  20. Battery testing at Argonne National Laboratory

    NASA Astrophysics Data System (ADS)

    Deluca, W. H.; Gillie, K. R.; Kulaga, J. E.; Smaga, J. A.; Tummillo, A. F.; Webster, C. E.

    1993-03-01

    Argonne National Laboratory's Analysis & Diagnostic Laboratory (ADL) tests advanced batteries under simulated electric and hybrid vehicle operating conditions. The ADL facilities also include a post-test analysis laboratory to determine, in a protected atmosphere if needed, component compositional changes and failure mechanisms. The ADL provides a common basis for battery performance characterization and life evaluations with unbiased application of tests and analyses. The battery evaluations and post-test examinations help identify factors that limit system performance and life and the most-promising R&D approaches for overcoming these limitations. Since 1991, performance characterizations and/or life evaluations have been conducted on eight battery technologies: Na/S, Li/S, Zn/Br, Ni/MH, Ni/Zn, Ni/Cd, Ni/Fe, and lead-acid. These evaluations were performed for the Department of Energy's. Office of Transportation Technologies, Electric and Hybrid Propulsion Division (DOE/OTT/EHP), and Electric Power Research Institute (EPRI) Transportation Program. The results obtained are discussed.

  1. Battery testing at Argonne National Laboratory

    NASA Astrophysics Data System (ADS)

    Deluca, W. H.; Gillie, K. R.; Kulaga, J. E.; Smaga, J. A.; Tummillo, A. F.; Webster, C. E.

    Advanced battery technology evaluations are performed under simulated electric-vehicle operating conditions at the Analysis & Diagnostic Laboratory (ADL) of Argonne National Laboratory. The ADL results provide insight into those factors that limit battery performance and life. The ADL facilities include a test laboratory to conduct battery experimental evaluations under simulated application conditions and a post-test analysis laboratory to determine, in a protected atmosphere if needed, component compositional changes and failure mechanisms. This paper summarizes the performance characterizations and life evaluations conducted during FY-92 on both single cells and multi-cell modules that encompass six battery technologies (Na/S, Li/FeS, Ni/Metal-Hydride, Ni/Zn, Ni/Cd, Ni/Fe). These evaluations were performed for the Department of Energy, Office of Transportation Technologies, Electric and Hybrid Propulsion Division, and the Electric Power Research Institute. The ADL provides a common basis for battery performance characterization and life evaluations with unbiased application of tests and analyses. The results help identify the most promising R&D approaches for overcoming battery limitations, and provide battery users, developers, and program managers with a measure of the progress being made in battery R&D programs, a comparison of battery technologies, and basic data for modeling.

  2. Battery testing at Argonne National Laboratory

    SciTech Connect

    DeLuca, W.H.; Gillie, K.R.; Kulaga, J.E.; Smaga, J.A.; Tummillo, A.F.; Webster, C.E.

    1993-03-25

    Argonne National Laboratory's Analysis Diagnostic Laboratory (ADL) tests advanced batteries under simulated electric and hybrid vehicle operating conditions. The ADL facilities also include a post-test analysis laboratory to determine, in a protected atmosphere if needed, component compositional changes and failure mechanisms. The ADL provides a common basis for battery performance characterization and life evaluations with unbiased application of tests and analyses. The battery evaluations and post-test examinations help identify factors that limit system performance and life, and the most-promising R D approaches for overcoming these limitations. Since 1991, performance characterizations and/or life evaluations have been conducted on eight battery technologies (Na/S, Li/S, Zn/Br, Ni/MH, Ni/Zn, Ni/Cd, Ni/Fe, and lead-acid). These evaluations were performed for the Department of Energy's. Office of Transportation Technologies, Electric and Hybrid Propulsion Division (DOE/OTT/EHP), and Electric Power Research Institute (EPRI) Transportation Program. The results obtained are discussed.

  3. Battery testing at Argonne National Laboratory

    SciTech Connect

    DeLuca, W.H.; Gillie, K.R.; Kulaga, J.E.; Smaga, J.A.; Tummillo, A.F.; Webster, C.E.

    1992-01-01

    Advanced battery technology evaluations are performed under simulated electric-vehicle operating conditions at the Analysis Diagnostic Laboratory (ADL) of Argonne National Laboratory. The ADL results provide insight into those factors that limit battery performance and life. The ADL facilities include a test laboratory to conduct battery experimental evaluations under simulated application conditions and a post-test analysis laboratory to determine, in a protected atmosphere if needed, component compositional changes and failure mechanisms. This paper summarizes the performance characterizations and life evaluations conducted during FY 1992 on both single cells and multi-cell modules that encompass six battery technologies [Na/S, Li/FeS, Ni/Metal-Hydride, Ni/Zn, Ni/Cd, Ni/Fe]. These evaluations were performed for the Department of Energy, Office of Transportation Technologies, Electric and Hybrid Propulsion Division, and the Electric Power Research Institute. The ADL provides a common basis for battery performance characterization and lie evaluations with unbiased application of tests and analyses. The results help identify the most promising R D approaches for overcoming battery limitations, and provide battery users, developers, and program managers with a measure of the progress being made in battery R D programs, a comparison of battery technologies, and basic data for modeling.

  4. Battery testing at Argonne National Laboratory

    SciTech Connect

    DeLuca, W.H.; Gillie, K.R.; Kulaga, J.E.; Smaga, J.A.; Tummillo, A.F.; Webster, C.E.

    1993-03-25

    Argonne National Laboratory`s Analysis & Diagnostic Laboratory (ADL) tests advanced batteries under simulated electric and hybrid vehicle operating conditions. The ADL facilities also include a post-test analysis laboratory to determine, in a protected atmosphere if needed, component compositional changes and failure mechanisms. The ADL provides a common basis for battery performance characterization and life evaluations with unbiased application of tests and analyses. The battery evaluations and post-test examinations help identify factors that limit system performance and life, and the most-promising R&D approaches for overcoming these limitations. Since 1991, performance characterizations and/or life evaluations have been conducted on eight battery technologies (Na/S, Li/S, Zn/Br, Ni/MH, Ni/Zn, Ni/Cd, Ni/Fe, and lead-acid). These evaluations were performed for the Department of Energy`s. Office of Transportation Technologies, Electric and Hybrid Propulsion Division (DOE/OTT/EHP), and Electric Power Research Institute (EPRI) Transportation Program. The results obtained are discussed.

  5. Nuclear reactor safety research since Three Mile Island

    SciTech Connect

    Mynatt, F.R.

    1982-04-09

    The Three Mile Island nuclear power plant accident has resulted in redirection of reactor safety research priorities. The small release to the environment of radioactive iodine-13 to 17 curies in a total radioactivity release of 2.4 million to 13 million curies-has led to a new emphasis on the physical chemistry of fission product behavior in accidents; the fact that the nuclear core was severely damaged but did not melt down has opened a new accident regime-that of the degraded core; the role of the operators in the progression and severity of the accident has shifted emphasis from equipment reliability to human reliability. As research progresses in these areas, the technical base for regulation and risk analysis will change substantially.

  6. Nuclear reactor safety research since three mile island.

    PubMed

    Mynatt, F R

    1982-04-09

    The Three Mile Island nuclear power plant accident has resulted in redirection of reactor safety research priorities. The small release to the environment of radioactive iodine-13 to 17 curies in a total radioactivity release of 2.4 million to 13 million curies-has led to a new emphasis on the physical chemistry of fission product behavior in accidents; the fact that the nuclear core was severely damaged but did not melt down has opened a new accident regime-that of the degraded core; the role of the operators in the progression and severity of the accident has shifted emphasis from equipment reliability to human reliability. As research progresses in these areas, the technical base for regulation and risk analysis will change substantially.

  7. REACTORS

    DOEpatents

    Spitzer, L. Jr.

    1961-10-01

    Thermonuclear reactors, methods, and apparatus are described for controlling and confining high temperature plasma. Main axial confining coils in combination with helical windings provide a rotational transform that avoids the necessity of a figure-eight shaped reactor tube. The helical windings provide a multipolar helical magnetic field transverse to the axis of the main axial confining coils so as to improve the effectiveness of the confining field by counteracting the tendency of the more central lines of force in the stellarator tube to exchange positions with the magnetic lines of force nearer the walls of the tube. (AEC)

  8. Reprocessing of research reactor fuel the Dounreay option

    SciTech Connect

    Cartwright, P.

    1997-08-01

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

  9. The Integral Fast Reactor

    SciTech Connect

    Chang, Y.I.

    1988-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. This paper describes the key features and potential advantages of the IFR concept, with emphasis on its safety characteristics. 3 refs., 4 figs., 1 tab.

  10. Unique educational opportunities at the Missouri University research reactor

    SciTech Connect

    Ketring, A.R.; Ross, F.K.; Spate, V.

    1997-12-01

    Since the Missouri University Research Reactor (MURR) went critical in 1966, it has been a center where students from many departments conduct their graduate research. In the past three decades, hundreds of graduate students from the MU departments of chemistry, physics, anthropology, nuclear engineering, etc., have received masters and doctoral degrees based on research using neutrons produced at MURR. More recently, the educational opportunities at MURR have been expanded to include undergraduate students and local high school students. Since 1989 MURR has participated in the National Science Foundation-funded Research Experience for Undergraduates (REU) program. As part of this program, undergraduate students from universities and colleges throughout the United States come to MURR and get hands-on research experience during the summer. Another program, started in 1994 by the Nuclear Analysis Program at MURR, allows students from a local high school to conduct a neutron activation analysis (NAA) experiment. We also conduct tours of the center, where we describe the research and educational programs at MURR to groups of elementary school children, high school science teachers, state legislators, professional organizations, and many other groups.

  11. Predicting Activation of Experiments Inside the Annular Core Research Reactor

    SciTech Connect

    Greenberg, Joseph Isaac

    2015-11-01

    The objective of this thesis is to create a program to quickly estimate the radioactivity and decay of experiments conducted inside of the Annular Core Research Reactor at Sandia National Laboratories and eliminate the need for users to write code. This is achieved by model the neutron fluxes in the reactor’s central cavity where experiments are conducted for 4 different neutron spectra using MCNP. The desired neutron spectrum, experiment material composition, and reactor power level are then input into CINDER2008 burnup code to obtain activation and decay information for every isotope generated. DREAD creates all of the files required for CINDER2008 through user selected inputs in a graphical user interface and executes the program for the user and displays the resulting estimation for dose rate at various distances. The DREAD program was validated by weighing and measuring various experiments in the different spectra and then collecting dose rate information after they were irradiated and comparing it to the dose rates that DREAD predicted. The program provides results with an average of 17% higher estimates than the actual values and takes seconds to execute.

  12. 75 FR 57080 - In the Matter of Aerotest Operations, Inc. (Aerotest Radiography and Research Reactor); Order...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-09-17

    ... COMMISSION In the Matter of Aerotest Operations, Inc. (Aerotest Radiography and Research Reactor); Order... which authorizes the possession, use, and operation of the Aerotest Radiography and Research Reactor... Regulations (10 CFR) Section 50.21(c) for research and development purposes. Aerotest is a wholly...

  13. PHYSICS AND SAFETY ANALYSIS FOR THE NIST RESEARCH REACTOR.

    SciTech Connect

    Cheng, L.; Diamond, D.; Xu, J.; Carew, J.; Rorer, D.

    2004-03-31

    Detailed reactor physics and safety analyses have been performed for the 20 MW D{sub 2}O moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analyses provide an update to the Final Safety Analysis Report (FSAR) and employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron and photon transport calculations were performed with the MCNP code to determine the safety parameters for the NBSR. The core depletion and determination of the fuel compositions were performed with MONTEBURNS. MCNP calculations were performed to determine the beginning, middle, and end-of-cycle power distributions, moderator temperature coefficient, and shim safety arm, beam tube and void reactivity worths. The calculational model included a plate-by-plate description of each fuel assembly, axial mid-plane water gap, beam tubes and the tubular geometry of the shim safety arms. The time-dependent analysis of the primary loop was determined with a RELAP5 transient analysis model that includes the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. The statistical analysis used to assure protection from critical heat flux (CHF) was performed using a Monte Carlo simulation of the uncertainties contributing to the CHF calculation. The power distributions used to determine the local fuel conditions and margin to CHF were determined with MCNP. Evaluations were performed for the following accidents: (1) the control rod withdrawal startup accident, (2) the maximum reactivity insertion accident, (3) loss-of-flow resulting from loss of electrical power, (4) loss-of-flow resulting from a primary pump seizure, (5) loss-of-flow resulting from inadvertent throttling of a flow control valve, (6) loss-of-flow resulting from failure of both shutdown cooling pumps and (7) misloading of a fuel element. In both the startup and maximum reactivity insertion accidents, the

  14. Plan for the future of neutron research on condensed matter: an Argonne National Laboratory report prepared in response to the Report of the Review Panel on Neutron Scattering

    SciTech Connect

    1981-01-27

    The Review Panel on Neutron Scattering has recommended an expanded budget to allow systematic development of the field. An alternative plan for the future of neutron research on condensed matter is presented here, in case it is not possible to fund the expanded budget. This plan leads, in a rational and logical way, to a world-class neutron source that will ensure the vitality of the field and exploit the many benefits that state-of-the-art neutron facilities can bring to programs in the materials and biological sciences. 2 tables. (RWR)

  15. The neutron texture diffractometer at the China Advanced Research Reactor

    NASA Astrophysics Data System (ADS)

    Li, Mei-Juan; Liu, Xiao-Long; Liu, Yun-Tao; Tian, Geng-Fang; Gao, Jian-Bo; Yu, Zhou-Xiang; Li, Yu-Qing; Wu, Li-Qi; Yang, Lin-Feng; Sun, Kai; Wang, Hong-Li; Santisteban, J. r.; Chen, Dong-Feng

    2016-03-01

    The first neutron texture diffractometer in China has been built at the China Advanced Research Reactor, due to strong demand for texture measurement with neutrons from the domestic user community. This neutron texture diffractometer has high neutron intensity, moderate resolution and is mainly applied to study texture in commonly used industrial materials and engineering components. In this paper, the design and characteristics of this instrument are described. The results for calibration with neutrons and quantitative texture analysis of zirconium alloy plate are presented. The comparison of texture measurements with the results obtained in HIPPO at LANSCE and Kowari at ANSTO illustrates the reliability of the texture diffractometer. Supported by National Nature Science Foundation of China (11105231, 11205248, 51327902) and International Atomic Energy Agency-TC program (CPR0012)

  16. Sodium fast reactor fuels and materials : research needs.

    SciTech Connect

    Denman, Matthew R.; Porter, Douglas; Wright, Art; Lambert, John; Hayes, Steven; Natesan, Ken; Ott, Larry J.; Garner, Frank; Walters, Leon; Yacout, Abdellatif

    2011-09-01

    An expert panel was assembled to identify gaps in fuels and materials research prior to licensing sodium cooled fast reactor (SFR) design. The expert panel considered both metal and oxide fuels, various cladding and duct materials, structural materials, fuel performance codes, fabrication capability and records, and transient behavior of fuel types. A methodology was developed to rate the relative importance of phenomena and properties both as to importance to a regulatory body and the maturity of the technology base. The technology base for fuels and cladding was divided into three regimes: information of high maturity under conservative operating conditions, information of low maturity under more aggressive operating conditions, and future design expectations where meager data exist.

  17. MCNP-model for the OAEP Thai Research Reactor

    SciTech Connect

    Gallmeier, F.X.; Tang, J.S.; Primm, R.T. III

    1998-06-01

    An MCNP input was prepared for the Thai Research Reactor, making extensive use of the MCNP geometry`s lattice feature that allows a flexible and easy rearrangement of the core components and the adjustment of the control elements. The geometry was checked for overdefined or undefined zones by two-dimensional plots of cuts through the core configuration with the MCNP geometry plotting capabilities, and by a three-dimensional view of the core configuration with the SABRINA code. Cross sections were defined for a hypothetical core of 67 standard fuel elements and 38 low-enriched uranium fuel elements--all filled with fresh fuel. Three test calculations were performed with the MCNP4B-code to obtain the multiplication factor for the cases with control elements fully inserted, fully withdrawn, and at a working position.

  18. Argonne scientist Cristina Negri talks about phytoremediation

    SciTech Connect

    Negri, Cristina

    2012-01-01

    Phytoremediation is the use of plants and trees to remove or neutralize contaminants in polluted soil or water. Argonne scientist M. Cristina Negri leads the phytotechnologies R&D activities at Argonne. Phytotechnologies encompass the treatment of environmental problems through the use of plants. She was the scientific lead in the deployment and monitoring of multi-acre field scale phytoremediation installations and for the development of a phyto- and bio-remediation researcha nd development project in Russia. Her interests also focus on input-efficient approaches to growing energy crops, water efficiency in growing biofuel crops, and on the advanced treatment and reuse of wastewater and other impaired water.

  19. Argonne scientist Cristina Negri talks about phytoremediation

    ScienceCinema

    Negri, Cristina

    2016-07-12

    Phytoremediation is the use of plants and trees to remove or neutralize contaminants in polluted soil or water. Argonne scientist M. Cristina Negri leads the phytotechnologies R&D activities at Argonne. Phytotechnologies encompass the treatment of environmental problems through the use of plants. She was the scientific lead in the deployment and monitoring of multi-acre field scale phytoremediation installations and for the development of a phyto- and bio-remediation researcha nd development project in Russia. Her interests also focus on input-efficient approaches to growing energy crops, water efficiency in growing biofuel crops, and on the advanced treatment and reuse of wastewater and other impaired water.

  20. Studies of acute and chronic radiation injury at the Biological and Medical Research Division, Argonne National Laboratory, 1953-1970: Description of individual studies, data files, codes, and summaries of significant findings

    SciTech Connect

    Grahn, D.; Fox, C.; Wright, B.J.; Carnes, B.A.

    1994-05-01

    Between 1953 and 1970, studies on the long-term effects of external x-ray and {gamma} irradiation on inbred and hybrid mouse stocks were carried out at the Biological and Medical Research Division, Argonne National Laboratory. The results of these studies, plus the mating, litter, and pre-experimental stock records, were routinely coded on IBM cards for statistical analysis and record maintenance. Also retained were the survival data from studies performed in the period 1943-1953 at the National Cancer Institute, National Institutes of Health, Bethesda, Maryland. The card-image data files have been corrected where necessary and refiled on hard disks for long-term storage and ease of accessibility. In this report, the individual studies and data files are described, and pertinent factors regarding caging, husbandry, radiation procedures, choice of animals, and other logistical details are summarized. Some of the findings are also presented. Descriptions of the different mouse stocks and hybrids are included in an appendix; more than three dozen stocks were involved in these studies. Two other appendices detail the data files in their original card-image format and the numerical codes used to describe the animal`s exit from an experiment and, for some studies, any associated pathologic findings. Tabular summaries of sample sizes, dose levels, and other variables are also given to assist investigators in their selection of data for analysis. The archive is open to any investigator with legitimate interests and a willingness to collaborate and acknowledge the source of the data and to recognize appropriate conditions or caveats.

  1. PHYSICS AND SAFETY ANALYSIS FOR THE NIST RESEARCH REACTOR.

    SciTech Connect

    Carew, J.; Hanson, A.; Xu, J.; Rorer, D.; Diamond, D.

    2003-08-26

    Detailed reactor physics and safety analyses have been performed for the 20 MW D{sub 2}O moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analyses provide an update to the Final Safety Analysis Report (FSAR) and employ state-of-the-art calculational methods. Three-dimensional MCNP Monte Carlo neutron and photon transport calculations were performed to determine the safety parameters for the NBSR. The core depletion and determination of the fuel compositions were performed with MONTEBURNS. MCNP calculations were performed to determine the beginning, middle, and end-of-cycle power distributions, moderator temperature coefficient, and shim arm, beam tube and void reactivity worths. The calculational model included a plate-by-plate description of each fuel assembly, axial mid-plane water gap, beam tubes and the tubular geometry of the shim arms. The time-dependent analysis of the primary loop was determined with a RELAP5 transient analysis model including the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. The statistical analysis used to assure protection from critical heat flux (CHF) was performed using a Monte Carlo simulation of the uncertainties contributing to the CHF calculation. The power distributions used to determine the local fuel conditions and margin to CHF were determined with MCNP. Evaluations were performed for the following accidents: (1) the control rod withdrawal startup accident, (2) the maximum reactivity insertion accident, (3) loss-of-flow resulting from loss of electrical power, (4) loss-of-flow resulting from a primary pump seizure, (5) loss-of-flow resulting from inadvertent throttling of a flow control valve, (6) loss-of-flow resulting from failure of both shutdown cooling pumps and (7) misloading of a fuel element. In both the startup and maximum reactivity insertion accidents, the core power transient is terminated

  2. A Look Inside Argonne's Center for Nanoscale Materials

    ScienceCinema

    Divan, Ralu; Rosenthal, Dan; Rose, Volker; Wai Hla, Saw; Liu, Yuzi

    2016-07-12

    At a very small, or "nano" scale, materials behave differently. The study of nanomaterials is much more than miniaturization - scientists are discovering how changes in size change a material's properties. From sunscreen to computer memory, the applications of nanoscale materials research are all around us. Researchers at Argonne's Center for Nanoscale Materials are creating new materials, methods and technologies to address some of the world's greatest challenges in energy security, lightweight but durable materials, high-efficiency lighting, information storage, environmental stewardship and advanced medical devices.

  3. A Look Inside Argonne's Center for Nanoscale Materials

    SciTech Connect

    Divan, Ralu; Rosenthal, Dan; Rose, Volker; Wai Hla, Saw; Liu, Yuzi

    2014-01-29

    At a very small, or "nano" scale, materials behave differently. The study of nanomaterials is much more than miniaturization - scientists are discovering how changes in size change a material's properties. From sunscreen to computer memory, the applications of nanoscale materials research are all around us. Researchers at Argonne's Center for Nanoscale Materials are creating new materials, methods and technologies to address some of the world's greatest challenges in energy security, lightweight but durable materials, high-efficiency lighting, information storage, environmental stewardship and advanced medical devices.

  4. Argonne's Laboratory Computing Resource Center 2009 annual report.

    SciTech Connect

    Bair, R. B.

    2011-05-13

    Now in its seventh year of operation, the Laboratory Computing Resource Center (LCRC) continues to be an integral component of science and engineering research at Argonne, supporting a diverse portfolio of projects for the U.S. Department of Energy and other sponsors. The LCRC's ongoing mission is to enable and promote computational science and engineering across the Laboratory, primarily by operating computing facilities and supporting high-performance computing application use and development. This report describes scientific activities carried out with LCRC resources in 2009 and the broad impact on programs across the Laboratory. The LCRC computing facility, Jazz, is available to the entire Laboratory community. In addition, the LCRC staff provides training in high-performance computing and guidance on application usage, code porting, and algorithm development. All Argonne personnel and collaborators are encouraged to take advantage of this computing resource and to provide input into the vision and plans for computing and computational analysis at Argonne. The LCRC Allocations Committee makes decisions on individual project allocations for Jazz. Committee members are appointed by the Associate Laboratory Directors and span a range of computational disciplines. The 350-node LCRC cluster, Jazz, began production service in April 2003 and has been a research work horse ever since. Hosting a wealth of software tools and applications and achieving high availability year after year, researchers can count on Jazz to achieve project milestones and enable breakthroughs. Over the years, many projects have achieved results that would have been unobtainable without such a computing resource. In fiscal year 2009, there were 49 active projects representing a wide cross-section of Laboratory research and almost all research divisions.

  5. Sodium fast reactor safety and licensing research plan. Volume I.

    SciTech Connect

    Sofu, Tanju; LaChance, Jeffrey L.; Bari, R.; Wigeland, Roald; Denman, Matthew R.; Flanagan, George F.

    2012-05-01

    This report proposes potential research priorities for the Department of Energy (DOE) with the intent of improving the licensability of the Sodium Fast Reactor (SFR). In support of this project, five panels were tasked with identifying potential safety-related gaps in available information, data, and models needed to support the licensing of a SFR. The areas examined were sodium technology, accident sequences and initiators, source term characterization, codes and methods, and fuels and materials. It is the intent of this report to utilize a structured and transparent process that incorporates feedback from all interested stakeholders to suggest future funding priorities for the SFR research and development. While numerous gaps were identified, two cross-cutting gaps related to knowledge preservation were agreed upon by all panels and should be addressed in the near future. The first gap is a need to re-evaluate the current procedures for removing the Applied Technology designation from old documents. The second cross-cutting gap is the need for a robust Knowledge Management and Preservation system in all SFR research areas. Closure of these and the other identified gaps will require both a reprioritization of funding within DOE as well as a re-evaluation of existing bureaucratic procedures within the DOE associated with Applied Technology and Knowledge Management.

  6. Application of a compact reactor to a submersible research vessel

    NASA Astrophysics Data System (ADS)

    Kusunoki, Tsuyoshi; Fujimoto, Hiromi; Nagata, Yutaka; Takahashi, Teruo; Ishida, Toshihisa

    Understanding of the global climate system is necessary to clarify the mechanism of the climate change and predict the fluctuations of the global climate and circumstance. Since the sign of these fluctuations can be predicted in the arctic zone where these fluctuations strongly affect the circumstance, the observation in this zone is very important. Observation data to be obtained there, however, are very few and limited due to difficulty caused by a thick ice covering on a wide-ranged area. A conceptual design of a submersible research vessel with a nuclear power source is studied to cope with this difficulty. The nuclear power is suitable for a submersible research vessel specialized to the undersea of the arctic zone, because it enables to operate for a long time without oxide or fuel supply. By taking account of working conditions in observation, the basic specifications of the vessel are decided; the total weight and the length of it are 500 t, 40 m, respectively, and the maximum ship speed is 12 knots. Two sets of nuclear compact reactor, SCR, light-weighted and of enhanced safety are installed in the vessel to supply the total electricity of 500 kW. This vessel is capable of providing a very rapid, fruitful activity in observation and research.

  7. Water Reactor Safety Research Division. Quarterly progress report, April 1-June 30, 1980

    SciTech Connect

    Abuaf, N.; Levine, M.M.; Saha, P.; van Rooyen, D.

    1980-08-01

    The Water Reactor Safety Research Programs quarterly report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the USNRC Division of Reactor Safety Research. The projects reported each quarter are the following: LWR Thermal Hydraulic Development, Advanced Code Evlauation, TRAC Code Assessment, and Stress Corrosion Cracking of PWR Steam Generator Tubing.

  8. Argonne National Laboratory contributions to the International Symposium on Fusion Nuclear Technology (ISFNT)

    SciTech Connect

    Not Available

    1988-10-01

    A total of sixteen papers with authors from Argonne National Laboratory were presented at the First International Symposium on Fusion Nuclear Technology (ISFNT), held in Tokyo, Japan, in April 1988. The papers cover the results of recent investigations in blanket design and analysis, fusion neutronics, materials experiments in liquid metal corrosion and solid breeders, tritium recovery analysis, experiments and analysis for liquid metal MHD, reactor safety and economic analysis, and transient electromagnetic analysis.

  9. High Technology School-to-Work Program at Argonne

    ScienceCinema

    None

    2016-07-12

    Argonne's High Technology School-to-Work Program for Chicago Public School Students. Supported by the Illinois Department of Commerce and Economic Opportunity, Chicago Public Schools, Argonne National Laboratory and the City of Chicago.

  10. A new safety channel based on ¹⁷N detection in research reactors.

    PubMed

    Seyfi, Somayye; Gharib, Morteza

    2015-10-01

    Tehran research reactor (TRR) is a representative of pool type research reactors using light water, as coolant and moderator. This reactor is chosen as a prototype to demonstrate and prove the feasibility of (17)N detection as a new redundant channel for reactor power measurement. In TRR, similar to other pool type reactors, neutron detectors are immersed in the pool around the core as the main power measuring devices. In the present article, a different approach, using out of water neutron detector, is employed to measure reactor power. This new method is based on (17)O (n,p) (17)N reaction taking place inside the core and subsequent measurement of delayed neutrons emitted due to (17)N disintegration. Count and measurement of neutrons around outlet water pipe provides a reliable redundant safety channel to measure reactor power. Results compared with other established channels indicate a good agreement and shows a linear interdependency with true thermal power. Safety of reactor operation is improved with installation & use of this new power measuring channel. The new approach may equally serve well as a redundant channel in all other types of reactors having coolant comprised of oxygen in its molecular constituents. Contrary to existing channels, this one is totally out of water and thus is an advantage over current instrumentations. It is proposed to employ the same idea on other reactors (nuclear power plants too) to improve safety criteria.

  11. U.S. Department of Energy Program of International Technical Cooperation for Research Reactor Utilization

    SciTech Connect

    Chong, D.; Manning, M.; Ellis, R.; Apt, K.; Flaim, S.; Sylvester, K.

    2004-10-03

    The U.S. Department of Energy, National Nuclear Security Administration (DOE/NNSA) has initiated collaborations with the national nuclear authorities of Egypt, Peru, and Romania for the purpose of advancing the commercial potential and utilization of their respective research reactors. Under its Office of International Safeguards ''Sister Laboratory'' program, DOE/NNSA has undertaken numerous technical collaborations over the past decade intended to promote peaceful applications of nuclear technology. Among these has been technical assistance in research reactor applications, such as neutron activation analysis, nuclear analysis, reactor physics, and medical radioisotope production. The current collaborations are intended to provide the subject countries with a methodology for greater commercialization of research reactor products and services. Our primary goal is the transfer of knowledge, both in administrative and technical issues, needed for the establishment of an effective business plan and utilization strategy for the continued operation of the countries' research reactors. Technical consultation, cooperation, and the information transfer provided are related to: identification, evaluation, and assessment of current research reactor capabilities for products and services; identification of opportunities for technical upgrades for new or expanded products and services; advice and consultation on research reactor upgrades and technical modifications; characterization of markets for reactor products and services; identification of competition and estimation of potential for market penetration; integration of technical constraints; estimation of cash flow streams; and case studies.

  12. Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program

    SciTech Connect

    Smith, Cyrus M; Nanstad, Randy K; Clayton, Dwight A; Matlack, Katie; Ramuhalli, Pradeep; Light, Glenn

    2012-09-01

    The Department of Energy s (DOE) Light Water Reactor Sustainability (LWRS) Program is a five year effort which works to develop the fundamental scientific basis to understand, predict, and measure changes in materials and systems, structure, and components as they age in environments associated with continued long-term operations of existing commercial nuclear power reactors. This year, the Materials Aging and Degradation (MAaD) Pathway of this program has placed emphasis on emerging Non-Destructive Evaluation (NDE) methods which support these objectives. DOE funded Research and Development (R&D) on emerging NDE techniques to support commercial nuclear reactor sustainability is expected to begin next year. This summer, the MAaD Pathway invited subject matter experts to participate in a series of workshops which developed the basis for the research plan of these DOE R&D NDE activities. This document presents the results of one of these workshops which are the DOE LWRS NDE R&D Roadmap for Reactor Pressure Vessels (RPV). These workshops made a substantial effort to coordinate the DOE NDE R&D with that already underway or planned by the Electric Power Research Institute (EPRI) and the Nuclear Regulatory Commission (NRC) through their representation at these workshops.

  13. Argonne Discovery Yields Self-Healing Diamond-Like Carbon

    DOE PAGES

    Cunningham, Greg; Jones, Katie Elyce

    2016-10-27

    We report that large-scale reactive molecular dynamics simulations carried out on the US Department of Energy’s IBM Blue Gene/Q Mira supercomputer at the Argonne Leadership Computing Facility, along with experiments conducted by researchers in Argonne’s Energy Systems Division, enabled the design of a “self-healing” anti-wear coating that drastically reduces friction and related degradation in engines and moving machinery. Now, the computational work advanced for this purpose is being used to identify the friction-fighting potential of other catalysts.

  14. Reactor

    DOEpatents

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  15. Leidos Biomed Teams with NCI, DOE, and Argonne National Lab to Support National X-Ray Resource | Poster

    Cancer.gov

    Scientists are making progress in understanding a bleeding disorder caused by prescription drug interactions, thanks to a high-tech research facility involving two federal national laboratories, Argonne and Frederick.

  16. Characterization of Novel Calorimeters in the Annular Core Research Reactor

    NASA Astrophysics Data System (ADS)

    Hehr, Brian D.; Parma, Edward J.; Peters, Curtis D.; Naranjo, Gerald E.; Luker, S. Michael

    2016-02-01

    A series of pulsed irradiation experiments have been performed in the central cavity of Sandia National Laboratories' Annular Core Research Reactor (ACRR) to characterize the responses of a set of elemental calorimeter materials including Si, Zr, Sn, Ta, W, and Bi. Of particular interest was the perturbing effect of the calorimeter itself on the ambient radiation field - a potential concern in dosimetry applications. By placing the calorimeter package into a neutron-thermalizing lead/polyethylene (LP) bucket and irradiating both with and without a cadmium wrapper, it was demonstrated that prompt capture gammas generated inside the calorimeters can be a significant contributor to the measured dose in the active disc region. An MCNP model of the experimental setup was shown to replicate measured dose responses to within 10%. The internal (n,γ) contribution was found to constitute as much as 50% of the response inside the LP bucket and up to 20% inside the nominal (unmodified) cavity environment, with Ta and W exhibiting the largest enhancement due to their sizable (n,γ) cross sections. Capture reactions in non-disc components of the calorimeter were estimated to be responsible for up to a few percent of the measured response. This work was supported by the United States Department of Energy under Contract DE-AC04-94AL85000. Sandia is a multiprogram laboratory operated by Sandia Corporation, a Lockheed Martin Company, for the United States Department of Energy.

  17. Integral Fast Reactor concept

    SciTech Connect

    Till, C.E.; Chang, Y.I.

    1986-01-01

    The Integral Fast Reactor (IFR) is an innovative LMR concept, being developed at Argonne National Laboratory, that fully exploits the inherent properties of liquid metal cooling and metallic fuel to achieve breakthroughs in economics and inherent safety. This paper describes key features and potential advantages of the IFR concept, technology development status, fuel cycle economics potential, and future development path.

  18. 78 FR 26811 - Dow Chemical Company, Dow TRIGA Research Reactor; License Renewal for the Dow Chemical TRIGA...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-05-08

    ... COMMISSION Dow Chemical Company, Dow TRIGA Research Reactor; License Renewal for the Dow Chemical TRIGA Research Reactor; Supplemental Information and Correction AGENCY: Nuclear Regulatory Commission. ACTION... Chemical TRIGA Research Reactor,'' to inform the public that the NRC is considering issuance of a...

  19. Argonne National Laboratory summary site environmental report for calendar year 2006.

    SciTech Connect

    Golchert, N. W.; ESH /QA Oversight

    2008-03-27

    This booklet is designed to inform the public about what Argonne National Laboratory is doing to monitor its environment and to protect its employees and neighbors from any adverse environmental impacts from Argonne research. The Downers Grove South Biology II class was selected to write this booklet, which summarizes Argonne's environmental monitoring programs for 2006. Writing this booklet also satisfies the Illinois State Education Standard, which requires that students need to know and apply scientific concepts to graduate from high school. This project not only provides information to the public, it will help students become better learners. The Biology II class was assigned to condense Argonne's 300-page, highly technical Site Environmental Report into a 16-page plain-English booklet. The site assessment relates to the class because the primary focus of the Biology II class is ecology and the environment. Students developed better learning skills by working together cooperatively, writing and researching more effectively. Students used the Argonne Site Environmental Report, the Internet, text books and information from Argonne scientists to help with their research on their topics. The topics covered in this booklet are the history of Argonne, groundwater, habitat management, air quality, Argonne research, Argonne's environmental non-radiological program, radiation, and compliance. The students first had to read and discuss the Site Environmental Report and then assign topics to focus on. Dr. Norbert Golchert and Mr. David Baurac, both from Argonne, came into the class to help teach the topics more in depth. The class then prepared drafts and wrote a final copy. Ashley Vizek, a student in the Biology class stated, 'I reviewed my material and read it over and over. I then took time to plan my paper out and think about what I wanted to write about, put it into foundation questions and started to write my paper. I rewrote and revised so I think the amount of

  20. Commercialization plan for Argonne's lubricious coatings

    SciTech Connect

    Skackson, R.H.

    1991-03-14

    This report summarizes an interview conducted with Dr. David S. Grummon, a specialist in ion beam assisted deposition (IBAD) and coatings at Michigan State University, to define the concept and components of an ion beam assisted deposition (IBAD) system; summarizes findings of a literature search conducted to identify prior and current efforts in developing lubricious coatings; and reports on the market potential for Argonne's IBAD lubricious coating process.

  1. Reactor Safety Research Programs Quarterly Report July - September 1981

    SciTech Connect

    Edler, S. K.

    1982-01-01

    This document summarizes the work performed by Pacific Northwest laboratory (PNL) from July 1 through September 30, 1981, for the Division of Accident Evaluation, U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR} steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, lspra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL), Idaho Falls, Idaho. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  2. Reactor Safety Research Programs Quarterly Report October - December 1981

    SciTech Connect

    Edler, S. K.

    1982-03-01

    This document summarizes the work performed by Pacific Northwest laboratory (PNL) from October 1 through December 31, 1981, for the Division of Accident Evaluation, U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where serviceinduced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and post accident coolability tests for the ESSOR reactor Super Sara Test Program, lspra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL), Idaho Falls, Idaho. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  3. Laboratory directed research and development

    SciTech Connect

    Not Available

    1991-11-15

    The purposes of Argonne's Laboratory Directed Research and Development (LDRD) Program are to encourage the development of novel concepts, enhance the Laboratory's R D capabilities, and further the development of its strategic initiatives. Among the aims of the projects supported by the Program are establishment of engineering proof-of-principle''; development of an instrumental prototype, method, or system; or discovery in fundamental science. Several of these project are closely associated with major strategic thrusts of the Laboratory as described in Argonne's Five Year Institutional Plan, although the scientific implications of the achieved results extend well beyond Laboratory plans and objectives. The projects supported by the Program are distributed across the major programmatic areas at Argonne. Areas of emphasis are (1) advanced accelerator and detector technology, (2) x-ray techniques in biological and physical sciences, (3) advanced reactor technology, (4) materials science, computational science, biological sciences and environmental sciences. Individual reports summarizing the purpose, approach, and results of projects are presented.

  4. Utilization of the Cornell University research reactors in support of the Nuclear Power Industry

    SciTech Connect

    Aderhold, A.C. )

    1993-01-01

    Cornell University is licensed to operate two research reactor facilities on its main campus in Ithaca, New York: a 500-kW pulsing TRIGA and a 100-W zero-power reactor (ZPR). The initial criticality of both reactors took place in 1962, and the utilization of each has been, and continues to be, dedicated to the teaching and research programs of Cornell's many academic departments. As the nation's nuclear power industry grew, the demand for services at research and test reactors increased. As a result, and in large part because of special design features of the TRIGA, Cornell responded to a few requests for reactor testing services while maintaining the policy that these services would not interfere with teaching and research programs. The frequency of service requests suddenly mushroomed in November of 1989, when the nation's major testing reactor was shut down for repairs. In spite of a small staff of two full-time reactor operators, a decision was made to support the nuclear industry to the fullest extent possible without jeopardizing Cornell's teaching and research programs. This turned into a monumental task of tight scheduling and meeting precise deadlines. It could only be accomplished by working late evenings and weekends and, on a number of occasions, staying at the facility for up to 5 days continuously.

  5. Metagenomes from Argonne's MG-RAST Metagenomics Analysis Server

    DOE Data Explorer

    MG-RAST has a large number of datasets that researchers have deposited for public use. As of July, 2014, the number of metagenomes represented by MG-RAST numbered more than 18,500, and the number of available sequences was more than 75 million! The public can browse the collection several different ways, and researchers can login to deposit new data. Researchers have the choice of keeping a dataset private so that it is viewable only by them when logged in, or they can choose to make a dataset public at any time with a simple click of a link. MG-RAST was launched in 2007 by the Mathematics and Computer Science Division at Argonne National Laboratory (ANL). It is part of the toolkit available to the Terragenomics project, which seeks to do a comprehensive metagenomics study of U.S. soil. The Terragenomics project page is located at http://www.mcs.anl.gov/research/projects/terragenomics/.

  6. Integrated Decision-Making Tool to Develop Spent Fuel Strategies for Research Reactors

    SciTech Connect

    Beatty, Randy L; Harrison, Thomas J

    2016-01-01

    IAEA Member States operating or having previously operated a Research Reactor are responsible for the safe and sustainable management and disposal of associated radioactive waste, including research reactor spent nuclear fuel (RRSNF). This includes the safe disposal of RRSNF or the corresponding equivalent waste returned after spent fuel reprocessing. One key challenge to developing general recommendations lies in the diversity of spent fuel types, locations and national/regional circumstances rather than mass or volume alone. This is especially true given that RRSNF inventories are relatively small, and research reactors are rarely operated at a high power level or duration typical of commercial power plants. Presently, many countries lack an effective long-term policy for managing RRSNF. This paper presents results of the International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) #T33001 on Options and Technologies for Managing the Back End of the Research Reactor Nuclear Fuel Cycle which includes an Integrated Decision Making Tool called BRIDE (Back-end Research reactor Integrated Decision Evaluation). This is a multi-attribute decision-making tool that combines the Total Estimated Cost of each life-cycle scenario with Non-economic factors such as public acceptance, technical maturity etc and ranks optional back-end scenarios specific to member states situations in order to develop a specific member state strategic plan with a preferred or recommended option for managing spent fuel from Research Reactors.

  7. Multipurpose epithermal neutron beam on new research station at MARIA research reactor in Swierk-Poland

    SciTech Connect

    Gryzinski, M.A.; Maciak, M.

    2015-07-01

    MARIA reactor is an open-pool research reactor what gives the chance to install uranium fission converter on the periphery of the core. It could be installed far enough not to induce reactivity of the core but close enough to produce high flux of fast neutrons. Special design of the converter is now under construction. It is planned to set the research stand based on such uranium converter in the near future: in 2015 MARIA reactor infrastructure should be ready (preparation started in 2013), in 2016 the neutron beam starts and in 2017 opening the stand for material and biological research or for medical training concerning BNCT. Unused for many years, horizontal channel number H2 at MARIA research rector in Poland, is going to be prepared as a part of unique stand. The characteristics of the neutron beam will be significant advantage of the facility. High flux of neutrons at the level of 2x10{sup 9} cm{sup -2}s{sup -1} will be obtainable by uranium neutron converter located 90 cm far from the reactor core fuel elements (still inside reactor core basket between so called core reflectors). Due to reaction of core neutrons with converter U{sub 3}Si{sub 2} material it will produce high flux of fast neutrons. After conversion neutrons will be collimated and moderated in the channel by special set of filters and moderators. At the end of H2 channel i.e. at the entrance to the research room neutron energy will be in the epithermal energy range with neutron intensity at least at the level required for BNCT (2x10{sup 9} cm{sup -2}s{sup -1}). For other purposes density of the neutron flux could be smaller. The possibility to change type and amount of installed filters/moderators which enables getting different properties of the beam (neutron energy spectrum, neutron-gamma ratio and beam profile and shape) is taken into account. H2 channel is located in separate room which is adjacent to two other empty rooms under the preparation for research laboratories (200 m2). It is

  8. IEA-R1 Nuclear Research Reactor: 58 Years of Operating Experience and Utilization for Research, Teaching and Radioisotopes Production

    SciTech Connect

    Cardenas, Jose Patricio Nahuel; Filho, Tufic Madi; Saxena, Rajendra; Filho, Walter Ricci

    2015-07-01

    IEA-R1 research reactor at the Instituto de Pesquisas Energeticas e Nucleares (Nuclear and Energy Research Institute) IPEN, Sao Paulo, Brazil is the largest power research reactor in Brazil, with a maximum power rating of 5 MWth. It is being used for basic and applied research in the nuclear and neutron related sciences, for the production of radioisotopes for medical and industrial applications, and for providing services of neutron activation analysis, real time neutron radiography, and neutron transmutation doping of silicon. IEA-R1 is a swimming pool reactor, with light water as the coolant and moderator, and graphite and beryllium as reflectors. The reactor was commissioned on September 16, 1957 and achieved its first criticality. It is currently operating at 4.5 MWth with a 60-hour cycle per week. In the early sixties, IPEN produced {sup 131}I, {sup 32}P, {sup 198}Au, {sup 24}Na, {sup 35}S, {sup 51}Cr and labeled compounds for medical use. During the past several years, a concerted effort has been made in order to upgrade the reactor power to 5 MWth through refurbishment and modernization programs. One of the reasons for this decision was to produce {sup 99}Mo at IPEN. The reactor cycle will be gradually increased to 120 hours per week continuous operation. It is anticipated that these programs will assure the safe and sustainable operation of the IEA-R1 reactor for several more years, to produce important primary radioisotopes {sup 99}Mo, {sup 125}I, {sup 131}I, {sup 153}Sm and {sup 192}Ir. Currently, all aspects of dealing with fuel element fabrication, fuel transportation, isotope processing, and spent fuel storage are handled by IPEN at the site. The reactor modernization program is slated for completion by 2015. This paper describes 58 years of operating experience and utilization of the IEA-R1 research reactor for research, teaching and radioisotopes production. (authors)

  9. Development of a mono-energetic positron beam line at the Kyoto University Research Reactor

    NASA Astrophysics Data System (ADS)

    Sato, K.; Xu, Q.; Yoshiie, T.; Sano, T.; Kawabe, H.; Nagai, Y.; Nagumo, K.; Inoue, K.; Toyama, T.; Oshima, N.; Kinomura, A.; Shirai, Y.

    2015-01-01

    Positron beam facilities are widely used for solid state physics and material science studies. A positron beam facility has been constructed at the Kyoto University Research Reactor (KUR) in order to expand its application range. The KUR is a light-water-moderated tank-type reactor operated at a rated thermal power of 5 MW. A positron beam has been transported successfully from the reactor to the irradiation chamber. The total moderated positron rate was greater than 1.4 × 106/s while the reactor operated at a reduced power of 1 MW. Special attention was paid for the design of the in-pile position source to prevent possible damage of the reactor in case of severe earthquakes.

  10. Modular Pebble-Bed Reactor Project: Laboratory-Directed Research and Development Program FY 2002 Annual Report

    SciTech Connect

    Petti, David Andrew; Dolan, Thomas James; Miller, Gregory Kent; Moore, Richard Leroy; Terry, William Knox; Ougouag, Abderrafi Mohammed-El-Ami; Oh, Chang H; Gougar, Hans D

    2002-11-01

    This report documents the results of our research in FY-02 on pebble-bed reactor technology under our Laboratory Directed Research and Development (LDRD) project entitled the Modular Pebble-Bed Reactor. The MPBR is an advanced reactor concept that can meet the energy and environmental needs of future generations under DOE’s Generation IV initiative. Our work is focused in three areas: neutronics, core design and fuel cycle; reactor safety and thermal hydraulics; and fuel performance.

  11. Production and release rate of (37)Ar from the UT TRIGA Mark-II research reactor.

    PubMed

    Johnson, Christine; Biegalski, Steven R; Artnak, Edward J; Moll, Ethan; Haas, Derek A; Lowrey, Justin D; Aalseth, Craig E; Seifert, Allen; Mace, Emily K; Woods, Vincent T; Humble, Paul

    2017-02-01

    Air samples were taken at various locations around The University of Texas at Austin's TRIGA Mark II research reactor and analyzed to determine the concentrations of (37)Ar, (41)Ar, and (133)Xe present. The measured ratio of (37)Ar/(41)Ar and historical records of (41)Ar releases were then utilized to estimate an annual average release rate of (37)Ar from the reactor facility. Using the calculated release rate, atmospheric transport modeling was performed in order to determine the potential impact of research reactor operations on nearby treaty verification activities. Results suggest that small research reactors (∼1 MWt) do not release (37)Ar in concentrations measurable by currently proposed OSI detection equipment.

  12. Argonne's Laboratory computing center - 2007 annual report.

    SciTech Connect

    Bair, R.; Pieper, G. W.

    2008-05-28

    Argonne National Laboratory founded the Laboratory Computing Resource Center (LCRC) in the spring of 2002 to help meet pressing program needs for computational modeling, simulation, and analysis. The guiding mission is to provide critical computing resources that accelerate the development of high-performance computing expertise, applications, and computations to meet the Laboratory's challenging science and engineering missions. In September 2002 the LCRC deployed a 350-node computing cluster from Linux NetworX to address Laboratory needs for mid-range supercomputing. This cluster, named 'Jazz', achieved over a teraflop of computing power (1012 floating-point calculations per second) on standard tests, making it the Laboratory's first terascale computing system and one of the 50 fastest computers in the world at the time. Jazz was made available to early users in November 2002 while the system was undergoing development and configuration. In April 2003, Jazz was officially made available for production operation. Since then, the Jazz user community has grown steadily. By the end of fiscal year 2007, there were over 60 active projects representing a wide cross-section of Laboratory expertise, including work in biosciences, chemistry, climate, computer science, engineering applications, environmental science, geoscience, information science, materials science, mathematics, nanoscience, nuclear engineering, and physics. Most important, many projects have achieved results that would have been unobtainable without such a computing resource. The LCRC continues to foster growth in the computational science and engineering capability and quality at the Laboratory. Specific goals include expansion of the use of Jazz to new disciplines and Laboratory initiatives, teaming with Laboratory infrastructure providers to offer more scientific data management capabilities, expanding Argonne staff use of national computing facilities, and improving the scientific reach and

  13. A Small-Animal Irradiation Facility for Neutron Capture Therapy Research at the RA-3 Research Reactor

    SciTech Connect

    Emiliano Pozzi; David W. Nigg; Marcelo Miller; Silvia I. Thorp; Amanda E. Schwint; Elisa M. Heber; Veronica A. Trivillin; Leandro Zarza; Guillermo Estryk

    2007-11-01

    The National Atomic Energy Commission of Argentina (CNEA) has constructed a thermal neutron source for use in Boron Neutron Capture Therapy (BNCT) applications at the RA-3 research reactor facility located in Buenos Aires. The Idaho National Laboratory (INL) and CNEA have jointly conducted some initial neutronic characterization measurements for one particular configuration of this source. The RA-3 reactor (Figure 1) is an open pool type reactor, with 20% enriched uranium plate-type fuel and light water coolant. A graphite thermal column is situated on one side of the reactor as shown. A tunnel penetrating the graphite structure enables the insertion of samples while the reactor is in normal operation. Samples up to 14 cm height and 15 cm width are accommodated.

  14. Leu conversion status of U.S. research reactors: September 1996

    SciTech Connect

    Matos, J.E.

    1996-09-01

    At the request of the Department of Energy, the RERTR Program has summarized the conversion status of research and test reactors in the United States and has made estimates of the uranium densities that would be needed to convert the reactors with power levels greater than or equal to 1 MW from Highly Enriched Uranium (HEU) (greater than or equal to 20% U-235) to Lightly Enriched Uranium (LEU) (less than 20% U-235) fuels. Detailed conversion studies for each of the reactors need to be completed in order to establish the feasibility of using LEU fuels.

  15. Spain-Chile and Spain-Ecuador cooperation in the field of research nuclear reactors

    SciTech Connect

    Avendano, G.; Rodriguez, M.L.; Manas, L.; Masalleras, E.; Montes, J.

    1981-01-01

    The Spanish Board of Nuclear Energy (JEN) has been cooperating for the last several years with the Comision Chilena de Energia Nuclear (Chilean Commission of Nuclear Energy (CCHEN)), on the one hand, and with the Comision Ecuatoriana de Energia Atomica (Ecuadorian Commission of Atomic Energy (CEEA)), on the other. The result of this cooperation has been the implementation of projects in both countries to create research centers around a nuclear reactor as the main working tool: the Lo Aguirre reactor in Chile and the Ruminahui reactor in Ecuador.

  16. Argonne`s Expedited Site Characterization: An integrated approach to cost- and time-effective remedial investigation

    SciTech Connect

    Burton, J.C.; Walker, J.L.; Aggarwal, P.K.; Meyer, W.T.

    1995-07-01

    Argonne National Laboratory has developed a methodology for remedial site investigation that has proven to be both technically superior to and more cost- and time-effective than traditional methods. This methodology is referred to as the Argonne Expedited Site Characterization (ESC). Quality is the driving force within the process. The Argonne ESC process is abbreviated only in time and cost and never in terms of quality. More usable data are produced with the Argonne ESC process than with traditional site characterization methods that are based on statistical-grid sampling and multiple monitoring wells. This paper given an overview of the Argonne ESC process and compares it with traditional methods for site characterization. Two examples of implementation of the Argonne ESC process are discussed to illustrate the effectiveness of the process in CERCLA (Comprehensive Environmental Response, Compensation, and Liability Act) and RCRA (Resource Conservation and Recovery Act) programs.

  17. MITR-III: Upgrade and relicensing studies for the MIT Research Reactor. Second annual report

    SciTech Connect

    Trosman, H.G.; Lanning, D.D.; Harling, O.K.

    1994-08-01

    The current operating license of the MIT research reactor will expire on May 7, 1996 or possibly a few years later if the US Nuclear Regulatory Commission agrees that the license period can start with the date of initial reactor operation. Driven by the imminent expiration of the operating license, a team of nuclear engineering staff and students have begun a study of the future options for the MIT Research Reactor. These options have included the range from a major rebuilding of the reactor to its decommissioning. This document reports the results of a two year intensive activity which has been supported by a $148,000 grant from the USDOE contract Number DEFG0293ER75859, approximately $100,000 of internal MIT funds and Nuclear Engineering Department graduate student fellowships as well as assistance from international visiting scientists and engineers.

  18. Photon spectrum behind biological shielding of the LVR-15 research reactor

    SciTech Connect

    Klupak, V.; Viererbl, L.; Lahodova, Z.; Marek, M.; Vins, M.

    2011-07-01

    The LVR-15 reactor is a light water research reactor situated at the Research Centre Rez, near Prague. It operates as a multipurpose facility with a maximum thermal power of 10 MW. The reactor core usually contains from 28 to 32 fuel assemblies with a total mass of {sup 235}U of about 5 kg. Emitted radiation from the fuel caused by fission is shielded by moderating water, a steel reactor vessel, and heavy concrete. This paper deals with measurement and analysis of the gamma spectrum near the outer surface of the concrete wall, behind biological shielding, mainly in the 3- to 10-MeV energy range. A portable HPGe detector with a portable multichannel analyzer was used to measure gamma spectra. The origin of energy lines in gamma detector spectra was identified. (authors)

  19. Reactor safety research programs. Quarterly report, April-June 1982

    SciTech Connect

    Edler, S.K.

    1982-11-01

    This document summarizes work performed by Pacific Northwest Laboratory (PNL) from April 1 through June 30, 1982, for the Division of Accident Evaluation and the Division of Engineering Technology, US Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities.

  20. Reactor safety research programs. Quarterly report, July-September 1983

    SciTech Connect

    Edler, S.K.

    1984-04-01

    Evaluations of nondestructive examination (NDE) techniques and instrumentation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, and examining NDE reliability and probabilistic fracture mechanics. Accelerated pellet-cladding interaction modeling is being conducted to predict the probability of fuel rod failure under normal operating conditions. Experimental data and analytical models are being provided to aid in decision making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Experimental data and validated models are being used to determine a method for evaluating the acceptance of welded or weld-repaired stainless steel piping. Thermal-hydraulic models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. High-temperature materials property tests are being conducted to provide data on severe core damage fuel behavior. Severe fuel damage accident tests are being conducted at the NRU reactor, Chalk River, Canada; and an instrumented fuel assembly irradiation program is being performed at Halden, Norway. Fuel assemblies and analytical support are being provided for experimental programs at other facilities, including the Super Sara Test Program, Ispra, Italy, and experimental programs at the Power Burst Facility.

  1. Reactor safety research programs. Quarterly report, January-March 1982

    SciTech Connect

    Edler, S.K.

    1982-07-01

    This document summarizes work performed by Pacific Northwest Laboratory (PNL) from January 1 through March 31, 1982, for the Division of Accident Evaluation and the Division of Engineering Technology, US Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities.

  2. Argonne Bubble Experiment Thermal Model Development II

    SciTech Connect

    Buechler, Cynthia Eileen

    2016-07-01

    This report describes the continuation of the work reported in “Argonne Bubble Experiment Thermal Model Development”. The experiment was performed at Argonne National Laboratory (ANL) in 2014. A rastered 35 MeV electron beam deposited power in a solution of uranyl sulfate, generating heat and radiolytic gas bubbles. Irradiations were performed at three beam power levels, 6, 12 and 15 kW. Solution temperatures were measured by thermocouples, and gas bubble behavior was observed. This report will describe the Computational Fluid Dynamics (CFD) model that was developed to calculate the temperatures and gas volume fractions in the solution vessel during the irradiations. The previous report described an initial analysis performed on a geometry that had not been updated to reflect the as-built solution vessel. Here, the as-built geometry is used. Monte-Carlo N-Particle (MCNP) calculations were performed on the updated geometry, and these results were used to define the power deposition profile for the CFD analyses, which were performed using Fluent, Ver. 16.2. CFD analyses were performed for the 12 and 15 kW irradiations, and further improvements to the model were incorporated, including the consideration of power deposition in nearby vessel components, gas mixture composition, and bubble size distribution. The temperature results of the CFD calculations are compared to experimental measurements.

  3. Boron neutron capture therapy and radiation synovectomy research at the Massachusetts Institute of Technology Research Reactor

    SciTech Connect

    Zamenhof, R.G.; Nwanguma, C.I.; Wazer, D.E.; Saris, S.; Madoc-Jones, H. ); Sledge, C.B.; Shortkroff, S. )

    1992-04-01

    In this paper, current research in boron neutron capture therapy (BNCT) and radiation synovectomy at the Massachusetts Institute of Technology Research Reactor is reviewed. In the last few years, major emphasis has been placed on the development of BNCT primarily for treatment of brain tumors. This has required a concerted effort in epithermal beam design and construction as well as the development of analytical capabilities for {sup 10}B analysis and patient treatment planning. Prompt gamma analysis and high-resolution track-etch autoradiography have been developed to meet the needs, respectively, for accurate bulk analysis and for quantitative imaging of {sup 10}B in tissue at subcellular resolutions. Monte Carlo-based treatment planning codes have been developed to ensure optimized and individualized patient treatments. In addition, the development of radiation synovectomy as an alternative therapy to surgical intervention is joints that are affected by rheumatoid arthritis is described.

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

    SciTech Connect

    Adams, T.M.; Dunsmuir, M.D.; Leduc, D.R.; Severynse, T.F.; Sindelar, R.L.; Moore, E.N.

    2013-07-01

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

  5. Safety Issues at the DOE Test and Research Reactors. A Report to the U.S. Department of Energy.

    ERIC Educational Resources Information Center

    National Academy of Sciences - National Research Council, Washington, DC. Commission on Physical Sciences, Mathematics, and Resources.

    This report provides an assessment of safety issues at the Department of Energy (DOE) test and research reactors. Part A identifies six safety issues of the reactors. These issues include the safety design philosophy, the conduct of safety reviews, the performance of probabilistic risk assessments, the reliance on reactor operators, the fragmented…

  6. Assessment of Sensor Technologies for Advanced Reactors

    SciTech Connect

    Korsah, Kofi; Ramuhalli, Pradeep; Vlim, R.; Kisner, Roger A.; Britton, Jr, Charles L.; Wootan, D. W.; Anheier, Jr, N. C.; Diaz, A. A.; Hirt, E. H.; Chien, H. T.; Sheen, S.; Bakhtiari, Sasan; Gopalsami, S.; Heifetz, A.; Tam, S. W.; Park, Y.; Upadhyaya, B. R.; Stanford, A.

    2016-10-01

    Sensors and measurement technologies provide information on processes, support operations and provide indications of component health. They are therefore crucial to plant operations and to commercialization of advanced reactors (AdvRx). This report, developed by a three-laboratory team consisting of Argonne National Laboratory (ANL), Oak Ridge National Laboratory (ORNL) and Pacific Northwest National Laboratory (PNNL), provides an assessment of sensor technologies and a determination of measurement needs for AdvRx. It provides the technical basis for identifying and prioritizing research targets within the instrumentation and control (I&C) Technology Area under the Department of Energy’s (DOE’s) Advanced Reactor Technology (ART) program and contributes to the design and implementation of AdvRx concepts.

  7. The Sodium Process Facility at Argonne National Laboratory-West

    SciTech Connect

    Michelbacher, J.A.; Henslee, S.P. McDermott, M.D.; Price, J.R.; Rosenberg, K.E.; Wells, P.B.

    1998-07-01

    Argonne National Laboratory-West (ANL-W) has approximately 680,000 liters of raw sodium stored in facilities on site. As mandated by the State of Idaho and the US Department of Energy (DOE), this sodium must be transformed into a stable condition for land disposal. To comply with this mandate, ANL-W designed and built the Sodium Process Facility (SPF) for the processing of this sodium into a dry, sodium carbonate powder. The major portion of the sodium stored at ANL-W is radioactively contaminated. The sodium will be processed in three separate and distinct campaigns: the 290,000 liters of Fermi-1 primary sodium, the 50,000 liters of the Experimental Breeder Reactor-II (EBR-II) secondary sodium, and the 330,000 liters of the EBR-II primary sodium. The Fermi-1 and the EBR-II secondary sodium contain only low-level of radiation, while the EBR-II primary sodium has radiation levels up to 0.5 mSv (50 mrem) per hour at 1 meter. The EBR-II primary sodium will be processed last, allowing the operating experience to be gained with the less radioactive sodium prior to reacting the most radioactive sodium. The sodium carbonate will be disposed of in 270 liter barrels, four to a pallet. These barrels are square in cross-section, allowing for maximum utilization of the space on a pallet, minimizing the required landfill space required for disposal.

  8. Fission energy: The integral fast reactor

    SciTech Connect

    Chang, Yoon I.

    1989-01-01

    The Integral Fast Reactor (IFR) is an innovative reactor concept being developed at Argonne National Laboratory as a such next- generation reactor concept. The IFR concept has a number of specific technical advantages that collectively address the potential difficulties facing the expansion of nuclear power deployment. In particular, the IFR concept can meet all three fundamental requirements needed in a next-generation reactor as discussed below. This document discusses these requirements.

  9. Quality management in BNCT at a nuclear research reactor.

    PubMed

    Sauerwein, Wolfgang; Moss, Raymond; Stecher-Rasmussen, Finn; Rassow, Jürgen; Wittig, Andrea

    2011-12-01

    Each medical intervention must be performed respecting Health Protection directives, with special attention to Quality Assurance (QA) and Quality Control (QC). This is the basis of safe and reliable treatments. BNCT must apply QA programs as required for performance and safety in (conventional) radiotherapy facilities, including regular testing of performance characteristics (QC). Furthermore, the well-established Quality Management (QM) system of the nuclear reactor used has to be followed. Organization of these complex QM procedures is offered by the international standard ISO 9001:2008.

  10. Argonne National Lab - Theory and Computing Sciences, Accelerating Scientific Discovery

    SciTech Connect

    Beckman, Pete

    2009-01-01

    Argonne's new TCS building houses all of Argonne's computing divisions, and is designed to foster collaboration of the Manhattan Project model "Getting the best people together and having them work on a problem with singular determination." More at http://www.anl.gov/Media_Center/News/2009/tcs0910.html

  11. Bureau of Indian Education Many Farms Training Program at Argonne

    SciTech Connect

    2010-01-01

    Bureau of Indian Education Many Farms Training Program for Renewable Energy at Argonne National Laboratory. Principal Contacts; Harold Myron (ANL), Anthony Dvorak (ANL), Freddie Cardenas (BIA). Supported by; United States Department of the Interior, Bureau of Indian Education, and Argonne National Laboratory

  12. Bureau of Indian Education Many Farms Training Program at Argonne

    ScienceCinema

    None

    2016-07-12

    Bureau of Indian Education Many Farms Training Program for Renewable Energy at Argonne National Laboratory. Principal Contacts; Harold Myron (ANL), Anthony Dvorak (ANL), Freddie Cardenas (BIA). Supported by; United States Department of the Interior, Bureau of Indian Education, and Argonne National Laboratory

  13. Argonne's program in heavy-ion fusion

    SciTech Connect

    Martin, R.L.; Arnold, R.; Burke, R.; Watson, J.

    1980-01-01

    Argonne is constructing the initial part of a xenon accelerator which could serve as the driver for an Inertial Confinement Fusion power plant. The project consists of a Xe source, a 1.5 Mv Dynamitron preaccelerator, independently phased rf cavities as a low ..beta.. linac, and Wideroe linac at 12.5 and 25 MHz. The beam will be injected at 220 MeV into a circular ring to its space charge limit, transferred to a second ring for multiturn injection after an x-y rotation, extracted, compressed in time duration, transported and focused onto foil targets. The project thus will demonstrate nearly all of the beam manipulations involved in a full scale ICF driver. The feasibility of accelerating in the accumulator ring to 10 GeV to achieve adequate beam energy to permit energy deposition experiments in hot plasma is being studied. The status and future plans of this project are discussed.

  14. Argonne potential and multi-neutron systems

    SciTech Connect

    Gridnev, D. K.; Gridnev, K. A.; Schramm, S.; Greiner, Walter

    2009-01-01

    Recently it was proved that the neutron matter interacting through Argonne V18 pair-potential plus modern variants of Urbana or Illinois three-body forces is unstable. For the energy of N neutrons E(N), which interact through these forces one has E(N) = −cN³⁺+O(N{sup 8/3}), where c > 0 is a constant. This means that: (i) the energy per particle and neutron density diverge rapidly for large neutron numbers; (ii) bound states of N neutrons exist for N large enough. The neutron matter collapse is possible due to the form of the repulsive core in three-body forces, which vanishes when three nucleons occupy the same site in space. The obtained results partly change the paradigm, in which the stability of neutron stars is attained through the Pauli principle; the strong repulsive core in the nucleon interactions is by no means less important.

  15. Using low-enriched uranium in research reactors: The RERTR program

    SciTech Connect

    Travelli, A.

    1994-05-01

    The goal of the RERTR program is to minimize and eventually eliminate use of highway enriched uranium (HEU) in research and test reactors. The program has been very successful, and has developed low-enriched uranium (LEU) fuel materials and designs which can be used effectively in approximately 90 percent of the research and test reactors which used HEU when the program began. This progress would not have been possible without active international cooperation among fuel developers, commercial vendors, and reactor operators. The new tasks which the RERTR program is undertaking at this time include development of new and better fuels that will allow use of LEU fuels in all research and test reactors; cooperation with Russian laboratories, which will make it possible to minimize and eventually eliminate use of HEU in research reactors throughout the world, irrespective of its origin; and development of an LEU-based process for the production of {sup 99}Mo. Continuation and intensification of international cooperation are essential to the achievement of the ultimate goals of the RERTR program.

  16. Environmental monitoring at Argonne National Laboratory. Annual report, 1981

    SciTech Connect

    Golchert, N.W.; Duffy, T.L.; Sedlet, J.

    1982-03-01

    The results of the environmental monitoring program at Argonne National Laboratory for 1981 are presented and discussed. To evaluate the effect of Argonne operations on the environment, measurements were made for a variety of radionuclides in air, surface water, soil, grass, bottom sediment, and milk; for a variety of chemical constituents in air, surface water, and Argonne effluent water; and of the environmental penetrating radiation dose. Sample collections and measurements were made at the site boundary and off the Argonne site for comparison purposes. Some on-site measurements were made to aid in the interpretation of the boundary and off-site data. The results of the program are interpreted in terms of the sources and origin of the radioactive and chemical substances (natural, fallout, Argonne, and other) and are compared with applicable environmental quality standards. The potential radiation dose to off-site population groups is also estimated.

  17. Environmental monitoring at Argonne National Laboratory. Annual report for 1980

    SciTech Connect

    Golchert, N. W.; Duffy, T. L.; Sedlet, J.

    1981-03-01

    The results of the environmental monitoring program at Argonne National Laboratory for 1980 are presented and discussed. To evaluate the effect of Argonne operations on the environment, measurements were made for a variety of radionuclides in air, surface water, soil, grass, bottom sediment, and foodstuffs; for a variety of chemical constituents in air, surface water, and Argonne effluent water; and of the environmental penetrating radiation dose. Sample collections and measurements were made at the site boundary and off the Argonne site for comparison purposes. Some on-site measurements were made to aid in the interpretation of the boundary and off-site data. The results of the program are interpreted in terms of the sources and origin of the radioactive and chemical substances (natural, fallout, Argonne, and other) and are compared with applicable environmental quality standards. The potential radiation dose to off-site population groups is also estimated.

  18. Environmental monitoring at Argonne National Laboratory. Annual report for 1984

    SciTech Connect

    Golchert, N.W.; Duffy, T.L.; Sedlet, J.

    1985-03-01

    The results of the environmental monitoring program at Argonne National Laboratory for 1984 are presented and discussed. To evaluate the effect of Argonne operations on the environment, measurements were made for a variety of radionuclides in air, surface water, ground water, soil, grass, bottom sediment, and milk; for a variety of chemical constituents in surface water, ground water, and Argonne effluent water; and of the environmental penetrating radiation dose. Sample collections and measurements were made on the site, at the site boundary, and off the Argonne site for comparison purposes. The potential radiation dose to off-site population groups is also estimated. The results of the program are interpreted in terms of the sources and origin of the radioactive and chemical substances (natural, fallout, Argonne, and other) and are compared with applicable environmental quality standards. 20 refs., 8 figs., 46 tabs.

  19. Environmental monitoring at Argonne National Laboratory. Annual report for 1982

    SciTech Connect

    Golchert, N.W.; Duffy, T.L.; Sedlet, J.

    1983-03-01

    The results of the environmental monitoring program at Argonne Ntaional Laboratory for 1982 are presented and discussed. To evaluate the effect of Argonne operations on the environment, measurements were made for a variety of radionuclides in air, surface water, soil, grass, bottom sediment, and milk; for a variety of chemical constituents in air, surface water, ground water, and Argonne effluent water; and of the environmental penetrating radiation dose. Sample collections and masurements were made at the site boundary and off the Argonne site for comparison purposes. Some on-site measurements were made to aid in the interpretation of the boundary and off-site data. The results of the program are interpreted in terms of the sources and origin of the radioactive and chemical substances (natural, fallout, Argonne, and other) and are compared with applicable environmental quality standards. The potential radiation dose to off-site population groups is also estimated.

  20. Environmental assessment related to the operation of Argonne National Laboratory, Argonne, Illinois

    SciTech Connect

    Not Available

    1982-08-01

    In order to evaluate the environmental impacts of Argonne National Laboratory (ANL) operations, this assessment includes a descriptive section which is intended to provide sufficient detail to allow the various impacts to be viewed in proper perspective. In particular, details are provided on site characteristics, current programs, characterization of the existing site environment, and in-place environmental monitoring programs. In addition, specific facilities and operations that could conceivably impact the environment are described at length. 77 refs., 16 figs., 47 tabs.

  1. Joint Assessment of ETRR-2 Research Reactor Operations Program, Capabilities, and Facilities

    SciTech Connect

    Bissani, M; O'Kelly, D S

    2006-05-08

    A joint assessment meeting was conducted at the Egyptian Atomic Energy Agency (EAEA) followed by a tour of Egyptian Second Research Reactor (ETRR-2) on March 22 and 23, 2006. The purpose of the visit was to evaluate the capabilities of the new research reactor and its operations under Action Sheet 4 between the U.S. DOE and the EAEA, ''Research Reactor Operation'', and Action Sheet 6, ''Technical assistance in The Production of Radioisotopes''. Preliminary Recommendations of the joint assessment are as follows: (1) ETRR-2 utilization should be increased by encouraging frequent and sustained operations. This can be accomplished in part by (a) Improving the supply-chain management for fresh reactor fuel and alleviating the perception that the existing fuel inventory should be conserved due to unreliable fuel supply; and (b) Promulgating a policy for sample irradiation priority that encourages the use of the reactor and does not leave the decision of when to operate entirely at the discretion of reactor operations staff. (2) Each experimental facility in operation or built for a single purpose should be reevaluated to focus on those that most meet the goals of the EAEA strategic business plan. Temporary or long-term elimination of some experimental programs might be necessary to provide more focused utilization. There may be instances of emerging reactor applications for which no experimental facility is yet designed or envisioned. In some cases, an experimental facility may have a more beneficial use than the purpose for which it was originally designed. For example, (a) An effective Boron Neutron Capture Therapy (BNCT) program requires nearby high quality medical facilities. These facilities are not available and are unlikely to be constructed near the Inshas site. Further, the BNCT facility is not correctly designed for advanced research and therapy programs using epithermal neutrons. (b) The ETRR-2 is frequently operated to provide color-enhanced gemstones but is

  2. The Need for Cyber-Informed Engineering Expertise for Nuclear Research Reactors

    SciTech Connect

    Anderson, Robert Stephen

    2015-12-01

    Engineering disciplines may not currently understand or fully embrace cyber security aspects as they apply towards analysis, design, operation, and maintenance of nuclear research reactors. Research reactors include a wide range of diverse co-located facilities and designs necessary to meet specific operational research objectives. Because of the nature of research reactors (reduced thermal energy and fission product inventory), hazards and risks may not have received the same scrutiny as normally associated with power reactors. Similarly, security may not have been emphasized either. However, the lack of sound cybersecurity defenses may lead to both safety and security impacts. Risk management methodologies may not contain the foundational assumptions required to address the intelligent adversary’s capabilities in malevolent cyber attacks. Although most research reactors are old and may not have the same digital footprint as newer facilities, any digital instrument and control function must be considered as a potential attack platform that can lead to sabotage or theft of nuclear material, especially for some research reactors that store highly enriched uranium. This paper will provide a discussion about the need for cyber-informed engineering practices that include the entire engineering lifecycle. Cyber-informed engineering as referenced in this paper is the inclusion of cybersecurity aspects into the engineering process. A discussion will consider several attributes of this process evaluating the long-term goal of developing additional cyber safety basis analysis and trust principles. With a culture of free information sharing exchanges, and potentially a lack of security expertise, new risk analysis and design methodologies need to be developed to address this rapidly evolving (cyber) threatscape.

  3. Research and development of an electrochemical biocide reactor

    NASA Technical Reports Server (NTRS)

    See, G. G.; Bodo, C. A.; Glennon, J. P.

    1975-01-01

    An alternate disinfecting process to chemical agents, heat, or radiation in an aqueous media has been studied. The process is called an electrochemical biocide and employs cyclic, low-level voltages at chemically inert electrodes to pass alternating current through water and, in the process, to destroy microorganisms. The paper describes experimental hardware, methodology, and results with a tracer microorganism (Escherichia coli). The results presented show the effects on microorganism kill of operating parameters, including current density (15 to 55 mA/sq cm (14 to 51 ASF)), waveform of applied electrical signal (square, triangular, sine), frequency of applied electrical signal (0.5 to 1.5 Hz), process water flow rate (100 to 600 cc/min (1.6 to 9.5 gph)), and reactor resident time (0 to 4 min). Comparisons are made between the disinfecting property of the electrochemical biocide and chlorine, bromine, and iodine.

  4. Estimation of Na-24 activity concentration in BAEC TRIGA Research Reactor

    NASA Astrophysics Data System (ADS)

    Ajijul Hoq, M.; Malek Soner, M. A.; Salam, M. A.; Khanom, Salma; Fahad, S. M.

    The Bangladesh Atomic Energy Commission (BAEC) TRIGA Research Reactor is a unique nuclear installation of the country generally implemented for a wide variety of research applications and serves as an excellent source of neutron. During reactor operation it is necessary to measure and control the activity concentration of the pool water for fuel element failure detection and for the determination of contamination. The present study deals with the estimation of activity concentration for Na-24 present in water coolant produced as a result of 23Na (n, γ) 24Na reaction. Several governing equations have been employed to estimate the Na-24 activity concentrations theoretically at different reactor power levels including maximum reactor power of 2.4 MW. From the obtained result it is ensured that the estimated Na-24 activity of 8.83 × 10-3 μCi /cm3 is not significant enough for any radiological hazard. Thus for ensuring radiological safety issues of the research reactor the assessment performed under the present study has an implication.

  5. Radioisotope radiotherapy research and achievements at the University of Missouri Research Reactor

    NASA Astrophysics Data System (ADS)

    Ehrhardt, G. J.; Ketring, A. R.; Cutler, C. S.

    2003-01-01

    The University of Missouri Research Reactor (MURR) in collaboration with faculty in other departments at the University of Missouri has been involved in developing new means of internal radioisotopic therapy for cancer for many years. These efforts have centered on methods of targeting radioisotopes such as brachytherapy, embolisation of liver tumors with radioactive microspheres, small-molecule-labelled chelates for the treatment of bone cancer, and various means of radioimmunotherapy or labelled receptor agent targeting. This work has produced two radioactive agents, Sm-153 Quadramet™ and Y-90 TheraSphere™, which have U.S. Food and Drug Administration approval for the palliation of bone cancer pain and treatment of inoperable liver cancer, respectively. MURR has also pioneered development of other beta-emitting isotopes for internal radiotherapy such as Re-186, Re-188, Rh-105, Ho-166, Lu-177, and Pm-149, many of which are in research and clinical trials throughout the U.S. and the world. This important work has been made possible by the very high neutron flux available at MURR combined with MURR's outstanding reliability of operation and flexibility in meeting the needs of researchers and the radiopharmaceutical industry.

  6. Radiation dosimetry for NCT facilities at the Brookhaven Medical Research Reactor

    SciTech Connect

    Holden, N.E.; Hu, J.P.; Greenberg, D.D.; Reciniello, R.N.

    1998-12-31

    Brookhaven Medical Research Reactor (BMRR) is a 3 mega-watt (MW) heterogeneous, tank-type, light water cooled and moderated, graphite reflected reactor, which was designed for medical and biological studies and became operational in 1959. Over time, the BMRR was modified to provide thermal and epithermal neutron beams suitable for research studies. NCT studies have been performed at both the epithermal neutron irradiation facility (ENIF) on the east side of the BMRR reactor core and the thermal neutron irradiation facility (TNIF) on the west side of the core. Neutron and gamma-ray dosimetry performed from 1994 to the present in both facilities are described and the results are presented and discussed.

  7. Characterization of the Annular Core Research Reactor (ACRR) Neutron Radiography System Imaging Plane

    NASA Astrophysics Data System (ADS)

    Kaiser, Krista; Chantel Nowlen, K.; DePriest, K. Russell

    2016-02-01

    The Annular Core Research Reactor (ACRR) at Sandia National Laboratories (SNL) is an epithermal pool-type research reactor licensed up to a thermal power of 2.4 MW. The ACRR facility has a neutron radiography facility that is used for imaging a wide range of items including reactor fuel and neutron generators. The ACRR neutron radiography system has four apertures (65:1, 125:1, 250:1, and 500:1) available to experimenters. The neutron flux and spectrum as well as the gamma dose rate were characterized at the imaging plane for the ACRR's neutron radiography system for the 65:1, 125:1 and 250:1 apertures.

  8. Initial verification and validation of RAZORBACK - A research reactor transient analysis code

    SciTech Connect

    Talley, Darren G.

    2015-09-01

    This report describes the work and results of the initial verification and validation (V&V) of the beta release of the Razorback code. Razorback is a computer code designed to simulate the operation of a research reactor (such as the Annular Core Research Reactor (ACRR)) by a coupled numerical solution of the point reactor kinetics equations, the energy conservation equation for fuel element heat transfer, and the mass, momentum, and energy conservation equations for the water cooling of the fuel elements. This initial V&V effort was intended to confirm that the code work to-date shows good agreement between simulation and actual ACRR operations, indicating that the subsequent V&V effort for the official release of the code will be successful.

  9. Determination of the optimal positions for installing gamma ray detection systems at Tehran Research Reactor

    NASA Astrophysics Data System (ADS)

    Sayyah, A.; Rahmani, F.; Khalafi, H.

    2015-09-01

    Dosimetric instruments must constantly monitor radiation dose levels in different areas of nuclear reactor. Tehran Research Reactor (TRR) has seven beam tubes for different research purposes. All the beam tubes extend from the reactor core to Beam Port Floor (BPF) of the reactor facility. During the reactor operation, the gamma rays exiting from each beam tube outlet produce a specific gamma dose rate field in the space of the BPF. To effectively monitor the gamma dose rates on the BPF, gamma ray detection systems must be installed in optimal positions. The selection of optimal positions is a compromise between two requirements. First, the installation positions must possess largest gamma dose rates and second, gamma ray detectors must not be saturated in these positions. In this study, calculations and experimental measurements have been carried out to identify the optimal positions of the gamma ray detection systems. Eight three dimensional models of the reactor core and related facilities corresponding to eight scenarios have been simulated using MCNPX Monte Carlo code to calculate the gamma dose equivalent rate field in the space of the BPF. These facilities are beam tubes, thermal column, pool, BPF space filled with air, facilities such as neutron radiography facility, neutron powder diffraction facility embedded in the beam tubes as well as biological shields inserted into the unused beam tubes. According to the analysis results of the combined gamma dose rate field, three positions on the north side and two positions on the south side of the BPF have been recognized as optimal positions for installing the gamma ray detection systems. To ensure the consistency of the simulation data, experimental measurements were conducted using TLDs (600 and 700) pairs during the reactor operation at 4.5 MW.

  10. A reload and startup plan for conversion of the NIST research reactor

    SciTech Connect

    D. J. Diamond

    2016-03-31

    The National Institute of Standards and Technology operates a 20 MW research reactor for neutron-based research. The heavy-water moderated and cooled reactor is fueled with high-enriched uranium (HEU) but a program to convert the reactor to low-enriched uranium (LEU) fuel is underway. Among other requirements, a reload and startup test plan must be submitted to the U.S. Nuclear Regulatory Commission (NRC) for their approval. The NRC provides guidance for what should be in the plan to ensure that the licensee has sufficient information to operate the reactor safely. Hence, a plan has been generated consisting of two parts. The reload portion of the plan specifies the fuel management whereby initially only two LEU fuel elements are in the core for eight fuel cycles. This is repeated until a point when the optimum approach is to place four fresh LEU elements into the reactor each cycle. This final transition is repeated and after eight cycles the reactor is completely fueled with LEU. By only adding two LEU fuel elements initially, the plan allows for the consumption of HEU fuel elements that are expected to be in storage at the time of conversion and provides additional qualification of production LEU fuel under actual operating conditions. Because the reload is to take place over many fuel cycles, startup tests will be done at different stages of the conversion. The tests, to be compared with calculations to show that the reactor will operate as planned, are the measurement of critical shim arm position and shim arm and regulating rod reactivity worths. An acceptance criterion for each test is specified based on technical specifications that relate to safe operation. Additional tests are being considered that have less safety significance but may be of interest to bolster the validation of analysis tools.

  11. Research Reactor Preparations for the Air Shipment of Highly Enriched Uranium from Romania

    SciTech Connect

    K. J. Allen; I. Bolshinsky; L. L. Biro; M. E. Budu; N. V. Zamfir; M. Dragusin; C. Paunoiu; M. Ciocanescu

    2010-03-01

    In June 2009 two air shipments transported both unirradiated (fresh) and irradiated (spent) Russian-origin highly enriched uranium (HEU) nuclear fuel from two research reactors in Romania to the Russian Federation for conversion to low enriched uranium. The Institute for Nuclear Research at Pitesti (SCN Pitesti) shipped 30.1 kg of HEU fresh fuel pellets to Dimitrovgrad, Russia and the Horia Hulubei National Institute of Physics and Nuclear Engineering (IFIN-HH) shipped 23.7 kilograms of HEU spent fuel assemblies from the VVR S research reactor at Magurele, Romania, to Chelyabinsk, Russia. Both HEU shipments were coordinated by the Russian Research Reactor Fuel Return Program (RRRFR) as part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), were managed in Romania by the National Commission for Nuclear Activities Control (CNCAN), and were conducted in cooperation with the Russian Federation State Corporation Rosatom and the International Atomic Energy Agency. Both shipments were transported by truck to and from respective commercial airports in Romania and the Russian Federation and stored at secure nuclear facilities in Russia until the material is converted into low enriched uranium. These shipments resulted in Romania becoming the 3rd country under the RRRFR program and the 14th country under the GTRI program to remove all HEU. This paper describes the research reactor preparations and license approvals that were necessary to safely and securely complete these air shipments of nuclear fuel.

  12. Recent accomplishments in neutron beam projects at the University of Texas Research Reactor

    SciTech Connect

    Uenlue, K.; Wehring, B.W.

    1994-12-31

    The design of a cold neutron source facility at the University of Texas TRIGA research reactor is described. The UT-TRIGA has 5 neutron beam ports. Because of the different characteristics of the ports, various research projects are being pursued. Among these projects, The Texas cold neutron source and neutron depth profiling are operational; neutron focusing, prompt gamma activation analysis, and neutron capture therapy research are progressing.

  13. Argonne National Laboratory summary site environmental report for calendar year 2007.

    SciTech Connect

    Golchert, N. W.

    2009-05-22

    This summary of Argonne National Laboratory's Site Environmental Report for calendar year 2007 was written by 20 students at Downers Grove South High School in Downers Grove, Ill. The student authors are classmates in Mr. Howard's Bio II course. Biology II is a research-based class that teaches students the process of research by showing them how the sciences apply to daily life. For the past seven years, Argonne has worked with Biology II students to create a short document summarizing the Site Environmental Report to provide the public with an easy-to-read summary of the annual 300-page technical report on the results of Argonne's on-site environmental monitoring program. The summary is made available online and given to visitors to Argonne, researchers interested in collaborating with Argonne, future employees, and many others. In addition to providing Argonne and the public with an easily understandable short summary of a large technical document, the participating students learn about professional environmental monitoring procedures, achieve a better understanding of the time and effort put forth into summarizing and publishing research, and gain confidence in their own abilities to express themselves in writing. The Argonne Summary Site Environmental Report fits into the educational needs for 12th grade students. Illinois State Educational Goal 12 states that a student should understand the fundamental concepts, principles, and interconnections of the life, physical, and earth/space sciences. To create this summary booklet, the students had to read and understand the larger technical report, which discusses in-depth many activities and programs that have been established by Argonne to maintain a safe local environment. Creating this Summary Site Environmental Report also helps students fulfill Illinois State Learning Standard 12B5a, which requires that students be able to analyze and explain biodiversity issues, and the causes and effects of extinction. The

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

    SciTech Connect

    Vinson, D.

    2010-07-11

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

  15. 75 FR 62892 - Massachusetts Institute of Technology Research Reactor Environmental Assessment and Finding of No...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-10-13

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Massachusetts Institute of Technology Research Reactor Environmental Assessment and Finding of No Significant Impact Correction In notice document 2010-24809 beginning on page 61220 in the issue of...

  16. Status of DOE efforts to renew acceptance of foreign research reactor spent nuclear fuel

    SciTech Connect

    Head, C.R.

    1997-08-01

    This presentation summarizes the efforts being made by the Department of Energy to renew acceptance of spent nuclear fuel shipments from foreign research reactors. The author reviews the actions undertaken in this process in a fairly chronological manner, through the present time, as well as the development of an environmental impact statement to support the proposed actions.

  17. Environmental Assessment of Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel

    SciTech Connect

    Not Available

    1994-04-01

    The Department of Energy has completed the Environmental Assessment (EA) of Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel and issued a Finding of No Significant Impact (FONSI) for the proposed action. The EA and FONSI are enclosed for your information. The Department has decided to accept a limited number of spent nuclear fuel elements (409 elements) containing uranium that was enriched in the United States from eight research reactors in Austria, Denmark, Germany, Greece, the Netherlands, Sweden, and Switzerland. This action is necessary to maintain the viability of a major US nuclear weapons nonproliferation program to limit or eliminate the use of highly enriched uranium in civil programs. The purpose of the EA is to maintain the cooperation of the foreign research reactor operators with the nonproliferation program while a more extensive Environmental Impact Statement (EIS) is prepared on a proposed broader policy involving the acceptance of up to 15,000 foreign research reactor spent fuel elements over a 10 to 15 year period. Based on an evaluation of transport by commercial container liner or chartered vessel, five eastern seaboard ports, and truck and train modes of transporting the spent fuel overland to the Savannah River Sits, the Department has concluded that no significant impact would result from any combination of port and made of transport. In addition, no significant impacts were found from interim storage of spent fuel at the Savannah River Site.

  18. Argonne National Laboratory operations during ASCOT 1991

    SciTech Connect

    Coulter, R.L.; Martin, T.J.

    1991-12-01

    The Atmospheric Studies in Complex Terrain (ASCOT) field study took place during 1991 in conjunction with a model verification exercise in and around the Rocky Flats Plant (RFP) northwest of Denver, Colorado, between January 29 and February 8. As part of this exercise, Argonne (ANL) operated a portable minisodar at several locations around REP during each of the experimental nights. In addition, ANL operated ``permanent`` network minisodars with enhanced time resolution at Coal Creek and the Bartlett Ranch for the duration of the field study. Real-time data from each minisodar are identical in format, consisting of (1) vertical profiles of horizontal wind speed along the pointing directions of each of two beams tilted from the vertical by about 17 deg, (2) vertical wind speed along a vertically pointed beam, (3) the standard deviation of the components along each of these direction, (4) the signal amplitude, and (5) the number of samples within each average that were accepted as ``good.`` The data output to the ASCOT data center consists of wind speed, wind direction, vertical wind speed, standard deviation of wind speed along the three pointing directions, and the vertical beam signal amplitude in arbitrary units. Maximum heights of the minisodars were generally limited to 300 m; however, lower heights were occasionally used to increase the sampling rate or because of limited signal strength. During this field study, operational periods were on selected nights from 20000 hr until 0500 hr the following morning.

  19. Argonne National Laboratory operations during ASCOT 1991

    SciTech Connect

    Coulter, R.L.; Martin, T.J.

    1991-12-01

    The Atmospheric Studies in Complex Terrain (ASCOT) field study took place during 1991 in conjunction with a model verification exercise in and around the Rocky Flats Plant (RFP) northwest of Denver, Colorado, between January 29 and February 8. As part of this exercise, Argonne (ANL) operated a portable minisodar at several locations around REP during each of the experimental nights. In addition, ANL operated permanent'' network minisodars with enhanced time resolution at Coal Creek and the Bartlett Ranch for the duration of the field study. Real-time data from each minisodar are identical in format, consisting of (1) vertical profiles of horizontal wind speed along the pointing directions of each of two beams tilted from the vertical by about 17 deg, (2) vertical wind speed along a vertically pointed beam, (3) the standard deviation of the components along each of these direction, (4) the signal amplitude, and (5) the number of samples within each average that were accepted as good.'' The data output to the ASCOT data center consists of wind speed, wind direction, vertical wind speed, standard deviation of wind speed along the three pointing directions, and the vertical beam signal amplitude in arbitrary units. Maximum heights of the minisodars were generally limited to 300 m; however, lower heights were occasionally used to increase the sampling rate or because of limited signal strength. During this field study, operational periods were on selected nights from 20000 hr until 0500 hr the following morning.

  20. Neutronic safety parameters and transient analyses for potential LEU conversion of the Budapest Research Reactor.

    SciTech Connect

    Pond, R. B.; Hanan, N. A.; Matos, J. E.; Maraczy, C.

    1999-09-27

    An initial safety study for potential LEU conversion of the Budapest Research Reactor was completed. The study compares safety parameters and example transients for reactor cores with HEU and LEU fuels. Reactivity coefficients, kinetic parameters and control rod worths were calculated for cores with HEU(36%) UAl alloy fuel and UO{sub 2}-Al dispersion fuel, and with LEU (19.75%)UO{sub 2}-Al dispersion fuel that has a uranium density of about 2.5 g/cm{sup 3}. A preliminary fuel conversion plan was developed for transition cores that would convert the BRR from HEU to LEU fuel after the process is begun.

  1. Argonne Fuel Cycle Facility ventilation system -- modeling and results

    SciTech Connect

    Mohr, D.; Feldman, E.E.; Danielson, W.F.

    1995-02-01

    This paper describes an integrated study of the Argonne-West Fuel Cycle Facility (FCF) interconnected ventilation systems during various operations. Analyses and test results include first a nominal condition reflecting balanced pressures and flows followed by several infrequent and off-normal scenarios. This effort is the first study of the FCF ventilation systems as an integrated network wherein the hydraulic effects of all major air systems have been analyzed and tested. The FCF building consists of many interconnected regions in which nuclear fuel is handled, transported and reprocessed. The ventilation systems comprise a large number of ducts, fans, dampers, and filters which together must provide clean, properly conditioned air to the worker occupied spaces of the facility while preventing the spread of airborne radioactive materials to clean am-as or the atmosphere. This objective is achieved by keeping the FCF building at a partial vacuum in which the contaminated areas are kept at lower pressures than the other worker occupied spaces. The ventilation systems of FCF and the EBR-II reactor are analyzed as an integrated totality, as demonstrated. We then developed the network model shown in Fig. 2 for the TORAC code. The scope of this study was to assess the measured results from the acceptance/flow balancing testing and to predict the effects of power failures, hatch and door openings, single-failure faulted conditions, EBR-II isolation, and other infrequent operations. The studies show that the FCF ventilation systems am very controllable and remain stable following off-normal events. In addition, the FCF ventilation system complex is essentially immune to reverse flows and spread of contamination to clean areas during normal and off-normal operation.

  2. Empirical correlation of residual gamma radiation resulting from operation of the Health Physics Research Reactor

    SciTech Connect

    Chou, T.L.; Ragan, G.E.; Sims, C.S.

    1985-04-01

    An empirical equation has been developed which gives gamma dose equivalent rate as a function of time, distance, and fission yield after a pulsed operation of Oak Ridge National Laboratory's (ORNL) unshielded Health Physics Research Reactor (HPRR). A related expression which is applicable to steady-state reactor operation has been mathematically derived from the aforementioned empirical equation. The two relations can be used to predict the gamma dose equivalent rate to within 25% for times between 1 minute and 90 minutes after reactor shutdown. Similar agreement is expected for up to several days. In most cases the relations are expected to overestimate the gamma dose equivalent rate. 5 refs., 4 figs., 1 tab.

  3. Status report on conversion of the Georgia Tech Research Reactor to low enrichment fuel

    SciTech Connect

    Karam, R.A. ); Matos, J.E.; Mo, S.C.; Woodruff, W.L. )

    1991-01-01

    The 5 MW Georgia Tech Research Reactor (GTRR) is a heterogeneous, heavy water moderated and cooled reactor, fueled with highly-enriched uranium aluminum alloy fuel plates. The GTRR is required to convert to low enrichment (LEU) fuel in accordance with USNRC policy. The US Department of Energy is funding a program to compare reactor performance with high and low enrichment fuels. The goals of the program are: (1) to amend the SAR and the Technical Specifications of the GTRR so that LEU U{sub 3}Si{sub 2}-Al dispersion fuel plates can replace the current HEU U-Al alloy fuel, and (2) to optimize the LEU core such that maximum value neutron beams can be extracted for possible neutron capture therapy application. This paper presents a status report on the LEU conversion effort.

  4. Status report on conversion of the Georgia Tech Research Reactor to low enrichment fuel

    SciTech Connect

    Karam, R.A.; Matos, J.E.; Mo, S.C.; Woodruff, W.L.

    1991-12-31

    The 5 MW Georgia Tech Research Reactor (GTRR) is a heterogeneous, heavy water moderated and cooled reactor, fueled with highly-enriched uranium aluminum alloy fuel plates. The GTRR is required to convert to low enrichment (LEU) fuel in accordance with USNRC policy. The US Department of Energy is funding a program to compare reactor performance with high and low enrichment fuels. The goals of the program are: (1) to amend the SAR and the Technical Specifications of the GTRR so that LEU U{sub 3}Si{sub 2}-Al dispersion fuel plates can replace the current HEU U-Al alloy fuel, and (2) to optimize the LEU core such that maximum value neutron beams can be extracted for possible neutron capture therapy application. This paper presents a status report on the LEU conversion effort.

  5. The Oak Ridge Research Reactor: safety analysis: Volume 2, supplement 2

    SciTech Connect

    Hurt, S.S.

    1986-11-01

    The Oak Ridge Research Reactor Safety Analysis was last updated via ORNL-4169, Vol. 2, Supplement 1, in May of 1978. Since that date, several changes have been effected through the change-memo system described below. While these changes have involved the cooling system, the electrical system, and the reactor instrumentation and controls, they have not, for the most part, presented new or unreviewed safety questions. However, some of the changes have been based on questions or recommendations stemming from safety reviews or from reactor events at other sites. This paper discusses those changes which were judged to be safety related and which include revisions to the syphon-break system and changes related to seismic considerations which were very recently completed. The maximum hypothetical accident postulated in the original safety analysis requires dynamic containment and filtered flow for compliance with 10CFR100 limits at the site boundary.

  6. The Annular Core Research Reactor (ACRR) postulated limiting event initial and building source terms

    SciTech Connect

    Restrepo, L F

    1992-08-01

    As part of the update of the Safety analysis Report (SAR) for the Annular Core Research Reactor (ACRR), operational limiting events under the category of inadvertent withdrawal of an experiment while at power or during a power pulse were determined to be the most limiting event(s) for this reactor. This report provides a summary of the assumptions, modeling, and results in evaluation of: Reactivity and thermal hydraulics analysis to determine the amount of fuel melt or fuel damage ratios; The reactor inventories following the limiting event; A literature review of post NUREG-0772 release fraction experiment results on severe fuel damages; Decontamination factors due to in-pool transport; and In-building transport modeling and building source term analysis.

  7. ReactorHealth Physics operations at the NIST center for neutron research.

    PubMed

    Johnston, Thomas P

    2015-02-01

    Performing health physics and radiation safety functions under a special nuclear material license and a research and test reactor license at a major government research and development laboratory encompasses many elements not encountered by industrial, general, or broad scope licenses. This article reviews elements of the health physics and radiation safety program at the NIST Center for Neutron Research, including the early history and discovery of the neutron, applications of neutron research, reactor overview, safety and security of radiation sources and radioactive material, and general health physics procedures. These comprise precautions and control of tritium, training program, neutron beam sample processing, laboratory audits, inventory and leak tests, meter calibration, repair and evaluation, radioactive waste management, and emergency response. In addition, the radiation monitoring systems will be reviewed including confinement building monitoring, ventilation filter radiation monitors, secondary coolant monitors, gaseous fission product monitors, gas monitors, ventilation tritium monitor, and the plant effluent monitor systems.

  8. The structure of decommissioning plan for VVR-S research reactor

    SciTech Connect

    Popescu, C.; Paunescu, A.; Garlea, I.; Garlea, C.

    1996-12-31

    This paper presents the activity in preparing the decommissioning plan for the VVR-S research reactor taking into account that there is no experience in this field in Hungary. VVR-S IPNE reactor, situated in Magurele Village, near Bucharest, was put into operation on the 27th of July 1957 and it has continuously been operating without major events, up to now. During this period, no modifications concerning the core and main circuits were made. The reactor is still operating with the original equipment and control instrumentation supplied by Ex-Soviets Union. The reactor was designed and built at the scientific and technological level of the 50`s. VVR-S IPNE reactor is a thermal one, of 2 MW nominal power and tank-type. It is light water cooled and moderated. The core operates with EK-10 type fuel assemblies (made of UO{sub 2} - MgO, with 10% U{sup 235} enrichment); after 1984 a mixed core with EK-10 and S-36 type assemblies was employed. The present fuel stock is S-36 type (36.6% U{sup 235} enrichment).

  9. Civilian nuclear power on the drawing board: the development of Experimental Breeder Reactor-II.

    SciTech Connect

    Westfall, C.

    2003-02-20

    On September 28, 2001 a symposium was held at Argonne National Laboratory as part of the festivities to mark the 100th birthday of Enrico Fermi. The symposium celebrated Fermi's ''contribution to the development of nuclear power'' and focused on one particular ''line of development'' resulting from Fermi's interest in power reactors: Argonne's fast reactor program. Symposium participants made many references to the ways in which the program was linked to Fermi, who led the team which created the world's first self-sustaining nuclear chain reaction. For example, one presentation featured an April, 1944 memo that described a meeting attended by Fermi and others. The memo came from the time when research on plutonium and the nuclear chain reaction at Chicago's WWII Metallurgical Laboratory was nearing its end. Even as other parts of the Manhattan Engineering Project were building on this effort to create the bombs that would end the war, Fermi and his colleagues were taking the first steps to plan the use of nuclear energy in the postwar era. After noting that Fermi ''viewed the use of [nuclear] power for the heating of cities with sympathy,'' the group outlined several power reactor designs. In the course of discussion, Fermi and his colleagues took the first steps in conjuring the vision that would later be brought to life with Experimental Breeder Reactor I (EBR-I) and Experimental Breeder Reactor II (EBR-II), the celebrated achievements of the Argonne fast reactor program. Group members considered various schemes for a breeder reactor in which the relatively abundant U-238 would be placed near a core of fissionable material. The reactor would be a fast reactor; that is, neutrons would not be moderated, as were most wartime reactors. Thus, the large number of neutrons emitted in fast neutron fission would hit the U-238 and create ''extra'' fissionable material, that is, more than ''invested,'' and at the same time produce power. The group identified the problem of

  10. High-temperature superconductor applications development at Argonne National Laboratory

    NASA Astrophysics Data System (ADS)

    Hull, J. R.; Poeppel, R. B.

    1992-02-01

    Developments at Argonne National Laboratory of near and intermediate term applications using high-temperature superconductors are discussed. Near-term applications of liquid-nitrogen depth sensors, current leads, and magnetic bearings are discussed in detail.

  11. High-temperature superconductor applications development at Argonne National Laboratory

    SciTech Connect

    Hull, J.R.; Poeppel, R.B.

    1992-02-09

    Developments at Argonne National Laboratory of near and intermediate term applications using high-temperature superconductors are discussed. Near-term applications of liquid-nitrogen depth sensors, current leads, and magnetic bearings are discussed in detail.

  12. Argonne National Laboratory Wins EPA Federal Green Challenge Award

    EPA Pesticide Factsheets

    CHICAGO -- The U.S. Environmental Protection Agency today announced that the Argonne National Laboratory has won Federal Green Challenge awards for waste reduction, transportation and electronics. The Challenge encourages federal agencies throughout the na

  13. Argonne Tau-charm factory collider design study

    SciTech Connect

    Teng, L.C.; Crosbie, E.A.; Norem, J.

    1995-12-01

    The design approach and design principles for a Tau-charm Factory at Argonne were studied. These studies led to a set of preliminary parameters and tentative component features as presented in this paper.

  14. Argonne National Laboratory site environmental report for calendar year 2007.

    SciTech Connect

    Golchert, N. W.; Davis, T. M.; Moos, L. P.; ESH /QA Oversight

    2008-09-09

    This report discusses the status and the accomplishments of the environmental protection program at Argonne National Laboratory for calendar year 2007. The status of Argonne environmental protection activities with respect to compliance with the various laws and regulations is discussed, along with the progress of environmental corrective actions and restoration projects. To evaluate the effects of Argonne operations on the environment, samples of environmental media collected on the site, at the site boundary, and off the Argonne site were analyzed and compared with applicable guidelines and standards. A variety of radionuclides were measured in air, surface water, on-site groundwater, and bottom sediment samples. In addition, chemical constituents in surface water, groundwater, and Argonne effluent water were analyzed. External penetrating radiation doses were measured, and the potential for radiation exposure to off-site population groups was estimated. Results are interpreted in terms of the origin of the radioactive and chemical substances (i.e., natural, fallout, Argonne, and other) and are compared with applicable environmental quality standards. A U.S. Department of Energy dose calculation methodology, based on International Commission on Radiological Protection recommendations and the U.S. Environmental Protection Agency's CAP-88 Version 3 (Clean Air Act Assessment Package-1988) computer code, was used in preparing this report.

  15. Argonne National Laboratory site environmental report for calendar year 2006.

    SciTech Connect

    Golchert, N. W.; ESH /QA Oversight

    2007-09-13

    This report discusses the status and the accomplishments of the environmental protection program at Argonne National Laboratory for calendar year 2006. The status of Argonne environmental protection activities with respect to compliance with the various laws and regulations is discussed, along with the progress of environmental corrective actions and restoration projects. To evaluate the effects of Argonne operations on the environment, samples of environmental media collected on the site, at the site boundary, and off the Argonne site were analyzed and compared with applicable guidelines and standards. A variety of radionuclides were measured in air, surface water, on-site groundwater, and bottom sediment samples. In addition, chemical constituents in surface water, groundwater, and Argonne effluent water were analyzed. External penetrating radiation doses were measured, and the potential for radiation exposure to off-site population groups was estimated. Results are interpreted in terms of the origin of the radioactive and chemical substances (i.e., natural, fallout, Argonne, and other) and are compared with applicable environmental quality standards. A U.S. Department of Energy dose calculation methodology, based on International Commission on Radiological Protection recommendations and the U.S. Environmental Protection Agency's CAP-88 Version 3 (Clean Air Act Assessment Package-1988) computer code, was used in preparing this report.

  16. Argonne National Laboratory site enviromental report for calendar year 2008.

    SciTech Connect

    Golchert, N. W.; Davis, T. M.; Moos, L. P.

    2009-09-02

    This report discusses the status and the accomplishments of the environmental protection program at Argonne National Laboratory for calendar year 2008. The status of Argonne environmental protection activities with respect to compliance with the various laws and regulations is discussed, along with the progress of environmental corrective actions and restoration projects. To evaluate the effects of Argonne operations on the environment, samples of environmental media collected on the site, at the site boundary, and off the Argonne site were analyzed and compared with applicable guidelines and standards. A variety of radionuclides were measured in air, surface water, on-site groundwater, and bottom sediment samples. In addition, chemical constituents in surface water, groundwater, and Argonne effluent water were analyzed. External penetrating radiation doses were measured, and the potential for radiation exposure to off-site population groups was estimated. Results are interpreted in terms of the origin of the radioactive and chemical substances (i.e., natural, fallout, Argonne, and other) and are compared with applicable environmental quality standards. A U.S. Department of Energy dose calculation methodology, based on International Commission on Radiological Protection recommendations and the U.S. Environmental Protection Agency's CAP-88 Version 3 (Clean Air Act Assessment Package-1988) computer code, was used in preparing this report.

  17. Argonne National Laboratory Site Environmental report for calendar year 2009.

    SciTech Connect

    Golchert, N. W.; Davis, T. M.; Moos, L. P.

    2010-08-04

    This report discusses the status and the accomplishments of the environmental protection program at Argonne National Laboratory for calendar year 2009. The status of Argonne environmental protection activities with respect to compliance with the various laws and regulations is discussed, along with the progress of environmental corrective actions and restoration projects. To evaluate the effects of Argonne operations on the environment, samples of environmental media collected on the site, at the site boundary, and off the Argonne site were analyzed and compared with applicable guidelines and standards. A variety of radionuclides were measured in air, surface water, on-site groundwater, and bottom sediment samples. In addition, chemical constituents in surface water, groundwater, and Argonne effluent water were analyzed. External penetrating radiation doses were measured, and the potential for radiation exposure to off-site population groups was estimated. Results are interpreted in terms of the origin of the radioactive and chemical substances (i.e., natural, Argonne, and other) and are compared with applicable environmental quality standards. A U.S. Department of Energy (DOE) dose calculation methodology, based on International Commission on Radiological Protection recommendations and the U.S. Environmental Protection Agency's (EPA) CAP-88 Version 3 (Clean Air Act Assessment Package-1988) computer code, was used in preparing this report.

  18. Argonne National Laboratory Site Environmental Report for Calendar Year 2013

    SciTech Connect

    Davis, T. M.; Gomez, J. L.; Moos, L. P.

    2014-09-02

    This report discusses the status and the accomplishments of the environmental protection program at Argonne National Laboratory for calendar year 2013. The status of Argonne environmental protection activities with respect to compliance with the various laws and regulations is discussed, along with environmental management, sustainability efforts, environmental corrective actions, and habitat restoration. To evaluate the effects of Argonne operations on the environment, samples of environmental media collected on the site, at the site boundary, and off the Argonne site were analyzed and compared with applicable guidelines and standards. A variety of radionuclides were measured in air, surface water, on-site groundwater, and bottom sediment samples. In addition, chemical constituents in surface water, groundwater, and Argonne effluent water were analyzed. External penetrating radiation doses were measured, and the potential for radiation exposure to off-site population groups was estimated. Results are interpreted in terms of the origin of the radioactive and chemical substances (i.e., natural, Argonne, and other) and are compared with applicable standards intended to protect human health and the environment. A U.S. Department of Energy (DOE) dose calculation methodology, based on International Commission on Radiological Protection (ICRP) recommendations and the U.S. Environmental Protection Agency’s (EPA) CAP-88 Version 3 computer code, was used in preparing this report.

  19. Further Development of Crack Growth Detection Techniques for US Test and Research Reactors

    SciTech Connect

    Kohse, Gordon; Carpenter, David M.; Ostrovsky, Yakov; Joseph Palmer, A.; Teysseyre, Sebastien P.; Davis, Kurt L.; Rempe, Joy L.

    2015-07-01

    One of the key issues facing Light Water Reactors (LWRs) in extending lifetimes beyond 60 years is characterizing the combined effect of irradiation and water chemistry on material degradation and failure. Irradiation Assisted Stress Corrosion Cracking (IASCC), in which a crack propagates in a susceptible material under stress in an aggressive environment, is a mechanism of particular concern. Full understanding of IASCC depends on real time crack growth data acquired under relevant irradiation conditions. Techniques to measure crack growth in actively loaded samples under irradiation have been developed outside the US - at the Halden Boiling Water Reactor, for example. Several types of IASCC tests have also been deployed at the MITR, including passively loaded crack growth measurements and actively loaded slow strain rate tests. However, there is not currently a facility available in the US to measure crack growth on actively loaded, pre-cracked specimens in LWR irradiation environments. A joint program between the Idaho National Laboratory (INL) and the Massachusetts Institute of Technology (MIT) Nuclear Reactor Laboratory (NRL) is currently underway to develop and demonstrate such a capability for US test and research reactors. Based on the Halden design, the samples will be loaded using miniature high pressure bellows and a compact loading mechanism, with crack length measured in real time using the switched Direct Current Potential Drop (DCPD) method. The basic design and initial mechanical testing of the load system and implementation of the DCPD method have been previously reported. This paper presents the results of initial autoclave testing at INL and the adaptation of the design for use in the high pressure, high temperature water loop at the MITR 6 MW research reactor, where an initial demonstration is planned in mid-2015. Materials considerations for the high pressure bellows are addressed. Design modifications to the loading mechanism required by the

  20. Nuclear Reactors. Revised.

    ERIC Educational Resources Information Center

    Hogerton, John F.

    This publication is one of a series of information booklets for the general public published by the United States Atomic Energy Commission. Among the topics discussed are: How Reactors Work; Reactor Design; Research, Teaching, and Materials Testing; Reactors (Research, Teaching and Materials); Production Reactors; Reactors for Electric Power…

  1. Fresh and Spent Nuclear Fuel Repatriation from the IRT-2000 Research Reactor Facility, Sofia, Bulgaria

    SciTech Connect

    K. J. Allen; T. G. Apostolov; I. S. Dimitrov

    2009-03-01

    The IRT 2000 research reactor, operated by the Bulgarian Institute for Nuclear Research and Nuclear Energy (INRNE), safely shipped all of their Russian-origin nuclear fuel from the Republic of Bulgaria to the Russian Federation beginning in 2003 and completing in 2008. These fresh and spent fuel shipments removed all highly enriched uranium (HEU) from Bulgaria. The fresh fuel was shipped by air in December 2003 using trucks and a commercial cargo aircraft. One combined spent fuel shipment of HEU and low enriched uranium (LEU) was completed in July 2008 using high capacity VPVR/M casks transported by truck, barge, and rail. The HEU shipments were assisted by the Russian Research Reactor Fuel Return Program (RRRFR) and the LEU spent fuel shipment was funded by Bulgaria. This report describes the work, approvals, organizations, equipment, and agreements required to complete these shipments and concludes with several major lessons learned.

  2. A risk characterization of safety research areas for Integral Fast Reactor program planning

    SciTech Connect

    Mueller, C.J.; Cahalan, J.E.; Hill, D.J.; Kramer, J.M.; Marchaterre, J.F.; Pedersen, D.R.; Sevy, R.H.; Tilbrook, R.W.; Wei, T.Y.; Wright, A.E.

    1988-01-01

    This paper characterizes the areas of Integral Fast Reactor (IFR) safety research in terms of their importance in addressing the risk of core disruption sequences for innovative designs. Such sequences have traditionally been determined to constitute the primary risk to public health and safety. All core disruption sequences are folded into four fault categories: classic unprotected (unscrammed) events; loss of decay heat; local fault propagation; and failure of critical reactor structures. Event trees are used to describe these sequences and the areas in the IFR Safety and related Base Technology research programs are discussed with respect to their relevance in addressing the key issues in preventing or delimiting core disruptive sequences. Thus a measure of potential for risk reduction is obtained for guidance in establishing research priorites.

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

    SciTech Connect

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

    2013-07-01

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

  4. Recent upgrades and new scientific infrastructure of MARIA research reactor, Otwock-Swierk, Poland

    SciTech Connect

    2015-07-01

    The MARIA reactor is open-pool type, water and beryllium moderated. It has two independent primary cooling systems: fuel and pool cooling system. Each fuel assembly is cooled down separately in pressurized channels with individual performances characterization. The fuel assemblies consist of five layers of bent plates or six concentric tubes. Currently it is one of the most powerful research reactors in Europe with operation availability at least up to 2030. Its nominal thermal power is 30 MW. It is characterized by high neutron flux density: up to 3x10{sup 14} n cm{sup -2} s{sup -1} in case of thermal neutrons, and up to 2x10{sup 13} n cm{sup -2} s{sup -1} in case of fast neutrons. The reactor is operated for ca. 4000 h per year. The reactor facility is equipped with fully equipped three hot cells with shielding up to 10{sup 15} Bq. Adjacent to the reactor facility, the radio-pharmaceutics plant (POLATOM) and Material Research Laboratory are located. They are equipped with a number of hot cells with instrumentation. The transport system of radioactive materials from reactor facility to Material Research Laboratory is available. During 2014 the MARIA reactor has been operated with three different types of fuel the same time: previous 36% enriched fuel, and two types of new LEU fuels. In the meantime, molybdenum irradiation programme has been developed. Maria is a multifunctional research tool, with a notable application in production of radioisotopes, radio-pharmaceutics manufacturing (ca. 600 TBq/y), {sup 99}Mo for medical scintigraphy (ca. 6000 TBq/y), neutron transmutation doping of silicon single crystals, wide scientific research based on neutron beams utilization. From the beginning MARIA reactor was intended for loop and fuel testing research activities. Currently it is used mostly as material testing and irradiation facility and for that reason it has wide experimental capabilities. There are eight horizontal irradiation channels from among whom six of them

  5. Argonne National Laboratory Physics Division annual report, January--December 1996

    SciTech Connect

    Thayer, K.J.

    1997-08-01

    The past year has seen several of the Physics Division`s new research projects reach major milestones with first successful experiments and results: the atomic physics station in the Basic Energy Sciences Research Center at the Argonne Advanced Photon Source was used in first high-energy, high-brilliance x-ray studies in atomic and molecular physics; the Short Orbit Spectrometer in Hall C at the Thomas Jefferson National Accelerator (TJNAF) Facility that the Argonne medium energy nuclear physics group was responsible for, was used extensively in the first round of experiments at TJNAF; at ATLAS, several new beams of radioactive isotopes were developed and used in studies of nuclear physics and nuclear astrophysics; the new ECR ion source at ATLAS was completed and first commissioning tests indicate excellent performance characteristics; Quantum Monte Carlo calculations of mass-8 nuclei were performed for the first time with realistic nucleon-nucleon interactions using state-of-the-art computers, including Argonne`s massively parallel IBM SP. At the same time other future projects are well under way: preparations for the move of Gammasphere to ATLAS in September 1997 have progressed as planned. These new efforts are imbedded in, or flowing from, the vibrant ongoing research program described in some detail in this report: nuclear structure and reactions with heavy ions; measurements of reactions of astrophysical interest; studies of nucleon and sub-nucleon structures using leptonic probes at intermediate and high energies; atomic and molecular structure with high-energy x-rays. The experimental efforts are being complemented with efforts in theory, from QCD to nucleon-meson systems to structure and reactions of nuclei. Finally, the operation of ATLAS as a national users facility has achieved a new milestone, with 5,800 hours beam on target for experiments during the past fiscal year.

  6. Corrosion Surveillance for Research Reactor Spent Nuclear Fuel in Wet Basin Storage

    SciTech Connect

    Howell, J.P.

    1998-10-16

    Foreign and domestic test and research reactor fuel is currently being shipped from locations over the world for storage in water filled basins at the Savannah River Site (SRS). The fuel was provided to many of the foreign countries as a part of the "Atoms for Peace" program in the early 1950's. In support of the wet storage of this fuel at the research reactor sites and at SRS, corrosion surveillance programs have been initiated. The International Atomic Energy Agency (IAEA) established a Coordinated Research Program (CRP) in 1996 on "Corrosion of Research Reactor Aluminum-Clad Spent Fuel in Water" and scientists from ten countries worldwide were invited to participate. This paper presents a detailed discussion of the IAEA sponsored CRP and provides the updated results from corrosion surveillance activities at SRS. In May 1998, a number of news articles around the world reported stories that microbiologically influenced corrosion (MIC) was active on the aluminum-clad spent fuel stored in the RBOF basin at SRS. This assessment was found to be in error with details presented in this paper. A biofilm was found on aluminum coupons, but resulted in no corrosion. Cracks seen on the surface were not caused by corrosion, but by stresses from the volume expansion of the oxide formed during pre-conditioning autoclaving. There has been no pitting caused by MIC or any other corrosion mechanism seen in the RBOF basin since initiation of the SRS Corrosion Surveillance Program in 1993.

  7. Theory Institute in Automated Reasoning held at Argonne National Laboratory, August 6--10, 1990

    SciTech Connect

    Wos, L.

    1990-08-01

    On August 6--10, 1990, Argonne National Laboratory hosted a Theory Institute in Automated Reasoning. The institute was organized by the Mathematics and Computer Science Division and was supported by special funding from Argonne's Physical Research Program Administration. The focus of the Institute was on the obstacles confronting the effective automation of reasoning. The objective was to lay the groundwork for formulating a theory governing the interrelationship of representation, inference rule, and strategy. Here we summarize the activities that took place during the week-long Institute. We also present an evaluation of the progress achieved-progress that includes the solution of challenge questions, the increasing use of both our database of problems and our automated reasoning program OTTER, and the discovery of new used for OTTER. 7 refs.

  8. Argonne's Michael Wang talks about the GREET Model for reducing vehicle emi

    SciTech Connect

    Michael Wang

    2012-07-25

    To fully evaluate energy and emission impacts of advanced vehicle technologies and new transportation fuels, the fuel cycle from wells to wheels and the vehicle cycle through material recovery and vehicle disposal need to be considered. Sponsored by the U.S. Department of Energy's Office of Energy Efficiency and Renewable Energy (EERE), Argonne has developed a full life-cycle model called GREET (Greenhouse gases, Regulated Emissions, and Energy use in Transportation). It allows researchers and analysts to evaluate various vehicle and fuel combinations on a full fuel-cycle/vehicle-cycle basis. The first version of GREET was released in 1996. Since then, Argonne has continued to update and expand the model. The most recent GREET versions are the GREET 1 2012 version for fuel-cycle analysis and GREET 2.7 version for vehicle-cycle analysis.

  9. Argonne's Michael Wang talks about the GREET Model for reducing vehicle emi

    ScienceCinema

    Michael Wang

    2016-07-12

    To fully evaluate energy and emission impacts of advanced vehicle technologies and new transportation fuels, the fuel cycle from wells to wheels and the vehicle cycle through material recovery and vehicle disposal need to be considered. Sponsored by the U.S. Department of Energy's Office of Energy Efficiency and Renewable Energy (EERE), Argonne has developed a full life-cycle model called GREET (Greenhouse gases, Regulated Emissions, and Energy use in Transportation). It allows researchers and analysts to evaluate various vehicle and fuel combinations on a full fuel-cycle/vehicle-cycle basis. The first version of GREET was released in 1996. Since then, Argonne has continued to update and expand the model. The most recent GREET versions are the GREET 1 2012 version for fuel-cycle analysis and GREET 2.7 version for vehicle-cycle analysis.

  10. Strategic Plan for Light Water Reactor Research and Development

    SciTech Connect

    2004-02-01

    The purpose of this strategic plan is to establish a framework that will allow the Department of Energy (DOE) and the nuclear power industry to jointly plan the nuclear energy research and development (R&D) agenda important to achieving the Nation's energy goals. This strategic plan has been developed to focus on only those R&D areas that will benefit from a coordinated government/industry effort. Specifically, this plan focuses on safely sustaining and expanding the electricity output from currently operating nuclear power plants and expanding nuclear capacity through the deployment of new plants. By focusing on R&D that addresses the needs of both current and future nuclear plants, DOE and industry will be able to take advantage of the synergism between these two technology areas, thus improving coordination, enhancing efficiency, and further leveraging public and private sector resources. By working together under the framework of this strategic plan, DOE and the nuclear industry reinforce their joint commitment to the future use of nuclear power and the National Energy Policy's goal of expanding its use in the United States. The undersigned believe that a public-private partnership approach is the most efficient and effective way to develop and transfer new technologies to the marketplace to achieve this goal. This Strategic Plan is intended to be a living document that will be updated annually.

  11. Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core.

    PubMed

    Lashkari, A; Khalafi, H; Kazeminejad, H

    2013-05-01

    In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR_PC package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change.

  12. Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core

    PubMed Central

    Lashkari, A.; Khalafi, H.; Kazeminejad, H.

    2013-01-01

    In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR_PC package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change. PMID:24976672

  13. NCTPlan application for neutron capture therapy dosimetric planning at MEPhI nuclear research reactor.

    PubMed

    Elyutina, A S; Kiger, W S; Portnov, A A

    2011-12-01

    The results of modeling of two therapeutic beams HEC-1 and HEC-4 at the NRNU "MEPhI" research nuclear reactor exploitable for preclinical treatments are reported. The exact models of the beams are constructed as an input to the NCTPlan code used for planning Neutron Capture Therapy (NCT) procedure. The computations are purposed to improve the accuracy of prediction of a dose absorbed in tissue with the account of all components of radiation.

  14. Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Main report

    SciTech Connect

    Konzek, G.J.; Ludwick, J.D.; Kennedy, W.E. Jr.; Smith, R.I.

    1982-03-01

    Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). The study results are presented in two volumes. Volume 1 (Main Report) contains the results in summary form.

  15. The Possible Use of Cermet Fuel in the Dido and Pluto Heavy-Water Research Reactors,

    DTIC Science & Technology

    1981-08-01

    tions, I.A.E.A., Vienna (September 1980). 5. TOULOUKIAN , Y.S., ’Thermophysical properties of high-temperature solid materials’, Purdue University...the implications of using lower-enrichment fuels in heavy-water-moderated research reactors. The uranium content of the U/Al alloy at present used...cannot be increased sufficiently to maintain the requisite UL3 5 content without undesirable effects on the physical properties of the alloy . A different

  16. Fire protection review revisit no. 2, Argonne National Laboratory, Argonne, Illinois

    NASA Astrophysics Data System (ADS)

    Dobson, P. H.; Earley, M. W.; Mattern, L. J.

    1985-05-01

    A fire protection survey was conducted at Argonne National Laboratory on April 1-5, 8-12, and April 29-May 2, 1985. The purpose was to review the facility fire protection program and to make recommendations or identify areas according to criteria established by the Department of Energy. There has been a substantial improvement in fire protection at this laboratory since the 1977 audit. Numerous areas which were previously provided with detection systems only have since been provided with automatic sprinkler protection. The following basic fire protection features are not properly controlled: (1) resealing wall and floor penetrations between fire areas after installation of services; (2) cutting and welding; and (3) housekeeping. The present Fire Department manpower level appears adequate to control a route fire. Their ability to adequately handle a high-challenge fire, or one involving injuries to personnel, or fire spread beyond the initial fire area is doubtful.

  17. Nuclear mass inventory, photon dose rate and thermal decay heat of spent research reactor fuel assemblies

    SciTech Connect

    Pond, R.B.; Matos, J.E.

    1996-12-31

    This document has been prepared to assist research reactor operators possessing spent fuel containing enriched uranium of United States origin to prepare part of the documentation necessary to ship this fuel to the United States. Data are included on the nuclear mass inventory, photon dose rate, and thermal decay heat of spent research reactor fuel assemblies. Isotopic masses of U, Np, Pu and Am that are present in spent research reactor fuel are estimated for MTR, TRIGA and DIDO-type fuel assembly types. The isotopic masses of each fuel assembly type are given as functions of U-235 burnup in the spent fuel, and of initial U-235 enrichment and U-235 mass in the fuel assembly. Photon dose rates of spent MTR, TRIGA and DIDO-type fuel assemblies are estimated for fuel assemblies with up to 80% U-235 burnup and specific power densities between 0.089 and 2.857 MW/kg[sup 235]U, and for fission product decay times of up to 20 years. Thermal decay heat loads are estimated for spent fuel based upon the fuel assembly irradiation history (average assembly power vs. elapsed time) and the spent fuel cooling time.

  18. Advanced Test Reactor National Scientific User Facility: Addressing advanced nuclear materials research

    SciTech Connect

    John Jackson; Todd Allen; Frances Marshall; Jim Cole

    2013-03-01

    The Advanced Test Reactor National Scientific User Facility (ATR NSUF), based at the Idaho National Laboratory in the United States, is supporting Department of Energy and industry research efforts to ensure the properties of materials in light water reactors are well understood. The ATR NSUF is providing this support through three main efforts: establishing unique infrastructure necessary to conduct research on highly radioactive materials, conducting research in conjunction with industry partners on life extension relevant topics, and providing training courses to encourage more U.S. researchers to understand and address LWR materials issues. In 2010 and 2011, several advanced instruments with capability focused on resolving nuclear material performance issues through analysis on the micro (10-6 m) to atomic (10-10 m) scales were installed primarily at the Center for Advanced Energy Studies (CAES) in Idaho Falls, Idaho. These instruments included a local electrode atom probe (LEAP), a field-emission gun scanning transmission electron microscope (FEG-STEM), a focused ion beam (FIB) system, a Raman spectrometer, and an nanoindentor/atomic force microscope. Ongoing capability enhancements intended to support industry efforts include completion of two shielded, irradiation assisted stress corrosion cracking (IASCC) test loops, the first of which will come online in early calendar year 2013, a pressurized and controlled chemistry water loop for the ATR center flux trap, and a dedicated facility intended to house post irradiation examination equipment. In addition to capability enhancements at the main site in Idaho, the ATR NSUF also welcomed two new partner facilities in 2011 and two new partner facilities in 2012; the Oak Ridge National Laboratory, High Flux Isotope Reactor (HFIR) and associated hot cells and the University California Berkeley capabilities in irradiated materials analysis were added in 2011. In 2012, Purdue University’s Interaction of Materials

  19. NUMERICAL SIMULATION FOR MECHANICAL BEHAVIOR OF U10MO MONOLITHIC MINIPLATES FOR RESEARCH AND TEST REACTORS

    SciTech Connect

    Hakan Ozaltun & Herman Shen

    2011-11-01

    This article presents assessment of the mechanical behavior of U-10wt% Mo (U10Mo) alloy based monolithic fuel plates subject to irradiation. Monolithic, plate-type fuel is a new fuel form being developed for research and test reactors to achieve higher uranium densities within the reactor core to allow the use of low-enriched uranium fuel in high-performance reactors. Identification of the stress/strain characteristics is important for understanding the in-reactor performance of these plate-type fuels. For this work, three distinct cases were considered: (1) fabrication induced residual stresses (2) thermal cycling of fabricated plates; and finally (3) transient mechanical behavior under actual operating conditions. Because the temperatures approach the melting temperature of the cladding during the fabrication and thermal cycling, high temperature material properties were incorporated to improve the accuracy. Once residual stress fields due to fabrication process were identified, solution was used as initial state for the subsequent simulations. For thermal cycling simulation, elasto-plastic material model with thermal creep was constructed and residual stresses caused by the fabrication process were included. For in-service simulation, coupled fluid-thermal-structural interaction was considered. First, temperature field on the plates was calculated and this field was used to compute the thermal stresses. For time dependent mechanical behavior, thermal creep of cladding, volumetric swelling and fission induced creep of the fuel foil were considered. The analysis showed that the stresses evolve very rapidly in the reactor. While swelling of the foil increases the stress of the foil, irradiation induced creep causes stress relaxation.

  20. Environment, Safety and Health Progress Assessment of the Argonne Illinois Site

    SciTech Connect

    Not Available

    1993-11-01

    This report documents the results of the US Department of Energy (DOE) Environment, Safety and Health (ES&H) Progress Assessment of the Argonne Illinois Site (AIS), near Chicago, Illinois, conducted from October 25 through November 9, 1993. During the Progress Assessment, activities included a selective review of the ES&H management systems and programs with principal focus on the DOE Office of Energy Research (ER); CH, which includes the Argonne Area Office; the University of Chicago; and the contractor`s organization responsible for operation of Argonne National Laboratory (ANL). The ES&H Progress Assessments are part of DOE`s continuing effort to institutionalize line management accountability and the self-assessment process throughout DOE and its contractor organizations. The purpose of the AIS ES&H Progress Assessment was to provide the Secretary of Energy, senior DOE managers, and contractor management with concise independent information on the following: change in culture and attitude related to ES&H activities; progress and effectiveness of the ES&H corrective actions resulting from the previous Tiger Team Assessment; adequacy and effectiveness of the ES&H self-assessment process of the DOE line organizations, the site management, and the operating contractor; and effectiveness of DOE and contractor management structures, resources, and systems to effectively address ES&H problems and new ES&H initiatives.

  1. The Ongoing Impact of the U.S. Fast Reactor Integral Experiments Program

    SciTech Connect

    John D. Bess; Michael A. Pope; Harold F. McFarlane

    2012-11-01

    The creation of a large database of integral fast reactor physics experiments advanced nuclear science and technology in ways that were unachievable by less capital intensive and operationally challenging approaches. They enabled the compilation of integral physics benchmark data, validated (or not) analytical methods, and provided assurance of future rector designs The integral experiments performed at Argonne National Laboratory (ANL) represent decades of research performed to support fast reactor design and our understanding of neutronics behavior and reactor physics measurements. Experiments began in 1955 with the Zero Power Reactor No. 3 (ZPR-3) and terminated with the Zero Power Physics Reactor (ZPPR, originally the Zero Power Plutonium Reactor) in 1990 at the former ANL-West site in Idaho, which is now part of the Idaho National Laboratory (INL). Two additional critical assemblies, ZPR-6 and ZPR-9, operated at the ANL-East site in Illinois. A total of 128 fast reactor assemblies were constructed with these facilities [1]. The infrastructure and measurement capabilities are too expensive to be replicated in the modern era, making the integral database invaluable as the world pushes ahead with development of liquid metal cooled reactors.

  2. Analysis of neutron flux distribution for the validation of computational methods for the optimization of research reactor utilization.

    PubMed

    Snoj, L; Trkov, A; Jaćimović, R; Rogan, P; Zerovnik, G; Ravnik, M

    2011-01-01

    In order to verify and validate the computational methods for neutron flux calculation in TRIGA research reactor calculations, a series of experiments has been performed. The neutron activation method was used to verify the calculated neutron flux distribution in the TRIGA reactor. Aluminium (99.9 wt%)-Gold (0.1 wt%) foils (disks of 5mm diameter and 0.2mm thick) were irradiated in 33 locations; 6 in the core and 27 in the carrousel facility in the reflector. The experimental results were compared to the calculations performed with Monte Carlo code MCNP using detailed geometrical model of the reactor. The calculated and experimental normalized reaction rates in the core are in very good agreement for both isotopes indicating that the material and geometrical properties of the reactor core are modelled well. In conclusion one can state that our computational model describes very well the neutron flux and reaction rate distribution in the reactor core. In the reflector however, the accuracy of the epithermal and thermal neutron flux distribution and attenuation is lower, mainly due to lack of information about the material properties of the graphite reflector surrounding the core, but the differences between measurements and calculations are within 10%. Since our computational model properly describes the reactor core it can be used for calculations of reactor core parameters and for optimization of research reactor utilization.

  3. Georgia Institute of Technology research on the Gas Core Actinide Transmutation Reactor (GCATR)

    NASA Technical Reports Server (NTRS)

    Clement, J. D.; Rust, J. H.; Schneider, A.; Hohl, F.

    1976-01-01

    The program reviewed is a study of the feasibility, design, and optimization of the GCATR. The program is designed to take advantage of initial results and to continue work carried out on the Gas Core Breeder Reactor. The program complements NASA's program of developing UF6 fueled cavity reactors for power, nuclear pumped lasers, and other advanced technology applications. The program comprises: (1) General Studies--Parametric survey calculations performed to examine the effects of reactor spectrum and flux level on the actinide transmutation for GCATR conditions. The sensitivity of the results to neutron cross sections are to be assessed. Specifically, the parametric calculations of the actinide transmutation are to include the mass, isotope composition, fission and capture rates, reactivity effects, and neutron activity of recycled actinides. (2) GCATR Design Studies--This task is a major thrust of the proposed research program. Several subtasks are considered: optimization criteria studies of the blanket and fuel reprocessing, the actinide insertion and recirculation system, and the system integration. A brief review of the background of the GCATR and ongoing research is presented.

  4. OVERVIEW OF CRITERIA FOR INTERIM WET & DRY STORAGE OF RESEARCH REACTOR SPENT NUCLEAR FUEL

    SciTech Connect

    Sindelar, R.; Vinson, D.; Iyer, N.; Fisher, D.

    2010-11-03

    Following discharge from research reactors, spent nuclear fuel may be stored 'wet' in water pools or basins, or it may be stored 'dry' in various configurations including non-sealed or sealed containers until retrieved for ultimate disposition. Interim safe storage practices are based on avoiding degradation to the fuel that would impact functions related to safety. Recommended practices including environmental controls with technical bases, are outlined for wet storage and dry storage of aluminum-clad, aluminum-based research reactor fuel. For wet storage, water quality must be maintained to minimize corrosion degradation of aluminum fuel. For dry storage, vented canister storage of aluminum fuel readily provides a safe storage configuration. For sealed dry storage, drying must be performed so as to minimize water that would cause additional corrosion and hydrogen generation. Consideration must also be given to the potential for radiolytically-generated hydrogen from the bound water in the attendant oxyhydroxides on aluminum fuel from reactor operation for dry storage systems.

  5. IGORR-1: Proceedings of the first meeting of the international group on research reactors

    SciTech Connect

    West, C.D.

    1990-05-01

    Many organizations, in several countries, are planning or implementing new or upgraded research reactor projects, but there has been no organized forum devoted entirely to discussion and exchange of information in this field. Over the past year or so, informal discussions resulted in widespread agreement that such a forum would serve a useful purpose. Accordingly, a proposal to form a group was submitted to the leading organizations known to be involved in projects to build or upgrade reactor facilities. Essentially all agreed to join in the formation of the International Group on Research Reactors (IGORR) and nominated a senior staff member to serve on its international organizing committee. The first IGORR meeting took place on February 28--March 2, 1990. It was very successful and well attended; some 52 scientists and engineers from 25 organizations in 10 countries participated in 2-1/2 days of open and informative presentations and discussions. Two workshop sessions offered opportunities for more detailed interaction among participants and resulted in identification of common R D needs, sources of data, and planned new facilities. Individual papers have been cataloged separately.

  6. Validation of deterministic and Monte Carlo codes for neutronics calculation of the IRT-type research reactor

    NASA Astrophysics Data System (ADS)

    Shchurovskaya, M. V.; Alferov, V. P.; Geraskin, N. I.; Radaev, A. I.

    2017-01-01

    The results of the validation of a research reactor calculation using Monte Carlo and deterministic codes against experimental data and based on code-to-code comparison are presented. The continuous energy Monte Carlo code MCU-PTR and the nodal diffusion-based deterministic code TIGRIS were used for full 3-D calculation of the IRT MEPhI research reactor. The validation included the investigations for the reactor with existing high enriched uranium (HEU, 90 w/o) fuel and low enriched uranium (LEU, 19.7 w/o, U-9%Mo) fuel.

  7. Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)

    SciTech Connect

    J. K. Wright; R. N. Wright

    2008-04-01

    The U.S. Department of Energy has selected the High Temperature Gas-cooled Reactor design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic, or pebble-bed reactor and use low-enriched uranium, Tri-Isotopic-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development Program is responsible for performing research and development on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. Studies of potential Reactor Pressure Vessel (RPV) steels have been carried out as part of the pre-conceptual design studies. These design studies generally focus on American Society of Mechanical Engineers (ASME) Code status of the steels, temperature limits, and allowable stresses. Three realistic candidate materials have been identified by this process: conventional light water reactor RPV steels A508/533, 2¼Cr-1Mo in the annealed condition, and modified 9Cr 1Mo ferritic martenistic steel. Based on superior strength and higher temperature limits, the modified 9Cr-1Mo steel has been identified by the majority of design engineers as the preferred choice for the RPV. All of the vendors have

  8. Reactor physics teaching and research in the Swiss nuclear engineering master

    SciTech Connect

    Chawla, R.

    2012-07-01

    Since 2008, a Master of Science program in Nuclear Engineering (NE) has been running in Switzerland, thanks to the combined efforts of the country's key players in nuclear teaching and research, viz. the Swiss Federal Inst.s of Technology at Lausanne (EPFL) and at Zurich (ETHZ), the Paul Scherrer Inst. (PSI) at Villigen and the Swiss Nuclear Utilities (Swissnuclear). The present paper, while outlining the academic program as a whole, lays emphasis on the reactor physics teaching and research training accorded to the students in the framework of the developed curriculum. (authors)

  9. The Advanced Neutron Source (ANS) project: A world-class research reactor facility

    SciTech Connect

    Thompson, P.B.; Meek, W.E.

    1993-07-01

    This paper provides an overview of the Advanced Neutron Source (ANS), a new research facility being designed at Oak Ridge National Laboratory. The facility is based on a 330 MW, heavy-water cooled and reflected reactor as the neutron source, with a thermal neutron flux of about 7.5{times}10{sup 19}m{sup {minus}2}{center_dot}sec{sup {minus}1}. Within the reflector region will be one hot source which will serve 2 hot neutron beam tubes, two cryogenic cold sources serving fourteen cold neutron beam tubes, two very cold beam tubes, and seven thermal neutron beam tubes. In addition there will be ten positions for materials irradiation experiments, five of them instrumented. The paper touches on the project status, safety concerns, cost estimates and scheduling, a description of the site, the reactor, and the arrangements of the facilities.

  10. Characterisation of the epithermal neutron irradiation facility at the Portuguese research reactor using MCNP.

    PubMed

    Beasley, D G; Fernandes, A C; Santos, J P; Ramos, A R; Marques, J G; King, A

    2015-05-01

    The radiation field at the epithermal beamline and irradiation chamber installed at the Portuguese Research Reactor (RPI) at the Campus Tecnológico e Nuclear of Instituto Superior Técnico was characterised in the context of Prompt Gamma Neutron Activation Analysis (PGNAA) applications. Radiographic films, activation foils and thermoluminescence dosimeters were used to measure the neutron fluence and photon dose rates in the irradiation chamber. A fixed-source MCNPX model of the beamline and chamber was developed and compared to measurements in the first step towards planning a new irradiation chamber. The high photon background from the reactor results in the saturation of the detector and the current facility configuration yields an intrinsic insensitivity to various elements of interest for PGNAA. These will be addressed in future developments.

  11. Production and modeling of radioactive gold nanoparticles in Tehran research reactor.

    PubMed

    Hosseini, Seyedeh Fatemeh; Sadeghi, Mahdi; Aboudzadeh, Mohammad Reza; Mohseni, Morteza

    2016-12-01

    Gold has two medically useful radioactive isotopes, (198)Au and (199)Au, for locally irradiating and killing tumor cells. (198)Au radionuclide has been produced through the irradiation of the pure gold via (197)Au(n,γ)(198)Au reaction in the Tehran Research Reactor at a thermal neutron flux of 4.5×10(13)ncm(-2)s(-1) for the different irradiation times. In this paper, the activity of (198)Au radionuclide has been determined using MCNPX-2.6 and TALYS-1.6 codes and also the theoretical approach. The calculated results were compared with the corresponding experimental values. The calculated results were in good agreement with the experimental data, thus the used codes can be used as a powerful tool to predict and optimize production conditions in reactor.

  12. Design of the cold neutron triple-axis spectrometer at the China Advanced Research Reactor

    NASA Astrophysics Data System (ADS)

    Cheng, P.; Zhang, Hongxia; Bao, W.; Schneidewind, A.; Link, P.; Grünwald, A. T. D.; Georgii, R.; Hao, L. J.; Liu, Y. T.

    2016-06-01

    The design of the first cold neutron triple-axis spectrometer at the China Advanced Research Reactor is presented. Based on the Monte Carlo simulations using neutron ray-tracing program McStas, the parameters of major neutron optics in this instrument are optimized. The neutron flux at sample position is estimated to be 5.6 ×107 n/cm2/s at neutron incident energy Ei=5 meV when the reactor operates normally at the designed 60 MW power. The performances of several neutron supermirror polarizing devices are compared and their critical parameters are optimized for this spectrometer. The polarization analysis will be realized with a flexible switch from the unpolarized experimental mode.

  13. Shielding calculation and criticality safety analysis of spent fuel transportation cask in research reactors.

    PubMed

    Mohammadi, A; Hassanzadeh, M; Gharib, M

    2016-02-01

    In this study, shielding calculation and criticality safety analysis were carried out for general material testing reactor (MTR) research reactors interim storage and relevant transportation cask. During these processes, three major terms were considered: source term, shielding, and criticality calculations. The Monte Carlo transport code MCNP5 was used for shielding calculation and criticality safety analysis and ORIGEN2.1 code for source term calculation. According to the results obtained, a cylindrical cask with body, top, and bottom thicknesses of 18, 13, and 13 cm, respectively, was accepted as the dual-purpose cask. Furthermore, it is shown that the total dose rates are below the normal transport criteria that meet the standards specified.

  14. RELAP5 Application to Accident Analysis of the NIST Research Reactor

    SciTech Connect

    Baek, J.; Cuadra Gascon, A.; Cheng, L.Y.; Diamond, D.

    2012-03-18

    Detailed safety analyses have been performed for the 20 MW D{sub 2}O moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The time-dependent analysis of the primary system is determined with a RELAP5 transient analysis model that includes the reactor vessel, the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. A post-processing of the simulation results has been conducted to evaluate minimum critical heat flux ratio (CHFR) using the Sudo-Kaminaga correlation. Evaluations are performed for the following accidents: (1) the control rod withdrawal startup accident and (2) the maximum reactivity insertion accident. In both cases the RELAP5 results indicate that there is adequate margin to CHF and no damage to the fuel will occur because of sufficient coolant flow through the fuel channels and the negative scram reactivity insertion.

  15. University Reactor Conversion Lessons Learned Workshop for the University of Florida

    SciTech Connect

    Eric C. Woolstenhulme; Dana M. Meyer

    2007-04-01

    The Department of Energy’s (DOE) Idaho National Laboratory (INL), under its programmatic responsibility for managing the University Research Reactor Conversions, has completed the conversion of the reactor at the University of Florida. This project was successfully completed through an integrated and collaborative effort involving the INL, Argonne National Laboratory (ANL), DOE (Headquarters and Field Office), the Nuclear Regulatory Commission, the Universities, and contractors involved in analyses, fuel design and fabrication, and SNF shipping and disposition. With the work completed with these two universities, and in anticipation of other impending conversion projects, INL convened and engaged the project participants in a structured discussion to capture lessons learned. The objectives of this meeting were to capture the observations, insights, issues, concerns, and ideas of those involved in the reactor conversions so that future efforts can be conducted with greater effectiveness, efficiency, and with fewer challenges.

  16. Performance model of the Argonne Voyager multimedia server

    SciTech Connect

    Disz, T.; Olson, R.; Stevens, R.

    1997-07-01

    The Argonne Voyager Multimedia Server is being developed in the Futures Lab of the Mathematics and Computer Science Division at Argonne National Laboratory. As a network-based service for recording and playing multimedia streams, it is important that the Voyager system be capable of sustaining certain minimal levels of performance in order for it to be a viable system. In this article, the authors examine the performance characteristics of the server. As they examine the architecture of the system, they try to determine where bottlenecks lie, show actual vs potential performance, and recommend areas for improvement through custom architectures and system tuning.

  17. Inspection methods for physical protection Task III review of other agencies' physical security activities for research reactors

    SciTech Connect

    1980-01-01

    In Task I of this project, the current Nuclear Regulatory Commission (NRC) position-on physical security practices and procedures at research reactors were reviewed. In the second task, a sampling of the physical security plans was presented and the three actual reactor sites described in the security plans were visited. The purpose of Task III is to review other agencies' physical security activities for research reactors. During this phase, the actions, procedures and policies of two domestic and two foreign agencies other than the NRC that relate to the research reactor community were examined. The agencies examined were: International Atomic Energy Agency; Canadian Atomic Energy Control Board; Department of Energy; and American Nuclear Insurers.

  18. International Atomic Energy Agency support of research reactor highly enriched uranium to low enriched uranium fuel conversion projects

    SciTech Connect

    Bradley, E.; Adelfang, P.; Goldman, I.N.

    2008-07-15

    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 assisted efforts to convert research reactors from HEU to LEU fuel. HEU to LEU fuel conversion projects differ significantly depending on several factors including the design of the reactor and fuel, technical needs of the member state, local nuclear infrastructure, and available resources. To support such diverse endeavours, the IAEA tailors each project to address the relevant constraints. This paper presents the different approaches taken by the IAEA to address the diverse challenges involved in research reactor HEU to LEU fuel conversion projects. Examples of conversion related projects in different Member States are fully detailed. (author)

  19. Decommissioning of German Research Reactors Under the Governance of the Federal Ministry of Education and Research - 12154

    SciTech Connect

    Weigl, M.

    2012-07-01

    Since 1956, nuclear research and development (R and D) in Germany has been supported by the Federal Government. The goal was to help German industry to become competitive in all fields of nuclear technology. National research centers were established and demonstration plants were built. In the meantime, all these facilities were shut down and are now in a state of decommissioning and dismantling (D and D). Meanwhile, Germany is one of the leading countries in the world in the field of D and D. Two big demonstration plants, the Niederaichbach Nuclear Power Plant (KKN) a heavy-water cooled pressure tube reactor with carbon-dioxide cooling and the Karlstein Superheated Steam Reactor (HDR) a boiling light water reactor with a thermal power of 100 MW, are totally dismantled and 'green field' is reached. Another big project was finished in 2008. The Forschungs-Reaktor Juelich 1 (FRJ1), a research reactor with a thermal power of 10 MW was completely dismantled and in September 2008 an oak tree was planted on a green field at the site, where the FRJ1 was standing before. This is another example for German success in the field of D and D. Within these projects a lot of new solutions and innovative techniques were tested, which were developed at German universities and in small and medium sized companies mostly funded by the Federal Ministry of Education and Research (BMBF). Some examples are underwater-cutting technologies like plasma arc cutting and contact arc metal cutting. This clearly shows that research on the field of D and D is important for the future. Moreover, these research activities are important to save the know-how in nuclear engineering in Germany and will enable enterprises to compete on the increasing market of D and D services. The author assumes that an efficient decommissioning of nuclear installations will help stabilize the credibility of nuclear energy. Some critics of nuclear energy are insisting that a return to 'green field sites' is not possible

  20. The Argonne Leadership Computing Facility 2010 annual report.

    SciTech Connect

    Drugan, C.

    2011-05-09

    Researchers found more ways than ever to conduct transformative science at the Argonne Leadership Computing Facility (ALCF) in 2010. Both familiar initiatives and innovative new programs at the ALCF are now serving a growing, global user community with a wide range of computing needs. The Department of Energy's (DOE) INCITE Program remained vital in providing scientists with major allocations of leadership-class computing resources at the ALCF. For calendar year 2011, 35 projects were awarded 732 million supercomputer processor-hours for computationally intensive, large-scale research projects with the potential to significantly advance key areas in science and engineering. Argonne also continued to provide Director's Discretionary allocations - 'start up' awards - for potential future INCITE projects. And DOE's new ASCR Leadership Computing (ALCC) Program allocated resources to 10 ALCF projects, with an emphasis on high-risk, high-payoff simulations directly related to the Department's energy mission, national emergencies, or for broadening the research community capable of using leadership computing resources. While delivering more science today, we've also been laying a solid foundation for high performance computing in the future. After a successful DOE Lehman review, a contract was signed to deliver Mira, the next-generation Blue Gene/Q system, to the ALCF in 2012. The ALCF is working with the 16 projects that were selected for the Early Science Program (ESP) to enable them to be productive as soon as Mira is operational. Preproduction access to Mira will enable ESP projects to adapt their codes to its architecture and collaborate with ALCF staff in shaking down the new system. We expect the 10-petaflops system to stoke economic growth and improve U.S. competitiveness in key areas such as advancing clean energy and addressing global climate change. Ultimately, we envision Mira as a stepping-stone to exascale-class computers that will be faster than petascale

  1. Laboratory directed research and development. FY 1991 program activities: Summary report

    SciTech Connect

    Not Available

    1991-11-15

    The purposes of Argonne`s Laboratory Directed Research and Development (LDRD) Program are to encourage the development of novel concepts, enhance the Laboratory`s R&D capabilities, and further the development of its strategic initiatives. Among the aims of the projects supported by the Program are establishment of engineering ``proof-of-principle``; development of an instrumental prototype, method, or system; or discovery in fundamental science. Several of these project are closely associated with major strategic thrusts of the Laboratory as described in Argonne`s Five Year Institutional Plan, although the scientific implications of the achieved results extend well beyond Laboratory plans and objectives. The projects supported by the Program are distributed across the major programmatic areas at Argonne. Areas of emphasis are (1) advanced accelerator and detector technology, (2) x-ray techniques in biological and physical sciences, (3) advanced reactor technology, (4) materials science, computational science, biological sciences and environmental sciences. Individual reports summarizing the purpose, approach, and results of projects are presented.

  2. Estimation of (41)Ar activity concentration and release rate from the TRIGA Mark-II research reactor.

    PubMed

    Hoq, M Ajijul; Soner, M A Malek; Rahman, A; Salam, M A; Islam, S M A

    2016-03-01

    The BAEC TRIGA research reactor (BTRR) is the only nuclear reactor in Bangladesh. Bangladesh Atomic Energy Regulatory Authority (BAERA) regulations require that nuclear reactor licensees undertake all reasonable precautions to protect the environment and the health and safety of persons, including identifying, controlling and monitoring the release of nuclear substances to the environment. The primary activation product of interest in terms of airborne release from the reactor is (41)Ar. (41)Ar is a noble gas readily released from the reactor stacks and most has not decayed by the time it moves offsite with normal wind speed. Initially (41)Ar is produced from irradiation of dissolved air in the primary water which eventually transfers into the air in the reactor bay. In this study, the airborne radioisotope (41)Ar generation concentration, ground level concentration and release rate from the BTRR bay region are evaluated theoretically during the normal reactor operation condition by several governing equations. This theoretical calculation eventually minimizes the doubt about radiological safety to determine the radiation level for (41)Ar activity whether it is below the permissible limit or not. Results show that the estimated activity for (41)Ar is well below the maximum permissible concentration limit set by the regulatory body, which is an assurance for the reactor operating personnel and general public. Thus the analysis performed within this paper is so much effective in the sense of ensuring radiological safety for working personnel and the environment.

  3. Current status of the development of high density LEU fuel for Russian research reactors

    SciTech Connect

    Vatulin, A.; Dobrikova, I.; Suprun, V.; Trifonov, Y.; Kartashev, E.; Lukichev, V.

    2008-07-15

    One of the main directions of the Russian RERTR program is to develop U-Mo fuel and fuel elements/FA with this fuel. The development is carried out both for existing reactors, and for new advanced designs of reactors. Many organizations in Russia, i.e. 'TVEL', RDIPE, RIAR, IRM, NPCC participate in the work. Two fuels are under development: dispersion and monolithic U-Mo fuel, as well two types of FA to use the dispersion U-Mo fuel: with tubular type fuel elements and with pin type fuel elements. The first stage of works was successfully completed. This stage included out-pile, in-pile and post irradiation examinations of U-Mo dispersion fuel in experimental tubular and pin fuel elements under parameters similar to operation conditions of Russian design pool-type research reactors. The results received both in Russia and abroad enabled to go on to the next stage of development which includes irradiation tests both of full-scale IRT pin-type and tube-type fuel assemblies with U-Mo dispersion fuel and of mini-fuel elements with modified U-Mo dispersion fuel and monolithic fuel. The paper gives a generalized review of the results of U-Mo fuel development accomplished by now. (author)

  4. Fuel cells for transportation R and D at Argonne National Laboratory

    SciTech Connect

    Kumar, R.; Ahmed, S.; Bloom, I.; Carter, J.D.; Doshi, R.; Kramarz, K.; Lee, S.H.D.; Krumpelt, M.; Myles, K.M.

    1997-10-01

    This paper describes the transportation fuel cell systems research at Argonne National Laboratory (ANL). Two areas of research are discussed: the development of a catalytic partial-oxidation reformer for conventional and alternative transportation fuels, and a novel approach for the removal of carbon monoxide from reformate for use in polymer electrolyte fuel cells. The objective of the first study is to develop reformers for converting liquid fuels (gasoline, ethanol, or methanol) to hydrogen gas for use with fuel cell systems in light-duty vehicles. The second study is investigating the use of acidic cuprous chloride (or other suitable sorbent) to chemically bind and thus remove the CO from the reformate.

  5. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    SciTech Connect

    Chang, Y.I.

    1992-01-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

  6. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    SciTech Connect

    Chang, Y.I.

    1992-07-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

  7. Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)

    SciTech Connect

    J. K. Wright; R. N. Wright

    2010-07-01

    The U.S. Department of Energy (DOE) has selected the High-Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production, with an outlet gas temperature in the range of 750°C, and a design service life of 60 years. The reactor design will be a graphite-moderated, helium-cooled, prismatic, or pebble bed reactor and use low-enriched uranium, Tri-Isotopic (TRISO)-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. This technology development plan details the additional research and development (R&D) required to design and license the NGNP RPV, assuming that A 508/A 533 is the material of construction. The majority of additional information that is required is related to long-term aging behavior at NGNP vessel temperatures, which are somewhat above those commonly encountered in the existing database from LWR experience. Additional data are also required for the anticipated NGNP environment. An assessment of required R&D for a Grade 91 vessel has been retained from the first revision of the R&D plan in Appendix B in somewhat less detail. Considerably more development is required for this steel compared to A 508/A 533 including additional irradiation testing for expected NGNP operating temperatures, high-temperature mechanical properties, and extensive studies of long-term microstructural stability.

  8. "Ask Argonne" - Charlie Catlett, Computer Scientist, Part 2

    ScienceCinema

    Catlett, Charlie

    2016-07-12

    A few weeks back, computer scientist Charlie Catlett talked a bit about the work he does and invited questions from the public during Part 1 of his "Ask Argonne" video set (http://bit.ly/1joBtzk). In Part 2, he answers some of the questions that were submitted. Enjoy!

  9. "Ask Argonne" - Charlie Catlett, Computer Scientist, Part 2

    SciTech Connect

    Catlett, Charlie

    2014-06-17

    A few weeks back, computer scientist Charlie Catlett talked a bit about the work he does and invited questions from the public during Part 1 of his "Ask Argonne" video set (http://bit.ly/1joBtzk). In Part 2, he answers some of the questions that were submitted. Enjoy!

  10. "Ask Argonne" - Robert Jacob, Climate Scientist, Part 2

    ScienceCinema

    Jacob, Robert

    2016-07-12

    Previously, climate scientist Robert Jacob talked a bit about the work he does and invited questions from the public during Part 1 of his "Ask Argonne" video set (http://bit.ly/1aK6WDv). In Part 2, he answers some of the questions that were submitted.

  11. Argonne National Laboratory Smart Grid Technology Interactive Model

    ScienceCinema

    Ted Bohn

    2016-07-12

    As our attention turns to new cars that run partially or completely on electricity, how can we redesign our electric grid to not only handle the new load, but make electricity cheap and efficient for everyone? Argonne engineer Ted Bohn explains a model of a "smart grid" that gives consumers the power to choose their own prices and sources of electricity.

  12. Organic structure components within the Argonne premium coal samples

    SciTech Connect

    Winans, R.E.

    1994-02-01

    Several key features are evident from examining the data (desorption electron impact high resolution mass spectrometry) from the Argonne coals. First, the aromatic clusters are not very large. Second, heteroatoms are associated with a majority of the clusters. Finally, we need more quantitative information on the links between clusters.

  13. Argonne Director Eric Isaacs addresses the National Press Club

    SciTech Connect

    Eric Isaccs

    2009-09-17

    Argonne Director Eric Isaacs addresses the National Press Club on 9/15/2009. To build a national economy based on sustainable energy, the nation must first "reignite its innovation ecology," he said. Issacs makes the case for investing in science to secure America's future.

  14. Argonne Director Eric Isaacs addresses the National Press Club

    ScienceCinema

    Eric Isaccs

    2016-07-12

    Argonne Director Eric Isaacs addresses the National Press Club on 9/15/2009. To build a national economy based on sustainable energy, the nation must first "reignite its innovation ecology," he said. Issacs makes the case for investing in science to secure America's future.

  15. Argonne Laboratory Computing Resource Center - FY2004 Report.

    SciTech Connect

    Bair, R.

    2005-04-14

    In the spring of 2002, Argonne National Laboratory founded the Laboratory Computing Resource Center, and in April 2003 LCRC began full operations with Argonne's first teraflops computing cluster. The LCRC's driving mission is to enable and promote computational science and engineering across the Laboratory, primarily by operating computing facilities and supporting application use and development. This report describes the scientific activities, computing facilities, and usage in the first eighteen months of LCRC operation. In this short time LCRC has had broad impact on programs across the Laboratory. The LCRC computing facility, Jazz, is available to the entire Laboratory community. In addition, the LCRC staff provides training in high-performance computing and guidance on application usage, code porting, and algorithm development. All Argonne personnel and collaborators are encouraged to take advantage of this computing resource and to provide input into the vision and plans for computing and computational analysis at Argonne. Steering for LCRC comes from the Computational Science Advisory Committee, composed of computing experts from many Laboratory divisions. The CSAC Allocations Committee makes decisions on individual project allocations for Jazz.

  16. Argonne National Laboratory Smart Grid Technology Interactive Model

    SciTech Connect

    Ted Bohn

    2009-10-13

    As our attention turns to new cars that run partially or completely on electricity, how can we redesign our electric grid to not only handle the new load, but make electricity cheap and efficient for everyone? Argonne engineer Ted Bohn explains a model of a "smart grid" that gives consumers the power to choose their own prices and sources of electricity.

  17. "Ask Argonne" - Robert Jacob, Climate Scientist, Part 2

    SciTech Connect

    Jacob, Robert

    2014-01-08

    Previously, climate scientist Robert Jacob talked a bit about the work he does and invited questions from the public during Part 1 of his "Ask Argonne" video set (http://bit.ly/1aK6WDv). In Part 2, he answers some of the questions that were submitted.

  18. Safety evaluation report related to the renewal of the facility license for the research reactor at the Dow Chemical Company

    SciTech Connect

    Not Available

    1989-04-01

    This safety evaluation report for the application filed by the Dow Chemical Company for renewal of facility Operating License R-108 to continue to operate its research reactor at an increased operating power level has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located on the grounds of the Michigan Division of the Dow Chemical Company in Midland, Michigan. The staff concludes that the Dow Chemical Company can continue to operate its reactor without endangering the health and safety of the public.

  19. Determination of fast neutron flux distribution in irradiation sites of the Malaysian Nuclear Agency research reactor.

    PubMed

    Yavar, A R; Sarmani, S B; Wood, A K; Fadzil, S M; Radir, M H; Khoo, K S

    2011-05-01

    Determination of thermal to fast neutron flux ratio (f(fast)) and fast neutron flux (ϕ(fast)) is required for fast neutron reactions, fast neutron activation analysis, and for correcting interference reactions. The f(fast) and subsequently ϕ(fast) were determined using the absolute method. The f(fast) ranged from 48 to 155, and the ϕ(fast) was found in the range 1.03×10(10)-4.89×10(10) n cm(-2) s(-1). These values indicate an acceptable conformity and applicable for installation of the fast neutron facility at the MNA research reactor.

  20. Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Appendices

    SciTech Connect

    Konzek, G.J.; Ludwick, J.D.; Kennedy, W.E. Jr.; Smith, R.I.

    1982-03-01

    Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and EMTOMB (entombment). The study results are presented in two volumes. Volume 2 (Appendices) contains the detailed data that support the results given in Volume 1, including unit-component data.

  1. Filtered epithermal quasi-monoenergetic neutron beams at research reactor facilities.

    PubMed

    Mansy, M S; Bashter, I I; El-Mesiry, M S; Habib, N; Adib, M

    2015-03-01

    Filtered neutron techniques were applied to produce quasi-monoenergetic neutron beams in the energy range of 1.5-133keV at research reactors. A simulation study was performed to characterize the filter components and transmitted beam lines. The filtered beams were characterized in terms of the optimal thickness of the main and additive components. The filtered neutron beams had high purity and intensity, with low contamination from the accompanying thermal emission, fast neutrons and γ-rays. A computer code named "QMNB" was developed in the "MATLAB" programming language to perform the required calculations.

  2. Modification of the radial beam port of ITU TRIGA Mark II research reactor for BNCT applications.

    PubMed

    Akan, Zafer; Türkmen, Mehmet; Çakir, Tahir; Reyhancan, İskender A; Çolak, Üner; Okka, Muhittin; Kiziltaş, Sahip

    2015-05-01

    This paper aims to describe the modification of the radial beam port of ITU (İstanbul Technical University) TRIGA Mark II research reactor for BNCT applications. Radial beam port is modified with Polyethylene and Cerrobend collimators. Neutron flux values are measured by neutron activation analysis (Au-Cd foils). Experimental results are verified with Monte Carlo results. The results of neutron/photon spectrum, thermal/epithermal neutron flux, fast group photon fluence and change of the neutron fluxes with the beam port length are presented.

  3. Dose Measurements at Epithermal Beams of Research Reactors with Fricke Gel and Thermoluminescence Detectors

    SciTech Connect

    Gambarini, Grazia; Artuso, Emanuele; Giove, Dario; Felisi, Marco; Agosteo, Stefano; Barcaglioni, Luca; Pola, Andrea; Garlati, Luisella; Borroni, Marta; Carrara, Mauro; Klupak, Vit; Viererbl, Ladislav; Vins, Miroslav; Marek, Milan

    2015-07-01

    Suitable dosimeter methods have to be developed to measure the different dose contributions in phantoms exposed to epithermal/thermal neutron beams of a research reactor. The method based on Fricke Xylenol Orange gel dosimeter in form of layers has shown to be very effective for achieving images of the various dose components in air or in phantoms exposed to epithermal/thermal neutron beams with very high fluence rate. Another useful method is based on the use of TLD-700 chips, from whose response the gamma dose and the thermal neutron fluence can be obtained by means of appropriate parameters of the glow curve. (authors)

  4. Evaluation of differential shim rod worth measurements in the Oak Ridge Research Reactor

    SciTech Connect

    Bretscher, M.M.

    1987-01-01

    Reasonable agreement between calculated and measured differential shim rod worths in the Oak Ridge Research Reactor (ORR) has been achieved by taking into account the combined effects of negative reactivity contributions from changing fuel-moderator temperatures and of delayed photoneutrons. A method has been developed for extracting the asymptotic period from the shape of the initial portion of the measured time-dependent neutron flux profile following a positive reactivity insertion. In this region of the curve temperature-related reactivity feedback effects are negligibly small. Results obtained by applying this technique to differential shim rod worth measurements made in a wide variety of ORR cores are presented.

  5. Secretary Chu visits Argonne—Groundbreaking ceremony for new Energy Sciences building

    ScienceCinema

    Isaacs, Eric

    2016-07-12

    U.S. Department of Energy Secretary Steven Chu, joined Senator Richard Durbin, University of Chicago President Robert Zimmer and Argonne Director Eric Isaacs to break ground for Argonne's new Energy and Sciences building.

  6. Volume Reduction of Solid Radioactive Waste From Research Reactor and Nuclear Laboratories - Industrial Experience

    SciTech Connect

    Singh, B.N.; Gandhi, K.G.; Chander, M.; Raj, K.

    2006-07-01

    Various research reactors and nuclear laboratories at Bhabha Atomic Research Centre, Mumbai, India generate approximately 600 m{sup 3} of radioactive solid waste annually. These wastes are categorized and segregated based on their radiation field, physical nature and radionuclides present. The low level waste is further segregated based on compactability criteria. The compactable wastes are packed in 200 litres carbon steel drums and pelletized to get a volume reduction factor of about five. The compaction system designed for Cat-I (Table-1) radioactive waste is having 200 Tons capacity hydraulic press, housed in a well-ventilated enclosure. Before pelletizing, the drum is assayed to estimate {beta},{gamma} activity. Further, the imaging of waste drum is also done so as to avoid any possibility of non-compactable material being taken for pelletizing. The pelletizing system comprises of conveying, pushing, indexing and compacting. All operations are controlled by programmable logic control (PLC) based control system. Apart from the drum palletising, the system is also equipped to compact the used Pre and HEPA filters, being generated from exhaust and supply air system of clean room, nuclear laboratories, research reactors, fuel reprocessing plants, high level waste management facility etc. The system is designed to handle about 5 drums or filters per hour. So far about 3000 number of each, HEPA filters and waste drums have been safely compacted and disposed. (authors)

  7. Special issue on the "Consortium for Advanced Simulation of Light Water Reactors Research and Development Progress"

    NASA Astrophysics Data System (ADS)

    Turinsky, Paul J.; Martin, William R.

    2017-04-01

    In this special issue of the Journal of Computational Physics, the research and development completed at the time of manuscript submission by the Consortium for Advanced Simulation of Light Water Reactors (CASL) is presented. CASL is the first of several Energy Innovation Hubs that have been created by the Department of Energy. The Hubs are modeled after the strong scientific management characteristics of the Manhattan Project and AT&T Bell Laboratories, and function as integrated research centers that combine basic and applied research with engineering to accelerate scientific discovery that addresses critical energy issues. Lifetime of a Hub is expected to be five or ten years depending upon performance, with CASL being granted a ten year lifetime.

  8. Supercritical Water Reactor (SCWR) - Survey of Materials Research and Development Needs to Assess Viability

    SciTech Connect

    Philip E. MacDonald

    2003-09-01

    Supercritical water-cooled reactors (SCWRs) are among the most promising advanced nuclear systems because of their high thermal efficiency [i.e., about 45% vs. 33% of current light water reactors (LWRs)] and considerable plant simplification. SCWRs achieve this with superior thermodynamic conditions (i.e., high operating pressure and temperature), and by reducing the containment volume and eliminating the need for recirculation and jet pumps, pressurizer, steam generators, steam separators and dryers. The reference SCWR design in the U.S. is a direct cycle, thermal spectrum, light-water-cooled and moderated reactor with an operating pressure of 25 MPa and inlet/outlet coolant temperature of 280/500 °C. The inlet flow splits, partly to a down-comer and partly to a plenum at the top of the reactor pressure vessel to flow downward through the core in special water rods to the inlet plenum. This strategy is employed to provide good moderation at the top of the core, where the coolant density is only about 15-20% that of liquid water. The SCWR uses a power conversion cycle similar to that used in supercritical fossil-fired plants: high- intermediate- and low-pressure turbines are employed with one moisture-separator re-heater and up to eight feedwater heaters. The reference power is 3575 MWt, the net electric power is 1600 MWe and the thermal efficiency is 44.8%. The fuel is low-enriched uranium oxide fuel and the plant is designed primarily for base load operation. The purpose of this report is to survey existing materials for fossil, fission and fusion applications and identify the materials research and development needed to establish the SCWR viabilitya with regard to possible materials of construction. The two most significant materials related factors in going from the current LWR designs to the SCWR are the increase in outlet coolant temperature from 300 to 500 °C and the possible compatibility issues associated with the supercritical water environment.

  9. Irradiation of electronic components and circuits at the Portuguese Research Reactor: Lessons learned

    SciTech Connect

    Marques, J.G.; Ramos, A.R.; Fernandes, A.C.; Santos, J.P.

    2015-07-01

    The behavior of electronic components and circuits under radiation is a concern shared by the nuclear industry, the space community and the high-energy physics community. Standard commercial components are used as much as possible instead of radiation hard components, since they are easier to obtain and allow a significant reduction of costs. However, these standard components need to be tested in order to determine their radiation tolerance. The Portuguese Research Reactor (RPI) is a 1 MW pool-type reactor, operating since 1961. The irradiation of electronic components and circuits is one area where a 1 MW reactor can be competitive, since the fast neutron fluences required for testing are in most cases well below 10{sup 16} n/cm{sup 2}. A program was started in 1999 to test electronics components and circuits for the LHC facility at CERN, initially using a dedicated in-pool irradiation device and later a beam line with tailored neutron and gamma filters. Neutron filters are essential to reduce the intensity of the thermal neutron flux, which does not produce significant defects in electronic components but produces unwanted radiation from activation of contacts and packages of integrated circuits and also of the printed circuit boards. In irradiations performed within the line-of-sight of the core of a fission reactor there is simultaneous gamma radiation which complicates testing in some cases. Filters can be used to reduce its importance and separate testing with a pure gamma radiation source can contribute to clarify some irradiation results. Practice has shown the need to introduce several improvements to the procedures and facilities over the years. We will review improvements done in the following areas: - Optimization of neutron and gamma filters; - Dosimetry procedures in mixed neutron / gamma fields; - Determination of hardness parameter and 1 MeV-equivalent neutron fluence; - Temperature measurement and control during irradiation; - Follow-up of reactor

  10. Argonne Leadership Computing Facility 2011 annual report : Shaping future supercomputing.

    SciTech Connect

    Papka, M.; Messina, P.; Coffey, R.; Drugan, C.

    2012-08-16

    The ALCF's Early Science Program aims to prepare key applications for the architecture and scale of Mira and to solidify libraries and infrastructure that will pave the way for other future production applications. Two billion core-hours have been allocated to 16 Early Science projects on Mira. The projects, in addition to promising delivery of exciting new science, are all based on state-of-the-art, petascale, parallel applications. The project teams, in collaboration with ALCF staff and IBM, have undertaken intensive efforts to adapt their software to take advantage of Mira's Blue Gene/Q architecture, which, in a number of ways, is a precursor to future high-performance-computing architecture. The Argonne Leadership Computing Facility (ALCF) enables transformative science that solves some of the most difficult challenges in biology, chemistry, energy, climate, materials, physics, and other scientific realms. Users partnering with ALCF staff have reached research milestones previously unattainable, due to the ALCF's world-class supercomputing resources and expertise in computation science. In 2011, the ALCF's commitment to providing outstanding science and leadership-class resources was honored with several prestigious awards. Research on multiscale brain blood flow simulations was named a Gordon Bell Prize finalist. Intrepid, the ALCF's BG/P system, ranked No. 1 on the Graph 500 list for the second consecutive year. The next-generation BG/Q prototype again topped the Green500 list. Skilled experts at the ALCF enable researchers to conduct breakthrough science on the Blue Gene system in key ways. The Catalyst Team matches project PIs with experienced computational scientists to maximize and accelerate research in their specific scientific domains. The Performance Engineering Team facilitates the effective use of applications on the Blue Gene system by assessing and improving the algorithms used by applications and the techniques used to implement those algorithms

  11. Status report on the Small Secure Transportable Autonomous Reactor (SSTAR) /Lead-cooled Fast Reactor (LFR) and supporting research and development.

    SciTech Connect

    Sienicki, J. J.; Moisseytsev, A.; Yang, W. S.; Wade, D. C.; Nikiforova, A.; Hanania, P.; Ryu, H. J.; Kulesza, K. P.; Kim, S. J.; Halsey, W. G.; Smith, C. F.; Brown, N. W.; Greenspan, E.; de Caro, M.; Li, N.; Hosemann, P.; Zhang, J.; Yu, H.; Nuclear Engineering Division; LLNL; LANL; Massachusetts Inst. of Tech.; Ecole des Mines de Paris; Oregon State Univ.; Univ.of California at Berkley

    2008-06-23

    This report provides an update on development of a pre-conceptual design for the Small Secure Transportable Autonomous Reactor (SSTAR) Lead-Cooled Fast Reactor (LFR) plant concept and supporting research and development activities. SSTAR is a small, 20 MWe (45 MWt), natural circulation, fast reactor plant for international deployment concept incorporating proliferation resistance for deployment in non-fuel cycle states and developing nations, fissile self-sufficiency for efficient utilization of uranium resources, autonomous load following making it suitable for small or immature grid applications, and a high degree of passive safety further supporting deployment in developing nations. In FY 2006, improvements have been made at ANL to the pre-conceptual design of both the reactor system and the energy converter which incorporates a supercritical carbon dioxide Brayton cycle providing higher plant efficiency (44 %) and improved economic competitiveness. The supercritical CO2 Brayton cycle technology is also applicable to Sodium-Cooled Fast Reactors providing the same benefits. One key accomplishment has been the development of a control strategy for automatic control of the supercritical CO2 Brayton cycle in principle enabling autonomous load following over the full power range between nominal and essentially zero power. Under autonomous load following operation, the reactor core power adjusts itself to equal the heat removal from the reactor system to the power converter through the large reactivity feedback of the fast spectrum core without the need for motion of control rods, while the automatic control of the power converter matches the heat removal from the reactor to the grid load. The report includes early calculations for an international benchmarking problem for a LBE-cooled, nitride-fueled fast reactor core organized by the IAEA as part of a Coordinated Research Project on Small Reactors without Onsite Refueling; the calculations use the same neutronics

  12. Neutron flux characterisation of the Pavia TRIGA Mark II research reactor for radiobiological and microdosimetric applications.

    PubMed

    Alloni, D; Prata, M; Salvini, A; Ottolenghi, A

    2015-09-01

    Nowadays the Pavia TRIGA reactor is available for national and international collaboration in various research fields. The TRIGA Mark II nuclear research reactor of the Pavia University offers different in- and out-core neutron irradiation channels, each characterised by different neutron spectra. In the last two years a campaign of measurements and simulations has been performed in order to guarantee a better characterisation of these different fluxes and to meet the demands of irradiations that require precise information on these spectra in particular for radiobiological and microdosimetric studies. Experimental data on neutron fluxes have been collected analysing and measuring the gamma activity induced in thin target foils of different materials irradiated in different TRIGA experimental channels. The data on the induced gamma activities have been processed with the SAND II deconvolution code and finally compared with the spectra obtained with Monte Carlo simulations. The comparison between simulated and measured spectra showed a good agreement allowing a more precise characterisation of the neutron spectra and a validation of the adopted method.

  13. Fundamental approaches for analysis thermal hydraulic parameter for Puspati Research Reactor

    SciTech Connect

    Hashim, Zaredah Lanyau, Tonny Anak Farid, Mohamad Fairus Abdul; Kassim, Mohammad Suhaimi; Azhar, Noraishah Syahirah

    2016-01-22

    The 1-MW PUSPATI Research Reactor (RTP) is the one and only nuclear pool type research reactor developed by General Atomic (GA) in Malaysia. It was installed at Malaysian Nuclear Agency and has reached the first criticality on 8 June 1982. Based on the initial core which comprised of 80 standard TRIGA fuel elements, the very fundamental thermal hydraulic model was investigated during steady state operation using the PARET-code. The main objective of this paper is to determine the variation of temperature profiles and Departure of Nucleate Boiling Ratio (DNBR) of RTP at full power operation. The second objective is to confirm that the values obtained from PARET-code are in agreement with Safety Analysis Report (SAR) for RTP. The code was employed for the hot and average channels in the core in order to calculate of fuel’s center and surface, cladding, coolant temperatures as well as DNBR’s values. In this study, it was found that the results obtained from the PARET-code showed that the thermal hydraulic parameters related to safety for initial core which was cooled by natural convection was in agreement with the designed values and safety limit in SAR.

  14. Nuclear mass inventory, photon dose rate and thermal decay heat of spent research reactor fuel assemblies

    SciTech Connect

    Pond, R.B.; Matos, J.E.

    1996-05-01

    As part of the Department of Energy`s spent nuclear fuel acceptance criteria, the mass of uranium and transuranic elements in spent research reactor fuel must be specified. These data are, however, not always known or readily determined. It is the purpose of this report to provide estimates of these data for some of the more common research reactor fuel assembly types. The specific types considered here are MTR, TRIGA and DIDO fuel assemblies. The degree of physical protection given to spent fuel assemblies is largely dependent upon the photon dose rate of the spent fuel material. These data also, are not always known or readily determined. Because of a self-protecting dose rate level of radiation (dose rate greater than 100 ren-x/h at I m in air), it is important to know the dose rate of spent fuel assemblies at all time. Estimates of the photon dose rate for spent MTR, TRIGA and DIDO-type fuel assemblies are given in this report.

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

    SciTech Connect

    Tolosa, S.C.; Marajofsky, A.

    2004-10-06

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

  16. The current state of the Russian reduced enrichment research reactors program

    SciTech Connect

    Aden, V.G.; Kartashov, E.F.; Lukichev, V.A.

    1997-08-01

    During the last year after the 16-th International Conference on Reducing Fuel Enrichment in Research Reactors held in October, 1993 in Oarai, Japan, the conclusive stage of the Program on reducing fuel enrichment (to 20% in U-235) in research reactors was finally made up in Russia. The Program was started late in 70th and the first stage of the Program was completed by 1986 which allowed to reduce fuel enrichment from 80-90% to 36%. The completion of the Program current stage, which is counted for 5-6 years, will exclude the use of the fuel enriched by more than 20% from RF to other countries such as: Poland, Czeck Republick, Hungary, Roumania, Bulgaria, Libya, Viet-Nam, North Korea, Egypt, Latvia, Ukraine, Uzbekistan and Kazakhstan. In 1994 the Program, approved by RF Minatom authorities, has received the status of an inter-branch program since it was admitted by the RF Ministry for Science and Technical Policy. The Head of RF Minatom central administrative division N.I.Ermakov was nominated as the Head of the Russian Program, V.G.Aden, RDIPE Deputy Director, was nominated as the scientific leader. The Program was submitted to the Commission for Scientific, Technical and Economical Cooperation between USA and Russia headed by Vice-President A. Gore and Prime Minister V. Chemomyrdin and was given support also.

  17. The Oak Ridge Research Reactor: Safety analysis: Volume 2, Supplement 3

    SciTech Connect

    Cook, D.H.; Hamrick, T.P.

    1987-06-29

    The Oak Ridge Research Reactor (ORR) was constructed in the mid 1950s. Since it is an older facility, the issue of life-limiting conditions or material deterioration resulting from prolonged exposure to the normal operating environment is an item that should be addressed in the safety analysis for the ORR. Life-limiting conditions were considered in the original design of ORR; but due to the limited data that were available at that time on material performance in research reactors, various studies were completed during the first 10 years of operation at ORR to verify the applicable life-limiting parameters. Based on today's knowledge of life limiting conditions and the previous 30 years of operating experience at the ORR facility, the three specific areas of concern are addressed in this supplement: (1) embrittlement of the structures due to radiation damage, which is described in Section 2; (2) fatigue due to the effects of both thermal cycling and vibration, which is addressed in Section 3; and (3) the effects of corrosion on the integrity of the primary system, which is described in Section 4. The purpose of this document is to provide a review of the applicable safety studies which have been performed, and to state the status of the ORR with regard to embrittlement, fatigue (due to thermal cycling and vibration), and corrosion.

  18. Fundamental approaches for analysis thermal hydraulic parameter for Puspati Research Reactor

    NASA Astrophysics Data System (ADS)

    Hashim, Zaredah; Lanyau, Tonny Anak; Farid, Mohamad Fairus Abdul; Kassim, Mohammad Suhaimi; Azhar, Noraishah Syahirah

    2016-01-01

    The 1-MW PUSPATI Research Reactor (RTP) is the one and only nuclear pool type research reactor developed by General Atomic (GA) in Malaysia. It was installed at Malaysian Nuclear Agency and has reached the first criticality on 8 June 1982. Based on the initial core which comprised of 80 standard TRIGA fuel elements, the very fundamental thermal hydraulic model was investigated during steady state operation using the PARET-code. The main objective of this paper is to determine the variation of temperature profiles and Departure of Nucleate Boiling Ratio (DNBR) of RTP at full power operation. The second objective is to confirm that the values obtained from PARET-code are in agreement with Safety Analysis Report (SAR) for RTP. The code was employed for the hot and average channels in the core in order to calculate of fuel's center and surface, cladding, coolant temperatures as well as DNBR's values. In this study, it was found that the results obtained from the PARET-code showed that the thermal hydraulic parameters related to safety for initial core which was cooled by natural convection was in agreement with the designed values and safety limit in SAR.

  19. The Chemical Technology Division at Argonne National Laboratory: Applying chemical innovation to environmental problems

    SciTech Connect

    1995-06-01

    The Chemical Technology Division is one of the largest technical divisions at Argonne National Laboratory, a leading center for research and development related to energy and environmental issues. Since its inception in 1948, the Division has pioneered in developing separations processes for the nuclear industry. The current scope of activities includes R&D on methods for disposing of radioactive and hazardous wastes and on energy conversion processes with improved efficiencies, lower costs, and reduced environmental impact. Many of the technologies developed by CMT can be applied to solve manufacturing as well as environmental problems of industry.

  20. IAEA Coordinated Research Project on HTGR Reactor Physics, Thermal-hydraulics and Depletion Uncertainty Analysis

    SciTech Connect

    Strydom, Gerhard; Bostelmann, F.

    2015-09-01

    The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of HTGR design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The predictive capability of coupled neutronics/thermal-hydraulics and depletion simulations for reactor design and safety analysis can be assessed with sensitivity analysis (SA) and uncertainty analysis (UA) methods. Uncertainty originates from errors in physical data, manufacturing uncertainties, modelling and computational algorithms. (The interested reader is referred to the large body of published SA and UA literature for a more complete overview of the various types of uncertainties, methodologies and results obtained). SA is helpful for ranking the various sources of uncertainty and error in the results of core analyses. SA and UA are required to address cost, safety, and licensing needs and should be applied to all aspects of reactor multi-physics simulation. SA and UA can guide experimental, modelling, and algorithm research and development. Current SA and UA rely either on derivative-based methods such as stochastic sampling methods or on generalized perturbation theory to obtain sensitivity coefficients. Neither approach addresses all needs. In order to benefit from recent advances in modelling and simulation and the availability of new covariance data (nuclear data uncertainties) extensive sensitivity and uncertainty studies are needed for quantification of the impact of different sources of uncertainties on the design and safety parameters of HTGRs. Only a parallel effort in advanced simulation and in nuclear data improvement will be able to provide designers with more robust and well validated calculation tools to meet design target accuracies. In February 2009, the Technical Working Group on Gas-Cooled Reactors (TWG-GCR) of the International Atomic Energy Agency (IAEA) recommended that the proposed Coordinated Research Program (CRP) on

  1. IAEA coordinated research projects on core physics benchmarks for high temperature gas-cooled reactors

    SciTech Connect

    Methnani, M.

    2006-07-01

    High-temperature Gas-Cooled Reactor (HTGR) designs present special computational challenges related to their core physics characteristics, in particular neutron streaming, double heterogeneities, impurities and the random distribution of coated fuel particles in the graphite matrix. In recent years, two consecutive IAEA Coordinated Research Projects (CRP 1 and CRP 5) have focused on code-to-code and code-to-experiment comparisons of representative benchmarks run by several participating international institutes. While the PROTEUS critical HTR experiments provided the test data reference for CRP-1, the more recent CRP-5 data has been made available by the HTTR, HTR-10 and ASTRA test facilities. Other benchmark cases are being considered for the GT-MHR and PBMR core designs. This paper overviews the scope and some sample results of both coordinated research projects. (authors)

  2. Recommendations concerning research and model evaluation needs to support breeder reactor environmental radiological assessments

    SciTech Connect

    Miller, C. W.; Dunning, Jr., D. E.; Etnier, E. L.; Kocher, D. C.; McDowell-Boyer, L. M.; Meyer, H. R.; Rohwer, P. S.

    1980-12-01

    Purpose of this report is to present recommendations concerning needs for model evaluations, environmental research, and biomedical research to support breeder reactor environmental radiological assessments. More data are needed to specify dry deposition velocities and to validate plume depletion models. More atmospheric dispersion data are required to characterize flow near buildings, in complex terrain, and for travel distances at 100 km or more. Field data are needed for terrestrial food chain transport models, especially those used to assess the impact of acute radionuclide releases. Efforts are needed to develop models for the estimation of dose from external exposure to photons from a finite, elevated plume resulting from an acute radionuclide release to the atmosphere. Estimates of doses to man from internally deposited radionuclides require scrutiny. Further study of tritium is needed to determine its dependence on dose and dose rate and to specify the relative toxicity of various physiochemical forms of tritium in the environment.

  3. DOE/NE University Program in robotics for advanced reactors research

    SciTech Connect

    Trivedi, M.M.

    1990-01-01

    The document presents the bimonthly progress reports published during 1990 regarding the US Department of Energy/NE-sponsored research at the University of Tennessee Knoxville under the DOE Robitics for Advanced Reactors Research Grant. Significant accomplishments are noted in the following areas: development of edge-segment based stereo matching algorithm; vision system integration in the CESAR laboratory; evaluation of algorithms for surface characterization from range data; comparative study of data fusion techniques; development of architectural framework, software, and graphics environment for sensor-based robots; algorithms for acquiring tactile images from planer surfaces; investigations in geometric model-based robotic manipulation; investigations of non-deterministic approaches to sensor fusion; and evaluation of sensor calibration techniques. (MB)

  4. CERL/ORNL research and development programs in support of prestressed concrete reactor vessel development

    SciTech Connect

    Hornby, I.W.; Naus, D.J.

    1984-01-01

    In support of the evolution of PCRV designs being developed both in the UK and USA, research and developments programmers are being conducted at the CEGB Central Electricity Research Laboratories (CERL) and the Oak Ridge National Laboratory (ORNL) respectively. In the UK, recent work has focused on elevated temperature effects on concrete properties and instrument systems for PCRVs. The concrete development program at ORNL consists of generic studies designed to provide technical support for ongoing prestressed concrete reactor vessel-related activities, to contribute to the technological data base, and to provide independent review and evaluation of the relevant technology. Recent activities have been related to the development of properties for high-strength concrete mix designs for the PCRV of a 2240 MW(t) HTGR-SC/C lead plant project, and the development of PCRV model testing techniques.

  5. Dual Shell Pressure Balanced Reactor Vessel cooperative research and development agreement with Innotek, Inc., Final report

    SciTech Connect

    Robertus, R.J.; Fassbender, A.G.; Deverman, G.S.

    1995-04-01

    The Department of Energy`s Office of Energy Research (OER) has provided support for the development of several chemical processes, including supercritical water oxidation, liquefaction, and aqueous hazardous waste destruction, where chemical and phase transformations are conducted at high pressure and temperature. These and many other commercial processes require a pressure vessel capable of operating in a corrosive environment where safety and economy are important requirements. This document details a cooperative research and development agreement for a novel Dual Shell Pressure Balanced Vessel (DSPBV) which could solve a number of these problems. The Technology could be immediately useful in continuing commercialization of an R&D 100 award-winning technology, Sludge-to-oil Reactor System (STORS), originally developed through funding by OER.

  6. Delayed Gamma Measurements in Different Nuclear Research Reactors Bringing Out the Importance of the Delayed Contribution in Gamma Flux Calculations

    SciTech Connect

    Fourmentel, D.; Radulovic, V.; Barbot, L.; Villard, J-F.; Zerovnik, G.; Snoj, L.; Tarchalski, M.; Pytel, K.; Malouch, F.

    2015-07-01

    Neutron and gamma flux levels are key parameters in nuclear research reactors. In Material Testing Reactors, such as the future Jules Horowitz Reactor, under construction at the French Alternative Energies and Atomic Energy Commission (CEA Cadarache, France), the expected gamma flux levels are very high (nuclear heating is of the order of 20 W/g at 100 MWth). As gamma rays deposit their energy in the reactor structures and structural materials it is important to take them into account when designing irradiation devices. There are only a few sensors which allow measurements of the nuclear heating ; a recent development at the CEA Cadarache allows measurements of the gamma flux using a miniature ionization chamber (MIC). The measured MIC response is often compared with calculation using modern Monte Carlo (MC) neutron and photon transport codes, such as TRIPOLI-4 and MCNP6. In these calculations only the production of prompt gamma rays in the reactor is usually modelled thus neglecting the delayed gamma rays. Hence calculations and measurements are usually in better accordance for the neutron flux than for the gamma flux. In this paper we study the contribution of delayed gamma rays to the total MIC signal in order to estimate the systematic error in gamma flux MC calculations. In order to experimentally determine the delayed gamma flux contributions to the MIC response, we performed gamma flux measurements with CEA developed MIC at three different research reactors: the OSIRIS reactor (MTR - 70 MWth at CEA Saclay, France), the TRIGA MARK II reactor (TRIGA - 250 kWth at the Jozef Stefan Institute, Slovenia) and the MARIA reactor (MTR - 30 MWth at the National Center for Nuclear Research, Poland). In order to experimentally assess the delayed gamma flux contribution to the total gamma flux, several reactor shut down (scram) experiments were performed specifically for the purpose of the measurements. Results show that on average about 30 % of the MIC signal is due to

  7. The RERTR (Reduced Enrichment Research and Test Reactor) Program: Progress and plans

    SciTech Connect

    Travelli, A.

    1987-01-01

    The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. After a brief summary of the results which the RERTR Program, in collaboration with its many international partners, had achieved by the end of 1986, the activities, results, and new developments which occurred in 1987 are reviewed. Irradiation of the second miniplate series, concentrating on U/sub 3/Si/sub 2/-Al and U/sub 3/Si-Al fuels, was completed and postirradiation examinations were performed on many of its miniplates. The whole-core ORR demonstration with U/sub 3/Si/sub 2/-Al fuel at 4.8 g U/cm/sup 3/ was completed at the end of March with excellent results and with 29 elements estimated to have reached at least 40% average burnup. Good progress was made in the area of LEU usage for the production of fission /sup 99/Mo, and in the coordination of safety evaluations related to LEU conversions of US university reactors. Planned activities include testing and demonstrating advanced fuels intended to allow use of reduced enrichment uranium in very-high-performance reactors. Two candidate fuels are U/sub 3/Si-Al with 19.75% enrichment and U/sub 3/Si/sub 2/-Al with 45% enrichment. Demonstration of these fuels will include irradiation of full-size elements and, possibly, a full-core demonstration. Achievement of the final program goals is still projected for 1990. This progress could not have been possible without the close international cooperation which has existed from the beginning, and which is essential to the ultimate success of the RERTR Program.

  8. Microcomputer-based equipment-control and data-acquisition system for fission-reactor reactivity-worth measurements

    SciTech Connect

    McDowell, W.P.; Bucher, R.G.

    1980-01-01

    Material reactivity-worth measurements are one of the major classes of experiments conducted on the Zero Power research reactors (ZPR) at Argonne National Laboratory. These measurements require the monitoring of the position of a servo control element as a sample material is positioned at various locations in a critical reactor configuration. In order to guarantee operational reliability and increase experimental flexibility for these measurements, the obsolete hardware-based control unit has been replaced with a microcomputer based equipment control and data acquisition system. This system is based on an S-100 bus, dual floppy disk computer with custom built cards to interface with the experimental system. To measure reactivity worths, the system accurately positions samples in the reactor core and acquires data on the position of the servo control element. The data are then analyzed to determine statistical adequacy. The paper covers both the hardware and software aspects of the design.

  9. CWDD accelerator at Argonne: Status and future opportunities

    SciTech Connect

    McMichael, G.; Carwardine, J.; Den Hartog, P.; Sagalovsky, L.; Yule, T.; Clarkson, I.; Papsco, R.; Pile, G.

    1994-09-01

    The Continuous Wave Deuterium Demonstrator (CWDD) accelerator, a cryogenically-cooled (26K) linac, was designed to accelerate 80 mA cw of D to 7.5 MeV. CWDD was being built to demonstrate the launching of a beam with characteristics suitable for a space-based neutral particle beam (NPB). A considerable amount of hardware was constructed and installed in the Argonne-based facility, and major performance milestones were achieved before program funding ended in October 1993. Existing assets have been turned over to Argonne for continuation under other sponsors. These include a fully functional 200 kV cw D injector and high power (1 MW) cw rf amplifier, a cw RFQ that has been tuned, leak checked and aligned, and a partially completed ramped-gradient DTL. Project status and achievements are reviewed and proposals for future use of the equipment are discussed.

  10. [Research on Cultivation and Stability of Nitritation Granular Sludge in Integrated ABR-CSTR Reactor].

    PubMed

    Wu, Kai-cheng; Wu, Peng; Shen, Yao-liang; Li, Yue-han; Wang, Han-fang; Xu, Yue-zhong

    2015-11-01

    Abstract: The last two compartments of the Anaerobic Baffled Readtor ( ABR) were altered into aeration tank and sedimentation tank respectively to get an integrated anaerobic-aerobic reactor, using anaerobic granular sludge in anaerobic zone and aerobic granular sludge in aerobic zone as seed sludge. The research explored the condition to cultivate nitritation granular sludge, under the condition of continuous flow. The C/N rate was decreased from 1 to 0.4 and the ammonia nitrogen volumetric loading rate was increased from 0.89 kg x ( m3 x d)(-1) to 2.23 kg x (m3 x d)(-1) while the setting time of 1 h was controlled in the aerobic zone. After the system was operated for 45 days, the mature nitritation granular sludge in aerobic zone showed a compact structure and yellow color while the nitrite accumulation rate was about 80% in the effluent. The associated inhibition of free ammonia (FA) and free nitrous acid (FNA) dominated the nitritation. Part of granules lost stability during the initial period of operation and flocs appeared in the aerobic zone. However, the flocs were transformed into newly generated small particles in the following reactor operation, demonstrating that organic carbon was benefit to granulation and the enrichment of slow-growing nitrifying played an important role in the stability of granules.

  11. LOSS-OF-COOLANT ACIDENT SIMULATIONS IN THE NATIONAL RESEARCH UNIVERSAL REACTOR

    SciTech Connect

    Bennett, W D; Goodman, R L; Heaberlin, S W; Hesson, G M; Nealley, C; Kirg, L L; Marshall, R K; McNair, G W; Meitzler, W D; Neally, G W; Parchen, L J; Pilger, J P; Rausch, W N; Russcher, G E; Schreiber, R E; Wildung, N J; Wilson, C L

    1981-02-01

    Pressurized water reactor loss-of-coolant accident (LOCA) phenomena are being simulated with a series of experiments in the U-2 loop of the National Research Universal Reactor at Chalk River, Ontario, Canada. The first of these experiments includes up to 45 parametric thermal-hydraulic tests to establish the relationship among the reflood delay time of emergency coolant, the reflooding rate, and the resultant fuel rod cladding peak temperature. Subsequent experiments establish the fuel rod failure characteristics at selected peak cladding temperatures. Fuel rod cladding pressurization simulates high burnup fission gas pressure levels of modern PWRs. This document contains both an experiment overview of the LOCA simulation program and a review of the safety analyses performed by Pacific Northwest Laboratory (PNL) to define the expected operating conditions as well as to evaluate the worst case operating conditions. The primary intent of this document is to supply safety information required by the Chalk River Nuclear Laboratories (CRNL), to establish readiness to proceed from one test phase to the next and to establish the overall safety of the experiment. A hazards review summarizes safety issues, normal operation and three worst case accidents that have been addressed during the development of the experiment plan.

  12. A neutronic feasibility study for LEU conversion of the WWR-SM research reactor in Uzbekistan.

    SciTech Connect

    Rakhmanov, A.

    1998-10-19

    The WWR-SM research reactor in Uzbekistan has operated at 10 MW since 1979, using Russian-supplied IRT-3M fuel assemblies containing 90% enriched uranium. Burnup tests of three full-sized IRT-3M FA with 36% enrichment were successfully completed to a burn up of about {approximately}50% in 1987-1989. In August 1998, four IRT-3M FA with 36% enriched uranium were loaded into the core to initiate conversion of the entire core to 36% enriched fuel. This paper presents the results of equilibrium fuel cycle comparisons of the reactor using HEU (90%) and HEU (36%) IRT-3M fuel and compares results with the performance of IRT-4M FA containing LEU (19.75%). The results show that an LEU (19.75%) density of 3.8 g/cm{sup 3} is required to match the cycle length of the HEU (90%) core and an LEU density 3.9 g/cm{sup 3} is needed to match the cycle length of the HEU (36%) core.

  13. A neutronic feasibility study for LEU conversion of the IR-8 research reactor.

    SciTech Connect

    Deen, J. R.

    1998-10-22

    Equilibrium fuel cycle comparisons for the IR-8 research reactor were made for HEU(90%), HEU(36%), and LEU (19.75%) fuel assembly (FA) designs using three dimensional multi-group diffusion theory models benchmarked to detailed Monte Carlo models of the reactor. Comparisons were made of changes in reactivity, cycle length, average {sup 235}U discharge burnup, thermal neutron flux, and control rod worths for the 90% and 36% enriched IRT-3M fuel assembly and the 19.75% enriched IRT-4M fuel assembly with the same fuel management strategy. The results of these comparisons showed that a uranium density of 3.5 g/cm{sup 3} in the fuel meat would be required in the LEU IRT-4M fuel assembly to match the cycle length of the HEU(90%) IRT-3M FA and an LEU density of 3.7 g/cm{sup 3} is needed to match the cycle length of the HEU(36%) IRT-3M FA.

  14. Site characteristics of Argonne National Laboratory in Illinois

    SciTech Connect

    Chang, Y.W.

    1995-01-01

    This report reviews the geology and topography of the Argonne National Laboratory, near Lemont, Illinois. It describes the thickness and stratigraphy of soils, glacial till, and bedrock in and adjacent to the laboratory and support facilities. Seismic surveys were also conducted through the area to help determine the values of seismic wave velocities in the glacial till which is important in determining the seismic hazard of the area. Borehole log descriptions are summarized along with information on area topography.

  15. Progress of the Argonne Pulsed Cable Test Facility

    SciTech Connect

    Kim, S.H.; Knott, M.J.; Krieger, C.I.; McGhee, D.G.

    1983-01-01

    Initial tests of the Pulsed Cable Test Facility (PCTF) at Argonne National Laboratory (ANL) have been completed. Additional components are prepared for the tests of developmental high-current cable conductors: a minicomputer based data acquisition system, a new 5.5 MW power supply for triangular or trapezoidal pulsing modes of the PCTF coil, and a pair of 25 kA current leads for the transport current of the testing conductors.

  16. Characterization of wastes in and around early reactors at the Hanford Site: The use of historical research

    SciTech Connect

    Gerber, M.S.

    1993-10-01

    This paper will present the waste characterization knowledge that has been gained in the first, ``Large-Scale Remediation Study`` to be performed on the reactor areas (100 Areas) of the Hanford Site. Undertaken throughout the past year, this research project has identified thousands of pieces of buried hardware, as well as the volumes of liquid wastes in burial sites in the reactor areas. The author of this landmark study, Dr. Michele Gerber, will discuss historical research as a safe and cost-effective characterization tool.

  17. Users guide to the Argonne SP scheduling system

    SciTech Connect

    Lifka, D.A. |; Henderson, M.W.; Rayl, K.

    1995-05-01

    During the past five years scientists discovered that modern UNIX workstations connected with ethernet and fiber networks could provide enough computational performance to compete with the supercomputers of the day. As this concept became increasingly popular, the need for distributed queuing and scheduling systems became apparent. Today, supercomputers, such as Argonne National Laboratory`s IBM SP system, can provide more CPU and networking speed than can be obtained from these networks of workstations. These modern supercomputers look like clusters of workstations, however, so developers felt that the scheduling systems that were previously used on clusters of workstations should still apply. After trying to apply some of these scheduling systems to Argonne`s SP environment, it became obvious that these two computer environments have very different scheduling needs. Recognizing this need and realizing that no one has addressed it, we developed a new scheduling system. The approach taken in creating this system was unique in that user input and interaction were encouraged throughout the development process. Thus, a scheduler was built that actually worked the way the users wanted it to work. This document serves a dual purpose. It is both a user`s guide and an administrator`s guide for the ANL SP scheduling system. Look for revisions to this guide that will be appearing.

  18. Nuclear Fuel Traces Definition in Storage Ponds of Research VVR-2 and OR Reactors in NRC 'Kurchatov Institute'

    SciTech Connect

    Stepanov, Alexey; Simirskii, Iurii; Stepanov, Vyacheslav; Semin, Ilya; Volkovich, Anatoly

    2015-07-01

    The Gas Plant complex is the experimental base of the Institute of Nuclear Reactors, which is part of the Kurchatov Institute. In 1954 the commissioning of the first Soviet water-cooled water-moderated research reactor VVR-2 on enriched uranium, and until 1983 the complex operated two research water-cooled water-moderated reactors 3 MW (VVR-2) and 300 kW (OR) capacity, which were dismantled in connection with the overall upgrades of the complex. The complex has three storage ponds in the reactor building. They are sub-surface vessels filled with water (the volume of water in each is about 6 m{sup 3}). In 2007-2013 the spent nuclear fuel from storages was removed for processing to 'Mayk'. Survey of Storage Ponds by Underwater Collimated Spectrometric System shows a considerable layer of slime on the bottom of ponds and traces of spent nuclear fuel in one of the storage. For determination qualitative and the quantitative composition of radionuclide we made complex α-, β-, γ- spectrometric research of water and bottom slimes from Gas Plant complex storage ponds. We found the spent nuclear fuel in water and bottom slime in all storage ponds. Specific activity of radionuclides in the bottom slime exceeded specific activity of radionuclides in the ponds water and was closed to levels of high radioactive waste. Analysis of the obtained data and data from earlier investigation of reactor MR storage ponds showed distinctions of specific activity of uranium and plutonium radionuclides. (authors)

  19. Irradiation behavior of metallic fast reactor fuels

    SciTech Connect

    Pahl, R.G.; Porter, D.L.; Crawford, D.C.; Walters, L.C.

    1991-01-01

    Metallic fuels were the first fuels chosen for liquid metal cooled fast reactors (LMR's). In the late 1960's world-wide interest turned toward ceramic LMR fuels before the full potential of metallic fuel was realized. However, during the 1970's the performance limitations of metallic fuel were resolved in order to achieve a high plant factor at the Argonne National Laboratory's Experimental Breeder Reactor II. The 1980's spawned renewed interest in metallic fuel when the Integral Fast Reactor (IFR) concept emerged at Argonne National Laboratory. A fuel performance demonstration program was put into place to obtain the data needed for the eventual licensing of metallic fuel. This paper will summarize the results of the irradiation program carried out since 1985.

  20. Evolution of the liquid metal reactor: The Integral Fast Reactor (IFR) concept

    SciTech Connect

    Till, C.E.; Chang, Y.I.

    1989-01-01

    The Integral Fast Reactor (IFR) concept has been under development at Argonne National Laboratory since 1984. A key feature of the IFR concept is the metallic fuel. Metallic fuel was the original choice in early liquid metal reactor development. Solid technical accomplishments have been accumulating year after year in all aspects of the IFR development program. But as we make technical progress, the ultimate potential offered by the IFR concept as a next generation advanced reactor becomes clearer and clearer. The IFR concept can meet all three fundamental requirements needed in a next generation reactor. This document discusses these requirements: breeding, safety, and waste management. 5 refs., 4 figs.

  1. Residual stress diffractometer KOWARI at the Australian research reactor OPAL: Status of the project

    NASA Astrophysics Data System (ADS)

    Brule, Alain; Kirstein, Oliver

    2006-11-01

    Neutron scattering using diffraction techniques is now recognized as the most precise and reliable method of mapping sub-surface residual stresses in materials or even components, which are not only of academic but also of industrial-economic relevance. The great potential of neutrons in the field of residual stresses was recognized by ANSTO and its external Beam Instrument Advisory Group for the new research reactor OPAL. The recommendation was to build the dedicated strain scanner KOWARI among the first suite of instruments available to users. We give an update on the overall project and present the current status of the diffractometer. It is anticipated that the instrument will be commissioned in mid 2006 and available to users at the end of the OPAL project.

  2. Isotope production target irradiation experience at the annular core research reactor

    SciTech Connect

    Talley, D.G.

    1997-02-01

    As a result of an Environmental Impact Statement (EIS) recently issued by the Department of Energy, Sandia National Laboratories (SNL) has been selected as the {open_quotes}most appropriate facility{close_quotes} for the production of {sup 99}Mo. The daughter product of {sup 99}Mo is {sup 99m}Tc, a radioisotope used in 36,000 medical procedures per day in the U.S.{close_quote} At SNL, the {sup 99}Mo would be created by the fission process in UO{sub 2} coated {open_quotes}targets{close_quotes} and chemically separated in the SNL Hot Cell Facility (HCF). SNL has recently completed the irradiation of five production targets at its Annular Core Research Reactor (ACRR). Following irradiation, four of the targets were chemically processed in the HCF using the Cintichem process.

  3. Fuel development activities of the US RERTR Program. [Reduced Enrichment Research and Test Reactor

    SciTech Connect

    Snelgrove, J.L.; Domagala, R.F.; Wiencek, T.C.; Copeland, G.L.

    1983-01-01

    Progress in the development and irradiation testing of high-density fuels for use with low-enriched uranium in research and test reactors is reported. Swelling and blister-threshold temperature data obtained from the examination of miniature fuel plates containing UAl/sub x/, U/sub 3/O/sub 8/, U/sub 3/Si/sub 2/, or U/sub 3/Si dispersed in an aluminum matrix are presented. Combined with the results of metallurgical examinations, these data show that these four fuel types will perform adequately to full burnup of the /sup 235/U contained in the low-enriched fuel. The exothermic reaction of the uranium-silicide fuels with aluminum has been found to occur at about the same temperature as the melting of the aluminum matrix and cladding and to be essentially quenched by the melting endotherm. A new series of miniature fuel plate irradiations is also discussed.

  4. Neutronic safety and transient analyses for potential LEU conversion of the IR-8 research reactor.

    SciTech Connect

    Deen, J. R.; Hanan, N. A.; Smith, R. S.; Matos, J. E.; Egorenkov, P. M.; Nasonov, V. A.

    1999-09-27

    Kinetic parameters, isothermal reactivity feedback coefficients and three transients for the IR-8 research reactor cores loaded with either HEU(90%), HEU(36%), or LEU (19.75%) fuel assemblies (FA) were calculated using three dimensional diffusion theory flux solutions, RELAP5/MOD3.2 and PARET. The prompt neutron generation time and effective delayed neutron fractions were calculated for fresh and beginning-of-equilibrium-cycle cores. Isothermal reactivity feedback coefficients were calculated for changes in coolant density, coolant temperature and fuel temperature in fresh and equilibrium cores. These kinetic parameters and reactivity coefficients were used in transient analysis models to predict power histories, and peak fuel, clad and coolant temperatures. The transients modeled were a rapid and slow loss-of-flow, a slow reactivity insertion, and a fast reactivity insertion.

  5. Report on the workshop "Decay spectroscopy at CARIBU: advanced fuel cycle applications, nuclear structure and astrophysics". 14-16 April 2011, Argonne National Laboratory, USA.

    SciTech Connect

    Kondev, F.; Carpenter, M.P.; Chowdhury, P.; Clark, J.A.; Lister, C.J.; Nichols, A.L.; Swewryniak, D.

    2011-10-06

    A workshop on 'Decay Spectroscopy at CARIBU: Advanced Fuel Cycle Applications, Nuclear Structure and Astrophysics' will be held at Argonne National Laboratory on April 14-16, 2011. The aim of the workshop is to discuss opportunities for decay studies at the Californium Rare Isotope Breeder Upgrade (CARIBU) of the ATLAS facility with emphasis on advanced fuel cycle (AFC) applications, nuclear structure and astrophysics research. The workshop will consist of review and contributed talks. Presentations by members of the local groups, outlining the status of relevant in-house projects and availabile equipment, will also be organized. time will also be set aside to discuss and develop working collaborations for future decay studies at CARIBU. Topics of interest include: (1) Decay data of relevance to AFC applications with emphasis on reactor decay heat; (2) Discrete high-resolution gamma-ray spectroscopy following radioactive decya and related topics; (3) Calorimetric studies of neutron-rich fission framgents using Total ABsorption Gamma-Ray Spectrometry (TAGS) technique; (4) Beta-delayed neutron emissions and related topics; and (5) Decay data needs for nuclear astrophysics.

  6. Past and Future Work on Radiobiology Mega-Studies: A Case Study At Argonne National Laboratory

    SciTech Connect

    Haley, Benjamin; Wang, Qiong; Wanzer, Beau; Vogt, Stefan; Finney, Lydia; Yang, Ping Liu; Paunesku, Tatjana; Woloschak, Gayle

    2011-09-06

    Between 1952 and 1992, more than 200 large radiobiology studies were conducted in research institutes throughout Europe, North America, and Japan to determine the effects of external irradiation and internal emitters on the lifespan and tissue toxicity development in animals. At Argonne National Laboratory, 22 external beam studies were conducted on nearly 700 beagle dogs and 50,000 mice between 1969 and 1992. These studies helped to characterize the effects of neutron and gamma irradiation on lifespan, tumorigenesis, and mutagenesis across a range of doses and dosing patterns. The records and tissues collected at Argonne during that time period have been carefully preserved and redisseminated. Using these archived data, ongoing statistical work has been done and continues to characterize quality of radiation, dose, dose rate, tissue, and gender-specific differences in the radiation responses of exposed animals. The ongoing application of newly-developed molecular biology techniques to the archived tissues has revealed gene-specific mutation rates following exposure to ionizing irradiation. The original and ongoing work with this tissue archive is presented here as a case study of a more general trend in the radiobiology megastudies. These experiments helped form the modern understanding of radiation responses in animals and continue to inform development of new radiation models. Recent archival efforts have facilitated open access to the data and materials produced by these studies, and so a unique opportunity exists to expand this continued research.

  7. Argonne simulation framework for intelligent transportation systems

    SciTech Connect

    Ewing, T.; Doss, E.; Hanebutte, U.; Canfield, T.; Brown-VanHoozer, A.; Tentner, A.

    1996-04-01

    A simulation framework has been developed which defines a high-level architecture for a large-scale, comprehensive, scalable simulation of an Intelligent Transportation System (ITS). The simulator is designed to run on parallel computers and distributed (networked) computer systems; however, a version for a stand alone workstation is also available. The ITS simulator includes an Expert Driver Model (EDM) of instrumented ``smart`` vehicles with in-vehicle navigation units. The EDM is capable of performing optimal route planning and communicating with Traffic Management Centers (TMC). A dynamic road map data base is sued for optimum route planning, where the data is updated periodically to reflect any changes in road or weather conditions. The TMC has probe vehicle tracking capabilities (display position and attributes of instrumented vehicles), and can provide 2-way interaction with traffic to provide advisories and link times. Both the in-vehicle navigation module and the TMC feature detailed graphical user interfaces that includes human-factors studies to support safety and operational research. Realistic modeling of variations of the posted driving speed are based on human factor studies that take into consideration weather, road conditions, driver`s personality and behavior and vehicle type. The simulator has been developed on a distributed system of networked UNIX computers, but is designed to run on ANL`s IBM SP-X parallel computer system for large scale problems. A novel feature of the developed simulator is that vehicles will be represented by autonomous computer processes, each with a behavior model which performs independent route selection and reacts to external traffic events much like real vehicles. Vehicle processes interact with each other and with ITS components by exchanging messages. With this approach, one will be able to take advantage of emerging massively parallel processor (MPP) systems.

  8. The development and application of k0-standardization method of neutron activation analysis at Es-Salam research reactor

    NASA Astrophysics Data System (ADS)

    Alghem, L.; Ramdhane, M.; Khaled, S.; Akhal, T.

    2006-01-01

    In recent years the k0-NAA method has been applied and developed at the 15 MW Es-Salam research reactor, which includes: (1) the detection efficiency calibration of γ-spectrometer used in k0-NAA, (2) the determination of reactor neutron spectrum parameters such as α and f factors in the irradiation channel, and (3) the validation of the developed k0-NAA procedure by analysing SRM, namely AIEA-Soil7 and CRM, namely IGGE-GSV4. The analysis results obtained by k0-NAA with 27 elements of Soil-7 standard and 14 elements of GSV-4 standard were compared with certified values. The analysis results showed that the deviations between experimental and certified values were mostly less than 10%. The k0-NAA procedure established at Es-Salam research reactor has been regarded as a reliable standardization method of NAA and as available for practical applications.

  9. Very High Temperature Reactor (VHTR) Survey of Materials Research and Development Needs to Support Early Deployment

    SciTech Connect

    Eric Shaber; G. Baccaglini; S. Ball; T. Burchell; B. Corwin; T. Fewell; M. Labar; P. MacDonald; P. Rittenhouse; Russ Vollam; F. Southworth

    2003-01-01

    The VHTR reference concept is a helium-cooled, graphite moderated, thermal neutron spectrum reactor with an outlet temperature of 1000 C or higher. It is expected that the VHTR will be purchased in the future as either an electricity producing plant with a direct cycle gas turbine or a hydrogen producing (or other process heat application) plant. The process heat version of the VHTR will require that an intermediate heat exchanger (IHX) and primary gas circulator be located in an adjoining power conversion vessel. A third VHTR mission - actinide burning - can be accomplished with either the hydrogen-production or gas turbine designs. The first ''demonstration'' VHTR will produce both electricity and hydrogen using the IHX to transfer the heat to either a hydrogen production plant or the gas turbine. The plant size, reactor thermal power, and core configuration will be designed to assure passive decay heat removal without fuel damage during accidents. The fuel cycle will be a once-through very high burnup low-enriched uranium fuel cycle. The purpose of this report is to identify the materials research and development needs for the VHTR. To do this, we focused on the plant design described in Section 2, which is similar to the GT-MHR plant design (850 C core outlet temperature). For system or component designs that present significant material challenges (or far greater expense) there may be some viable design alternatives or options that can reduce development needs or allow use of available (cheaper) materials. Nevertheless, we were not able to assess those alternatives in the time allotted for this report and, to move forward with this material research and development assessment, the authors of this report felt that it was necessary to use a GT-MHR type design as the baseline design.

  10. Zero Power Reactor Simulation

    SciTech Connect

    2010-01-01

    Ever wanted to see a nuclear reactor core in action? Here's a detailed simulation of the Zero Power Reactor experiment, run by Argonne's unique "UNIC" code. Here, we use VisIt to visualize a numerical model of the ZPR-6 Assembly 6a experiment simulated using the Argonne UNIC code. 0:00-0:06: The fuel and other plates are dropped into an example drawer, and the drawer is inserted into the matrix tube. 0:06-0:21: The two matrix halves are brought together to make a critical (self-sustaining) assembly. 0:21-0:35: The fission power is revealed to be centralized in the thin, enriched Uranium plates. 0:35-0:53: Returning to our example drawer, we show the detailed local plate powers and their relation to the drawer composition. 0:53-1:11: Our model shows that each plate can have widely-varying local changes in the space and energy neutron densities. Read more at http://www.anl.gov/Media_Center/News/2010/news100121.html

  11. Dismantling Structures and Equipment of the MR Reactor and its Loop Facilities at the National Research Center 'Kurchatov Institute' - 12051

    SciTech Connect

    Volkov, V.G.; Danilovich, A.S.; Zverkov, Yu. A.; Ivanov, O.P.; Kolyadin, V.I.; Lemus, A.V.; Muzrukova, V.D.; Pavlenko, V.I.; Semenov, S.G.; Fadin, S.Yu.; Shisha, A.D.; Chesnokov, A.V.

    2012-07-01

    In 2008 a design of decommissioning of research reactors MR and RFT has been developed in the National research Center 'Kurchatov institute'. The design has been approved by Russian State Authority in July 2009 year and has received the positive conclusion of ecological expertise. In 2009-2010 a preparation for decommissioning of reactors MR and RFT was spent. Within the frames of a preparation a characterization, sorting and removal of radioactive objects, including the irradiated fuel, from reactor storage facilities and pool have been executed. During carrying out of a preparation on removal of radioactive objects from reactor sluice pool water treating has been spent. For these purposes modular installation for clearing and processing of a liquid radioactive waste 'Aqua - Express' was used. As a result of works it was possible to lower volume activity of water on three orders in magnitude that has allowed improving essentially of radiating conditions in a reactor hall. Auxiliary systems of ventilation, energy and heat supplies, monitoring systems of radiating conditions of premises of the reactor and its loop-back installations are reconstructed. In 2011 the license for a decommissioning of the specified reactors has been received and there are begun dismantling works. Within the frames of works under the design the armature and pipelines are dismantled in a under floor space of a reactor hall where a moving and taking away pipelines of loop facilities and the first contour of the MR reactor were replaced. A dismantle of the main equipment of loop facility with the gas coolant has been spent. Technologies which were used on dismantle of the radioactive contaminated equipment are presented, the basic works on reconstruction of systems of maintenance of on the decommissioning works are described, the sequence of works on the decommissioning of reactors MR and RFT is shown. Dismantling works were carried out with application of means of a dust suppression that, in

  12. AREVA Back-End Possibilities for the Used Fuel of Research Test Reactors

    SciTech Connect

    Auziere, P.; Emin, J.L.; Louvet, T.; Ohayon, D.; Hunter, I.

    2006-07-01

    One of the major issues faced by the Research and Test Reactor (RTR) operators is the back end management of the used fuel elements. RTR used fuel for both HEU and LEU types are problematic for storing and disposal as their Aluminium cladding degrades leading to activity release, possible loss of containment and criticality concerns. Thus, direct disposal of RTR used fuel, (without prior treatment and conditioning) is in this respect hardly suitable. In the same manner, long term interim storage of RTR used fuel has to take into account the issue of fuel corrosion. Treating RTR used fuel allows separating the content into recyclable materials and residues. It offers many advantages as compared to direct disposal such as the retrieval of valuable fissile material, the reduction of radio-toxicity and a very significant reduction of the volume of the ultimate waste package (reduction factor between 30 and 50). In addition, the vitrification of the residues provides a package that has been specifically designed to ensure long term durability for long term interim storage as well as final disposal (99% of the activity is encapsulated into a stable matrix). RTR fuel treatment process was developed several decades ago by AREVA with now thirty years of experience at an industrial level. The treatment process consists in dissolving the whole assembly (including the Al cladding) in nitric acid and then diluting it with standard Uranium Oxide fuel dissolution liquor prior to treatment with the nominal Tributylphosphate solvent extraction process. A wide range of RTR spent fuel has already been treated in the AREVA facilities. First, at the Marcoule plant over 18 tons of U-Al type RTR fuel from 21 reactors in 11 countries was processed. The treatment activities are now undertaken at the La Hague plant where 17 tons of RTR used fuel from Australia Belgium, and France aligned for treatment. In June 2005, AREVA started to treat at La Hague ANSTO's Australian RTR used fuel from

  13. Proc. of the sixteenth symposium on energy engineering sciences, May 13-15, 1998, Argonne, IL.

    SciTech Connect

    1998-05-13

    This Proceedings Volume includes the technical papers that were presented during the Sixteenth Symposium on Energy Engineering Sciences on May 13--15, 1998, at Argonne National Laboratory, Argonne, Illinois. The Symposium was structured into eight technical sessions, which included 30 individual presentations followed by discussion and interaction with the audience. A list of participants is appended to this volume. The DOE Office of Basic Energy Sciences (BES), of which Engineering Research is a component program, is responsible for the long-term, mission-oriented research in the Department. The Office has prime responsibility for establishing the basic scientific foundation upon which the Nation's future energy options will be identified, developed, and built. BES is committed to the generation of new knowledge necessary to solve present and future problems regarding energy exploration, production, conversion, and utilization, while maintaining respect for the environment. Consistent with the DOE/BES mission, the Engineering Research Program is charged with the identification, initiation, and management of fundamental research on broad, generic topics addressing energy-related engineering problems. Its stated goals are to improve and extend the body of knowledge underlying current engineering practice so as to create new options for enhancing energy savings and production, prolonging the useful life of energy-related structures and equipment, and developing advanced manufacturing technologies and materials processing. The program emphasis is on reducing costs through improved industrial production and performance and expanding the nation's store of fundamental knowledge for solving anticipated and unforeseen engineering problems in energy technologies. To achieve these goals, the Engineering Research Program supports approximately 130 research projects covering a broad spectrum of topics that cut across traditional engineering disciplines. The program focuses on

  14. Nuclear Research Reactor IEA-R1 - A Study of the Preparing for Decommissioning

    SciTech Connect

    Lopes, Valdir Maciel; Filho, Tufic Madi; Ricci, Walter; Martins, Mauro Onofre; Pereira, Maria da Conceicao

    2015-07-01

    The Institute of Energy and Nuclear Research (IPEN), Sao Paulo, according to the assignments given by the National Commission of Nuclear Energy (CNEN), enabled the development of this study, especially operational reports about refurbishing carried out on 2013, involving the production of radioisotopes and research in the areas of Radiochemistry and Nuclear Physics. These reports are made in accordance with established standard procedures to meet the requirements of CNEN (National Nuclear Energy Commission, the regulator the nuclear area activities in Brazil) and IAEA (International Atomic Energy Agency). This study presents an assessment of the procedures and methods of treatments for decontamination of the refrigeration primary circuit and changes parts, equipment and tubes of the of the IEA-R1 nuclear research reactor, pool type, power between 3,5 and 4,5 MW. In order to have a sequence in the work, the well-known contaminant radioisotopes were evaluated firstly, using Geiger- Muller equipment. In the second phase, the decontamination was done manually together with the ultrasound cleaning and washing equipment. From the several water solutions of citric acid assessment, the concentration with better confidence was obtained; in order to achieve the best results for decontamination. This study intends to define the best process for decontamination with low taxes of waste and without expensive costs. (authors)

  15. Substantiation of parameters of the geometric model of the research reactor core for the calculation using the Monte Carlo method

    NASA Astrophysics Data System (ADS)

    Radaev, A. I.; Schurovskaya, M. V.

    2015-12-01

    The choice of the spatial nodalization for the calculation of the power density and burnup distribution in a research reactor core with fuel assemblies of the IRT-3M and VVR-KN type using the program based on the Monte Carlo code is described. The influence of the spatial nodalization on the results of calculating basic neutronic characteristics and calculation time is investigated.

  16. Substantiation of parameters of the geometric model of the research reactor core for the calculation using the Monte Carlo method

    SciTech Connect

    Radaev, A. I. Schurovskaya, M. V.

    2015-12-15

    The choice of the spatial nodalization for the calculation of the power density and burnup distribution in a research reactor core with fuel assemblies of the IRT-3M and VVR-KN type using the program based on the Monte Carlo code is described. The influence of the spatial nodalization on the results of calculating basic neutronic characteristics and calculation time is investigated.

  17. Analysis of subcritical experiments using fresh and spent research reactor fuel assemblies

    NASA Astrophysics Data System (ADS)

    Zino, John Frederick

    1999-11-01

    This research investigated the concepts associated with crediting the burnup of spent nuclear fuel assemblies for the purposes of criticality safety. To accomplish this, a collaborative experimental research program was undertaken between Westinghouse, the University of Missouri Research Reactor (MURR) facility and Oak Ridge National Laboratory (ORNL). The purpose of the program was to characterize the subcritical behavior of a small array of fresh and spent MURR fuel assemblies using the 252Cf Source-driven noise technique. An aluminum test rig was built which was capable of holding up to four, highly enriched (93.15 wt.% 235U) MURR fuel assemblies in a 2 x 2 array. The rig was outfitted with one source and four detector drywells which allowed researchers to perform active neutron noise measurements on the array of fuel assemblies. The 1 atmosphere gas 3He neutron detectors used to perform the measurements were quenched with CF4 gas to allow improved discrimination of the neutron signals in the very high gamma-ray fields associated with spent fuel (˜8000 R/hr). In addition, the detector drywells were outfitted with 1″ lead collars to provide additional gamma-ray shielding from the spent fuel. Reactivity changes were induced in the subcritical lattice by replacing individual fresh assemblies (in a 4-assembly array) with spent assemblies of known, maximum burnup (143 Mw-D). The absolute and relative measured reactivity changes were then compared to those predicted by three-dimensional Monte Carlo calculations. The purpose of these comparisons was to investigate the accuracy of modern transport theory depletion calculations to accurately simulate the reactivity effects of burnup in spent nuclear fuel. A total of seven subcritical measurements were performed at the MURR reactor facility on July 20th and 27th, 1998. These measurements generated several estimates of prompt neutron decay constants (alpha) and ratios of spectral densities through frequency correlations

  18. Integral fast reactor concept inherent safety features

    SciTech Connect

    Marchaterre, J.F.; Sevy, R.H.; Cahalan, J.E.

    1987-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFT development effort are improved economics and enhanced safety. The design features that together fulfill these goals are: 1) a liquid metal (sodium) coolant, 2) a pool-type reactor primary system configuration, 3) an advanced ternary alloy metallic fuel, and 4) an integral fuel cycle. This paper reviews the design features that contribute to the safety margins inherent to the IFR concept. Special emphasis is placed on the ability of the IFR design to accommodate anticipated transients without scram (ATWS).

  19. Integral Fast Reactor concept inherent safety features

    SciTech Connect

    Marchaterre, J.F.; Sevy, R.H.; Cahalan, J.E.

    1986-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFR development effort are improved economics and enhanced safety. The design features that together fulfill these goals are: (1) a liquid metal (sodium) coolant, (2) a pool-type reactor primary system configuration, (3) an advanced ternary alloy metallic fuel, and (4) an integral fuel cycle. This paper reviews the design features that contribute to the safety margins inherent to the IFR concept. Special emphasis is placed on the ability of the IFR design to accommodate anticipated transients without scram (ATWS).

  20. Training courses on neutron detection systems on the ISIS research reactor: on-site and through internet training

    SciTech Connect

    Lescop, B.; Badeau, G.; Ivanovic, S.; Foulon, F.

    2015-07-01

    Today, ISIS research reactor is an essential tool for Education and Training programs organized by the National Institute for Nuclear Science and Technology (INSTN) from CEA. In the field of nuclear instrumentation, the INSTN offers both, theoretical courses and training courses on the use of neutron detection systems taking advantage of the ISIS research reactor for the supply of a wide range of neutron fluxes. This paper describes the content of the training carried out on the use of neutron detectors and detection systems, on-site or remote. The ISIS reactor is a 700 kW open core pool type reactor. The facility is very flexible since neutron detectors can be inserted into the core or its vicinity, and be used at different levels of power according to the needs of the course. Neutron fluxes, typically ranging from 1 to 10{sup 12} n/cm{sup 2}.s, can be obtained for the characterisation of the neutron detectors and detection systems. For the monitoring of the neutron density at low level of power, the Instrumentation and Control (I and C) system of the reactor is equipped with two detection systems, named BN1 and BN2. Each way contains a fission chamber, type CFUL01, connected to an electronic system type SIREX.The system works in pulse mode and exhibits two outputs: the counting rate and the doubling time. For the high level of power, the I and C is equipped with two detection systems HN1 and HN2.Each way contain a boron ionization chamber (type CC52) connected to an electronics system type SIREX. The system works in current mode and has two outputs: the current and the doubling time. For each mode, the trainees can observe and measure the signal at the different stages of the electronic system, with an oscilloscope. They can understand the role of each component of the detection system: detector, cable and each electronic block. The limitation of the detection modes and their operating range can be established from the measured signal. The trainees can also

  1. Criticality safety strategy for the Fuel Cycle Facility electrorefiner at Argonne National Laboratory, West

    SciTech Connect

    Mariani, R.D.; Benedict, R.W.; Lell, R.M.; Turski, R.B.; Fujita, E.K.

    1993-09-01

    The Integral Fast Reactor being developed by Argonne National Laboratory (ANL) combines the advantages of metal-fueled, liquid-metal-cooled reactors and a closed fuel cycle. Presently, the Fuel Cycle Facility (FCF) at ANL-West in Idaho Falls, Idaho is being modified to recycle spent metallic fuel from Experimental Breeder Reactor II as part of a demonstration project sponsored by the Department of Energy. A key component of the FCF is the electrorefiner (ER) in which the actinides are separated from the fission products. In the electrorefining process, the metal fuel is anodically dissolved into a high-temperature molten salt and refined uranium or uranium/plutonium products are deposited at cathodes. In this report, the criticality safety strategy for the FCF ER is summarized. FCF ER operations and processes formed the basis for evaluating criticality safety and control during actinide metal fuel refining. In order to show criticality safety for the FCF ER, the reference operating conditions for the ER had to be defined. Normal operating envelopes (NOES) were then defined to bracket the important operating conditions. To keep the operating conditions within their NOES, process controls were identified that can be used to regulate the actinide forms and content within the ER. A series of operational checks were developed for each operation that wig verify the extent or success of an operation. The criticality analysis considered the ER operating conditions at their NOE values as the point of departure for credible and incredible failure modes. As a result of the analysis, FCF ER operations were found to be safe with respect to criticality.

  2. Neutron Environment Characterization of the Central Cavity in the Annular Core Research Reactor

    NASA Astrophysics Data System (ADS)

    Parma, Edward J.; Naranjo, Gerald E.; Lippert, Lance L.; Vehar, David W.

    2016-02-01

    Characterization of the neutron environment in the central cavity of the Sandia National Laboratories' Annular Core Research Reactor (ACRR) is important in order to provide experimenters with the most accurate spectral information and maintain a high degree of fidelity in performing reactor experiments. Characterization includes both modeling and experimental efforts. Building accurate neutronic models of the ACRR and the central cavity "bucket" environments that can be used by experimenters is important in planning and designing experiments, as well as assessing the experimental results and quantifying uncertainties. Neutron fluence characterizations of two bucket environments, LB44 and PLG, are presented. These two environments are used frequently and represent two extremes in the neutron spectrum. The LB44 bucket is designed to remove the thermal component of the neutron spectrum and significantly attenuate the gamma-ray fluence. The PLG bucket is designed to enhance the thermal component of the neutron spectrum and attenuate the gamma-ray fluence. The neutron characterization for each bucket was performed by irradiating 20 different activation foil types, some of which were cadmium covered, resulting in 37 different reactions at the peak axial flux location in each bucket. The dosimetry results were used in the LSL-M2 spectrum adjustment code with a 640-energy group MCNP-generated trial spectrum, self-shielding correction factors, the SNLRML or IRDFF dosimetry cross-section library, trial spectrum uncertainty, and trial covariance matrix, to generate a least-squares adjusted neutron spectrum, spectrum uncertainty, and covariance matrix. Both environment character-izations are well documented and the environments are available for use by experimenters. Work supported by the United States Department of Energy at Sandia National Laboratories. Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned

  3. Deployment of Smart 3D Subsurface Contaminant Characterization at the Brookhaven Graphite Research Reactor

    SciTech Connect

    Sullivan, T.; Heiser, J.; Kalb, P.; Milian, L.; Newson, C.; Lilimpakas, M.; Daniels, T.

    2002-02-26

    The Brookhaven Graphite Research Reactor (BGRR) Historical Site Assessment (BNL 1999) identified contamination inside the Below Grade Ducts (BGD) resulting from the deposition of fission and activation products from the pile on the inner carbon steel liner during reactor operations. Due to partial flooding of the BGD since shutdown, some of this contamination may have leaked out of the ducts into the surrounding soils. The baseline remediation plan for cleanup of contaminated soils beneath the BGD involves complete removal of the ducts, followed by surveying the underlying and surrounding soils, then removing soil that has been contaminated above cleanup goals. Alternatively, if soil contamination around and beneath the BGD is either non-existent/minimal (below cleanup goals) or is very localized and can be ''surgically removed'' at a reasonable cost, the BGD can be decontaminated and left in place. The focus of this Department of Energy Accelerated Site Technology Deployment (DOE ASTD) project was to determine the extent (location, type, and level) of soil contamination surrounding the BGD and to present this data to the stakeholders as part of the Engineering Evaluation/Cost Analysis (EE/CA) process. A suite of innovative characterization tools was used to complete the characterization of the soil surrounding the BGD in a cost-effective and timely fashion and in a manner acceptable to the stakeholders. The tools consisted of a tracer gas leak detection system that was used to define the gaseous leak paths out of the BGD and guide soil characterization studies, a small-footprint Geoprobe to reach areas surrounding the BGD that were difficult to access, two novel, field-deployed, radiological analysis systems (ISOCS and BetaScint) and a three-dimensional (3D) visualization system to facilitate data analysis/interpretation. All of the technologies performed as well or better than expected and the characterization could not have been completed in the same time or at

  4. Interim status report on lead-cooled fast reactor (LFR) research and development.

    SciTech Connect

    Tzanos, C. P.; Sienicki, J. J.; Moisseytsev, A.; Smith, C. F.; de Caro, M.; Halsey, W. G.; Li, N.; Hosemann, P.; Zhang, J.; Bolind, A.; LLNL; LANL; Univ. of Illinois

    2008-03-31

    This report discusses the status of Lead-Cooled Fast Reactor (LFR) research and development carried out during the first half of FY 2008 under the U.S. Department of Energy Generation IV Nuclear Energy Systems Initiative. Lead-Cooled Fast Reactor research and development has recently been transferred from Generation IV to the Reactor Campaign of the Global Nuclear Energy Partnership (GNEP). Another status report shall be issued at the end of FY 2008 covering all of the LFR activities carried out in FY 2008 for both Generation IV and GNEP. The focus of research and development in FY 2008 is an initial investigation of a concept for a LFR Advanced Recycling Reactor (ARR) Technology Pilot Plant (TPP)/demonstration test reactor (demo) incorporating features and operating conditions of the European Lead-cooled SYstem (ELSY) {approx} 600 MWe lead (Pb)-cooled LFR preconceptual design for the transmutation of waste and central station power generation, and which would enable irradiation testing of advanced fuels and structural materials. Initial scoping core concept development analyses have been carried out for a 100 MWt core composed of sixteen open-lattice 20 by 20 fuel assemblies largely similar to those of the ELSY preconceptual fuel assembly design incorporating fuel pins with mixed oxide (MOX) fuel, central control rods in each fuel assembly, and cooled with Pb coolant. For a cycle length of three years, the core is calculated to have a conversion ratio of 0.79, an average discharge burnup of 108 MWd/kg of heavy metal, and a burnup reactivity swing of about 13 dollars. With a control rod in each fuel assembly, the reactivity worth of an individual rod would need to be significantly greater than one dollar which is undesirable for postulated rod withdrawal reactivity insertion events. A peak neutron fast flux of 2.0 x 10{sup 15} (n/cm{sup 2}-s) is calculated. For comparison, the 400 MWt Fast Flux Test Facility (FFTF) achieved a peak neutron fast flux of 7.2 x 10{sup

  5. Advance Liquid Metal Reactor Discrete Dynamic Event Tree/Bayesian Network Analysis and Incident Management Guidelines (Risk Management for Sodium Fast Reactors)

    SciTech Connect

    Denman, Matthew R.; Groth, Katrina M.; Cardoni, Jeffrey N.; Wheeler, Timothy A.

    2015-04-01

    Accident management is an important component to maintaining risk at acceptable levels for all complex systems, such as nuclear power plants. With the introduction of self-correcting, or inherently safe, reactor designs the focus has shifted from management by operators to allowing the system's design to manage the accident. Inherently and passively safe designs are laudable, but nonetheless extreme boundary conditions can interfere with the design attributes which facilitate inherent safety, thus resulting in unanticipated and undesirable end states. This report examines an inherently safe and small sodium fast reactor experiencing a beyond design basis seismic event with the intend of exploring two issues : (1) can human intervention either improve or worsen the potential end states and (2) can a Bayesian Network be constructed to infer the state of the reactor to inform (1). ACKNOWLEDGEMENTS The authors would like to acknowledge the U.S. Department of Energy's Office of Nuclear Energy for funding this research through Work Package SR-14SN100303 under the Advanced Reactor Concepts program. The authors also acknowledge the PRA teams at Argonne National Laboratory, Oak Ridge National Laboratory, and Idaho National Laboratory for their continue d contributions to the advanced reactor PRA mission area.

  6. Petascale algorithms for reactor hydrodynamics.

    SciTech Connect

    Fischer, P.; Lottes, J.; Pointer, W. D.; Siegel, A.

    2008-01-01

    We describe recent algorithmic developments that have enabled large eddy simulations of reactor flows on up to P = 65, 000 processors on the IBM BG/P at the Argonne Leadership Computing Facility. Petascale computing is expected to play a pivotal role in the design and analysis of next-generation nuclear reactors. Argonne's SHARP project is focused on advanced reactor simulation, with a current emphasis on modeling coupled neutronics and thermal-hydraulics (TH). The TH modeling comprises a hierarchy of computational fluid dynamics approaches ranging from detailed turbulence computations, using DNS (direct numerical simulation) and LES (large eddy simulation), to full core analysis based on RANS (Reynolds-averaged Navier-Stokes) and subchannel models. Our initial study is focused on LES of sodium-cooled fast reactor cores. The aim is to leverage petascale platforms at DOE's Leadership Computing Facilities (LCFs) to provide detailed information about heat transfer within the core and to provide baseline data for less expensive RANS and subchannel models.

  7. Argonne's Laboratory computing resource center : 2006 annual report.

    SciTech Connect

    Bair, R. B.; Kaushik, D. K.; Riley, K. R.; Valdes, J. V.; Drugan, C. D.; Pieper, G. P.

    2007-05-31

    Argonne National Laboratory founded the Laboratory Computing Resource Center (LCRC) in the spring of 2002 to help meet pressing program needs for computational modeling, simulation, and analysis. The guiding mission is to provide critical computing resources that accelerate the development of high-performance computing expertise, applications, and computations to meet the Laboratory's challenging science and engineering missions. In September 2002 the LCRC deployed a 350-node computing cluster from Linux NetworX to address Laboratory needs for mid-range supercomputing. This cluster, named 'Jazz', achieved over a teraflop of computing power (10{sup 12} floating-point calculations per second) on standard tests, making it the Laboratory's first terascale computing system and one of the 50 fastest computers in the world at the time. Jazz was made available to early users in November 2002 while the system was undergoing development and configuration. In April 2003, Jazz was officially made available for production operation. Since then, the Jazz user community has grown steadily. By the end of fiscal year 2006, there were 76 active projects on Jazz involving over 380 scientists and engineers. These projects represent a wide cross-section of Laboratory expertise, including work in biosciences, chemistry, climate, computer science, engineering applications, environmental science, geoscience, information science, materials science, mathematics, nanoscience, nuclear engineering, and physics. Most important, many projects have achieved results that would have been unobtainable without such a computing resource. The LCRC continues to foster growth in the computational science and engineering capability and quality at the Laboratory. Specific goals include expansion of the use of Jazz to new disciplines and Laboratory initiatives, teaming with Laboratory infrastructure providers to offer more scientific data management capabilities, expanding Argonne staff use of national

  8. Optimization of irradiation conditions for {sup 177}Lu production at the LVR-15 research reactor

    SciTech Connect

    Lahodova, Z.; Viererbl, L.; Klupak, V.; Srank, J.

    2012-07-01

    The use of lutetium in medicine has been increasing over the last few years. The {sup 177}Lu radionuclide is commercially available for research and test purposes as a diagnostic and radiotherapy agent in the treatment of several malignant tumours. The yield of {sup 177}Lu from the {sup 176}Lu(n,{gamma}){sup 177}Lu nuclear reaction depends significantly on the thermal neutron fluence rate. The capture cross-sections of both reaction {sup 176}Lu(n,{gamma}){sup 177}Lu and reaction {sup 177}Lu(n,{gamma}){sup 178}Lu are very high. Therefore a burn-up of target and product nuclides should be taken into account when calculating {sup 177}Lu activity. The maximum irradiation time, when the activity of the {sup 177}Lu radionuclide begins to decline, was found for different fluence rates. Two vertical irradiation channels at the LVR-15 nuclear research reactor were compared in order to choose the channel with better irradiation conditions, such as a higher thermal neutron fluence rate in the irradiation volume. In this experiment, lutetium was irradiated in a titanium capsule. The influence of the Ti capsule on the neutron spectrum was monitored using activation detectors. The choice of detectors was based on requirements for irradiation time and accurate determination of thermal neutrons. The following activation detectors were selected for measurement of the neutron spectrum: Ti, Fe, Ni, Co, Ag and W. (authors)

  9. International Atomic Energy Agency (IAEA) Coordinated Research Projects on Structural Integrity of Reactor Pressure Vessels

    SciTech Connect

    Server, W. L.; Nanstad, Randy K

    2009-01-01

    The International Atomic Energy Agency (IAEA) has conducted a series of Coordinated Research Projects (CRPs) that have focused on irradiated reactor pressure vessel (RPV) steel fracture toughness properties and approaches for assuring structural integrity of RPVs throughout operating life. A series of nine CRPs have been sponsored by the IAEA, starting in the early 1970s, focused on neutron radiation effects on RPV steels. The purpose of the CRPs was to develop comparisons and correlations to test the uniformity of irradiated results through coordinated international research studies and data sharing. Consideration of dose rate effects, effects of alloying (nickel, manganese, silicon, etc.) and residual elements (eg., copper and phosphorus), and drop in upper shelf toughness are also important for assessing neutron embrittlement effects. The ultimate use of embrittlement understanding is assuring structural integrity of the RPV under current and future operation and accident conditions. Material fracture toughness is the key ingredient needed for this assessment, and many of the CRPs have focused on measurement and application of irradiated fracture toughness. This paper presents an overview of the progress made since the inception of the CRPs in the early 1970s. The chronology and importance of each CRP have been reviewed and put into context for continued and long-term safe operation of RPVs.

  10. Neutronic and thermal-hydraulic analysis of new irradiation channels inside the Moroccan TRIGA Mark II research reactor core.

    PubMed

    Chham, E; El Bardouni, T; Benaalilou, K; Boukhal, H; El Bakkari, B; Boulaich, Y; El Younoussi, C; Nacir, B

    2016-10-01

    This study was conducted to improve the capacity of radioisotope production in the Moroccan TRIGA Mark II research reactor, which is considered as one of the most important applications of research reactors. The aim of this study is to enhance the utilization of TRIGA core in the field of neutron activation and ensure an economic use of the fuel. The main idea was to create an additional irradiation channel (IC) inside the core. For this purpose, three new core configurations are proposed, which differ according to the IC position in the core. Thermal neutron flux distribution and other neutronic safety parameters such as power peaking factors, excess reactivity, and control rods worth reactivity were calculated using the Monte Carlo N-Particle Transport (MCNP) code and neutron cross-section library based on ENDF/B-VII evaluation. The calculated thermal flux in the central thimble (CT) and in the added IC for the reconfigured core is compared with the thermal flux in the CT of the existing core, which is taken as a reference. The results show that all the obtained fluxes in CTs are very close to the reference value, while a remarkable difference is observed between the fluxes in the new ICs and reference. This difference depends on the position of IC in the reactor core. To demonstrate that the Moroccan TRIGA reactor could safely operate at 2MW, with new configurations based on new ICs, different safety-related thermal-hydraulic parameters were investigated. The PARET model was used in this study to verify whether the safety margins are met despite the new modifications of the core. The results show that it is possible to introduce new ICs safely in the reactor core, because the obtained values of the parameters are largely far from compromising the safety of the reactor.

  11. 1985 annual site environmental report for Argonne National Laboratory

    SciTech Connect

    Golchert, N.W.; Duffy, T.L.; Sedlet, J.

    1986-03-01

    This is one in a series of annual reports prepared to provide DOE, environmental agencies, and the public with information on the level of radioactive and chemical pollutants in the environment and on the amounts of such substances, if any, added to the environment as a result of Argonne operations. Included in this report are the results of measurements obtained in 1985 for a number of radionuclides in air, surface water, ground water, soil, grass, bottom sediment, and milk; for a variety of chemical constituents in surface and subsurface water; and for the external penetrating radiation dose.

  12. Alkaloid-derived molecules in low rank Argonne premium coals.

    SciTech Connect

    Winans, R. E.; Tomczyk, N. A.; Hunt, J. E.

    2000-11-30

    Molecules that are probably derived from alkaloids have been found in the extracts of the subbituminous and lignite Argonne Premium Coals. High resolution mass spectrometry (HRMS) and liquid chromatography mass spectrometry (LCMS) have been used to characterize pyridine and supercritical extracts. The supercritical extraction used an approach that has been successful for extracting alkaloids from natural products. The first indication that there might be these natural products in coals was the large number of molecules found containing multiple nitrogen and oxygen heteroatoms. These molecules are much less abundant in bituminous coals and absent in the higher rank coals.

  13. Small Liquid Metal Cooled Reactor Safety Study

    SciTech Connect

    Minato, A; Ueda, N; Wade, D; Greenspan, E; Brown, N

    2005-11-02

    The Small Liquid Metal Cooled Reactor Safety Study documents results from activities conducted under Small Liquid Metal Fast Reactor Coordination Program (SLMFR-CP) Agreement, January 2004, between the Central Research Institute of the Electric Power Industry (CRIEPI) of Japan and the Lawrence Livermore National Laboratory (LLNL)[1]. Evaluations were completed on topics that are important to the safety of small sodium cooled and lead alloy cooled reactors. CRIEPI investigated approaches for evaluating postulated severe accidents using the CANIS computer code. The methods being developed are improvements on codes such as SAS 4A used in the US to analyze sodium cooled reactors and they depend on calibration using safety testing of metal fuel that has been completed in the TREAT facility. The 4S and the small lead cooled reactors in the US are being designed to preclude core disruption from all mechanistic scenarios, including selected unprotected transients. However, postulated core disruption is being evaluated to support the risk analysis. Argonne National Laboratory and the University of California Berkeley also supported LLNL with evaluation of cores with small positive void worth and core designs that would limit void worth. Assessments were also completed for lead cooled reactors in the following areas: (1) continuing operations with cladding failure, (2) large bubbles passing through the core and (3) recommendations concerning reflector control. The design approach used in the US emphasizes reducing the reactivity in the control mechanisms with core designs that have essentially no, or a very small, reactivity change over the core life. This leads to some positive void worth in the core that is not considered to be safety problem because of the inability to identify scenarios that would lead to voiding of lead. It is also believed that the void worth will not dominate the severe accident analysis. The approach used by 4S requires negative void worth throughout

  14. Technical aspects of boron neutron capture therapy at the BNL Medical Research Reactor

    SciTech Connect

    Holden, N.E.; Rorer, D.C.; Patti, F.J.; Liu, H.B.; Reciniello, R.; Chanana, A.D.

    1997-07-01

    The Brookhaven Medical Research Reactor, BMRR, is a 3 MW heterogeneous, tank-type, light water cooled and moderated, graphite reflected reactor, which was designed for biomedical studies. Early BNL work in Boron Neutron Capture Therapy (BNCT) used a beam of thermal neutrons for experimental treatment of brain tumors. Research elsewhere and at BNL indicated that higher energy neutrons would be required to treat deep seated brain tumors. Epithermal neutrons would be thermalized as they penetrated the brain and peak thermal neutron flux densities would occur at the depth of brain tumors. One of the two BMRR thermal port shutters was modified in 1988 to include plates of aluminum and aluminum oxide to provide an epithermal port. Lithium carbonate in polyethylene was added in 1991 around the bismuth port to reduce the neutron flux density coming from outside the port. To enhance the epithermal neutron flux density, the two vertical thimbles A-3 (core edge) and E-3 (in core) were replaced with fuel elements. There are now four fuel elements of 190 grams each and 28 fuel elements of 140 grams each for a total of 4.68 kg of {sup 235}U in the core. The authors have proposed replacing the epithermal shutter with a fission converter plate shutter. It is estimated that the new shutter would increase the epithermal neutron flux density by a factor of seven and the epithermal/fast neutron ratio by a factor of two. The modifications made to the BMRR in the past few years permit BNCT for brain tumors without the need to reflect scalp and bone flaps. Radiation workers are monitored via a TLD badge and a self-reading dosimeter during each experiment. An early concern was raised about whether workers would be subject to a significant dose rate from working with patients who have been irradiated. The gamma ray doses for the representative key personnel involved in the care of the first 12 patients receiving BNCT are listed. These workers did not receive unusually high exposures.

  15. INTERIM STORAGE AND LONG TERM DISPOSAL OF RESEARCH REACTOR SPENT FUEL

    SciTech Connect

    Vinson, D

    2006-08-22

    Aluminum clad research reactor spent nuclear fuel (SNF) is currently being consolidated in wet storage basins (pools). Approximately 20 metric tons (heavy metal) of aluminum-based spent nuclear fuel (Al-SNF) is being consolidated for treatment, packaging, interim storage, and preparation for ultimate disposal in a geologic repository. The storage and disposal of Al-SNF are subject to requirements that provide for safety and acceptable radionuclide release. The options studied for interim storage of SNF include wet storage and dry storage. Two options have also been studied to develop the technical basis for the qualification and repository disposal of aluminum spent fuel. The two options studied include Direct Disposal and Melt-Dilute treatment. The implementation of these options present relative benefits and challenges. Both the Direct Disposal and the Melt-Dilute treatment options have been developed and their technical viability assessed. Adaptation of the melt-dilute technology for the treatment of spent fuel offers the benefits of converting the spent fuel into a proliferation resistant form and/or significantly reducing the volume of the spent fuel. A Mobile Melt-Dilute system concept has emerged to realize these benefits and a prototype system developed. The application of the melt-dilute technology for the treatment of legacy nuclear materials has been evaluated and also offers the promise for the safe disposal of these materials.

  16. Lewis Research Center's coal-fired, pressurized, fluidized-bed reactor test facility

    NASA Technical Reports Server (NTRS)

    Kobak, J. A.; Rollbuhler, R. J.

    1981-01-01

    A 200-kilowatt-thermal, pressurized, fluidized-bed (PFB) reactor, research test facility was designed, constructed, and operated as part of a NASA-funded project to assess and evaluate the effect of PFB hot-gas effluent on aircraft turbine engine materials that might have applications in stationary-power-plant turbogenerators. Some of the techniques and components developed for this PFB system are described. One of the more important items was the development of a two-in-one, gas-solids separator that removed 95+ percent of the solids in 1600 F to 1900 F gases. Another was a coal and sorbent feed and mixing system for injecting the fuel into the pressurized combustor. Also important were the controls and data-acquisition systems that enabled one person to operate the entire facility. The solid, liquid, and gas sub-systems all had problems that were solved over the 2-year operating time of the facility, which culminated in a 400-hour, hot-gas, turbine test.

  17. Analysis of LOCA Scenarios in the NIST Research Reactor Before and After Fuel Conversion

    SciTech Connect

    Baek, J. S.; Cheng, L. Y.; Diamond, D.

    2015-08-30

    An analysis has been done of hypothetical loss-of-coolant-accidents (LOCAs) in the research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The purpose of the analysis is to determine if the peak clad temperature remains below the Safety Limit, which is the blister temperature for the fuel. The configuration of the NBSR considered in the analysis is that projected for the future when changes will be made so that shutdown pumps do not operate when a LOCA signal is detected. The analysis was done for the present core with high-enriched uranium (HEU) fuel and with the proposed low-enriched uranium (LEU) fuel that would be used when the NBSR is converted from one to the other. The analysis consists of two parts. The first examines how the water would drain from the primary system following a break and the possibility for the loss of coolant from within the fuel element flow channels. This work is performed using the TRACE system thermal-hydraulic code. The second looks at the fuel clad temperature as a function of time given that the water may have drained from many of the flow channels and the water in the vessel is in a quasi-equilibrium state. The temperature behavior is investigated using the three-dimensional heat conduction code HEATING7.3. The results in all scenarios considered for both HEU and LEU fuel show that the peak clad temperature remains below the blister temperature.

  18. Characterization of commercially pure aluminum powder for research reactor fuel plates

    SciTech Connect

    Downs, V.D.; Wiencek, T.C.

    1992-11-01

    Aluminum powder is used as the matrix material in the production of uranium aluminide, oxide, and silicide dispersion fuel plates for research and test reactors. variability in the characteristics of the aluminum powder, such as moisture content and particle-size distribution, influences blending and compacting of the aluminum/fuel powder. A detailed study was performed to characterize the physical properties of three aluminum powder lots. An angle-of-shear test was devised to characterize the cohesiveness of the aluminum powder. Flow-rate measurements, apparent density determination, subsieve analysis, surface area measurements, and scanning electron microscopy were also used in the study. It was found that because of the various types of commercially available powders, proper specification of powder variables will ensure the receipt of consistent raw materials. Improved control of the initial powder will reduce the variability of fuel-plate production and will improve overall plate reproducibility. It is recommended that a standard specification be written for the aluminum powder and silicide fuel.

  19. Characterization of commercially pure aluminum powder for research reactor fuel plates

    SciTech Connect

    Downs, V.D. ); Wiencek, T.C. )

    1992-01-01

    Aluminum powder is used as the matrix material in the production of uranium aluminide, oxide, and silicide dispersion fuel plates for research and test reactors. variability in the characteristics of the aluminum powder, such as moisture content and particle-size distribution, influences blending and compacting of the aluminum/fuel powder. A detailed study was performed to characterize the physical properties of three aluminum powder lots. An angle-of-shear test was devised to characterize the cohesiveness of the aluminum powder. Flow-rate measurements, apparent density determination, subsieve analysis, surface area measurements, and scanning electron microscopy were also used in the study. It was found that because of the various types of commercially available powders, proper specification of powder variables will ensure the receipt of consistent raw materials. Improved control of the initial powder will reduce the variability of fuel-plate production and will improve overall plate reproducibility. It is recommended that a standard specification be written for the aluminum powder and silicide fuel.

  20. Utilization of coal fly ash in solidification of liquid radioactive waste from research reactor.

    PubMed

    Osmanlioglu, Ahmet Erdal

    2014-05-01

    In this study, the potential utilization of fly ash was investigated as an additive in solidification process of radioactive waste sludge from research reactor. Coal formations include various percentages of natural radioactive elements; therefore, coal fly ash includes various levels of radioactivity. For this reason, fly ashes have to be evaluated for potential environmental implications in case of further usage in any construction material. But for use in solidification of radioactive sludge, the radiological effects of fly ash are in the range of radioactive waste management limits. The results show that fly ash has a strong fixing capacity for radioactive isotopes. Specimens with addition of 5-15% fly ash to concrete was observed to be sufficient to achieve the target compressive strength of 20 MPa required for near-surface disposal. An optimum mixture comprising 15% fly ash, 35% cement, and 50% radioactive waste sludge could provide the solidification required for long-term storage and disposal. The codisposal of radioactive fly ash with radioactive sludge by solidification decreases the usage of cement in solidification process. By this method, radioactive fly ash can become a valuable additive instead of industrial waste. This study supports the utilization of fly ash in industry and the solidification of radioactive waste in the nuclear industry.

  1. The Conceptual Design for a Fuel Assembly of a New Research Reactor

    SciTech Connect

    Ryu, J-S.; Cho, Y-G.; Yoon, D-B.; Dan, H-J.; Chae, H-T.; Park, C.

    2004-10-06

    A new Research Reactor (ARR) has been under design by KAERI since 2002. In this work, as a first step for the design of the fuel assembly of the ARR, the conceptual design has been carried out. The vibration characteristics of the tubular fuel model and the locking performance of the preliminary designed locking devices were investigated. In order to investigate the effects of the stiffener on the vibration characteristics of the tubular fuel, a modal analysis was performed for the finite element models of the tubular fuels with stiffeners and without stiffeners. The analysis results show that the vibration characteristics of the tubular fuel with stiffeners are better than those of the tubular fuel without stiffeners. To investigate the locking performance of the preliminary designed locking devices for the fuel assembly of the ARR, the elements of the locking devices were fabricated. Then the torsional resistance, fixing status and vibration characteristics of the locking devices were tested. The test results show that using the locking device with fins on the bottom guide can prevent the torsional motion of the fuel assembly, and that additional springs or guides on the top of the fuel assembly are needed to suppress the lateral motion of the fuel assembly. Based on the modal analysis and experimental results, the fuel assembly and locking devices of the ARR were designed and its prototype was fabricated. The locking performance, pressure drop characteristics and vibration characteristics of the newly designed fuel assembly will be tested in the near future.

  2. Lewis Research Center's coal-fired, pressurized, fluidized-bed reactor test facility

    NASA Astrophysics Data System (ADS)

    Kobak, J. A.; Rollbuhler, R. J.

    1981-10-01

    A 200-kilowatt-thermal, pressurized, fluidized-bed (PFB) reactor, research test facility was designed, constructed, and operated as part of a NASA-funded project to assess and evaluate the effect of PFB hot-gas effluent on aircraft turbine engine materials that might have applications in stationary-power-plant turbogenerators. Some of the techniques and components developed for this PFB system are described. One of the more important items was the development of a two-in-one, gas-solids separator that removed 95+ percent of the solids in 1600 F to 1900 F gases. Another was a coal and sorbent feed and mixing system for injecting the fuel into the pressurized combustor. Also important were the controls and data-acquisition systems that enabled one person to operate the entire facility. The solid, liquid, and gas sub-systems all had problems that were solved over the 2-year operating time of the facility, which culminated in a 400-hour, hot-gas, turbine test.

  3. High-density reduced-enrichment fuels for Research and Test Reactors

    SciTech Connect

    Snelgrove, J.L.; Hofman, G.L.; Copeland, G.L.

    1983-01-01

    Development and irradiation testing of high-density fuels have been conducted by the US RERTR Program in order to provide the technical means to reduce the enrichment of fuels for research and test reactors. The traditional aluminum dispersion fuel technology has been extended to include the highest practical loadings of uranium-aluminide (UAl/sub x/, 2.3 MgU/m/sup 3/), uranium-oxide (U/sub 3/O/sub 8/, 3.2 MgU/m/sup 3/), and uranium-silicide (U/sub 3/Si/sub 2/, 5.5 MgU/m/sup 3/; U/sub 3/Si, 7.0 MgU/m/sup 3/) fuels. A third uranium-silicide alloy, U/sub 3/SiAl (U + 3.5 wt % Si + 1.5 wt % Al) has been found to perform poorly at high burnup. Testing of miniature fuel plates and full-sized fuel elements is at an advanced stage for the highest loadings of the aluminide and oxide fuels and intermediate loadings of the silicide fuels, and good results have been obtained for low-enriched uranium. The data obtained to date are discussed. 1 reference, 3 figures, 1 table.

  4. A feasibility study of the Tehran research reactor as a neutron source for BNCT.

    PubMed

    Kasesaz, Yaser; Khalafi, Hossein; Rahmani, Faezeh; Ezati, Arsalan; Keyvani, Mehdi; Hossnirokh, Ashkan; Shamami, Mehrdad Azizi; Monshizadeh, Mahdi

    2014-08-01

    Investigation on the use of the Tehran Research Reactor (TRR) as a neutron source for Boron Neutron Capture Therapy (BNCT) has been performed by calculating and measuring energy spectrum and the spatial distribution of neutrons in all external irradiation facilities, including six beam tubes, thermal column, and the medical room. Activation methods with multiple foils and a copper wire have been used for the mentioned measurements. The results show that (1) the small diameter and long length beam tubes cannot provide sufficient neutron flux for BNCT; (2) in order to use the medical room, the TRR core should be placed in the open pool position, in this situation the distance between the core and patient position is about 400 cm, so neutron flux cannot be sufficient for BNCT; and (3) the best facility which can be adapted for BNCT application is the thermal column, if all graphite blocks can be removed. The epithermal and fast neutron flux at the beginning of this empty column are 4.12×10(9) and 1.21×10(9) n/cm(2)/s, respectively, which can provide an appropriate neutron beam for BNCT by designing and constructing a proper Beam Shaping Assembly (BSA) structure.

  5. Characterization of radioactive contaminants and water treatment trials for the Taiwan Research Reactor's spent fuel pool.

    PubMed

    Huang, Chun-Ping; Lin, Tzung-Yi; Chiao, Ling-Huan; Chen, Hong-Bin

    2012-09-30

    There were approximately 926 m(3) of water contaminated by fission products and actinides in the Taiwan Research Reactor's spent fuel pool (TRR SFP). The solid and ionic contaminants were thoroughly characterized using radiochemical analyses, scanning electron microscopy equipped with an energy dispersive spectrometer (SEM-EDS), and inductively coupled plasma optical emission spectrometry (ICP-OES) in this study. The sludge was made up of agglomerates contaminated by spent fuel particles. Suspended solids from spent ion-exchange resins interfered with the clarity of the water. In addition, the ionic radionuclides such as (137)Cs, (90)Sr, U, and α-emitters, present in the water were measured. Various filters and cation-exchange resins were employed for water treatment trials, and the results indicated that the solid and ionic contaminants could be effectively removed through the use of <0.9 μm filters and cation exchange resins, respectively. Interestingly, the removal of U was obviously efficient by cation exchange resin, and the ceramic depth filter composed of diatomite exhibited the properties of both filtration and adsorption. It was found that the ceramic depth filter could adsorb β-emitters, α-emitters, and uranium ions. The diatomite-based ceramic depth filter was able to simultaneously eliminate particles and adsorb ionic radionuclides from water.

  6. Design and construction of a pipeline for transfer of radioactive sodium at Argonne National Laboratory-West.

    SciTech Connect

    Baily, C. E.

    1998-02-25

    Experimental Breeder Reactor-II (EBR-II), an experimental sodium cooled fast breeder reactor located at Argonne National Laboratory-West (ANL-W), was shut down in 1994, and has since been defueled in preparation for final plant closure. Approximately 100,000 gallons of liquid sodium is contained in the primary and secondary cooling systems of the EBR-II plant. The liquid sodium must be drained from the reactor systems during closure of the plant to place the reactor plant in an industrially and radiologically safe condition for long term storage or dismantlement. Because the liquid sodium is a listed waste under the Resource Conservation Recovery Act (RCRA), it is not suitable for disposal. It therefore must be transferred to the Sodium Process Facility (SPF), which is located approximately nine hundred feet from the reactor complex, where it will be processed into a non-reactive form, suitable for land disposal in Idaho. To facilitate this transfer, a heated pipeline for carrying liquid sodium metal from EBR-II to the SPF was designed and installed. The SPF was originally designed and built to process primary sodium from the Fermi-1 reactor. The sodium is stored at ANL-W in 55 gallon drums. Design of the SPF did not originally accommodate processing of EBR-II sodium. Therefore, no method of introducing the EBR-II sodium into the process existed. As part of modifying the SPF for processing EBR-II sodium, it was necessary to design a method for transferring sodium from the EBR-H complex to the SPF and introducing it into the existing system. This requirement has been fulfilled by design and installation of a pipeline between the two facilities.

  7. Integral fast reactor safety features

    SciTech Connect

    Cahalan, J.E.; Kramer, J.M.; Marchaterre, J.F.; Mueller, C.J.; Pedersen, D.R.; Sevy, R.H.; Wade, D.C.; Wei, T.Y.C.

    1988-01-01

    The Integral Fast Reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFR development effort are improved economics and enhanced safety. In addition to liquid metal cooling, the principal design features that distinguish the IFR are: (1) a pool-type primary system, (2) an advanced ternary alloy metallic fuel, and (3) an integral fuel cycle with on-site fuel reprocessing and fabrication. This paper focuses on the technical aspects of the improved safety margins available in the IFR concept. This increased level of safety is made possible by (1) the liquid metal (sodium) coolant and pool-type primary system layout, which together facilitate passive decay heat removal, and (2) a sodium-bonded metallic fuel pin design with thermal and neutronic properties that provide passive core responses which control and mitigate the consequences of reactor accidents.

  8. Soil bed reactor work of the Environmental Research Lab. of the University of Arizona in support of the research and development of Biosphere 2

    NASA Technical Reports Server (NTRS)

    Frye, Robert

    1990-01-01

    Research at the Environmental Research Lab in support of Biosphere 2 was both basic and applied in nature. One aspect of the applied research involved the use of biological reactors for the scrubbing of trace atmospheric organic contaminants. The research involved a quantitative study of the efficiency of operation of Soil Bed Reactors (SBR) and the optimal operating conditions for contaminant removal. The basic configuration of a SBR is that air is moved through a living soil that supports a population of plants. Upon exposure to the soil, contaminants are either passively adsorbed onto the surface of soil particles, chemically transformed in the soil to usable compounds that are taken up by the plants or microbes as a metabolic energy source and converted to CO2 and water.

  9. Light Water Reactor Sustainability Research and Development Program Plan -- Fiscal Year 2009–2013

    SciTech Connect

    Idaho National Laboratory

    2009-12-01

    Nuclear power has reliably and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. By the year 2030, domestic demand for electrical energy is expected to grow to levels of 16 to 36% higher than 2007 levels. At the same time, most currently operating nuclear power plants will begin reaching the end of their 60-year operating licenses. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline—even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary this year. U.S. regulators have begun considering extended operations of nuclear power plants and the research needed to support long-term operations. The Light Water Reactor Sustainability (LWRS) Research and Development (R&D) Program, developed and sponsored by the Department of Energy, is performed in close collaboration with industry R&D programs. The purpose of the LWRS R&D Program is to provide technical foundations for licensing and managing long-term, safe and economical operation of the current operating nuclear power plants. The LWRS R&D Program vision is captured in the following statements: Existing operating nuclear power plants will continue to safely provide clean and economic electricity well beyond their first license- extension period, significantly contributing to reduction of United States and global carbon emissions, enhancement of national energy security, and protection of the environment. There is a comprehensive technical basis for licensing and managing the long-term, safe, economical operation of nuclear power plants. Sustaining the existing operating U.S. fleet also will improve its international engagement

  10. Results of Coupling a Thermal-Hydraulic Test Loop and University Research Reactor

    SciTech Connect

    Cetiner, Sacit M.; Edwards, Robert M.

    2002-07-01

    The coupling of a university thermal-hydraulic test loop and a simulated reactor is presented. The thermal-hydraulic test loop used in this work is a one-half height scaled version of General Electric's Simplified Boiling Water Reactor (SBWR). The digitally simulated reactor exploits modal neutron kinetics equations up to the first harmonic, and governing equations are not linearized. The preserved nonlinearity makes the simulated reactor behave more realistically, and eigenfunction expansion to the first order lets half of the core be represented independently. A series of experiments are performed with the hybrid system including simulated control rod reactivity insertion/withdrawal, cross-mode interaction, etc. The experimental results are compared with the theoretical expectations. (authors)

  11. Advanced reactor safety research quarterly report, October-December 1982. Volume 24

    SciTech Connect

    1984-04-01

    This report describes progress in a number of activities dealing with current safety issues relevant to both light water reactors (LWRs) and breeder reactors. The work includes a broad range of experiments to simulate accidental conditions to provide the required data base to understand important accident sequences and to serve as a basis for development and verification of the complex computer simulation models and codes used in accident analysis and licensing reviews. Such a program must include the development of analytical models, verified by experiment, which can be used to predict reactor and safety system performance under a broad variety of abnormal conditions. Current major emphasis is focused on providing information to NRC relevant to (1) its deliberations and decisions dealing with severe LWR accidents and (2) its safety evaluation of the proposed Clinch River Breeder Reactor.

  12. Research on acceleration method of reactor physics based on FPGA platforms

    SciTech Connect

    Li, C.; Yu, G.; Wang, K.

    2013-07-01

    The physical designs of the new concept reactors which have complex structure, various materials and neutronic energy spectrum, have greatly improved the requirements to the calculation methods and the corresponding computing hardware. Along with the widely used parallel algorithm, heterogeneous platforms architecture has been introduced into numerical computations in reactor physics. Because of the natural parallel characteristics, the CPU-FPGA architecture is often used to accelerate numerical computation. This paper studies the application and features of this kind of heterogeneous platforms used in numerical calculation of reactor physics through practical examples. After the designed neutron diffusion module based on CPU-FPGA architecture achieves a 11.2 speed up factor, it is proved to be feasible to apply this kind of heterogeneous platform into reactor physics. (authors)

  13. Safety evaluation report related to the evaluation of low-enriched uranium silicide-aluminum dispersion fuel for use in non-power reactors

    SciTech Connect

    Not Available

    1988-07-01

    Low-enriched uranium silicide-aluminum dispersion plate-type fuels have been extensively researched and developed under the international program, Reduced Enrichment in Research and Test Reactors. The international effort was led by Argonne National Laboratory (ANL) in the United States. This evaluation is based primarily on reports issued by ANL that discuss and summarize the developmental tests and experiments, including postirradiation examinations, of both miniature and full-sized plates of prototypical fuel compositions. This evaluation concludes that plate-type fuels suitable and acceptable for use in research and test reactors can be fabricated with U/sub 3/Si/sub 2/-Al dispersion compacts with uranium densities up to 4.8 g/cm/sup 3/. 4 refs., 1 fig.

  14. [Research on change process of nitrosation granular sludge in continuous stirred-tank reactor].

    PubMed

    Yin, Fang-Fang; Liu, Wen-Ru; Wang, Jian-Fang; Wu, Peng; Shen, Yao-Liang

    2014-11-01

    In order to investigate the effect of different types of reactors on the nitrosation granular sludge, a continuous stirred-tank reactor (CSTR) was studied, using mature nitrosation granular sludge cultivated in sequencing batch reactor (SBR) as seed sludge. Results indicated that the change of reactor type and influent mode could induce part of granules to lose stability with gradual decrease in sludge settling ability during the initial period of operation. However, the flocs in CSTR achieved fast granulation in the following reactor operation. In spite of the changes of particle size distribution, e. g. the decreasing number of granules with diameter larger than 2.5 mm and the increasing number of granules with diameter smaller than 0.3 mm, granular sludge held the absolute predominance of sludge morphology in CSTR during the entire experimental period. Moreover, results showed that the change of reactor type and influent mode didn't affect the nitrite accumulation rate which was still kept at about 85% in effluent. Additionally, the average activity of the sludge in CSTR was stronger than that of the seed sludge, because the newly generated small particles in CSTR had higher specific reactive activity than the larger granules.

  15. Development of refined MCNPX-PARET multi-channel model for transient analysis in research reactors

    SciTech Connect

    Kalcheva, S.; Koonen, E.; Olson, A. P.

    2012-07-01

    Reactivity insertion transients are often analyzed (RELAP, PARET) using a two-channel model, representing the hot assembly with specified power distribution and an average assembly representing the remainder of the core. For the analysis of protected by the reactor safety system transients and zero reactivity feedback coefficients this approximation proves to give adequate results. However, a more refined multi-channel model representing the various assemblies, coupled through the reactivity feedback effects to the whole reactor core is needed for the analysis of unprotected transients with excluded over power and period trips. In the present paper a detailed multi-channel PARET model has been developed which describes the reactor core in different clusters representing typical BR2 fuel assemblies. The distribution of power and reactivity feedback in each cluster of the reactor core is obtained from a best-estimate MCNPX calculation using the whole core geometry model of the BR2 reactor. The sensitivity of the reactor response to power, temperature and energy distributions is studied for protected and unprotected reactivity insertion transients, with zero and non-zero reactivity feedback coefficients. The detailed multi-channel model is compared vs. simplified fewer-channel models. The sensitivities of transient characteristics derived from the different models are tested on a few reactivity insertion transients with reactivity feedback from coolant temperature and density change. (authors)

  16. Argonne's Laboratory Computing Resource Center : 2005 annual report.

    SciTech Connect

    Bair, R. B.; Coghlan, S. C; Kaushik, D. K.; Riley, K. R.; Valdes, J. V.; Pieper, G. P.

    2007-06-30

    Argonne National Laboratory founded the Laboratory Computing Resource Center in the spring of 2002 to help meet pressing program needs for computational modeling, simulation, and analysis. The guiding mission is to provide critical computing resources that accelerate the development of high-performance computing expertise, applications, and computations to meet the Laboratory's challenging science and engineering missions. The first goal of the LCRC was to deploy a mid-range supercomputing facility to support the unmet computational needs of the Laboratory. To this end, in September 2002, the Laboratory purchased a 350-node computing cluster from Linux NetworX. This cluster, named 'Jazz', achieved over a teraflop of computing power (10{sup 12} floating-point calculations per second) on standard tests, making it the Laboratory's first terascale computing system and one of the fifty fastest computers in the world at the time. Jazz was made available to early users in November 2002 while the system was undergoing development and configuration. In April 2003, Jazz was officially made available for production operation. Since then, the Jazz user community has grown steadily. By the end of fiscal year 2005, there were 62 active projects on Jazz involving over 320 scientists and engineers. These projects represent a wide cross-section of Laboratory expertise, including work in biosciences, chemistry, climate, computer science, engineering applications, environmental science, geoscience, information science, materials science, mathematics, nanoscience, nuclear engineering, and physics. Most important, many projects have achieved results that would have been unobtainable without such a computing resource. The LCRC continues to improve the computational science and engineering capability and quality at the Laboratory. Specific goals include expansion of the use of Jazz to new disciplines and Laboratory initiatives, teaming with Laboratory infrastructure providers to develop

  17. Evaluate the radioactivity along the central thimble hole of a decommissioned heavy water research reactor using TLD approach.

    PubMed

    Lee, Lun-Hui; Sher, Hai-Feng; Lu, I-Hsin; Pan, Lung-Kwang

    2012-04-01

    The radioactivity along the central thimble hole of a decommissioned heavy water research reactor, TRR, was evaluated using TLD approach. The decay radionuclide was verified to be Co-60. The dose along the TRR central thimble hole was detected and revised by performing an unfolding analysis. The revised data reduced to 70-90% of the original data (for example, the maximum dose rate was reduced from 6447 to 4831 mSv/h,) and were more reliable.

  18. The Development of Mobile Melt-Dilute Technology for the Treatment of Former Soviet Union Research Reactor Fuel

    SciTech Connect

    Adams, T.

    2003-10-09

    On-site application of the MMD process offers an economical method for converting weapons usable Former Soviet Union (FSU) High Enriched Uranium (HEU) research reactor fuel to a safe and secure Low Enriched Uranium (LEU) ingot. The objective of the MMD Project is to develop the mobile melt and dilute technology in preparation for active equipment deployment in the Newly Independent States (NIS) of the FSU.

  19. Argonne Out Loud: Computation, Big Data, and the Future of Cities

    SciTech Connect

    Catlett, Charlie

    2014-09-23

    Charlie Catlett, a Senior Computer Scientist at Argonne and Director of the Urban Center for Computation and Data at the Computation Institute of the University of Chicago and Argonne, talks about how he and his colleagues are using high-performance computing, data analytics, and embedded systems to better understand and design cities.

  20. Argonne Out Loud: Computation, Big Data, and the Future of Cities

    ScienceCinema

    Catlett, Charlie

    2016-07-12

    Charlie Catlett, a Senior Computer Scientist at Argonne and Director of the Urban Center for Computation and Data at the Computation Institute of the University of Chicago and Argonne, talks about how he and his colleagues are using high-performance computing, data analytics, and embedded systems to better understand and design cities.

  1. Development of a New Transportation/Storage Cask System for Use by the DOE Russian Research Reactor Fuel Return Program

    SciTech Connect

    Michael J. Tyacke; Frantisek Svitak; Jiri Rychecky; Miroslav Picek; Alexey Smirnov; Sergey Komarov; Edward Bradley; Alexander Dudchenko; Konstantin Golubkin

    2007-10-01

    The United States, the Russian Federation, and the International Atomic Energy Agency (IAEA) have been working together on a program called the Russian Research Reactor Fuel Return (RRRFR) Program. The purpose of this program is to return Soviet or Russian-supplied high-enriched uranium (HEU) fuel, currently stored at Russian-designed research reactors throughout the world, to Russia. To accommodate transport of the HEU spent nuclear fuel (SNF), a new large-capacity transport/storage cask system was specially designed for handling and operations under the unique conditions at these research reactor facilities. This new cask system is named the ŠKODA VPVR/M cask. The design, licensing, testing, and delivery of this new cask system result from a significant international cooperative effort by several countries and involved numerous private and governmental organizations. This paper contains the following sections: 1) Introduction; 2) VPVR/M Cask Description; 3) Ancillary Equipment, 4) Cask Licensing; 5) Cask Demonstration and Operations; 6) IAEA Procurement, Quality Assurance Inspections, Fabrication, and Delivery; and, 7) Conclusions.

  2. Advanced Test Reactor Tour

    SciTech Connect

    Miley, Don

    2011-01-01

    The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored. For more information about INL research, visit http://www.facebook.com/idahonationallaboratory.

  3. Advanced Test Reactor Tour

    ScienceCinema

    Miley, Don

    2016-07-12

    The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored. For more information about INL research, visit http://www.facebook.com/idahonationallaboratory.

  4. Users Handbook for the Argonne Premium Coal Sample Program

    SciTech Connect

    Vorres, K.S.

    1993-10-01

    This Users Handbook for the Argonne Premium Coal Samples provides the recipients of those samples with information that will enhance the value of the samples, to permit greater opportunities to compare their work with that of others, and aid in correlations that can improve the value to all users. It is hoped that this document will foster a spirit of cooperation and collaboration such that the field of basic coal chemistry may be a more efficient and rewarding endeavor for all who participate. The different sections are intended to stand alone. For this reason some of the information may be found in several places. The handbook is also intended to be a dynamic document, constantly subject to change through additions and improvements. Please feel free to write to the editor with your comments and suggestions.

  5. Accident Analysis for the NIST Research Reactor Before and After Fuel Conversion

    SciTech Connect

    Baek J.; Diamond D.; Cuadra, A.; Hanson, A.L.; Cheng, L-Y.; Brown, N.R.

    2012-09-30

    Postulated accidents have been analyzed for the 20 MW D2O-moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analysis has been carried out for the present core, which contains high enriched uranium (HEU) fuel and for a proposed equilibrium core with low enriched uranium (LEU) fuel. The analyses employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron transport calculations were performed with the MCNPX code to determine homogenized fuel compositions in the lower and upper halves of each fuel element and to determine the resulting neutronic properties of the core. The accident analysis employed a model of the primary loop with the RELAP5 code. The model includes the primary pumps, shutdown pumps outlet valves, heat exchanger, fuel elements, and flow channels for both the six inner and twenty-four outer fuel elements. Evaluations were performed for the following accidents: (1) control rod withdrawal startup accident, (2) maximum reactivity insertion accident, (3) loss-of-flow accident resulting from loss of electrical power with an assumption of failure of shutdown cooling pumps, (4) loss-of-flow accident resulting from a primary pump seizure, and (5) loss-of-flow accident resulting from inadvertent throttling of a flow control valve. In addition, natural circulation cooling at low power operation was analyzed. The analysis shows that the conversion will not lead to significant changes in the safety analysis and the calculated minimum critical heat flux ratio and maximum clad temperature assure that there is adequate margin to fuel failure.

  6. Pressure Loss Predictions of the Reactor Simulator Subsystem at NASA Glenn Research Center

    NASA Technical Reports Server (NTRS)

    Reid, Terry V.

    2016-01-01

    Testing of the Fission Power System (FPS) Technology Demonstration Unit (TDU) is being conducted at NASA Glenn Research Center. The TDU consists of three subsystems: the reactor simulator (RxSim), the Stirling Power Conversion Unit (PCU), and the heat exchanger manifold (HXM). An annular linear induction pump (ALIP) is used to drive the working fluid. A preliminary version of the TDU system (which excludes the PCU for now) is referred to as the "RxSim subsystem" and was used to conduct flow tests in Vacuum Facility 6 (VF 6). In parallel, a computational model of the RxSim subsystem was created based on the computer-aided-design (CAD) model and was used to predict loop pressure losses over a range of mass flows. This was done to assess the ability of the pump to meet the design intent mass flow demand. Measured data indicates that the pump can produce 2.333 kg/sec of flow, which is enough to supply the RxSim subsystem with a nominal flow of 1.75 kg/sec. Computational predictions indicated that the pump could provide 2.157 kg/sec (using the Spalart-Allmaras (S?A) turbulence model) and 2.223 kg/sec (using the k- turbulence model). The computational error of the predictions for the available mass flow is ?0.176 kg/sec (with the S-A turbulence model) and -0.110 kg/sec (with the k- turbulence model) when compared to measured data.

  7. Quantitative Imaging Evaluation of Corrosion in Oak Ridge Research Reactor POOL

    SciTech Connect

    Jang, P.R.; Arunkumar, R.; Long, Z.; Mott, M.A.; Okhuysen, W.P.; Su, Y.; Monts, D.L.; Kirk, P.G.; Ettien, J.

    2006-07-01

    The Oak Ridge Research Reactor (ORRR) was operated as an isotope production and irradiation facility from March 1958 until March 1987. The US Department of Energy permanently shut down and removed the fuel from the ORRR in 1987. The water level must be maintained in the ORRR pool as shielding for radioactive components still located in the pool. The U.S. Department of Energy's Environmental Management (DOE EM) needs to decontaminate and demolish the ORRR as part of the Oak Ridge cleanup program. In February 2004, increased pit corrosion was noted in the pool's 6-mm (1/4'')-thick aluminum liner in the section nearest where the radioactive components are stored. If pit corrosion has significantly penetrated the aluminum liner, then DOE EM must accelerate its decontamination and demolition (D and D) efforts. The goal of Mississippi State University's Diagnostic Instrumentation and Analysis Laboratory (DIAL) is to provide a determination of the extent and depth of corrosion. Results from the work will facilitate ORNL in making reliable disposition decisions. DIAL's inspection approach is to quantitatively estimate the amount of corrosion by using Fourier-transform profilometry (FTP). FTP is a non-contact 3-D shape measurement technique. By projecting a fringe pattern onto a target surface and observing its deformation due to surface irregularities from a different view angle, FTP is capable of determining the height (depth) distribution of the target surface, thus reproducing the profile of the target accurately. DIAL has previously demonstrated that its FTP system can quantitatively estimate the volume and depth of removed and residual material to high accuracy. (authors)

  8. Expansion of the radioactive ion beam program at Argonne

    NASA Astrophysics Data System (ADS)

    Clark, J. A.

    2011-01-01

    The Argonne Tandem Linear Accelerator System (ATLAS) at Argonne National Laboratory (ANL) provides a wide range of stable ion beams and radioactive beams which have contributed to our understanding of nuclear structure and reactions. Until now, most radioactive ion beams at ATLAS were produced in flight using light-ion reactions such as (p, n), (d, n), (d, p), (d,3He), and (3He,n). Within the next few months, the radioactive ion beam program at ATLAS will acquire much extended, new capabilities with the commissioning of a new facility: the CAlifornium Rare Isotope Breeder Upgrade (CARIBU). CARIBU will supply ion beams of 252Cf fission fragments, which are thermalized in a gas catcher. The singly- and doubly-charged ions extracted from the gas catcher will be mass-separated and either delivered to a low-energy experimental area, or charge bred with a modified ECR source and subsequently reaccelerated by the ATLAS facility. Properties of hundreds of these neutron-rich nuclides will be investigated using ion traps, decay stations, the newly commissioned HELical Orbit Spectrometer (HELIOS), and other available experimental equipment such as Gammasphere and the FMA. HELIOS was constructed to take advantage of rare ion beams, such as those provided by CARIBU, through light-ion transfer reactions in inverse kinematics, and represents a new approach to the study of direct reactions in inverse kinematics which avoids kinematic broadening. Experiments are currently being conducted with HELIOS, and first results with the d(28Si,p) and d(12B,p) reactions have shown excellent energy resolution.

  9. How Argonne's Intense Pulsed Neutron Source came to life and gained its niche : the view from an ecosystem perspective.

    SciTech Connect

    Westfall, C.; Office of The Director

    2008-02-25

    At first glance the story of the Intense Pulsed Neutron Source (IPNS) at Argonne National Laboratory (ANL) appears to have followed a puzzling course. When researchers first proposed their ideas for an accelerator-driven neutron source for exploring the structure of materials through neutron scattering, the project seemed so promising that both Argonne managers and officials at the laboratory's funding agency, the Department of Energy (DOE), suggested that it be made larger and more expensive. But then, even though prototype building, testing, and initial construction went well a group of prominent DOE reviewers recommended in fall 1980 that it be killed, just months before it had been slated to begin operation, and DOE promptly accepted the recommendation. In response, Argonne's leadership declared the project was the laboratory's top priority and rallied to save it. In late 1982, thanks to another review panel led by the same scientist who had chaired the panel that had delivered the death sentence, the project was granted a reprieve. However, by the late 1980s, the IPNS was no longer top priority within the international materials science community, at Argonne, or within the DOE budget because prospects for another, larger materials science accelerator emerged. At just this point, the facility started to produce exciting scientific results. For the next two decades, the IPNS, its research, and its experts became valued resources at Argonne, within the U.S. national laboratory system, and within the international materials science community. Why did this Argonne project prosper and then almost suffer premature death, even though it promised (and later delivered) good science? How was it saved and how did it go on to have a long, prosperous life for more than a quarter of a century? In particular, what did an expert assessment of the quality of IPNS science have to do with its fate? Getting answers to such questions is important. The U.S. government spends a lot

  10. Reactor-safety research programs. Quarterly report, July-September 1982

    SciTech Connect

    Edler, S.K.

    1983-03-01

    Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions.

  11. Reactor-safety research programs. Quarterly report, October-December 1982. Volume 4

    SciTech Connect

    Edler, S.K.

    1983-04-01

    Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized-water-reactor steam-generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models being developed to provide better digital codes to compute the bahavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities.

  12. Thermal Hydraulic Characteristics of Fuel Defects in Plate Type Nuclear Research Reactors

    SciTech Connect

    Bodey, Isaac T

    2014-05-01

    Turbulent flow coupled with heat transfer is investigated for a High Flux Isotope Reactor (HFIR) fuel plate. The Reynolds Averaged Navier-Stokes Models are used for fluid dynamics and the transfer of heat from a thermal nuclear fuel plate using the Multi-physics code COMSOL. Simulation outcomes are compared with experimental data from the Advanced Neutron Source Reactor Thermal Hydraulic Test Loop. The computational results for the High Flux Isotope Reactor core system provide a more physically accurate simulation of this system by modeling the turbulent flow field in conjunction with the diffusion of thermal energy within the solid and fluid phases of the model domain. Recommendations are made regarding Nusselt number correlations and material properties for future thermal hydraulic modeling efforts

  13. Safety evaluation report related to the renewal of the operating license for the research reactor at North Carolina State University

    SciTech Connect

    1997-04-01

    This safety evaluation report (SER) summarizes the findings of a safety review conducted by the staff of the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Reactor Regulation (NRR). The staff conducted this review in response to a timely application filed by North Carolina State University (the licensee or NCSU) for a 20-year renewal of Facility Operating License R-120 to continue to operate the NCSU PULSTAR research reactor. The facility is located in the Burlington Engineering Laboratory complex on the NCSU campus in Raleigh, North Carolina. In its safety review, the staff considered information submitted by the licensee (including past operating history recorded in the licensee`s annual reports to the NRC), as well as inspection reports prepared by NRC Region H personnel and first-hand observations. On the basis of this review, the staff concludes that NCSU can continue to operate the PULSTAR research reactor, in accordance with its application, without endangering the health and safety of the public. 16 refs., 31 figs., 7 tabs.

  14. Collaborative Physical and Biological Dosimetry Studies for Neutron Capture Therapy at the RA-1 Research Reactor Facility

    SciTech Connect

    Nigg, D.W.; Schwint, A.E.; Hartwell, J.K.; Heber, E.M.; Trivillin, V.; Castillo, J.; Wentzeis, L.; Sloan, P.; Wemple, C.A.

    2004-10-04

    Initial physical dosimetry measurements have been completed using activation spectrometry and thermoluminiscent dosimeters to characterize the BNCT irradiation facility developed at the RA-1 research reactor operated by the Argentine National Atomic Energy Commission in Buenos Aires. Some biological scoping irradiations have also been completed using a small-animal (hamster) oral mucosa tumor model. Results indicate that the RA-1 neutron source produces useful dose rates but that some improvements in the initial configuration will be needed to optimize the spectrum for thermal-neutron BNCT research applications.

  15. Collaborative Physical and Biological Dosimetry Studies for Neutron Capture Therapy at the RA-1 Research Reactor Facility

    SciTech Connect

    David W. Nigg; Amanda E. Schwint; John K. Hartwell; Elisa M. Heber; Veronica Trivillin; Jorge Castillo; Luis Wentzeis; Patrick Sloan; Charles A. Wemple

    2004-10-01

    Initial physical dosimetry measurements have been completed using activation spectrometry and thermoluminiscent dosimeters to characterize the BNCT irradiation facility developed at the RA-1 research reactor operated by the Argentine National Atomic Energy Commission in Buenos Aires. Some biological scoping irradiations have also been completed using a small-animal (hamster) oral mucosa tumor model. Results indicate that the RA-1 neutron source produces useful dose rates but that some improvements in the initial configuration will be needed to optimize the spectrum for thermal-neutron BNCT research applications.

  16. Neutron spectrum measurements at a radial beam port of the NUR research reactor using a Bonner spheres spectrometer.

    PubMed

    Mazrou, H; Nedjar, A; Seguini, T

    2016-08-01

    This paper describes the measurement campaign held around the neutron radiography (NR) facility of the Algerian 1MW NUR research reactor. The main objective of this work is to characterize accurately the neutron beam provided at one of the radial channels of the NUR research reactor taking benefit of the acquired CRNA Bonner spheres spectrometer (BSS). The specific objective was to improve the image quality of the NR facility. The spectrometric system in use is based on a central spherical (3)He thermal neutron proportional counter combined with high density polyethylene spheres of different diameters ranging from 3 to 12in. This counting system has good gamma ray discrimination and is able to cover an energy range from thermal to 20MeV. The measurements were performed at the sample distance of 0.6m from the beam port and at a height of 1.2m from the facility floor. During the BSS measurements, the reactor was operating at low power (100W) to avoid large dead times, pulse pileup and high level radiation exposures, in particular, during spheres handling. Thereafter, the neutron spectrum at the sample position was unfolded by means of GRAVEL and MAXED computer codes. The thermal, epithermal and fast neutron fluxes, the total neutron flux, the mean energy and the Cadmium ratio (RCd) were provided. A sensitivity analysis was performed taking into account various defaults spectra and ultimately a different response functions in the unfolding procedure. Overall, from the obtained results it reveals, unexpectedly, that the measured neutron spectrum at the sample position of the neutron radiography of the NUR reactor is being harder with a predominance of fast neutrons (>100keV) by about 60%. Finally, those results were compared to previous and more recent measurements obtained by activation foils detectors. The agreement was fairly good highlighting thereby the consistency of our findings.

  17. Gaseous-fuel nuclear reactor research for multimegawatt power in space

    NASA Technical Reports Server (NTRS)

    Thom, K.; Schneider, R. T.; Helmick, H. H.

    1977-01-01

    In the gaseous-fuel reactor concept, the fissile material is contained in a moderator-reflector cavity and exists in the form of a flowing gas or plasma separated from the cavity walls by means of fluid mechanical forces. Temperatures in excess of structural limitations are possible for low-specific-mass power and high-specific-impulse propulsion in space. Experiments have been conducted with a canister filled with enriched UF6 inserted into a beryllium-reflected cavity. A theoretically predicted critical mass of 6 kg was measured. The UF6 was also circulated through this cavity, demonstrating stable reactor operation with the fuel in motion. Because the flowing gaseous fuel can be continuously processed, the radioactive waste in this type of reactor can be kept small. Another potential of fissioning gases is the possibility of converting the kinetic energy of fission fragments directly into coherent electromagnetic radiation, the nuclear pumping of lasers. Numerous nuclear laser experiments indicate the possibility of transmitting power in space directly from fission energy. The estimated specific mass of a multimegawatt gaseous-fuel reactor power system is from 1 to 5 kg/kW while the companion laser-power receiver station would be much lower in specific mass.

  18. [Experimental research on combined water and air backwashing reactor technology for biological activated carbon].

    PubMed

    Xie, Zhi-Gang; Qiu, Xue-Min; Zhao, Yan-Ling

    2012-01-01

    To proper control the backwashing process of biological activated carbon (BAC) reactor and improve the overall operation performance, the evaluative indexes such as backwashing wastewater turbidity, organic pollutants removal rate of pre and post-backwashing, and the variation of biomass and biological activity in carbon column are used to compare and analyze the effect of three different combined water and air backwashing methods on the operation of BAC reactor. The result shows that intermittent combined water and air backwashing method is most suitable to BAC reactor. The biological activaty obviously increases by 62.5% after intermittent combined water and air backwashing process. While, the biological activaty using the backwashing method of air plus water and the backwashing method of water and air compounded plus water washing increases by 55.6%, 38.5%, respectively. After backwashing 308h, the reactor recovered to its normal function after intermittent combined water and air backwashing process with the removal rate of UV254 reaching to 60.0%. The fulvic-like fluorescence peak of backwashing water are very weak, and are characterized by low-excitation wavelength tryptophan like (peak S) and high excitation wavelength of tryptophan (peak T), which are caused by the microbial debris washed down. The three-dimensional fluorescence spectra also show that microbial fragments are easy to be washed clean with intermittent combined water and air backwashing.

  19. 77 FR 42771 - License Renewal for the Dow Chemical TRIGA Research Reactor

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-07-20

    ... nuclide found in the gaseous effluent stream is Argon-41. The licensee performed calculations to estimate the production level of Argon-41 associated with operation of the reactor at high power. The licensee's calculations indicated that annual Argon-41 release would result in an offsite concentration of...

  20. Advanced reactor safety research. Quarterly report, April-June 1982. Volume 22

    SciTech Connect

    1983-10-01

    Overall objective of this work is to provide NRC a comprehensive data base essential to (1) defining key safety issues, (2) understanding risk-significant accident sequences, (3) developing and verifying models used in safety assessments, and (4) assuring the public that power reactor systems will not be licensed and placed in commercial service in the United States without appropriate consideration being given to their effects on health and safety. This report describes progress in a number of activities dealing with current safety issues relevant to both light water and breeder reactors. The work includes a broad range of experiments to simulate accidental conditions to provide the required data base to understand important accident sequences and to serve as a basis for development and verification of the complex computer simulation models and codes used in accident analysis and licensing reviews. Such a program must include the development of analytical models, verified by experiment, which can be used to predict reactor and safety system performance under a broad variety of abnormal conditions. Current major emphasis is focused on providing information to NRC relevant to (1) its deliberations and decisions dealing with severe LWR accidents, and (2) its safety evaluation of the proposed Clinch River Breeder Reactor.